Application of computational fluid dynamics methods to improve thermal hydraulic code analysis
Sentell, Dennis Shannon, Jr.
A computational fluid dynamics code is used to model the primary natural circulation loop of a proposed small modular reactor for comparison to experimental data and best-estimate thermal-hydraulic code results. Recent advances in computational fluid dynamics code modeling capabilities make them attractive alternatives to the current conservative approach of coupled best-estimate thermal hydraulic codes and uncertainty evaluations. The results from a computational fluid dynamics analysis are benchmarked against the experimental test results of a 1:3 length, 1:254 volume, full pressure and full temperature scale small modular reactor during steady-state power operations and during a depressurization transient. A comparative evaluation of the experimental data, the thermal hydraulic code results and the computational fluid dynamics code results provides an opportunity to validate the best-estimate thermal hydraulic code's treatment of a natural circulation loop and provide insights into expanded use of the computational fluid dynamics code in future designs and operations. Additionally, a sensitivity analysis is conducted to determine those physical phenomena most impactful on operations of the proposed reactor's natural circulation loop. The combination of the comparative evaluation and sensitivity analysis provides the resources for increased confidence in model developments for natural circulation loops and provides for reliability improvements of the thermal hydraulic code.
Assessment of uncertainties of the models used in thermal-hydraulic computer codes
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.
Directory of Open Access Journals (Sweden)
JUN YEOB LEE
2014-10-01
Full Text Available During the development process of a thermal-hydraulic system code, a non-regression test (NRT must be performed repeatedly in order to prevent software regression. The NRT process, however, is time-consuming and labor-intensive. Thus, automation of this process is an ideal solution. In this study, we have developed a program to support an efficient NRT for the SPACE code and demonstrated its usability. This results in a high degree of efficiency for code development. The program was developed using the Visual Basic for Applications and designed so that it can be easily customized for the NRT of other computer codes.
SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis
Energy Technology Data Exchange (ETDEWEB)
Basehore, K.L.; Todreas, N.E.
1980-08-01
Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries.
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Proskuryakov, K.N.; Bogomazov, D.N.; Poliakov, N. [Moscow Power Engineering Institute (Technical University), Moscow (Russian Federation)
2007-07-01
The new special module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation is worked out. The Russian computer code Rainbow has been selected for joint use with a developed module. This code system provides the possibility of EFOCP (Eigen Frequencies of Oscillations of the Coolant Pressure) calculations in any coolant acoustical elements of primary circuits of NPP. EFOCP values have been calculated for transient and for stationary operating. The calculated results for nominal operating were compared with results of measured EFOCP. For example, this comparison was provided for the system: 'pressurizer + surge line' of a WWER-1000 reactor. The calculated result 0.58 Hz practically coincides with the result of measurement (0.6 Hz). The EFOCP variations in transients are also shown. The presented results are intended to be useful for NPP vibration-acoustical certification. There are no serious difficulties for using this module with other computer codes.
Development of a computer code for thermal hydraulics of reactors (THOR). [BWR and PWR
Energy Technology Data Exchange (ETDEWEB)
Wulff, W
1975-01-01
The purpose of the advanced code development work is to construct a computer code for the prediction of thermohydraulic transients in water-cooled nuclear reactor systems. The fundamental formulation of fluid dynamics is to be based on the one-dimensional drift flux model for non-homogeneous, non-equilibrium flows of two-phase mixtures. Particular emphasis is placed on component modeling, automatic prediction of initial steady state conditions, inclusion of one-dimensional transient neutron kinetics, freedom in the selection of computed spatial detail, development of reliable constitutive descriptions, and modular code structure. Numerical solution schemes have been implemented to integrate simultaneously the one-dimensional transient drift flux equations. The lumped-parameter modeling analyses of thermohydraulic transients in the reactor core and in the pressurizer have been completed. The code development for the prediction of the initial steady state has been completed with preliminary representation of individual reactor system components. A program has been developed to predict critical flow expanding from a dead-ended pipe; the computed results have been compared and found in good agreement with idealized flow solutions. Transport properties for liquid water and water vapor have been coded and verified.
Gudoshnikov, A. N.; Migrov, Yu. A.
2008-11-01
Calculations to verify the Russian computer code KORSAR were carried out for the B4.1 experimental operating conditions, in which nitrogen was supplied to the reactor coolant (primary) circuit of a reactor plant model, and which were simulated at the PKL III integral test facility. It is shown that dissolution of gases in coolant has an essential effect on the thermal-hydraulic processes during long-term passive removal of heat from the primary to secondary coolant circuit of the reactor plant model under the conditions of natural circulation.
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Bandini, B.R. [Los Alamos National Lab., NM (United States)
1990-05-01
No present light water reactor accident analysis code employs both high state of the art neutronics and thermal-hydraulics computational algorithms. Adding a modern three-dimensional neutron kinetics model to the present TRAC-PFI/MOD2 code would create a fully up to date pressurized water reactor accident evaluation code. After reviewing several options, it was decided that the Nodal Expansion Method would best provide the basis for this multidimensional transient neutronic analysis capability. Steady-state and transient versions of the Nodal Expansion Method were coded in both three-dimensional Cartesian and cylindrical geometries. In stand-alone form this method of solving the few group neutron diffusion equations was shown to yield efficient and accurate results for a variety of steady-state and transient benchmark problems. The Nodal Expansion Method was then incorporated into TRAC-PFl/MOD2. The combined NEM/TRAC code results agreed well with the EPRI-ARROTTA core-only transient analysis code when modelling a severe PWR control rod ejection accident.
COBRA-SFS (Spent Fuel Storage): A thermal-hydraulic analysis computer code: Volume 2, User's manual
Energy Technology Data Exchange (ETDEWEB)
Rector, D.R.; Cuta, J.M.; Lombardo, N.J.; Michener, T.E.; Wheeler, C.L.
1986-11-01
COBRA-SFS (Spent Fuel Storage) is a general thermal-hydraulic analysis computer code used to predict temperatures and velocities in a wide variety of systems. The code was refined and specialized for spent fuel storage system analyses for the US Department of Energy's Commercial Spent Fuel Management Program. The finite-volume equations governing mass, momentum, and energy conservation are written for an incompressible, single-phase fluid. The flow equations model a wide range of conditions including natural circulation. The energy equations include the effects of solid and fluid conduction, natural convection, and thermal radiation. The COBRA-SFS code is structured to perform both steady-state and transient calculations; however, the transient capability has not yet been validated. This volume contains the input instructions for COBRA-SFS and an auxiliary radiation exchange factor code, RADX-1. It is intended to aid the user in becoming familiar with the capabilities and modeling conventions of the code.
Energy Technology Data Exchange (ETDEWEB)
Virtanen, E.
1995-12-31
In the study three loss-of-feedwater type experiments which were preformed with the PACTEL facility has been calculated with two computer codes. The purpose of the experiments was to gain information about the behaviour of horizontal steam generator in a situation where the water level on the secondary side of the steam generator is decreasing. At the same time data that can be used in the assessment of thermal-hydraulic computer codes was assembled. The purpose of the work was to study the capabilities of two computer codes, APROS version 2.11 and RELAP5/MOD3.1, to calculate the phenomena in horizontal steam generator. In order to make the comparison of the calculation results easier the same kind of model of the steam generator was made for both codes. Only the steam generator was modelled, the rest of the facility was given for the codes as a boundary condition. (23 refs.).
Tiyapun, K.; Wetchagarun, S.
2017-06-01
The neutronic analysis of TRIGA Mark II reactor has been performed. A detailed model of the reactor core was conducted including standard fuel elements, fuel follower control rods, and irradiation devices. As the approach to safety nuclear design are based on determining the criticality (keff), reactivity worth, reactivity excess, hot rod power factor and power peaking of the reactor, the MCNPX code had been used to calculate the nuclear parameters for different core configuration designs. The thermal-hydraulic model has been developed using COOLOD-N2 for steady state, using the nuclear parameters and power distribution results from MCNPX calculation. The objective of the thermal-hydraulic model is to determine the thermal safety margin and to ensure that the fuel integrity is maintained during steady state as well as during abnormal condition at full power. The hot channel fuel centerline temperature, fuel surface temperature, cladding surface temperature, the departure from nucleate boiling (DNB) and DNB ratio were determined. The good agreement between experimental data and simulation concerning reactor criticality proves the reliability of the methodology of analysis from neutronic and thermal hydraulic perspective.
Energy Technology Data Exchange (ETDEWEB)
Suikkanen, P.
2009-01-15
The objective of the Masters thesis was to study guidelines and procedures for scaling of thermal hydraulic test facilities and to compare results from two test facility models and from EPR model. Aim was to get an impression of how well the studied test facilities describe the behaviour in power plant scale during accident scenarios with computer codes. Models were used to determine the influence of primary circuit mass inventory on the behaviour of the circuit. The data from test facility models represent the same phenomena as the data from EPR model. The results calculated with PKL model were also compared against PKL test facility data. They showed good agreement. Test facility data is used to validate computer codes, which are used in nuclear safety analysis. The scale of the facility has effect on the behaviour of the phenomena and therefore special care must be taken in using the data. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.; Stuart, D.S.; Thompson, S.L. [Sandia National Labs., Albuquerque, NM (United States); Hodge, S.A.; Hyman, C.R.; Sanders, R.L. [Oak Ridge National Lab., TN (United States)
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, and 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.
Energy Technology Data Exchange (ETDEWEB)
Takata, Takashi; Yamaguchi, Akira [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center
2002-12-01
A multi-component and multi-phase numerical analysis method is developed to investigate a mechanism of sodium-water reaction phenomena, which occur when pressurized water leaks from failed heat transfer tubes in the steam generator of a fast reactor. It is named SERAPHIM: Sodium-watEr Reaction Analysis PHysics of Interdisciplinary Multi-phase flow. In this code, the surface reaction model and the gas phase reaction model are implemented as a sodium-water reaction mechanism. The HSMAC method is adopted for numerical solution. A validation for compressible multi-phase flow analysis is carried out in the present paper. Two-dimensional analyses of the sodium-water reaction are also carried out and it is demonstrated that the numerical quantification of a sodium-water reaction accident by the SERAPHIM code is practicable. (author)
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
Thermal-hydraulic code selection for modular high temperature gas-cooled reactors
Energy Technology Data Exchange (ETDEWEB)
Komen, E.M.J.; Bogaard, J.P.A. van den
1995-06-01
In order to study the transient thermal-hydraulic system behaviour of modular high temperature gas-cooled reactors, the thermal-hydraulic computer codes RELAP5, MELCOR, THATCH, MORECA, and VSOP are considered at the Netherlands Energy Research Foundation ECN. This report presents the selection of the most appropriate codes. To cover the range of relevant accidents, a suite of three codes is recommended for analyses of HTR-M and MHTGR reactors. (orig.).
Current and anticipated uses of thermal hydraulic codes in Korea
Energy Technology Data Exchange (ETDEWEB)
Kim, Kyung-Doo; Chang, Won-Pyo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1997-07-01
In Korea, the current uses of thermal hydraulic codes are categorized into 3 areas. The first application is in designing both nuclear fuel and NSSS. The codes have usually been introduced based on the technology transfer programs agreed between KAERI and the foreign vendors. Another area is in the supporting of the plant operations and licensing by the utility. The third category is research purposes. In this area assessments and some applications to the safety issue resolutions are major activities using the best estimate thermal hydraulic codes such as RELAP5/MOD3 and CATHARE2. Recently KEPCO plans to couple thermal hydraulic codes with a neutronics code for the design of the evolutionary type reactor by 2004. KAERI also plans to develop its own best estimate thermal hydraulic code, however, application range is different from KEPCO developing code. Considering these activities, it is anticipated that use of the best estimate hydraulic analysis code developed in Korea may be possible in the area of safety evaluation within 10 years.
The analysis of thermal-hydraulic models in MELCOR code
Energy Technology Data Exchange (ETDEWEB)
Kim, M. H.; Hur, C.; Kim, D. K.; Cho, H. J. [POhang Univ., of Science and TECHnology, Pohang (Korea, Republic of)
1996-07-15
The objective of the present work is to verify the prediction and analysis capability of MELCOR code about the progression of severe accidents in light water reactor and also to evaluate appropriateness of thermal-hydraulic models used in MELCOR code. Comparing the results of experiment and calculation with MELCOR code is carried out to achieve the above objective. Specially, the comparison between the CORA-13 experiment and the MELCOR code calculation was performed.
INL Experimental Program Roadmap for Thermal Hydraulic Code Validation
Energy Technology Data Exchange (ETDEWEB)
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 of 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
Network coding for computing: Linear codes
Appuswamy, Rathinakumar; Karamchandani, Nikhil; Zeger, Kenneth
2011-01-01
In network coding it is known that linear codes are sufficient to achieve the coding capacity in multicast networks and that they are not sufficient in general to achieve the coding capacity in non-multicast networks. In network computing, Rai, Dey, and Shenvi have recently shown that linear codes are not sufficient in general for solvability of multi-receiver networks with scalar linear target functions. We study single receiver networks where the receiver node demands a target function of the source messages. We show that linear codes may provide a computing capacity advantage over routing only when the receiver demands a `linearly-reducible' target function. % Many known target functions including the arithmetic sum, minimum, and maximum are not linearly-reducible. Thus, the use of non-linear codes is essential in order to obtain a computing capacity advantage over routing if the receiver demands a target function that is not linearly-reducible. We also show that if a target function is linearly-reducible,...
A new thermal hydraulics code coupled to agent for light water reactor analysis
Eklund, Matthew Deric
A new numerical model for coupling a thermal hydraulics method based on the Drift Flux and Homogeneous Equilibrium Mixture (HEM) models, with a deterministic neutronics code system AGENT (Arbitrary Geometry Neutron Transport), is developed. Named the TH thermal hydraulics code, it is based on the mass continuity, momentum, and energy equations integrated with appropriate relations for liquid and vapor phasic velocities. The modified conservation equations are then evaluated in one-dimensional (1D) steady-state conditions for LWR coolant subchannel in the axial direction. This permits faster computation times without sacrificing significant accuracy, as compared to other three-dimensional (3D) codes such as RELAP5/TRACE. AGENT is a deterministic neutronics code system based on the Method of Characteristics to solve the 2D/3D neutron transport equation in current and future reactor systems. The coupling scheme between the TH and AGENT codes is accomplished by computing the normalized fission rate profile in the LWR fuel elements by AGENT. The normalized fission rate profile is then transferred to the TH thermal hydraulics code for computing the reactor coolant properties. In conjunction with the 1D axial TH code, a separate 1D radial heat transfer model within the TH code is used to determine the average fuel temperature at each node where coolant properties are calculated. These properties then are entered into Scale 6.1, a criticality analysis code, to recalculate fuel pin neutron interaction cross sections based on thermal feedback. With updated fuel neutron interaction cross sections, the fission rate profile is recalculated in AGENT, and the cycle continues until convergence is reached. The TH code and coupled AGENT-TH code are benchmarked against the TRACE reactor analysis software, showing required agreement in evaluating the basic reactor parameters.
Shapiro, Wilbur
1996-01-01
This is an overview of new and updated industrial codes for seal design and testing. GCYLT (gas cylindrical seals -- turbulent), SPIRALI (spiral-groove seals -- incompressible), KTK (knife to knife) Labyrinth Seal Code, and DYSEAL (dynamic seal analysis) are covered. CGYLT uses G-factors for Poiseuille and Couette turbulence coefficients. SPIRALI is updated to include turbulence and inertia, but maintains the narrow groove theory. KTK labyrinth seal code handles straight or stepped seals. And DYSEAL provides dynamics for the seal geometry.
Energy Technology Data Exchange (ETDEWEB)
Shamasundar, B.I.; Fehrenbach, M.E.
1981-05-01
The KUGEL computer code is designed to perform thermal/hydraulic analysis and coated-fuel particle performance calculations for axisymmetric pebble bed reactor (PBR) cores. This computer code was developed as part of a Department of Energy (DOE)-funded study designed to verify the published core performance data on PBRs. The KUGEL code is designed to interface directly with the 2DB code, a two-dimensional neutron diffusion code, to obtain distributions of thermal power, fission rate, fuel burnup, and fast neutron fluence, which are needed for thermal/hydraulic and fuel performance calculations. The code is variably dimensioned so that problem size can be easily varied. An interpolation routine allows variable mesh size to be used between the 2DB output and the two-dimensional thermal/hydraulic calculations.
Current and anticipated uses of thermal-hydraulic codes in Germany
Energy Technology Data Exchange (ETDEWEB)
Teschendorff, V.; Sommer, F.; Depisch, F.
1997-07-01
In Germany, one third of the electrical power is generated by nuclear plants. ATHLET and S-RELAP5 are successfully applied for safety analyses of the existing PWR and BWR reactors and possible future reactors, e.g. EPR. Continuous development and assessment of thermal-hydraulic codes are necessary in order to meet present and future needs of licensing organizations, utilities, and vendors. Desired improvements include thermal-hydraulic models, multi-dimensional simulation, computational speed, interfaces to coupled codes, and code architecture. Real-time capability will be essential for application in full-scope simulators. Comprehensive code validation and quantification of uncertainties are prerequisites for future best-estimate analyses.
Thermal-hydraulic interfacing code modules for CANDU reactors
Energy Technology Data Exchange (ETDEWEB)
Liu, W.S.; Gold, M.; Sills, H. [Ontario Hydro Nuclear, Toronto (Canada)] [and others
1997-07-01
The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.
Energy Technology Data Exchange (ETDEWEB)
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. Velocity 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.
Energy Technology Data Exchange (ETDEWEB)
Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.
2016-08-01
To be able to simulate the behavior of nuclear fuel under operating conditions, it is required to include all the representative loads, including the lateral hydraulic forces which were not included traditionally because of the difficulty of calculating them in a reliable way. Thanks to the advance in CFD codes, now it is possible to assess them. This study calculates the local lateral hydraulic forces, caused by the contraction and expansion of the flow due to the bow of the surrounding fuel assemblies, on of fuel assembly under typical operating conditions from a three loop Westinghouse PWR reactor. (Author)
Current and anticipated uses of the thermal hydraulics codes at the NRC
Energy Technology Data Exchange (ETDEWEB)
Caruso, R.
1997-07-01
The focus of Thermal-Hydraulic computer code usage in nuclear regulatory organizations has undergone a considerable shift since the codes were originally conceived. Less work is being done in the area of {open_quotes}Design Basis Accidents,{close_quotes}, and much more emphasis is being placed on analysis of operational events, probabalistic risk/safety assessment, and maintenance practices. All of these areas need support from Thermal-Hydraulic computer codes to model the behavior of plant fluid systems, and they all need the ability to perform large numbers of analyses quickly. It is therefore important for the T/H codes of the future to be able to support these needs, by providing robust, easy-to-use, tools that produce easy-to understand results for a wider community of nuclear professionals. These tools need to take advantage of the great advances that have occurred recently in computer software, by providing users with graphical user interfaces for both input and output. In addition, reduced costs of computer memory and other hardware have removed the need for excessively complex data structures and numerical schemes, which make the codes more difficult and expensive to modify, maintain, and debug, and which increase problem run-times. Future versions of the T/H codes should also be structured in a modular fashion, to allow for the easy incorporation of new correlations, models, or features, and to simplify maintenance and testing. Finally, it is important that future T/H code developers work closely with the code user community, to ensure that the code meet the needs of those users.
A Computational Model of Hydraulic Volume Displacement Drive
Directory of Open Access Journals (Sweden)
V. N. Pil'gunov
2014-01-01
Full Text Available The paper offers a computational model of industrial-purpose hydraulic drive with two hydraulic volume adjustable working chamber machines (pump and motor. Adjustable pump equipped with the pressure control unit can be run together with several adjustable hydraulic motors on the principle of three-phase hydraulic socket-outlet with high-pressure lines, drain, and drainage system. The paper considers the pressure-controlled hydrostatic transmission with hydraulic motor as an output link. It shows a possibility to create a saving hydraulic drive using a functional tie between the adjusting parameters of the pump and hydraulic motor through the pressure difference, torque, and angular rate of the hydraulic motor shaft rotation. The programmable logic controller can implement such tie. The Coulomb and viscous frictions are taken into consideration when developing a computational model of the hydraulic volume displacement drive. Discharge balance considers external and internal leakages in equivalent clearances of hydraulic machines, as well as compression loss volume caused by hydraulic fluid compressibility and deformation of pipe walls. To correct dynamic properties of hydraulic drive, the paper offers that in discharge balance are included the additional regulated external leakages in the open circuit of hydraulic drive and regulated internal leakages in the closed-loop circuit. Generalized differential equations having functional multipliers and multilinked nature have been obtained to describe the operation of hydraulic positioning and speed drive with two hydraulic volume adjustable working chamber machines. It is shown that a proposed computational model of hydraulic drive can be taken into consideration in development of LS («Load-Sensing» drives, in which the pumping pressure is tuned to the value required for the most loaded slave motor to overcome the load. Results attained can be used both in designing the industrial-purpose heavy
Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2
Energy Technology Data Exchange (ETDEWEB)
Coryell, E.W.; Siefken, L.J.; Harvego, E.A. [Idaho National Engineering Lab., Idaho Falls, ID (United States)] [and others
1997-07-01
The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The code is the result of merging the RELAP5, SCDAP, and COUPLE codes. The RELAP5 portion of the code calculates the overall reactor coolant system, thermal-hydraulics, and associated reactor system responses. The SCDAP portion of the code describes the response of the core and associated vessel structures. The COUPLE portion of the code describes response of lower plenum structures and debris and the failure of the lower head. The code uses a modular approach with the overall structure, input/output processing, and data structures following the pattern established for RELAP5. The code uses a building block approach to allow the code user to easily represent a wide variety of systems and conditions through a powerful input processor. The user can represent a wide variety of experiments or reactor designs by selecting fuel rods and other assembly structures from a range of representative core component models, and arrange them in a variety of patterns within the thermalhydraulic network. The COUPLE portion of the code uses two-dimensional representations of the lower plenum structures and debris beds. The flow of information between the different portions of the code occurs at each system level time step advancement. The RELAP5 portion of the code describes the fluid transport around the system. These fluid conditions are used as thermal and mass transport boundary conditions for the SCDAP and COUPLE structures and debris beds.
THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code
Energy Technology Data Exchange (ETDEWEB)
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 in 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.
Implementation of CFD module in the KORSAR thermal-hydraulic system code
Energy Technology Data Exchange (ETDEWEB)
Yudov, Yury V.; Danilov, Ilia G.; Chepilko, Stepan S. [Alexandrov Research Inst. of Technology (NITI), Sosnovy Bor (Russian Federation)
2015-09-15
The Russian KORSAR/GP (hereinafter KORSAR) computer code was developed by a joint team from Alexandrov NITI and OKB ''Gidropress'' for VVER safety analysis and certified by the Rostechnadzor of Russia in 2009. The code functionality is based on a 1D two-fluid model for calculation of two-phase flows. A 3D CFD module in the KORSAR computer code is being developed by Alexandrov NITI for representing 3D effects in the downcomer and lower plenum during asymmetrical loop operation. The CFD module uses Cartesian grid method with cut cell approach. The paper presents a numerical algorithm for coupling 1D and 3D thermal- hydraulic modules in the KORSAR code. The combined pressure field is calculated by the multigrid method. The performance efficiency of the algorithm for coupling 1D and 3D modules was demonstrated by solving the benchmark problem of mixing cold and hot flows in a T-junction.
Current and anticipated uses of thermal-hydraulic codes in NFI
Energy Technology Data Exchange (ETDEWEB)
Tsuda, K. [Nuclear Fuel Industries, Ltd., Tokyo (Japan); Takayasu, M. [Nuclear Fuel Industries, Ltd., Sennann-gun (Japan)
1997-07-01
This paper presents the thermal-hydraulic codes currently used in NFI for the LWR fuel development and licensing application including transient and design basis accident analyses of LWR plants. The current status of the codes are described in the context of code capability, modeling feature, and experience of code application related to the fuel development and licensing. Finally, the anticipated use of the future thermal-hydraulic code in NFI is briefly given.
Comparative study of Thermal Hydraulic Analysis Codes for Pressurized Water Reactors
Energy Technology Data Exchange (ETDEWEB)
Kim, Yang Hoon; Jang, Mi Suk; Han, Kee Soo [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)
2015-05-15
Various codes are used for the thermal hydraulic analysis of nuclear reactors. The use of some codes among these is limited by user and some codes are not even open to general person. Thus, the use of alternative code is considered for some analysis. In this study, simple thermal hydraulic behaviors are analyzed using three codes to show that alternative codes are possible for the analysis of nuclear reactors. We established three models of the simple u-tube manometer using three different codes. RELAP5 (Reactor Excursion and Leak Analysis Program), SPACE (Safety and Performance Analysis CodE for nuclear power Plants), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are selected for this analysis. RELAP5 is widely used codes for the analysis of system behavior of PWRs. SPACE has been developed based on RELAP5 for the analysis of system behavior of PWRs and licensing of the code is in progress. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. The internal behavior of u-tube manometer was analyzed by RELAP5, SPACE and GOTHIC codes. The general transient behavior was similar among 3 codes. However, the stabilized status of the transient period analyzed by REPAP5 was different from the other codes. It would be resulted from the different physical models used in the other codes, which is specialized for the multi-phase thermal hydraulic behavior analysis.
Interface requirements to couple thermal-hydraulic codes to severe accident codes: ATHLET-CD
Energy Technology Data Exchange (ETDEWEB)
Trambauer, K. [GRS, Garching (Germany)
1997-07-01
The system code ATHLET-CD is being developed by GRS in cooperation with IKE and IPSN. Its field of application comprises the whole spectrum of leaks and large breaks, as well as operational and abnormal transients for LWRs and VVERs. At present the analyses cover the in-vessel thermal-hydraulics, the early phases of core degradation, as well as fission products and aerosol release from the core and their transport in the Reactor Coolant System. The aim of the code development is to extend the simulation of core degradation up to failure of the reactor pressure vessel and to cover all physically reasonable accident sequences for western and eastern LWRs including RMBKs. The ATHLET-CD structure is highly modular in order to include a manifold spectrum of models and to offer an optimum basis for further development. The code consists of four general modules to describe the reactor coolant system thermal-hydraulics, the core degradation, the fission product core release, and fission product and aerosol transport. Each general module consists of some basic modules which correspond to the process to be simulated or to its specific purpose. Besides the code structure based on the physical modelling, the code follows four strictly separated steps during the course of a calculation: (1) input of structure, geometrical data, initial and boundary condition, (2) initialization of derived quantities, (3) steady state calculation or input of restart data, and (4) transient calculation. In this paper, the transient solution method is briefly presented and the coupling methods are discussed. Three aspects have to be considered for the coupling of different modules in one code system. First is the conservation of masses and energy in the different subsystems as there are fluid, structures, and fission products and aerosols. Second is the convergence of the numerical solution and stability of the calculation. The third aspect is related to the code performance, and running time.
COMPUTATIONAL FLOW RATE FEEDBACK AND CONTROL METHOD IN HYDRAULIC ELEVATORS
Institute of Scientific and Technical Information of China (English)
Xu Bing; Ma Jien; Lin Jianjie
2005-01-01
The computational flow rate feedback and control method, which can be used in proportional valve controlled hydraulic elevators, is discussed and analyzed. In a hydraulic elevator with this method, microprocessor receives pressure information from the pressure transducers and computes the flow rate through the proportional valve based on pressure-flow conversion real time algorithm. This hydraulic elevator is of lower cost and energy consumption than the conventional closed loop control hydraulic elevator whose flow rate is measured by a flow meter. Experiments are carried out on a test rig which could simulate the load of hydraulic elevator. According to the experiment results, the means to modify the pressure-flow conversion algorithm are pointed out.
Evaluation of the RELAP4/MOD6 thermal-hydraulic code. [PWR
Energy Technology Data Exchange (ETDEWEB)
Haigh, W.S.; Margolis, S.G.; Rice, R.E.
1978-01-01
The NRC RELAP4/MOD6 computer code was recently released to the public for use in thermal-hydraulic analysis. This code has a unique new capability permitting analysis of both the blowdown and reflood portions of a postulated pressurized water reactor (PWR) loss-of-coolant accident (LOCA). A principal code evaluation objective is to assess the accuracy of the code for computing LOCA behavior over a wide range of system sizes and scaling concepts. The scales of interest include all LOCA experiments and will ultimately encompass full-sized PWR systems for which no experiments or data are available. Quantitative assessment of the accuracy of the code when it is applied to large PWR systems is still in the future. With RELAP4/MOD6, however, a technique has been demonstrated for using results derived from small-scale blowdown and reflood experiments to predict the accuracy of calculations for similar experiments of significantly different scale or component size. This demonstration is considered a first step in establishing confidence levels for the accuracy of calculations of a postulated LOCA.
Physics codes on parallel computers
Energy Technology Data Exchange (ETDEWEB)
Eltgroth, P.G.
1985-12-04
An effort is under way to develop physics codes which realize the potential of parallel machines. A new explicit algorithm for the computation of hydrodynamics has been developed which avoids global synchronization entirely. The approach, called the Independent Time Step Method (ITSM), allows each zone to advance at its own pace, determined by local information. The method, coded in FORTRAN, has demonstrated parallelism of greater than 20 on the Denelcor HEP machine. ITSM can also be used to replace current implicit treatments of problems involving diffusion and heat conduction. Four different approaches toward work distribution have been investigated and implemented for the one-dimensional code on the Denelcor HEP. They are ''self-scheduled'', an ASKFOR monitor, a ''queue of queues'' monitor, and a distributed ASKFOR monitor. The self-scheduled approach shows the lowest overhead but the poorest speedup. The distributed ASKFOR monitor shows the best speedup and the lowest execution times on the tested problems. 2 refs., 3 figs.
Um, Jeong-Gi; Han, Jisu; Lee, Dahye; Cho, Taechin
2017-04-01
A computer program code was developed to estimate the hydraulic head distribution through the 2-D DFN(discrete fracture network) blocks considering hydraulic aperture of the individual fractures, and to determine flow quantity, directional block hydraulic conductivity and principal hydraulic conductivity tensor according to fracture geometry such as orientation, frequency and size of the fracture network systems. The generated stochastic DFN system is assumed to have a network structure in which the equivalent flow pipe composed linear fractures is complexly connected. DFN systems often include individual or group of sub-network that are isolated from a network that can act as fluid flow passages from one flow boundary to another, and the fluid flow is completely blocked due to lack of connectivity. Fractures that are completely or partially isolated in the DFN system do not contribute to the overall fluid flow through the DFN system and add to the burden of numerical computation. This sometimes leads to numerical instability and failure to provide a solution. In this study, geometric and mathematical routines were designed and implemented to classify and eliminate such sub-networks. The developed program code can calculate the total head at each node connected to the flow path with various aperture as well as hydraulic conductivity of the individual flow pipe using the SOR method. Numerical experiments have been carried out to explore the applicability of the developed program code. A total of 108 stochastic 2-D DFN blocks of 20 m×20 m with various hydraulic aperture were prepared using two joint sets with fixed input parameters of fracture orientation, frequency and size distribution. The hydraulic anisotropy and the chance for equivalent continuum behavior of the DFN system were found to depend on the variability of fracture aperture.
Extensive use of computational fluid dynamics in the upgrading of hydraulic turbines
Energy Technology Data Exchange (ETDEWEB)
Sabourin, M.; Eremeef, R.; De Henau, V.
1995-12-31
Computational fluid dynamics codes, based on turbulent Navier-Stokes equations, allow evaluation of the hydraulic losses of each turbine component with precision. Using those codes with the new generation of computers enables a wide variety of component geometries to be modelled and compared to the original designs under flow conditions obtained from testing, at a reasonable cost and in a relatively short time. This paper reviews the actual method used in the design of a solution to a turbine rehabilitation project involving runner replacement, redesign of upstream components (stay vanes and wicket gates), and downstream components (draft tubes and runner outlets). The paper shows how computational fluid dynamics can help hydraulic engineers to obtain valuable information not only on performance enhancement but also on the phenomena that produce the enhancement, and to reduce the variety of modifications to be tested.
3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors
Energy Technology Data Exchange (ETDEWEB)
Langenbuch, S.; Velkov, K. [GRS, Garching (Germany); Lizorkin, M. [Kurchatov-Institute, Moscow (Russian Federation)] [and others
1997-07-01
This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.
Institute of Scientific and Technical Information of China (English)
FAN Liwei; Hai Reti; WANG Wenxing; LU Zexiang; YANG Zhiming
2008-01-01
A subsurface flow wetland (SSFW) was simulated using a commercial computational fluid dynamic (CFD) code. The constructed media was simulated using porous media and the liquid resident time distribution (RTD) in the SSFW was obtained using the particle trajectory model. The effect of wetland configuration and operating conditions on the hydraulic performance of the SSFW were investigated. The results indicated that the hydraulic performance of the SSFW was predominantly affected by the wetland configuration. The hydraulic efficiency of the SSFW with an inlet at the middle edge of the upper media was 0.584 and the best among the SSFWs with an inlet at the top, the middle, and the bottom edge of the upper media. The constructed media affected the hydraulic performance by the ratio (K) of the upper and lower media resistance. The selection of appropriate media resistance in the protection layer can improve the hydraulic efficiency. When the viscous resistance coefficient of the media in the protection layer changed from 2.315×105 to 1.200×108, the hydraulic efficiency of the SSFW increased from 0.301 to 0.751. However, the effect of operating conditions on the hydraulic efficiency of the SSFW was slight.
Energy Technology Data Exchange (ETDEWEB)
Ebert, D.
1997-07-01
This is a report on the CSNI Workshop on Transient Thermal-Hydraulic and Neutronic Codes Requirements held at Annapolis, Maryland, USA November 5-8, 1996. This experts` meeting consisted of 140 participants from 21 countries; 65 invited papers were presented. The meeting was divided into five areas: (1) current and prospective plans of thermal hydraulic codes development; (2) current and anticipated uses of thermal-hydraulic codes; (3) advances in modeling of thermal-hydraulic phenomena and associated additional experimental needs; (4) numerical methods in multi-phase flows; and (5) programming language, code architectures and user interfaces. The workshop consensus identified the following important action items to be addressed by the international community in order to maintain and improve the calculational capability: (a) preserve current code expertise and institutional memory, (b) preserve the ability to use the existing investment in plant transient analysis codes, (c) maintain essential experimental capabilities, (d) develop advanced measurement capabilities to support future code validation work, (e) integrate existing analytical capabilities so as to improve performance and reduce operating costs, (f) exploit the proven advances in code architecture, numerics, graphical user interfaces, and modularization in order to improve code performance and scrutibility, and (g) more effectively utilize user experience in modifying and improving the codes.
Development of An Automatic Verification Program for Thermal-hydraulic System Codes
Energy Technology Data Exchange (ETDEWEB)
Lee, J. Y.; Ahn, K. T.; Ko, S. H.; Kim, Y. S.; Kim, D. W. [Pusan National University, Busan (Korea, Republic of); Suh, J. S.; Cho, Y. S.; Jeong, J. J. [System Engineering and Technology Co., Daejeon (Korea, Republic of)
2012-05-15
As a project activity of the capstone design competitive exhibition, supported by the Education Center for Green Industry-friendly Fusion Technology (GIFT), we have developed a computer program which can automatically perform non-regression test, which is needed repeatedly during a developmental process of a thermal-hydraulic system code, such as the SPACE code. A non-regression test (NRT) is an approach to software testing. The purpose of the non-regression testing is to verify whether, after updating a given software application (in this case, the code), previous software functions have not been compromised. The goal is to prevent software regression, whereby adding new features results in software bugs. As the NRT is performed repeatedly, a lot of time and human resources will be needed during the development period of a code. It may cause development period delay. To reduce the cost and the human resources and to prevent wasting time, non-regression tests need to be automatized. As a tool to develop an automatic verification program, we have used Visual Basic for Application (VBA). VBA is an implementation of Microsoft's event-driven programming language Visual Basic 6 and its associated integrated development environment, which are built into most Microsoft Office applications (In this case, Excel)
Scherer, W.; Brockmann, H.; Haas, K. A.; Rütten, H. J.
2005-01-01
V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to HTRs and to two spatial dimensions. The...
Energy Technology Data Exchange (ETDEWEB)
Suh, Kune Yull; Yoon, Sang Hyuk; Noh, Sang Woo; Lee, Il Suk [Seoul National University, Seoul (Korea)
2002-03-01
This study is concerned with developing a multidimensional flow model required for the system analysis code MARS to more mechanistically simulate a variety of thermal hydraulic phenomena in the nuclear stem supply system. The capability of the MARS code as a thermal hydraulic analysis tool for optimized system design can be expanded by improving the current calculational methods and adding new models. In this study the relevant literature was surveyed on the multidimensional flow models that may potentially be applied to the multidimensional analysis code. Research items were critically reviewed and suggested to better predict the multidimensional thermal hydraulic behavior and to identify test requirements. A small-scale preliminary test was performed in the downcomer formed by two vertical plates to analyze multidimensional flow pattern in a simple geometry. The experimental result may be applied to the code for analysis of the fluid impingement to the reactor downcomer wall. Also, data were collected to find out the controlling parameters for the one-dimensional and multidimensional flow behavior. 22 refs., 40 figs., 7 tabs. (Author)
Characterizing Video Coding Computing in Conference Systems
Tuquerres, G.
2000-01-01
In this paper, a number of coding operations is provided for computing continuous data streams, in particular, video streams. A coding capability of the operations is expressed by a pyramidal structure in which coding processes and requirements of a distributed information system are represented. Th
Steady state thermal hydraulic analysis of LMR core using COBRA-K code
Energy Technology Data Exchange (ETDEWEB)
Kim, Eui Kwang; Kim, Young Gyun; Kim Young In; Kim Young Cheol
1997-02-01
A thermal hydraulics analysis code COBRA-K is being developed by the KAERI LMR core design technology development team. COBRA-K is a part of the integrated computation system for LMR core design and analysis, the K-CORE system. COBRA-K is supposed to predict the flow and temperature distributions in LMR core. COBRA-K is an extension of the previously published COBRA-IV-I code with several functional improvements. Specially COBRA-K has been improved to analyze single and multi-assembly, and whole-core in the transient condition. This report describes the overall features of COBRA-K and gives general input descriptions. The 19 pin assembly experimental data of ORNL were used to verify the accuracy of this code for the steady state analysis. The comparative results show good agreements between the calculated and the measured data. And COBRA-K can be used to predict flow and temperature distributions for the LMR core design. (author). 7 refs., 6 tabs., 13 figs.
Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes
Energy Technology Data Exchange (ETDEWEB)
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 and 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.
Decker, Robert L.; Kirby, Klane
This curriculum guide contains a course in hydraulics to train entry-level workers for automotive mechanics and other fields that utilize hydraulics. The module contains 14 instructional units that cover the following topics: (1) introduction to hydraulics; (2) fundamentals of hydraulics; (3) reservoirs; (4) lines, fittings, and couplers; (5)…
Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis
Energy Technology Data Exchange (ETDEWEB)
Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E. [Sandia National Labs., Albuquerque, NM (United States); Tills, J. [J. Tills and Associates, Inc., Sandia Park, NM (United States)
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 the 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.
Computer Code for Nanostructure Simulation
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.
Cloud Computing for Complex Performance Codes.
Energy Technology Data Exchange (ETDEWEB)
Appel, Gordon John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Klein, Brandon Thorin [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Miner, John Gifford [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
2017-02-01
This report describes the use of cloud computing services for running complex public domain performance assessment problems. The work consisted of two phases: Phase 1 was to demonstrate complex codes, on several differently configured servers, could run and compute trivial small scale problems in a commercial cloud infrastructure. Phase 2 focused on proving non-trivial large scale problems could be computed in the commercial cloud environment. The cloud computing effort was successfully applied using codes of interest to the geohydrology and nuclear waste disposal modeling community.
Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures
Directory of Open Access Journals (Sweden)
Alessandro Petruzzi
2008-01-01
Full Text Available In the last four decades, large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Whereas the first system codes, developed at the beginning of the 1970s, utilized the homogenous equilibrium model with three balance equations to describe the two-phase flow, nowadays the more advanced system codes are based on the so-called “two-fluid model” with separation of the water and vapor phases, resulting in systems with at least six balance equations. The wide experimental campaign, constituted by the integral and separate effect tests, conducted under the umbrella of the OECD/CSNI was at the basis of the development and validation of the thermal-hydraulic system codes by which they have reached the present high degree of maturity. However, notwithstanding the huge amounts of financial and human resources invested, the results predicted by the code are still affected by errors whose origins can be attributed to several reasons as model deficiencies, approximations in the numerical solution, nodalization effects, and imperfect knowledge of boundary and initial conditions. In this context, the existence of qualified procedures for a consistent application of qualified thermal-hydraulic system code is necessary and implies the drawing up of specific criteria through which the code-user, the nodalization, and finally the transient results are qualified.
Development of thermal hydraulic models for the reliable regulatory auditing code
Energy Technology Data Exchange (ETDEWEB)
Chung, B. D.; Song, C. H.; Lee, Y. J.; Kwon, T. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
2003-04-15
The objective of this project is to develop thermal hydraulic models for use in improving the reliability of the regulatory auditing codes. The current year fall under the first step of the 3 year project, and the main researches were focused on identifying the candidate thermal hydraulic models for improvement and to develop prototypical model development. During the current year, the verification calculations submitted for the APR 1400 design certification have been reviewed, the experimental data from the MIDAS DVI experiment facility in KAERI have been analyzed and evaluated, candidate thermal hydraulic models for improvement have been identified, prototypical models for the improved thermal hydraulic models have been developed, items for experiment in connection with the model development have been identified, and preliminary design of the experiment has been carried out.
Energy Technology Data Exchange (ETDEWEB)
Banerjee, S.; Hassan, Y.A. [Texas A& M Univ., College Station, TX (United States)
1995-09-01
Condensation in the presence of noncondensible gases plays an important role in the nuclear industry. The RELAP5/MOD3 thermal hydraulic code was used to study the ability of the code to predict this phenomenon. Two separate effects experiments were simulated using this code. These were the Massachusetts Institute of Technology`s (MIT) Pressurizer Experiment, the MIT Single Tube Experiment. A new iterative approach to calculate the interface temperature and the degraded heat transfer coefficient was developed and implemented in the RELAP5/MOD3 thermal hydraulic code. This model employs the heat transfer simultaneously. This model was found to perform much better than the reduction factor approach. The calculations using the new model were found to be in much better agreement with the experimental values.
Development of thermal hydraulic models for the reliable regulatory auditing code
Energy Technology Data Exchange (ETDEWEB)
Chung, B. D.; Song, C. H.; Lee, Y. J.; Kwon, T. S.; Lee, S. W. [Korea Automic Energy Research Institute, Taejon (Korea, Republic of)
2004-02-15
The objective of this project is to develop thermal hydraulic models for use in improving the reliability of the regulatory auditing codes. The current year fall under the second step of the 3 year project, and the main researches were focused on the development of downcorner boiling model. During the current year, the bubble stream model of downcorner has been developed and installed in he auditing code. The model sensitivity analysis has been performed for APR1400 LBLOCA scenario using the modified code. The preliminary calculation has been performed for the experimental test facility using FLUENT and MARS code. The facility for air bubble experiment has been installed. The thermal hydraulic phenomena for VHTR and super critical reactor have been identified for the future application and model development.
Energy Technology Data Exchange (ETDEWEB)
Page, R.; Jones, J.R.
1997-07-01
Ensuring that safety analysis needs are met in the future is likely to lead to the development of new codes and the further development of existing codes. It is therefore advantageous to define standards for data interfaces and to develop software interfacing techniques which can readily accommodate changes when they are made. Defining interface standards is beneficial but is necessarily restricted in application if future requirements are not known in detail. Code interfacing methods are of particular relevance with the move towards automatic grid frequency response operation where the integration of plant dynamic, core follow and fault study calculation tools is considered advantageous. This paper describes the background and features of a new code TALINK (Transient Analysis code LINKage program) used to provide a flexible interface to link the RELAP5 thermal hydraulics code with the PANTHER neutron kinetics and the SIBDYM whole plant dynamic modelling codes used by Nuclear Electric. The complete package enables the codes to be executed in parallel and provides an integrated whole plant thermal-hydraulics and neutron kinetics model. In addition the paper discusses the capabilities and pedigree of the component codes used to form the integrated transient analysis package and the details of the calculation of a postulated Sizewell `B` Loss of offsite power fault transient.
4C code analysis of thermal-hydraulic transients in the KSTAR PF1 superconducting coil
Savoldi Richard, L.; Bonifetto, R.; Chu, Y.; Kholia, A.; Park, S. H.; Lee, H. J.; Zanino, R.
2013-01-01
The KSTAR tokamak, in operation since 2008 at the National Fusion Research Institute in Korea, is equipped with a full superconducting magnet system including the central solenoid (CS), which is made of four symmetric pairs of coils PF1L/U-PF4L/U. Each of the CS coils is pancake wound using Nb3Sn cable-in-conduit conductors with a square Incoloy jacket. The coils are cooled with supercritical He in forced circulation at nominal 4.5 K and 5.5 bar inlet conditions. During different test campaigns the measured temperature increase due to AC losses turned out to be higher than expected, which motivates the present study. The 4C code, already validated against and applied to different types of thermal-hydraulic transients in different superconducting coils, is applied here to the thermal-hydraulic analysis of a full set of trapezoidal current pulses in the PF1 coils, with different ramp rates. We find the value of the coupling time constant nτ that best fits, at each current ramp rate, the temperature increase up to the end of the heating at the coil outlet. The agreement between computed results and the whole set of measured data, including temperatures, pressures and mass flow rates, is then shown to be very good both at the inlet and at the outlet of the coil. The nτ values needed to explain the experimental results decrease at increasing current ramp rates, consistently with the results found in the literature.
Gender codes why women are leaving computing
Misa, Thomas J
2010-01-01
The computing profession is facing a serious gender crisis. Women are abandoning the computing field at an alarming rate. Fewer are entering the profession than anytime in the past twenty-five years, while too many are leaving the field in mid-career. With a maximum of insight and a minimum of jargon, Gender Codes explains the complex social and cultural processes at work in gender and computing today. Edited by Thomas Misa and featuring a Foreword by Linda Shafer, Chair of the IEEE Computer Society Press, this insightful collection of essays explores the persisting gender imbalance in computing and presents a clear course of action for turning things around.
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.
Interface requirements to couple thermal-hydraulic codes to 3D neutronic codes
Energy Technology Data Exchange (ETDEWEB)
Langenbuch, S.; Austregesilo, H.; Velkov, K. [GRS, Garching (Germany)] [and others
1997-07-01
The present situation of thermalhydraulics codes and 3D neutronics codes is briefly described and general considerations for coupling of these codes are discussed. Two different basic approaches of coupling are identified and their relative advantages and disadvantages are discussed. The implementation of the coupling for 3D neutronics codes in the system ATHLET is presented. Meanwhile, this interface is used for coupling three different 3D neutronics codes.
Summary of papers on current and anticipated uses of thermal-hydraulic codes
Energy Technology Data Exchange (ETDEWEB)
Caruso, R.
1997-07-01
The author reviews a range of recent papers which discuss possible uses and future development needs for thermal/hydraulic codes in the nuclear industry. From this review, eight common recommendations are extracted. They are: improve the user interface so that more people can use the code, so that models are easier and less expensive to prepare and maintain, and so that the results are scrutable; design the code so that it can easily be coupled to other codes, such as core physics, containment, fission product behaviour during severe accidents; improve the numerical methods to make the code more robust and especially faster running, particularly for low pressure transients; ensure that future code development includes assessment of code uncertainties as integral part of code verification and validation; provide extensive user guidelines or structure the code so that the `user effect` is minimized; include the capability to model multiple fluids (gas and liquid phase); design the code in a modular fashion so that new models can be added easily; provide the ability to include detailed or simplified component models; build on work previously done with other codes (RETRAN, RELAP, TRAC, CATHARE) and other code validation efforts (CSAU, CSNI SET and IET matrices).
Energy Technology Data Exchange (ETDEWEB)
Akimoto, Hajime; Kukita; Ohnuki, Akira [Japan Atomic Energy Research Institute, Ibaraki (Japan)
1997-07-01
The Japan Atomic Energy Research Institute (JAERI) is conducting several research programs related to thermal-hydraulic and neutronic behavior of light water reactors (LWRs). These include LWR safety research projects, which are conducted in accordance with the Nuclear Safety Commission`s research plan, and reactor engineering projects for the development of innovative reactor designs or core/fuel designs. Thermal-hydraulic and neutronic codes are used for various purposes including experimental analysis, nuclear power plant (NPP) safety analysis, and design assessment.
Measurement of Fracture Geometry for Accurate Computation of Hydraulic Conductivity
Chae, B.; Ichikawa, Y.; Kim, Y.
2003-12-01
mechanical aperture from hydraulic aperture due to rough geometry of fracture walls. The hydraulic conductivity did not follow the cubic law. It verifies that a parallel plate model is not suitable to express the hydraulic conductivity including local fracture geometry. The measurement results are used to compute hydraulic conductivity along a rock fracture based on the homogenization theory.
Interface requirements to couple thermal hydraulics codes to severe accident codes: ICARE/CATHARE
Energy Technology Data Exchange (ETDEWEB)
Camous, F.; Jacq, F.; Chatelard, P. [IPSN/DRS/SEMAR CE-Cadarache, St Paul Lez Durance (France)] [and others
1997-07-01
In order to describe with the same code the whole sequence of severe LWR accidents, up to the vessel failure, the Institute of Protection and Nuclear Safety has performed a coupling of the severe accident code ICARE2 to the thermalhydraulics code CATHARE2. The resulting code, ICARE/CATHARE, is designed to be as pertinent as possible in all the phases of the accident. This paper is mainly devoted to the description of the ICARE2-CATHARE2 coupling.
Energy Technology Data Exchange (ETDEWEB)
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 code 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.
Energy Technology Data Exchange (ETDEWEB)
Uwaba, Tomoyuki, E-mail: uwaba.tomoyuki@jaea.go.jp; Ohshima, Hiroyuki; Ito, Masahiro
2017-06-15
Highlights: • The coupled computational code system allowed for mechanical and thermal-hydraulic analyses in a fast reactor fuel subassembly. • In this system interactive calculations between flow area deformations and coolant temperature changes are repeated to their convergence state. • Effects on bundle-duct interaction on coolant temperature distributions were investigated by using the code system. - Abstract: The coupled numerical analysis of mechanical and thermal-hydraulic behaviors was performed for a wire-wrapped fuel pin bundle subassembly irradiated in a fast reactor. For the analysis, the fuel pin bundle deformation analysis code BAMBOO and the thermal-hydraulic analysis code ASFRE exchanged the deformation and temperature analysis results through the iterative calculations to attain convergence corresponding to the static balance between deformation and temperature. The analysis by the coupled code system showed that the radial distribution of coolant temperature in the subassembly tended to flatten as a result of the fuel pin bundle deformation governed by cladding void swelling and irradiation creep. Such flattening of temperature distribution was slightly observed as a result of fuel pin bowings due to the cladding-wire interaction even when no bundle-duct interaction occurred. The effect of the spacer wire-pitch on deformation and thermal-hydraulics was also investigated in this study.
Thermal hydraulic codes for LWR safety analysis - present status and future perspective
Energy Technology Data Exchange (ETDEWEB)
Staedtke, H. [Commission of the European Union, Ispra (Italy)
1997-07-01
The aim of the present paper is to give a review on the current status and future perspective of present best-estimate Thermal Hydraulic codes. Reference is made to internationally well-established codes which have reached a certain state of maturity. The first part of the paper deals with the common basic code features with respect to the physical modelling and their numerical methods used to describe complex two-phase flow and heat transfer processes. The general predictive capabilities are summarized identifying some remaining code deficiencies and their underlying limitations. The second part discusses various areas including physical modelling, numerical techniques and informatic structure where the codes could be substantially improved.
Computer Security: is your code sane?
Stefan Lueders, Computer Security Team
2015-01-01
How many of us write code? Software? Programs? Scripts? How many of us are properly trained in this and how well do we do it? Do we write functional, clean and correct code, without flaws, bugs and vulnerabilities*? In other words: are our codes sane? Figuring out weaknesses is not that easy (see our quiz in an earlier Bulletin article). Therefore, in order to improve the sanity of your code, prevent common pit-falls, and avoid the bugs and vulnerabilities that can crash your code, or – worse – that can be misused and exploited by attackers, the CERN Computer Security team has reviewed its recommendations for checking the security compliance of your code. “Static Code Analysers” are stand-alone programs that can be run on top of your software stack, regardless of whether it uses Java, C/C++, Perl, PHP, Python, etc. These analysers identify weaknesses and inconsistencies including: employing undeclared variables; expressions resu...
Mathematical model and computer code for the analysis of advanced fast reactor dynamics
Energy Technology Data Exchange (ETDEWEB)
Schukin, N.V. (Moscow Engineering Physics Inst. (Russian Federation)); Korsun, A.S. (Moscow Engineering Physics Inst. (Russian Federation)); Vitruk, S.G. (Moscow Engineering Physics Inst. (Russian Federation)); Zimin, V.G. (Moscow Engineering Physics Inst. (Russian Federation)); Romanin, S.D. (Moscow Engineering Physics Inst. (Russian Federation))
1993-04-01
Efficient algorithms for mathematical modeling of 3-D neutron kinetics and thermal hydraulics are described. The model and appropriate computer code make it possible to analyze a variety of transient events ranging from normal operational states to catastrophic accident excursions. To verify the code, a number of calculations of different kind of transients was carried out. The results of the calculations show that the model and the computer code could be used for conceptual design of advanced liquid metal reactors. The detailed description of calculations of TOP WS accident is presented. (orig./DG)
Development of best estimate auditing code for CANDU thermal hydraulic safety analysis
Energy Technology Data Exchange (ETDEWEB)
Chung, B. D.; Lee, W. J.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1998-04-15
The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is first step of the whole project, thus focus to the establishment of improvement area. The study was performed by reconsideration of the previous code assessment works and investigation of AECL design analysis tools. In order to identify the thermal hydraulic phenomena for events, the whole system of CANDU plant was divided into main functional systems and subcomponents. Each phenomena was addressed to the each subcomponent. FinaIly improvement areas of model development for auditing tool were established based on the identified phenomena.
Development of best estimate auditing code for CANDU thermal hydraulic safety analysis
Energy Technology Data Exchange (ETDEWEB)
Chung, B. D.; Lee, W. J.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejeon (Korea)
1998-04-01
The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is first step of the whole project, thus focus to the establishment of improvement area. The study was performed by reconsideration of the previous code assessment works and investigation of AECL design analysis tools. In order to identify the thermal hydraulic phenomena for events, the whole system of CANDU plant was divided into main functional systems and subcomponents. Each phenomena was addressed to the each subcomponent. Finally improvement areas of model development for auditing tool were established based on the identified phenomena. 8 refs., 21 figs., 19 tabs. (Author)
Energy Technology Data Exchange (ETDEWEB)
Hainoun, A., E-mail: pscientific2@aec.org.sy [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Doval, A. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400 S.C de Bariloche, Rio Negro (Argentina); Umbehaun, P. [Centro de Engenharia Nuclear – CEN, IPEN-CNEN/SP, Av. Lineu Prestes 2242-Cidade Universitaria, CEP-05508-000 São Paulo, SP (Brazil); Chatzidakis, S. [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States); Ghazi, N. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Park, S. [Research Reactor Design and Engineering Division, Basic Science Project Operation Dept., Korea Atomic Energy Research Institute (Korea, Republic of); Mladin, M. [Institute for Nuclear Research, Campului Street No. 1, P.O. Box 78, 115400 Mioveni, Arges (Romania); Shokr, A. [Division of Nuclear Installation Safety, Research Reactor Safety Section, International Atomic Energy Agency, A-1400 Vienna (Austria)
2014-12-15
Highlights: • A set of advanced system thermal hydraulic codes are benchmarked against IFA of IEA-R1. • Comparative safety analysis of IEA-R1 reactor during LOFA by 7 working teams. • This work covers both experimental and calculation effort and presents new out findings on TH of RR that have not been reported before. • LOFA results discrepancies from 7% to 20% for coolant and peak clad temperatures are predicted conservatively. - Abstract: In the framework of the IAEA Coordination Research Project on “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal hydraulic computational methods and tools for operation and safety analysis of research reactors” the Brazilian research reactor IEA-R1 has been selected as reference facility to perform benchmark calculations for a set of thermal hydraulic codes being widely used by international teams in the field of research reactor (RR) deterministic safety analysis. The goal of the conducted benchmark is to demonstrate the application of innovative reactor analysis tools in the research reactor community, validation of the applied codes and application of the validated codes to perform comprehensive safety analysis of RR. The IEA-R1 is equipped with an Instrumented Fuel Assembly (IFA) which provided measurements for normal operation and loss of flow transient. The measurements comprised coolant and cladding temperatures, reactor power and flow rate. Temperatures are measured at three different radial and axial positions of IFA summing up to 12 measuring points in addition to the coolant inlet and outlet temperatures. The considered benchmark deals with the loss of reactor flow and the subsequent flow reversal from downward forced to upward natural circulation and presents therefore relevant phenomena for the RR safety analysis. The benchmark calculations were performed independently by the participating teams using different thermal hydraulic and safety
An assessment of the CORCON-MOD3 code. Part 1: Thermal-hydraulic calculations
Energy Technology Data Exchange (ETDEWEB)
Strizhov, V.; Kanukova, V.; Vinogradova, T.; Askenov, E. [Russian Academy of Sciences, Moscow (Russian Federation). Inst. of Nuclear Safety; Nikulshin, V. [Russian Research Center, Moscow (Russian Federation). Kurchatov Inst.
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 from 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.
FX2-TH: a two-dimensional nuclear reactor kinetics code with thermal-hydraulic feedback
Energy Technology Data Exchange (ETDEWEB)
Shober, R.A.; Daly, T.A.; Ferguson, D.R.
1978-10-01
FX2-TH is a two-dimensional, time-dependent nuclear reactor kinetics program with thermal and hydraulic feedback. The neutronics model used is multigroup neutron diffusion theory. The following geometry options are available: x, r, x-y, r-z, theta-r, and triangular. FX2-TH contains two basic thermal and hydraulic models: a simple adiabatic fuel temperature calculation, and a more detailed model consisting of an explicit representation of a fuel pin, gap, clad, and coolant. FX2-TH allows feedback effects from both fuel temperature (Doppler) and coolant temperature (density) changes. FX2-TH will calculate a consistent set of steady state conditions by iterating between the neutronics and thermal-hydraulics until convergence is reached. The time-dependent calculation is performed by the use of the improved quasistatic method. A disk editing capability is available. FX2-TH is operational on IBM system 360 or 370 computers and on the CDC 7600.
Incompressible face seals: Computer code IFACE
Artiles, Antonio
1994-01-01
Capabilities of the computer code IFACE are given in viewgraph format. These include: two dimensional, incompressible, isoviscous flow; rotation of both rotor and housing; roughness in both rotor and housing; arbitrary film thickness distribution, including steps, pockets, and tapers; three degrees of freedom; dynamic coefficients; prescribed force and moments; pocket pressures or orifice size; turbulence, Couette and Poiseuille flow; cavitation; and inertia pressure drops at inlets to film.
Computing Challenges in Coded Mask Imaging
Skinner, Gerald
2009-01-01
This slide presaentation reviews the complications and challenges in developing computer systems for Coded Mask Imaging telescopes. The coded mask technique is used when there is no other way to create the telescope, (i.e., when there are wide fields of view, high energies for focusing or low energies for the Compton/Tracker Techniques and very good angular resolution.) The coded mask telescope is described, and the mask is reviewed. The coded Masks for the INTErnational Gamma-Ray Astrophysics Laboratory (INTEGRAL) instruments are shown, and a chart showing the types of position sensitive detectors used for the coded mask telescopes is also reviewed. Slides describe the mechanism of recovering an image from the masked pattern. The correlation with the mask pattern is described. The Matrix approach is reviewed, and other approaches to image reconstruction are described. Included in the presentation is a review of the Energetic X-ray Imaging Survey Telescope (EXIST) / High Energy Telescope (HET), with information about the mission, the operation of the telescope, comparison of the EXIST/HET with the SWIFT/BAT and details of the design of the EXIST/HET.
New developments in the Saphire computer codes
Energy Technology Data Exchange (ETDEWEB)
Russell, K.D.; Wood, S.T.; Kvarfordt, K.J. [Idaho Engineering Lab., Idaho Falls, ID (United States)] [and others
1996-03-01
The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) refers to a suite of computer programs that were developed to create and analyze a probabilistic risk assessment (PRA) of a nuclear power plant. Many recent enhancements to this suite of codes have been made. This presentation will provide an overview of these features and capabilities. The presentation will include a discussion of the new GEM module. This module greatly reduces and simplifies the work necessary to use the SAPHIRE code in event assessment applications. An overview of the features provided in the new Windows version will also be provided. This version is a full Windows 32-bit implementation and offers many new and exciting features. [A separate computer demonstration was held to allow interested participants to get a preview of these features.] The new capabilities that have been added since version 5.0 will be covered. Some of these major new features include the ability to store an unlimited number of basic events, gates, systems, sequences, etc.; the addition of improved reporting capabilities to allow the user to generate and {open_quotes}scroll{close_quotes} through custom reports; the addition of multi-variable importance measures; and the simplification of the user interface. Although originally designed as a PRA Level 1 suite of codes, capabilities have recently been added to SAPHIRE to allow the user to apply the code in Level 2 analyses. These features will be discussed in detail during the presentation. The modifications and capabilities added to this version of SAPHIRE significantly extend the code in many important areas. Together, these extensions represent a major step forward in PC-based risk analysis tools. This presentation provides a current up-to-date status of these important PRA analysis tools.
SALE: Safeguards Analytical Laboratory Evaluation computer code
Energy Technology Data Exchange (ETDEWEB)
Carroll, D.J.; Bush, W.J.; Dolan, C.A.
1976-09-01
The Safeguards Analytical Laboratory Evaluation (SALE) program implements an industry-wide quality control and evaluation system aimed at identifying and reducing analytical chemical measurement errors. Samples of well-characterized materials are distributed to laboratory participants at periodic intervals for determination of uranium or plutonium concentration and isotopic distributions. The results of these determinations are statistically-evaluated, and each participant is informed of the accuracy and precision of his results in a timely manner. The SALE computer code which produces the report is designed to facilitate rapid transmission of this information in order that meaningful quality control will be provided. Various statistical techniques comprise the output of the SALE computer code. Assuming an unbalanced nested design, an analysis of variance is performed in subroutine NEST resulting in a test of significance for time and analyst effects. A trend test is performed in subroutine TREND. Microfilm plots are obtained from subroutine CUMPLT. Within-laboratory standard deviations are calculated in the main program or subroutine VAREST, and between-laboratory standard deviations are calculated in SBLV. Other statistical tests are also performed. Up to 1,500 pieces of data for each nuclear material sampled by 75 (or fewer) laboratories may be analyzed with this code. The input deck necessary to run the program is shown, and input parameters are discussed in detail. Printed output and microfilm plot output are described. Output from a typical SALE run is included as a sample problem.
Fletcher, C. D.
The capability to perform thermal-hydraulic analyses of a space reactor using the ATHENA computer code is demonstrated. The fast reactor, liquid-lithium coolant loops, and lithium-filled heat pipes of the preliminary General electric SP-100 design were modeled with ATHENA. Two demonstration transient calculations were performed simulating accident conditions. Calculated results are available for display using the Nuclear Plant Analyzer color graphics analysis tool in addition to traditional plots. ATHENA-calculated results appear reasonable, both for steady state full power conditions, and for the two transients. This analysis represents the first known transient thermal-hydraulic simulation using an integral space reactor system model incorporating heat pipes.
Neutron spectrum unfolding using computer code SAIPS
Karim, S
1999-01-01
The main objective of this project was to study the neutron energy spectrum at rabbit station-1 in Pakistan Research Reactor (PARR-I). To do so, multiple foils activation method was used to get the saturated activities. The computer code SAIPS was used to unfold the neutron spectra from the measured reaction rates. Of the three built in codes in SAIPS, only SANDI and WINDOWS were used. Contribution of thermal part of the spectra was observed to be higher than the fast one. It was found that the WINDOWS gave smooth spectra while SANDII spectra have violet oscillations in the resonance region. The uncertainties in the WINDOWS results are higher than those of SANDII. The results show reasonable agreement with the published results.
Energy Technology Data Exchange (ETDEWEB)
Arai, Kenji; Ebata, Shigeo [Toshiba Corp., Yokohama (Japan)
1997-07-01
This paper summarizes the current and anticipated use of the thermal-hydraulic and neutronic codes for the BWR transient and accident analyses in Japan. The codes may be categorized into the licensing codes and the best estimate codes for the BWR transient and accident analyses. Most of the licensing codes have been originally developed by General Electric. Some codes have been updated based on the technical knowledge obtained in the thermal hydraulic study in Japan, and according to the BWR design changes. The best estimates codes have been used to support the licensing calculations and to obtain the phenomenological understanding of the thermal hydraulic phenomena during a BWR transient or accident. The best estimate codes can be also applied to a design study for a next generation BWR to which the current licensing model may not be directly applied. In order to rationalize the margin included in the current BWR design and develop a next generation reactor with appropriate design margin, it will be required to improve the accuracy of the thermal-hydraulic and neutronic model. In addition, regarding the current best estimate codes, the improvement in the user interface and the numerics will be needed.
V.S.O.P. (99/05) computer code system
Energy Technology Data Exchange (ETDEWEB)
Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Scherer, W.
2005-11-01
V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to HTRs and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. V.S.O.P.(99 / 05) represents the further development of V.S.O.P. (99). Compared to its precursor, the code system has been improved in many details. Major improvements and extensions have been included concerning the neutron spectrum calculation, the 3-d neutron diffusion options, and the thermal hydraulic section with respect to 'multi-pass'-fuelled pebblebed cores. This latest code version was developed and tested under the WINDOWS-XP - operating system. The storage requirement for the executables and the basic libraries associated with the code amounts to about 15 MB. Another 5 MB are required - if desired - for storage of the source code ({approx}65000 Fortran statements). (orig.)
V.S.O.P. (99/05) computer code system
Energy Technology Data Exchange (ETDEWEB)
Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Scherer, W.
2005-11-01
V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to HTRs and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. V.S.O.P.(99 / 05) represents the further development of V.S.O.P. (99). Compared to its precursor, the code system has been improved in many details. Major improvements and extensions have been included concerning the neutron spectrum calculation, the 3-d neutron diffusion options, and the thermal hydraulic section with respect to 'multi-pass'-fuelled pebblebed cores. This latest code version was developed and tested under the WINDOWS-XP - operating system. The storage requirement for the executables and the basic libraries associated with the code amounts to about 15 MB. Another 5 MB are required - if desired - for storage of the source code ({approx}65000 Fortran statements). (orig.)
Rütten, H. J.; Haas, K. A.; Brockmann, H.; Ohlig, U.; Scherer, W.
2000-01-01
V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to HTRs and to two spatial dimensions. The...
Rütten, H. J.; Haas, K. A.; Brockmann, H.; Ohlig, U.; Scherer, W.
1998-01-01
V.S.O.P. (97) is a computer code system for the comprehensive numerical simulation ofthe physics of thermal reactors. It implies processing ofcross sections, the setup ofthe reactor and ofthe fuel element, repeated neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to EM and to two spatial dimensi...
From Coding to Computational Thinking and Back
DePryck, K.
2016-01-01
Presentation of Dr. Koen DePryck in the Computational Thinking Session in TEEM 2016 Conference, held in the University of Salamanca (Spain), Nov 2-4, 2016. Introducing coding in the curriculum at an early age is considered a long-term investment in bridging the skills gap between the technology demands of the labour market and the availability of people to fill them. The keys to success include moving from mere literacy to active control – not only at the level of learners but also ...
Spiking network simulation code for petascale computers
Kunkel, Susanne; Schmidt, Maximilian; Eppler, Jochen M.; Plesser, Hans E.; Masumoto, Gen; Igarashi, Jun; Ishii, Shin; Fukai, Tomoki; Morrison, Abigail; Diesmann, Markus; Helias, Moritz
2014-01-01
Brain-scale networks exhibit a breathtaking heterogeneity in the dynamical properties and parameters of their constituents. At cellular resolution, the entities of theory are neurons and synapses and over the past decade researchers have learned to manage the heterogeneity of neurons and synapses with efficient data structures. Already early parallel simulation codes stored synapses in a distributed fashion such that a synapse solely consumes memory on the compute node harboring the target neuron. As petaflop computers with some 100,000 nodes become increasingly available for neuroscience, new challenges arise for neuronal network simulation software: Each neuron contacts on the order of 10,000 other neurons and thus has targets only on a fraction of all compute nodes; furthermore, for any given source neuron, at most a single synapse is typically created on any compute node. From the viewpoint of an individual compute node, the heterogeneity in the synaptic target lists thus collapses along two dimensions: the dimension of the types of synapses and the dimension of the number of synapses of a given type. Here we present a data structure taking advantage of this double collapse using metaprogramming techniques. After introducing the relevant scaling scenario for brain-scale simulations, we quantitatively discuss the performance on two supercomputers. We show that the novel architecture scales to the largest petascale supercomputers available today. PMID:25346682
Spiking network simulation code for petascale computers
Directory of Open Access Journals (Sweden)
Susanne eKunkel
2014-10-01
Full Text Available Brain-scale networks exhibit a breathtaking heterogeneity in the dynamical properties and parameters of their constituents. At cellular resolution, the entities of theory are neurons and synapses and over the past decade researchers have learned to manage the heterogeneity of neurons and synapses with efficient data structures. Already early parallel simulation codes stored synapses in a distributed fashion such that a synapse solely consumes memory on the compute node harboring the target neuron. As petaflop computers with some 100,000 nodes become increasingly available for neuroscience, new challenges arise for neuronal network simulation software: Each neuron contacts on the order of 10,000 other neurons and thus has targets only on a fraction of all compute nodes; furthermore, for any given source neuron, at most a single synapse is typically created on any compute node. From the viewpoint of an individual compute node, the heterogeneity in the synaptic target lists thus collapses along two dimensions: the dimension of the types of synapses and the dimension of the number of synapses of a given type. Here we present a data structure taking advantage of this double collapse using metaprogramming techniques. After introducing the relevant scaling scenario for brain-scale simulations, we quantitatively discuss the performance on two supercomputers. We show that the novel architecture scales to the largest petascale supercomputers available today.
Directory of Open Access Journals (Sweden)
M. Avramova
2013-01-01
Full Text Available The international OECD/NRC PSBT benchmark has been established to provide a test bed for assessing the capabilities of thermal-hydraulic codes and to encourage advancement in the analysis of fluid flow in rod bundles. The benchmark was based on one of the most valuable databases identified for the thermal-hydraulics modeling developed by NUPEC, Japan. The database includes void fraction and departure from nucleate boiling measurements in a representative PWR fuel assembly. On behalf of the benchmark team, PSU in collaboration with US NRC has performed supporting calculations using the PSU in-house advanced thermal-hydraulic subchannel code CTF and the US NRC system code TRACE. CTF is a version of COBRA-TF whose models have been continuously improved and validated by the RDFMG group at PSU. TRACE is a reactor systems code developed by US NRC to analyze transient and steady-state thermal-hydraulic behavior in LWRs and it has been designed to perform best-estimate analyses of LOCA, operational transients, and other accident scenarios in PWRs and BWRs. The paper presents CTF and TRACE models for the PSBT void distribution exercises. Code-to-code and code-to-data comparisons are provided along with a discussion of the void generation and void distribution models available in the two codes.
Current and anticipated uses of thermal-hydraulic codes in Spain
Energy Technology Data Exchange (ETDEWEB)
Pelayo, F.; Reventos, F. [Consejo de Seguridad Nuclear, Barcelona (Spain)
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 the 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.
Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis
Energy Technology Data Exchange (ETDEWEB)
Chung, B.D.; Lee, W.J.; Lim, H.S. [Korea Atomic Energy Research Institute, Taejon (Korea)
1999-04-01
The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is a second step of the whole project, and focus to the implementation of CANDU models based on the previous study. FORTRAN 90 language have been used for the development of RELAP5.MOD3/CANDU PC version. For the convenience of the previous Workstation users, the FOTRAN 77 version has been coded also and implanted into the original RELAP5 source file. The verification of model implementation has been performed through the simple verification calculations using the CANDU version. 6 refs., 15 figs., 7 tabs. (Author)
Directory of Open Access Journals (Sweden)
Esch Markus
2014-01-01
Full Text Available For future high temperature reactor projects, e. g., for electricity production or nuclear process heat applications, the steam generator is a crucial component. A typical design is a helical coil steam generator consisting of several tubes connected in parallel forming cylinders of different diameters. This type of steam generator was a significant component used at the thorium high temperature reactor. In the work presented the temperature profile is being analyzed by the nodal thermal hydraulics code TRACE for the thorium high temperature reactor steam generator. The influence of the nodalization is being investigated within the scope of this study and compared to experimental results from the past. The results of the standard TRACE code are compared to results using a modified Nusselt number for the primary side. The implemented heat transfer correlation was developed within the past German HTR program. This study shows that both TRACE versions are stable and provides a discussion of the nodalization requirements.
Development of system analysis code for thermal-hydraulic simulation of integral reactor, Rex-10
Energy Technology Data Exchange (ETDEWEB)
NONE
2010-10-15
Rex-10 is an environment-friendly and economical small-scale nuclear reactor to provide the energy for district heating as well as the electric power in micro-grid. This integral reactor comprises several innovative concepts supported by advanced primary circuit components, low coolant parameters and natural circulation cooling. To evaluate the system performance and thermal-hydraulic behavior of the reactor, a system analysis code is being developed so that the new designs and technologies adopted in Rex-10 can be reflected. The research efforts are absorbed in programming the simple and fast-running thermal-hydraulic analysis software. The details of hydrodynamic governing equations component models and numerical solution scheme used in this code are presented in this paper. On the basis of one-dimensional momentum integral model, the models of point reactor neutron kinetics for thorium-fueled core, physical processes in the steam-gas pressurizer, and heat transfers in helically coiled steam generator are implemented to the system code. Implicit numerical scheme is employed to momentum and energy equations to assure the numerical stability. The accuracy of simulation is validated by applying the solution method to the Rex-10 test facility. Calculated natural circulation flow rate and coolant temperature at steady-state are compared to the experimental data. The validation is also carried out for the transients in which the sudden reduction in the core power or the feedwater flow takes place. The code's capability to predict the steady-state flow by natural convection and the qualitative behaviour of the primary system in the transients is confirmed. (Author)
Superimposed Code Theorectic Analysis of DNA Codes and DNA Computing
2010-03-01
Bounds for DNA Codes Based on Fibonacci Ensembles of DNA Sequences ”, 2008 IEEE Proceedings of International Symposium on Information Theory, pp. 2292...5, June 2008, pp. 525-34. 32 28. A. Macula, et al., “Random Coding Bounds for DNA Codes Based on Fibonacci Ensembles of DNA Sequences ”, 2008...combinatorial method of bio-memory design and detection that encodes item or process information as numerical sequences represented in DNA. ComDMem is a
Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis
Energy Technology Data Exchange (ETDEWEB)
Hwang, M. K.; Lee, W. J. [Korea Atomic Energy Research Institute, Taejeon (Korea)
2001-03-01
The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a fourth step of the whole project, applying the RELAP5/MOD3/CANDU+ version for the real CANDU plant LOCA Analysis and D2O leakage incident. There are three main models under investigation, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs, especially when CANDU LOCA is tested. Also, for Wolsung unit 1 D2O leakage incident analysis, the plant behavior is predicted with the newly developed version for the first 1000 seconds after onset of the incident, with the main interest aiming for system pressure, level control system, and thermal hydraulic transient behavior of the secondary system. The model applied for this particular application includes heat transfer model of nuclear fuel assembly, decay heat model, and MOV (Motor Operated Valve) model. Finally, the code maintenance work, mainly correcting the known errors, is presented. 12 refs., 26 figs., 3 tabs. (Author)
Verification of combined thermal-hydraulic and heat conduction analysis code FLOWNET/TRUMP
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.
Validation of coupled neutronic / thermal-hydraulic codes for VVER reactors. Final report
Energy Technology Data Exchange (ETDEWEB)
Mittag, S.; Grundmann, U.; Kliem, S.; Kozmenkov, Y.; Rindelhardt, U.; Rohde, U.; Weiss, F.-P.; Langenbuch, S.; Krzykacz-Hausmann, B.; Schmidt, K.-D.; Vanttola, T.; Haemaelaeinen, A.; Kaloinen, E.; Kereszturi, A.; Hegyi, G.; Panka, I.; Hadek, J.; Strmensky, C.; Darilek, P.; Petkov, P.; Stefanova, S.; Kuchin, A.; Khalimonchuk, V.; Hlbocky, P.; Sico, D.; Danilin, S.; Ionov, V.; Nikonov, S.; Powney, D.
2004-08-01
In recent years, the simulation methods for the safety analysis of nuclear power plants have been continuously improved to perform realistic calculations. Therefore in VALCO work package 2 (WP 2), the usual application of coupled neutron-kinetic / thermal-hydraulic codes to VVER has been supplemented by systematic uncertainty and sensitivity analyses. A comprehensive uncertainty analysis has been carried out. The GRS uncertainty and sensitivity method based on the statistical code package SUSA was applied to the two transients studied earlier in SRR-1/95: A load drop of one turbo-generator in Loviisa-1 (VVER-440), and a switch-off of one feed water pump in Balakovo-4 (VVER-1000). The main steps of these analyses and the results obtained by applying different coupled code systems (SMABRE - HEXTRAN, ATHLET - DYN3D, ATHLET - KIKO3D, ATHLET - BIPR-8) are described in this report. The application of this method is only based on variations of input parameter values. No internal code adjustments are needed. An essential result of the analysis using the GRS SUSA methodology is the identification of the input parameters, such as the secondary-circuit pressure, the control-assembly position (as a function of time), and the control-assembly efficiency, that most sensitively affect safety-relevant output parameters, like reactor power, coolant heat-up, and primary pressure. Uncertainty bands for these output parameters have been derived. The variation of potentially uncertain input parameter values as a consequence of uncertain knowledge can activate system actions causing quite different transient evolutions. This gives indications about possible plant conditions that might be reached from the initiating event assuming only small disturbances. In this way, the uncertainty and sensitivity analysis reveals the spectrum of possible transient evolutions. Deviations of SRR-1/95 coupled code calculations from measurements also led to the objective to separate neutron kinetics from
Development of probabilistic internal dosimetry computer code
Noh, Siwan; Kwon, Tae-Eun; Lee, Jai-Ki
2017-02-01
Internal radiation dose assessment involves biokinetic models, the corresponding parameters, measured data, and many assumptions. Every component considered in the internal dose assessment has its own uncertainty, which is propagated in the intake activity and internal dose estimates. For research or scientific purposes, and for retrospective dose reconstruction for accident scenarios occurring in workplaces having a large quantity of unsealed radionuclides, such as nuclear power plants, nuclear fuel cycle facilities, and facilities in which nuclear medicine is practiced, a quantitative uncertainty assessment of the internal dose is often required. However, no calculation tools or computer codes that incorporate all the relevant processes and their corresponding uncertainties, i.e., from the measured data to the committed dose, are available. Thus, the objective of the present study is to develop an integrated probabilistic internal-dose-assessment computer code. First, the uncertainty components in internal dosimetry are identified, and quantitative uncertainty data are collected. Then, an uncertainty database is established for each component. In order to propagate these uncertainties in an internal dose assessment, a probabilistic internal-dose-assessment system that employs the Bayesian and Monte Carlo methods. Based on the developed system, we developed a probabilistic internal-dose-assessment code by using MATLAB so as to estimate the dose distributions from the measured data with uncertainty. Using the developed code, we calculated the internal dose distribution and statistical values ( e.g. the 2.5th, 5th, median, 95th, and 97.5th percentiles) for three sample scenarios. On the basis of the distributions, we performed a sensitivity analysis to determine the influence of each component on the resulting dose in order to identify the major component of the uncertainty in a bioassay. The results of this study can be applied to various situations. In cases of
2013-01-01
The international OECD/NRC PSBT benchmark has been established to provide a test bed for assessing the capabilities of thermal-hydraulic codes and to encourage advancement in the analysis of fluid flow in rod bundles. The benchmark was based on one of the most valuable databases identified for the thermal-hydraulics modeling developed by NUPEC, Japan. The database includes void fraction and departure from nucleate boiling measurements in a representative PWR fuel assembly. On behalf of the be...
Development of best estimate auditing code for CANDU thermal-hydraulic safety analysis
Energy Technology Data Exchange (ETDEWEB)
Chung, Bub Dong; Lee, Won Jae; Hwang, Moon Kyu; Lim, Hong Sik [Korea Atomic Energy Research Institute, Taejeon (Korea)
2002-04-01
The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3.The study was performed by reconsideration of the previous code assessment works and phenomena identification for essential accident scenario. Improvement areas of model development for auditing tool were identified based on the code comparison and PIRT results. Nine models have been improved significantly for the analysis of LOCA and Mon LOCA event. Conceptual problem or separate effect assessment have been performed to verify the model improvement. The linking calculation with CONTAIN 2.0 has been also enabled to establish the unified auditing code system. Analysis for the CANDU plant real transient and hypothetical LOCA bas been performed using the improved version. It has been concluded that the developed version can be utilized for the auditing analysis of LOCA and non-LOCA event for the CANDU reactor. 25 refs., 84 figs., 36 tabs. (Author)
Interface design of VSOP'94 computer code for safety analysis
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.
A computer code for analysis of severe accidents in LWRs
Energy Technology Data Exchange (ETDEWEB)
NONE
2001-07-01
The ICARE2 computer code, developed and validated since 1988 at IPSN (nuclear safety and protection institute), calculates in a mechanistic way the physical and chemical phenomena involved in the core degradation process during possible severe accidents in LWR's. The coupling between ICARE2 and the best-estimate thermal-hydraulics code CATHARE2 was completed at IPSN and led to the release of a first ICARE/CATHARE V1 version in 1999, followed by 2 successive revisions in 2000 and 2001. This documents gathers all the contributions presented at the first international ICARE/CATHARE users'club seminar that took place in November 2001. This seminar was characterized by a high quality and variety of the presentations, showing an increase of reactor applications and user needs in this area (2D/3D aspects, reflooding, corium slumping into the lower head,...). 2 sessions were organized. The first one was dedicated to the applications of ICARE2 V3mod1 against small-scale experiments such as PHEBUS FPT2 and FPT3 tests, PHEBUS AIC, QUENCH experiments, NRU-FLHT-5 test, ACRR-MP1 and DC1 experiments, CORA-PWR tests, and PBF-SFD1.4 test. The second session involved ICARE/CATHARE V1mod1 reactor applications and users'guidelines. Among reactor applications we found: code applicability to high burn-up fuel rods, simulation of the TMI-2 transient, simulation of a PWR-900 high pressure severe accident sequence, and the simulation of a VVER-1000 large break LOCA scenario. (A.C.)
ICAN Computer Code Adapted for Building Materials
Murthy, Pappu L. N.
1997-01-01
The NASA Lewis Research Center has been involved in developing composite micromechanics and macromechanics theories over the last three decades. These activities have resulted in several composite mechanics theories and structural analysis codes whose applications range from material behavior design and analysis to structural component response. One of these computer codes, the Integrated Composite Analyzer (ICAN), is designed primarily to address issues related to designing polymer matrix composites and predicting their properties - including hygral, thermal, and mechanical load effects. Recently, under a cost-sharing cooperative agreement with a Fortune 500 corporation, Master Builders Inc., ICAN was adapted to analyze building materials. The high costs and technical difficulties involved with the fabrication of continuous-fiber-reinforced composites sometimes limit their use. Particulate-reinforced composites can be thought of as a viable alternative. They are as easily processed to near-net shape as monolithic materials, yet have the improved stiffness, strength, and fracture toughness that is characteristic of continuous-fiber-reinforced composites. For example, particlereinforced metal-matrix composites show great potential for a variety of automotive applications, such as disk brake rotors, connecting rods, cylinder liners, and other hightemperature applications. Building materials, such as concrete, can be thought of as one of the oldest materials in this category of multiphase, particle-reinforced materials. The adaptation of ICAN to analyze particle-reinforced composite materials involved the development of new micromechanics-based theories. A derivative of the ICAN code, ICAN/PART, was developed and delivered to Master Builders Inc. as a part of the cooperative activity.
A surface code quantum computer in silicon.
Hill, Charles D; Peretz, Eldad; Hile, Samuel J; House, Matthew G; Fuechsle, Martin; Rogge, Sven; Simmons, Michelle Y; Hollenberg, Lloyd C L
2015-10-01
The exceptionally long quantum coherence times of phosphorus donor nuclear spin qubits in silicon, coupled with the proven scalability of silicon-based nano-electronics, make them attractive candidates for large-scale quantum computing. However, the high threshold of topological quantum error correction can only be captured in a two-dimensional array of qubits operating synchronously and in parallel-posing formidable fabrication and control challenges. We present an architecture that addresses these problems through a novel shared-control paradigm that is particularly suited to the natural uniformity of the phosphorus donor nuclear spin qubit states and electronic confinement. The architecture comprises a two-dimensional lattice of donor qubits sandwiched between two vertically separated control layers forming a mutually perpendicular crisscross gate array. Shared-control lines facilitate loading/unloading of single electrons to specific donors, thereby activating multiple qubits in parallel across the array on which the required operations for surface code quantum error correction are carried out by global spin control. The complexities of independent qubit control, wave function engineering, and ad hoc quantum interconnects are explicitly avoided. With many of the basic elements of fabrication and control based on demonstrated techniques and with simulated quantum operation below the surface code error threshold, the architecture represents a new pathway for large-scale quantum information processing in silicon and potentially in other qubit systems where uniformity can be exploited.
Thermal-hydraulic analysis of SWAMUP facility using ATHLET-SC code
Directory of Open Access Journals (Sweden)
Zidi eWang
2015-03-01
Full Text Available During the loss of coolant accident (LOCA of supercritical water cooled reactor (SCWR, the pressure in the reactor system will undergo a rapid decrease from the supercritical pressure to the subcritical condition. This process is called trans-critical transients, which is of crucial importance for the LOCA analysis of SCWR. In order to simulate the trans-critical transient, a number of system codes for SCWR have been developed up to date. However, the validation work for the trans-critical models in these codes is still missing. The test facility Supercritical WAter MUltiPurpose loop (SWAMUP with 2×2 rod bundle in Shanghai Jiao Tong University (SJTU will be applied to provide test data for code validation. Some pre-test calculations are important and necessary to show the feasibility of the experiment. In this study, trans-critical transient analysis is performed for the SWAMUP facility with the system code ATHLET-SC, which is modified in SJTU, for supercritical water system. This paper presents the system behavior e.g. system pressure, coolant mass flow, cladding temperature during the depressurization. The effects of some important parameters such as heating power, depressurization rate on the system characteristics are also investigated in this paper. Additionally, some sensitivities study of the code models, e.g. heat transfer coefficient, CHF correlation, are analyzed and discussed. The results indicate that the revised system code ATHLET-SC is capable of simulating thermal hydraulic behavior during the trans-critical transient. According to the results, the cladding temperature during the transient is kept at a low value. However, the pressure difference of the heat exchanger after depressurization could reach 6 MPa, which should be considered in the experiment.
Energy Technology Data Exchange (ETDEWEB)
Di Marcello, Valentino, E-mail: valentino.marcello@kit.edu; Escalante, Javier Jimenez; Espinoza, Victor Sanchez
2015-07-15
Highlights: • Simulation of BFBT-BWR steady-state and transient tests with ATHLET. • Validation of thermal-hydraulic models based on pressure drops and void fraction measurements. • TRACE system code is used for the comparative study. • Predictions result in a good agreement with the experiments. • Discrepancies are smaller or comparable with respect to the measurements uncertainty. - Abstract: Validation and qualification of thermal-hydraulic system codes based on separate effect tests are essential for the reliability of numerical tools when applied to nuclear power plant analyses. To this purpose, the Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT) is involved in various validation and qualification activities of different CFD, sub-channel and system codes. In this paper, the capabilities of the thermal-hydraulic code ATHLET are assessed based on the experimental results provided within the NUPEC BFBT benchmark related to key Boiling Water Reactors (BWR) phenomena. Void fraction and pressure drops measurements in the BFBT bundle performed under steady-state and transient conditions which are representative for e.g. turbine trip and recirculation pump trip events, are compared with the numerical results of ATHLET. The comparison of code predictions with the BFBT data has shown good agreement given the experimental uncertainty and the results are consistent with the trends obtained with similar thermal-hydraulic codes.
Assessment of the computer code COBRA/CFTL
Energy Technology Data Exchange (ETDEWEB)
Baxi, C. B.; Burhop, C. J.
1981-07-01
The COBRA/CFTL code has been developed by Oak Ridge National Laboratory (ORNL) for thermal-hydraulic analysis of simulated gas-cooled fast breeder reactor (GCFR) core assemblies to be tested in the core flow test loop (CFTL). The COBRA/CFTL code was obtained by modifying the General Atomic code COBRA*GCFR. This report discusses these modifications, compares the two code results for three cases which represent conditions from fully rough turbulent flow to laminar flow. Case 1 represented fully rough turbulent flow in the bundle. Cases 2 and 3 represented laminar and transition flow regimes. The required input for the COBRA/CFTL code, a sample problem input/output and the code listing are included in the Appendices.
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.
Energy Technology Data Exchange (ETDEWEB)
Wang, Guanghuai; Peng, Changhong; Guo, Yun, E-mail: guoyun79@ustc.edu.cn
2016-11-01
Highlights: • The superheated steam and PWR schemes are analyzed by RELAP5 code. • The influence of non-uniform heating sources is include. • A supposed slow flow decrease case is discussed and the PWR scheme is better. - Abstract: Water cooled blanket (WCB) is very important in the conceptual design and energy transfer in future fusion power plant. One conceptual design of WCB is under computational testing. RELAP5 code, which is mature and often used in transient analysis in Pressurizer water reactor (PWR), is selected as the simulation tool. The complex inner flow channels and heat sources are simplified according to its thermal–hydraulic characteristics. Then the nodal model for REALP5 is built for approximating the conceptual design. Two typical operating plans, superheated steam scheme and PWR scheme, are analyzed. After some adjustments of the inlet flow resistance coefficients of some flow channels, the reasonable stable conditions of both operation plans can be obtained. The stable fluid and wall temperature distributions and pressure drops are studied. At last, a supposed slow flow decreasing is discussed under two operating conditions separately. According to present results, the superheated steam scheme still needs to be further optimized. The PWR scheme shows a very good safety feature.
Energy Technology Data Exchange (ETDEWEB)
Eslinger, Paul W.; Aaberg, Rosanne L.; Lopresti, Charles A.; Miley, Terri B.; Nichols, William E.; Strenge, Dennis L.
2004-09-14
This document contains detailed user instructions for a suite of utility codes developed for Rev. 1 of the Systems Assessment Capability. The suite of computer codes for Rev. 1 of Systems Assessment Capability performs many functions.
Development of a steady thermal-hydraulic analysis code for the China Advanced Research Reactor
Institute of Scientific and Technical Information of China (English)
TIAN Wenxi; QIU Suizheng; GUO Yun; SU Guanghui; JIA Dounan; LIU Tiancai; ZHANG Jianwei
2007-01-01
A multi-channel model steady-state thermalhydraulic analysis code was developed for the China Advanced Research Reactor (CARR). By simulating the whole reactor core, the detailed mass flow distribution in the core was obtained. The result shows that structure size plays the most important role in mass flow distribution, and the influence of core power could be neglected under singlephase flow. The temperature field of the fuel element under unsymmetrical cooling condition was also obtained, which is necessary for further study such as stress analysis, etc. Of the fuel element. At the same time, considering the hot channel effect including engineering factor and nuclear factor, calculation of the mean and hot channel was carried out and it is proved that all thermal-hydraulic parameters satisfy the "Safety design regulation of CARR".
Development of best estimate auditing code for CANDU thermal hydraulic safety analysis
Energy Technology Data Exchange (ETDEWEB)
Chung, B. D.; Lee, W. J.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
2000-03-15
The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model if existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is a third step of the whole project, and expand the RELAP5/MOD3/CANDU version for implementation of LOCA analysis. There are three main area of model development, i.e. moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs. Newly developed version, namely RELAP5/MOD3/CANDU+ is applicable to CANDU plant analysis with keeping the function of light water reactor analysis. The limited validations of model installation were performed. Assessment of CHF model using AECL separated effect test and calculation for Wolsong 2 plant were performed also for the applicability test of the developed version.
Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis
Energy Technology Data Exchange (ETDEWEB)
Chung, B.D.; Lee, W.J.; Lim, H.S. [Korea Atomic Energy Research Institute, Taejeon (Korea)
2000-03-01
The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a third step of the whole project, and expand the RELAP5/MOD3/CANDU version for implementation of LOCA Analysis. There are three main area of model development, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs. Newly developed version, namely RELAP5/MOD3/CANDU+ is applicable to CANDU plant analysis with keeping the function of light water reactor analysis. The limited validations of model installation were performed. Assessment of CHF model using AECL separated effect test and calculation for Wolsong 2 plant were performed also for the applicability test of the developed version. 15 refs., 37 figs., 8 tabs. (Author)
40 CFR 194.23 - Models and computer codes.
2010-07-01
... 40 Protection of Environment 24 2010-07-01 2010-07-01 false Models and computer codes. 194.23... General Requirements § 194.23 Models and computer codes. (a) Any compliance application shall include: (1... obtain stable solutions; (iv) Computer models accurately implement the numerical models; i.e.,...
Energy Technology Data Exchange (ETDEWEB)
Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok [Korea Atomic Energy Research Institute, Taejon (Korea)
1998-06-01
A multi-dimensional realistic thermal-hydraulic system analysis code, MARS version 1.3 has been developed. Main purpose of MARS 1.3 development is to have the realistic analysis capability of transient two-phase thermal-hydraulics of Pressurized Water Reactors (PWRs) especially during Large Break Loss of Coolant Accidents (LBLOCAs) where the multi-dimensional phenomena domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, three-dimensional (3D) reactor vessel analysis code, and RELAP5/MOD3.2.1.2, one-dimensional (1D) reactor system analysis code., Developmental requirements for MARS are chosen not only to best utilize the existing capability of the codes but also to have the enhanced capability in code maintenance, user accessibility, user friendliness, code portability, code readability, and code flexibility. For the maintenance of existing codes capability and the enhancement of code maintenance capability, user accessibility and user friendliness, MARS has been unified to be a single code consisting of 1D module (RELAP5) and 3D module (COBRA-TF). This is realized by implicitly integrating the system pressure matrix equations of hydrodynamic models and solving them simultaneously, by modifying the 1D/3D calculation sequence operable under a single Central Processor Unit (CPU) and by unifying the input structure and the light water property routines of both modules. In addition, the code structure of 1D module is completely restructured using the modular data structure of standard FORTRAN 90, which greatly improves the code maintenance capability, readability and portability. For the code flexibility, a dynamic memory management scheme is applied in both modules. MARS 1.3 now runs on PC/Windows and HP/UNIX platforms having a single CPU, and users have the options to select the 3D module to model the 3D thermal-hydraulics in the reactor vessel or other
Simulation of research loop LOBI-MOD2 with RELAP5/MOD3.3 code for LOBI thermo hydraulic test A1-93
Energy Technology Data Exchange (ETDEWEB)
Pesaran, Farshad; Barati, Ramin [Islamic Azad Univ., Shiraz (Iran, Islamic Republic of). Dept. of Electrical Engineering
2016-06-15
RELAP5/MOD3.3 is one of the used computer codes for the simulation of event thermal-hydraulics of nuclear power plants. The LOBI test facility is a full-power high-pressure integral system test facility, representing an approximately 1: 700 scale model of a 4-loop, 1300 MWe PWR. A new simulation of the small break LOCA test A1-93 has been carried out in a LOBI/Mod2 facility for reaching good agreement and to evaluate the performance of the RELAP5/MOD3.3 code. Good agreement was obtained in general between the code predictions and the experimental data in transient state.
Energy Technology Data Exchange (ETDEWEB)
Kalin, J., E-mail: jan.kalin@zag.s [Slovenian National Building and Civil Engineering Institute, Dimiceva 12, SI-1000 Ljubljana (Slovenia); Petkovsek, B., E-mail: borut.petkovsek@zag.s [Slovenian National Building and Civil Engineering Institute, Dimiceva 12, SI-1000 Ljubljana (Slovenia); Montarnal, Ph., E-mail: philippe.montarnal@cea.f [CEA/Saclay, DM2S/SFME/LSET, Gif-sur-Yvette, 91191 cedex (France); Genty, A., E-mail: alain.genty@cea.f [CEA/Saclay, DM2S/SFME/LSET, Gif-sur-Yvette, 91191 cedex (France); Deville, E., E-mail: estelle.deville@cea.f [CEA/Saclay, DM2S/SFME/LSET, Gif-sur-Yvette, 91191 cedex (France); Krivic, J., E-mail: jure.krivic@geo-zs.s [Geological Survey of Slovenia, Dimiceva 14, SI-1000 Ljubljana (Slovenia); Ratej, J., E-mail: joze.ratej@geo-zs.s [Geological Survey of Slovenia, Dimiceva 14, SI-1000 Ljubljana (Slovenia)
2011-04-15
In the past years the Slovenian Performance Analysis/Safety Assessment team has performed many generic studies for the future Slovenian low and intermediate level waste repository, most recently a Special Safety Analysis for the Krsko site. The modelling approach was to split the problem into three parts: near-field (detailed model of the repository), far-field (i.e., geosphere) and biosphere. In the Special Safety Analysis the code used to perform the near-field calculations was Hydrus2D. Recently the team has begun a cooperation with the French Commisariat al'Energie Atomique/Saclay (CEA/Saclay) and, as a part of this cooperation, began investigations into using the Alliances numerical platform for near-field calculations in order to compare the overall approach and calculated results. The article presents the comparison between these two codes for a silo-type repository that was considered in the Special Safety Analysis. The physical layout and characteristics of the repository are presented and a hydraulic and transport model of the repository is developed and implemented in Alliances. Some analysis of sensitivity to mesh fineness and to simulation timestep has been preformed and is also presented. The compared quantity is the output flux of radionuclides on the boundary of the model. Finally the results from Hydrus2D and Alliances are compared and the differences and similarities are commented.
CATHARE-3: A new system code for thermal-hydraulics in the context of the NEPTUNE project
Energy Technology Data Exchange (ETDEWEB)
Emonot, P., E-mail: philippe.emonot@cea.fr [CEA DEN/DER/SSTH, 17 rue des Martyrs, 38054 Grenoble Cedex 9 (France); Souyri, A., E-mail: annick.souyri@edf.fr [EDF R and D/MFEE, 6 Quai Watier, 78401 Chatou Cedex (France); Gandrille, J.L., E-mail: jeanluc.gandrille@areva.com [AREVA-NP, Tour Areva, 92084 Paris La Defense Cedex (France); Barre, F., E-mail: francois.barre@irsn.fr [IRSN DPAM, BP 3, 13115 Saint-Paul-Lez-Durance Cedex (France)
2011-11-15
After a thorough analysis of the industrial needs and of the limitations of current simulation tools, EDF and CEA (Commissariat a l'Energie Atomique) launched the NEPTUNE Project in 2001 (see) with the support of AREVA-NP and IRSN. The NEPTUNE activities include software development, research in physical modeling and numerical methods, development of advanced instrumentation techniques and new experimental programs. Four different simulation scales were addressed including DNS (Direct Numerical Simulation), CFD in open medium (Computational Fluid Dynamics), component (subchannel-type analysis) and system (reactor modeling) scales. In 2006 CEA, EDF, AREVA-NP and IRSN defined the strategy for the system scale of NEPTUNE and the CATHARE-3 development was launched. The main objectives are: Bullet advanced physical modeling of two-phases flows, mainly by using multi-field and turbulence models, Bullet improved 3D modeling by the use of fine and non conforming structured meshes, Bullet generalized coupling capabilities with other thermal-hydraulic scales and with other disciplines (core physics, structural mechanics, Horizontal-Ellipsis), Bullet extension of the applicability to new Gen IV reactors (Sodium Cooled Fast Breeder Reactors, Gas Cooled Reactors, Supercritical Light Water Reactors), Bullet a true object-oriented code architecture. At the same time CATHARE-3 is in continuity with the CATHARE-2 code which is the current industrial version of CATHARE and internationally used for nuclear power plant safety analysis, in simulators and in coupled simulation tools. The road map of these two codes will allow a smooth transition from CATHARE-2 to CATHARE-3 for all users. This paper gives an overview of the choices made for the development of CATHARE-3 including new physical models, validation strategy and experimental programs, numerical improvements, enhanced coupling capability and software architecture evolution. The current status of the project as well as the
Hanford Meteorological Station computer codes: Volume 6, The SFC computer code
Energy Technology Data Exchange (ETDEWEB)
Andrews, G.L.; Buck, J.W.
1987-11-01
Each hour the Hanford Meteorological Station (HMS), operated by Pacific Northwest Laboratory, records and archives weather observations. Hourly surface weather observations consist of weather phenomena such as cloud type and coverage; dry bulb, wet bulb, and dew point temperatures; relative humidity; atmospheric pressure; and wind speed and direction. The SFC computer code is used to archive those weather observations and apply quality assurance checks to the data. This code accesses an input file, which contains the previous archive's date and hour and an output file, which contains surface observations for the current day. As part of the program, a data entry form consisting of 24 fields must be filled in. The information on the form is appended to the daily file, which provides an archive for the hourly surface observations.
ANTEO+: A subchannel code for thermal-hydraulic analysis of liquid metal cooled systems
Energy Technology Data Exchange (ETDEWEB)
Lodi, F., E-mail: francesco.lodi5@unibo.it [DIN – Laboratory of Montecuccolino, University of Bologna, Via dei Colli 16, 40136 Bologna (Italy); Grasso, G., E-mail: giacomo.grasso@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Mattioli, D., E-mail: davide.mattioli@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Sumini, M., E-mail: marco.sumini@unibo.it [DIN – Laboratory of Montecuccolino, University of Bologna, Via dei Colli 16, 40136 Bologna (Italy)
2016-05-15
Highlights: • The code structure is presented in detail. • The performed validation is outlined. • Results are critically discussed assessing code accuracy. • Conclusions are drawn and ground for future work identified. - Abstract: Liquid metal cooled fast reactors are promising options for achieving the high degrees of safety and sustainability demanded by the Generation IV paradigm. Among the critical aspects to be addressed in the design process, thermal-hydraulics is one of the most challenging; in order to embed safety in the core conceptualization, these aspects are to be considered at the very beginning of the design process, and translated in a design perspective. For achieving these objectives the subchannel code ANTEO+ has been conceived, able to simulate pin bundle arrangements cooled by liquid metals. The main purposes of ANTEO+ are simplifying the problem description maintaining the required accuracy, enabling a more transparent interface with the user, and having a clear and identifiable application domain, in order to help the user interpreting the results and, mostly, defining their confidence. Since ANTEO+ relies on empirical correlations, the validation phase is of paramount importance along with a clear discussion on the simplifications adopted in modeling the conservation equations. In the present work a detailed description of ANTEO+ structure is given along with a thorough validation of the main models implemented for flow split, pressure drops and subchannel temperatures. The analysis confirmed the ability of ANTEO+ in reproducing experimental data in its anticipated validity domain, with a relatively high degree of accuracy when compared to other classical subchannel tools like ENERGY-II, COBRA-IV-I-MIT and BRS-TVS.
Energy Technology Data Exchange (ETDEWEB)
Stroh, K.R.
1979-03-01
The pebble bed reactor's cylindrical core volume contains a random bed of small, spherical fuel-moderator elements. These graphite spheres, containing a central region of dispersed coated-particle fissile and fertile material, are cooled by high pressure helium flowing through the connected interstitial voids. A mathematical model and numerical solution technique have been developed which allow calculation of macroscopic values of thermal-hydraulic variables in an axisymmetric pebble bed nuclear reactor core. The computer program PEBBLE is based on a mathematical model which treats the bed macroscopically as a generating, conducting porous medium. The steady-state model uses a nonlinear Forchheimer-type relation between the coolant pressure gradient and mass flux, with newly derived coefficients for the linear and quadratic resistance terms. The remaining equations in the model make use of mass continuity, and thermal energy balances for the solid and fluid phases.
Private Computing and Mobile Code Systems
Cartrysse, K.
2005-01-01
This thesis' objective is to provide privacy to mobile code. A practical example of mobile code is a mobile software agent that performs a task on behalf of its user. The agent travels over the network and is executed at different locations of which beforehand it is not known whether or not these ca
Energy Technology Data Exchange (ETDEWEB)
Chung, Ji Bum [Institute for Advanced Engineering, Yongin (Korea, Republic of); Park, Jong Woon [Korea Electric Power Research Institute, Taejon (Korea, Republic of)
1998-12-31
In order to enhance the dynamic and interactive simulation capability of a system thermal hydraulic code for nuclear power plant, applicability of flow network models in SINDA/FLUINT{sup TM} has been tested by modeling feedwater system and coupling to DSNP which is one of a system thermal hydraulic simulation code for a pressurized heavy water reactor. The feedwater system is selected since it is one of the most important balance of plant systems with a potential to greatly affect the behavior of nuclear steam supply system. The flow network model of this feedwater system consists of condenser, condensate pumps, low and high pressure heaters, deaerator, feedwater pumps, and control valves. This complicated flow network is modeled and coupled to DSNP and it is tested for several normal and abnormal transient conditions such turbine load maneuvering, turbine trip, and loss of class IV power. The results show reasonable behavior of the coupled code and also gives a good dynamic and interactive simulation capabilities for the several mild transient conditions. It has been found that coupling system thermal hydraulic code with a flow network code is a proper way of upgrading simulation capability of DSNP to mature nuclear plant analyzer (NPA). 5 refs., 10 figs. (Author)
Reducing Computational Overhead of Network Coding with Intrinsic Information Conveying
DEFF Research Database (Denmark)
Heide, Janus; Zhang, Qi; Pedersen, Morten V.;
is RLNC (Random Linear Network Coding) and the goal is to reduce the amount of coding operations both at the coding and decoding node, and at the same time remove the need for dedicated signaling messages. In a traditional RLNC system, coding operation takes up significant computational resources and adds......This paper investigated the possibility of intrinsic information conveying in network coding systems. The information is embedded into the coding vector by constructing the vector based on a set of predefined rules. This information can subsequently be retrieved by any receiver. The starting point...
Energy Technology Data Exchange (ETDEWEB)
Seiler, N.; Ruyer, P.; Biton, B., E-mail: nathalie.seiler@irsn.fr, E-mail: pierre.ruyer@irsn.fr [IRSN/DPAM/SEMCA/LEMAR, CE Cadarache, Saint Paul lez Durance (France)
2011-07-01
This study focuses on thermal-hydraulic simulations, at sub-channel scale, of a damaged PWR reactor core during a Loss Of Coolant Accident (LOCA). The aim of this study is to accurately simulate the thermal-hydraulics to provide the thermal-mechanical code DRACCAR with an accurate wall heat transfer law. This latter code is developed by the French Safety Institute “Institut de Radioprotection et de Surete Nucleaire” (IRSN) to evaluate the thermics and deformations of fuel assemblies within the core. The present paper first describes the methodology considered to evaluate the capabilities of existing codes CATHARE-3 and CESAR to simulate dispersed droplet flows at a sub-channel scale and then provides some first evaluations of them. (author)
Validation of physics and thermalhydraulic computer codes for advanced Candu reactor applications
Energy Technology Data Exchange (ETDEWEB)
Wren, D.J.; Popov, N.; Snell, V.G. [Atomic Energy of Canada Ltd, (Canada)
2004-07-01
Atomic Energy of Canada Ltd. (AECL) is developing an Advanced Candu Reactor (ACR) that is an evolutionary advancement of the currently operating Candu 6 reactors. The ACR is being designed to produce electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The ACR retains the modular Candu concept of horizontal fuel channels surrounded by a heavy water moderator. However, ACR uses slightly enriched uranium fuel compared to the natural uranium used in Candu 6. This achieves the twin goals of improved economics (via large reductions in the heavy water moderator volume and replacement of the heavy water coolant with light water coolant) and improved safety. AECL has developed and implemented a software quality assurance program to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. Since the basic design of the ACR is equivalent to that of the Candu 6, most of the key phenomena associated with the safety analyses of ACR are common, and the Candu industry standard tool-set of safety analysis codes can be applied to the analysis of the ACR. A systematic assessment of computer code applicability addressing the unique features of the ACR design was performed covering the important aspects of the computer code structure, models, constitutive correlations, and validation database. Arising from this assessment, limited additional requirements for code modifications and extensions to the validation databases have been identified. This paper provides an outline of the AECL software quality assurance program process for the validation of computer codes used to perform physics and thermal-hydraulics safety analyses of the ACR. It describes the additional validation work that has been identified for these codes and the planned, and ongoing, experimental programs to extend the code validation as required to address specific ACR design
A thermal-hydraulic code for transient analysis in a channel with a rod bundle
Energy Technology Data Exchange (ETDEWEB)
Khodjaev, I.D. [Research & Engineering Centre of Nuclear Plants Safety, Electrogorsk (Russian Federation)
1995-09-01
The paper contains the model of transient vapor-liquid flow in a channel with a rod bundle of core of a nuclear power plant. The computer code has been developed to predict dryout and post-dryout heat transfer in rod bundles of nuclear reactor core under loss-of-coolant accidents. Economizer, bubble, dispersed-annular and dispersed regimes are taken into account. The computer code provides a three-field representation of two-phase flow in the dispersed-annular regime. Continuous vapor, continuous liquid film and entrained liquid drops are three fields. For the description of dispersed flow regime two-temperatures and single-velocity model is used. Relative droplet motion is taken into account for the droplet-to-vapor heat transfer. The conservation equations for each of regimes are solved using an effective numerical technique. This technique makes it possible to determine distribution of the parameters of flows along the perimeter of fuel elements. Comparison of the calculated results with the experimental data shows that the computer code adequately describes complex processes in a channel with a rod bundle during accident.
Energy Technology Data Exchange (ETDEWEB)
Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Scherer, W.
2000-10-01
V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to HTRs and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. V.S.O.P.(99) represents the further development of V.S.O.P. (97). Compared to its precursor, the code system has been improved in many details. Major extensions have been included concerning the thermal hydraulic sections. Beyond that, the many modules of the code-system have been condensed to only 2 executables in the ''99''-release of V.S.O.P., to be comfortably handled on a WINDOWS-PC or a UNIX-computer. The necessary data input as well as the handling and book-keeping of intermediate data sets has been condensed and simplified. A 64 MB memory should be available for the execution of the code. The hard disk requirement for the executables and the basic libraries associated with the code amounts to about 7 MB. (orig.)
Computer codes for birds of North America
US Fish and Wildlife Service, Department of the Interior — Purpose of paper was to provide a more useful way to provide codes for all North American species, thus making the list useful for virtually all projects concerning...
Thermal-hydraulic analysis of a heavy-water reactor moderator tank using the CUPID Code
Energy Technology Data Exchange (ETDEWEB)
Choi, Su Ryong; Jeong, Jae Jun [Pusan National Univ., Busan (Korea, Republic of); Kim, Hyoung Tae; Yoon, Han Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2014-05-15
In this study, a preliminary analysis is performed for the CANDU moderator tank. The calculation results using the basic case input showed a unrealistic, thermal stratification in the upper region, which was caused by the lack of the momentum of the cooling water from the inlet nozzle. To increase the flow momentum from the inlet nozzle, the cross-section area of each inlet nozzle was reduced by half and, then, the calculation showed very realistic results. It is clear that the modeling of the inlet nozzle affects the calculation result significantly. Further studies are needed for a realistic and efficient simulation of the flow in the Calandria tank. When the core cooling system fails to remove the decay heat from the fuel channels during a loss of coolant accident (LOCA), the pressure tube (PT) could strain to contact its surrounding Calandria tube (CT), which leads to sustained CTs dry out, finally resulting in damages to nuclear fuel. This situation can occur when the degree of the subcooling of the moderator inside the Calandria vessel is insufficient. In this regard, to estimate the local subcooling of the moderator inside the Calandria vessel is very important. However, the local temperature is measured at the inlet and outlet of the vessel only. Therefore, we need to accurately predict the local temperature inside the Calandria vessel.In this study, the thermal-hydraulic analysis of the real-scale heavy-water reactor moderator is carried out using the CUPID code. The applicability of the CUPID code to the analysis of the flow in the Calandria vessel has been assessed in the previous studies.
Numerical Study of Thermal Hydraulic behavior of Pressurizer for PLCS Scenario by CUPID Code
Energy Technology Data Exchange (ETDEWEB)
Lee, Jae Ryong; Yoon, Han Young [KAERI, Daejeon (Korea, Republic of); Yoon, Bo Kam; Kim, Jeong Ju; Park, Jong Cheol; Lee, Gyu Cheon [KEPCO, Daejeon (Korea, Republic of)
2016-05-15
For a malfunction of a pressurizer level control system, a chemical and volume control system (CVCS) charging flowrate becomes a maximum level and a letdown flowrate a minimum level as well. Consequently, a water level and pressure of pressurizer is abnormally increased, which causes a pilot operated relief valve (POSRV) opened. It becomes important to investigate that a mixture from the POSRV becomes single-phase gas or two-phase mixture. In this study, the three-dimensional thermal-hydraulic behavior inside the pressurizer is numerically investigated by the CUPID code. The flow fields highly depend on some parameters such as subcooling of the stored water, interfacial drag model and POSRV opening. Thus, these parameters are examined in this study. These parameters are examined in this study. Less subcooling temperature makes the flow behavior unstable and flashing occur. The two-phase mixture is discharged through the POSRV. Moreover, less flow area delays a discharging flow rate. A sensitivity for the other parameters such critical flow model should be examined for the future work.
Energy Technology Data Exchange (ETDEWEB)
Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my [Nuclear Energy Department, Tenaga Nasional Berhad, Level 32, Dua Sentral, 50470 Kuala Lumpur (Malaysia); Roslan, Ridha [Nuclear Installation Division, Atomic Energy Licensing Board, Batu 24, Jalan Dengkil, 43800 Dengkil, Selangor (Malaysia); Ibrahim, Mohd Rizal Mamat [Technical Support Division, Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)
2014-02-12
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.
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.
Reduced gravity boiling and condensing experiments simulated with the COBRA/TRAC computer code
Cuta, Judith M.; Krotiuk, William
1988-01-01
A series of reduced-gravity two-phase flow experiments has been conducted with a boiler/condenser apparatus in the NASA KC-135 aircraft in order to obtain basic thermal-hydraulic data applicable to analytical design tools. Several test points from the KC-135 tests were selected for simulation by means of the COBRA/TRAC two-fluid, three-field thermal-hydraulic computer code; the points were chosen for a 25-90 percent void-fraction range. The possible causes for the lack of agreement noted between simulations and experiments are explored, with attention to the physical characteristics of two-phase flow in one-G and near-zero-G conditions.
PORPST: A statistical postprocessor for the PORMC computer code
Energy Technology Data Exchange (ETDEWEB)
Eslinger, P.W.; Didier, B.T. (Pacific Northwest Lab., Richland, WA (United States))
1991-06-01
This report describes the theory underlying the PORPST code and gives details for using the code. The PORPST code is designed to do statistical postprocessing on files written by the PORMC computer code. The data written by PORMC are summarized in terms of means, variances, standard deviations, or statistical distributions. In addition, the PORPST code provides for plotting of the results, either internal to the code or through use of the CONTOUR3 postprocessor. Section 2.0 discusses the mathematical basis of the code, and Section 3.0 discusses the code structure. Section 4.0 describes the free-format point command language. Section 5.0 describes in detail the commands to run the program. Section 6.0 provides an example program run, and Section 7.0 provides the references. 11 refs., 1 fig., 17 tabs.
Capabilities of the ATHENA computer code for modeling the SP-100 space reactor concept
Fletcher, C. D.
1985-09-01
The capability to perform thermal-hydraulic analyses of an SP-100 space reactor was demonstrated using the ATHENA computer code. The preliminary General Electric SP-100 design was modeled using Athena. The model simulates the fast reactor, liquid-lithium coolant loops, and lithium-filled heat pipes of this design. Two ATHENA demonstration calculations were performed simulating accident scenarios. A mask for the SP-100 model and an interface with the Nuclear Plant Analyzer (NPA) were developed, allowing a graphic display of the calculated results on the NPA.
Energy Technology Data Exchange (ETDEWEB)
Adoo, N.A., E-mail: nanakwame10@hotmail.com [School of Nuclear and Allied Sciences, University of Ghana, Legon, P.O. Box AE 1, Atomic, Accra (Ghana); Nyarko, B.J.B.; Akaho, E.H.K. [School of Nuclear and Allied Sciences, University of Ghana, Legon, P.O. Box AE 1, Atomic, Accra (Ghana); National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O. Box LG 80, Accra (Ghana); Alhassan, E.; Agbodemegbe, V.Y.; Bansah, C.Y.; Della, R. [School of Nuclear and Allied Sciences, University of Ghana, Legon, P.O. Box AE 1, Atomic, Accra (Ghana)
2011-12-15
The PARET/ANL code has been adapted by the IAEA for testing transient behaviour in research reactors since it provides a coupled thermal hydrodynamic and point kinetics capability for estimating thermal hydraulic margins. A two-channel power peaking profile of the Ghana Research Reactor-1 (GHARR-1) has been developed for the PARET/ANL (Version 7.3; 2007) using the Monte Carlo N-Particle code (MCNP) to determine the thermal hydraulic data for reactivity insertion transients in the range of 2.0 Multiplication-Sign 10{sup -3} {Delta}k/k to 5.5 Multiplication-Sign 10{sup -3} {Delta}k/k. Peak clad and coolant temperatures ranged from 59.18 Degree-Sign C to 112.36 Degree-Sign C and 42.95 Degree-Sign C to 79.42 Degree-Sign C respectively. Calculated safety margins (DNBR) satisfied the MNSR thermal hydraulic design criteria for which no boiling occurs in the reactor core. The generated thermal hydraulic data demonstrated a high inherent safety feature of GHARR-1 for which the high negative reactivity feedback of the moderator limits power excursion and consequently the escalation of the clad temperature.
Optimization of KINETICS Chemical Computation Code
Donastorg, Cristina
2012-01-01
NASA JPL has been creating a code in FORTRAN called KINETICS to model the chemistry of planetary atmospheres. Recently there has been an effort to introduce Message Passing Interface (MPI) into the code so as to cut down the run time of the program. There has been some implementation of MPI into KINETICS; however, the code could still be more efficient than it currently is. One way to increase efficiency is to send only certain variables to all the processes when an MPI subroutine is called and to gather only certain variables when the subroutine is finished. Therefore, all the variables that are used in three of the main subroutines needed to be investigated. Because of the sheer amount of code that there is to comb through this task was given as a ten-week project. I have been able to create flowcharts outlining the subroutines, common blocks, and functions used within the three main subroutines. From these flowcharts I created tables outlining the variables used in each block and important information about each. All this information will be used to determine how to run MPI in KINETICS in the most efficient way possible.
Codes of Ethics for Computing at Russian Institutions and Universities.
Pourciau, Lester J.; Spain, Victoria, Ed.
1997-01-01
To determine the degree to which Russian institutions and universities have formulated and promulgated codes of ethics or policies for acceptable computer use, the author examined Russian institution and university home pages. Lists home pages examined, 10 commandments for computer ethics from the Computer Ethics Institute, and a policy statement…
Energy Technology Data Exchange (ETDEWEB)
Mays, Brian [AREVA Federal Services, Lynchburg, VA (United States); Jackson, R. Brian [TerraPower, Bellevue, WA (United States)
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. 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.
Energy Technology Data Exchange (ETDEWEB)
Fanselau, R.W.; Thakkar, J.G.; Hiestand, J.W.; Cassell, D.
1981-03-01
The Comparative Thermal-Hydraulic Evaluation of Steam Generators program represents an analytical investigation of the thermal-hydraulic characteristics of four PWR steam generators. The analytical tool utilized in this investigation is the CALIPSOS code, a three-dimensional flow distribution code. This report presents the steady state thermal-hydraulic characteristics on the secondary side of a Westinghouse Model 51 steam generator. Details of the CALIPSOS model with accompanying assumptions, operating parameters, and transport correlations are identified. Comprehensive graphical and numerical results are presented to facilitate the desired comparison with other steam generators analyzed by the same flow distribution code.
First vapor explosion calculations performed with MC3D thermal-hydraulic code
Energy Technology Data Exchange (ETDEWEB)
Brayer, C.; Berthoud, G. [CEA Centre d`Etudes de Grenoble, 38 (France). Direction des Reacteurs Nucleaires
1998-01-01
This paper presents the first calculations performed with the `explosion` module of the multiphase computer code MC3D, which is devoted to the fine fragmentation and explosion phase of a fuel coolant interaction. A complete description of the physical laws included in this module is given. The fragmentation models, taking into account two fragmentation mechanisms, a thermal one and an hydrodynamic one, are also developed here. Results to some calculations to test the numerical behavior of MC3D and to test the explosion models in 1D or 2D are also presented. (author)
Continuous Materiality: Through a Hierarchy of Computational Codes
Directory of Open Access Journals (Sweden)
Jichen Zhu
2008-01-01
Full Text Available The legacy of Cartesian dualism inherent in linguistic theory deeply influences current views on the relation between natural language, computer code, and the physical world. However, the oversimplified distinction between mind and body falls short of capturing the complex interaction between the material and the immaterial. In this paper, we posit a hierarchy of codes to delineate a wide spectrum of continuous materiality. Our research suggests that diagrams in architecture provide a valuable analog for approaching computer code in emergent digital systems. After commenting on ways that Cartesian dualism continues to haunt discussions of code, we turn our attention to diagrams and design morphology. Finally we notice the implications a material understanding of code bears for further research on the relation between human cognition and digital code. Our discussion concludes by noticing several areas that we have projected for ongoing research.
An engineering based approach for hydraulic computations in river flows
Di Francesco, S.; Biscarini, C.; Pierleoni, A.; Manciola, P.
2016-06-01
This paper presents an engineering based approach for hydraulic risk evaluation. The aim of the research is to identify a criteria for the choice of the simplest and appropriate model to use in different scenarios varying the characteristics of main river channel. The complete flow field, generally expressed in terms of pressure, velocities, accelerations can be described through a three dimensional approach that consider all the flow properties varying in all directions. In many practical applications for river flow studies, however, the greatest changes occur only in two dimensions or even only in one. In these cases the use of simplified approaches can lead to accurate results, with easy to build and faster simulations. The study has been conducted taking in account a dimensionless parameter of channels (ratio of curvature radius and width of the channel (R/B).
Masuzawa, Toru; Ohta, Akiko; Tanaka, Nobuatu; Qian, Yi; Tsukiya, Tomonori
2009-01-01
The effect of the hydraulic force on magnetically levitated (maglev) pumps should be studied carefully to improve the suspension performance and the reliability of the pumps. A maglev centrifugal pump, developed at Ibaraki University, was modeled with 926 376 hexahedral elements for computational fluid dynamics (CFD) analyses. The pump has a fully open six-vane impeller with a diameter of 72.5 mm. A self-bearing motor suspends the impeller in the radial direction. The maximum pressure head and flow rate were 250 mmHg and 14 l/min, respectively. First, a steady-state analysis was performed using commercial code STAR-CD to confirm the model's suitability by comparing the results with the real pump performance. Second, transient analysis was performed to estimate the hydraulic force on the levitated impeller. The impeller was rotated in steps of 1 degrees using a sliding mesh. The force around the impeller was integrated at every step. The transient analysis revealed that the direction of the radial force changed dynamically as the vane's position changed relative to the outlet port during one circulation, and the magnitude of this force was about 1 N. The current maglev pump has sufficient performance to counteract this hydraulic force. Transient CFD analysis is not only useful for observing dynamic flow conditions in a centrifugal pump but is also effective for obtaining information about the levitation dynamics of a maglev pump.
Directory of Open Access Journals (Sweden)
Enrico Zio
2008-01-01
Full Text Available In the present work, the uncertainties affecting the safety margins estimated from thermal-hydraulic code calculations are captured quantitatively by resorting to the order statistics and the bootstrap technique. The proposed framework of analysis is applied to the estimation of the safety margin, with its confidence interval, of the maximum fuel cladding temperature reached during a complete group distribution blockage scenario in a RBMK-1500 nuclear reactor.
Structural Computer Code Evaluation. Volume I
1976-11-01
Rivlin model for large strains. Other exanmples are given in Reference 5. Hypoelasticity A hypoelastic material is one in which the components of...remains is the application of these codes to specific rocket nozzle problems and the evaluation of their capabilities to model modern nozzle mraterial...behavior. Further work may also require the development of appropriate material property data or new material models to adequately characterize these
Quantum computation with Turaev-Viro codes
Koenig, Robert; Reichardt, Ben W
2010-01-01
The Turaev-Viro invariant for a closed 3-manifold is defined as the contraction of a certain tensor network. The tensors correspond to tetrahedra in a triangulation of the manifold, with values determined by a fixed spherical category. For a manifold with boundary, the tensor network has free indices that can be associated to qudits, and its contraction gives the coefficients of a quantum error-correcting code. The code has local stabilizers determined by Levin and Wen. For example, applied to the genus-one handlebody using the Z_2 category, this construction yields the well-known toric code. For other categories, such as the Fibonacci category, the construction realizes a non-abelian anyon model over a discrete lattice. By studying braid group representations acting on equivalence classes of colored ribbon graphs embedded in a punctured sphere, we identify the anyons, and give a simple recipe for mapping fusion basis states of the doubled category to ribbon graphs. We explain how suitable initial states can ...
Lattice Boltzmann method fundamentals and engineering applications with computer codes
Mohamad, A A
2014-01-01
Introducing the Lattice Boltzmann Method in a readable manner, this book provides detailed examples with complete computer codes. It avoids the most complicated mathematics and physics without scarifying the basic fundamentals of the method.
Tuning complex computer code to data
Energy Technology Data Exchange (ETDEWEB)
Cox, D.; Park, J.S.; Sacks, J.; Singer, C.
1992-01-01
The problem of estimating parameters in a complex computer simulator of a nuclear fusion reactor from an experimental database is treated. Practical limitations do not permit a standard statistical analysis using nonlinear regression methodology. The assumption that the function giving the true theoretical predictions is a realization of a Gaussian stochastic process provides a statistical method for combining information from relatively few computer runs with information from the experimental database and making inferences on the parameters.
Computer aided power flow software engineering and code generation
Energy Technology Data Exchange (ETDEWEB)
Bacher, R. [Swiss Federal Inst. of Tech., Zuerich (Switzerland)
1996-02-01
In this paper a software engineering concept is described which permits the automatic solution of a non-linear set of network equations. The power flow equation set can be seen as a defined subset of a network equation set. The automated solution process is the numerical Newton-Raphson solution process of the power flow equations where the key code parts are the numeric mismatch and the numeric Jacobian term computation. It is shown that both the Jacobian and the mismatch term source code can be automatically generated in a conventional language such as Fortran or C. Thereby one starts from a high level, symbolic language with automatic differentiation and code generation facilities. As a result of this software engineering process an efficient, very high quality newton-Raphson solution code is generated which allows easier implementation of network equation model enhancements and easier code maintenance as compared to hand-coded Fortran or C code.
Computer aided power flow software engineering and code generation
Energy Technology Data Exchange (ETDEWEB)
Bacher, R. [Swiss Federal Inst. of Tech., Zuerich (Switzerland)
1995-12-31
In this paper a software engineering concept is described which permits the automatic solution of a non-linear set of network equations. The power flow equation set can be seen as a defined subset of a network equation set. The automated solution process is the numerical Newton-Raphson solution process of the power flow equations where the key code parts are the numeric mismatch and the numeric Jacobian term computation. It is shown that both the Jacobian and the mismatch term source code can be automatically generated in a conventional language such as Fortran or C. Thereby one starts from a high level, symbolic language with automatic differentiation and code generation facilities. As a result of this software engineering process an efficient, very high quality Newton-Raphson solution code is generated which allows easier implementation of network equation model enhancements and easier code maintenance as compared to hand-coded Fortran or C code.
Energy Technology Data Exchange (ETDEWEB)
Petruzzi, A.; Auria, F. [Pisa Universita, Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Pisa (Italy); Ivanov, K. [Pennsylvania State University, Department of Mechanical and Nuclear Engineering, PA (Italy)
2003-07-01
The paper stresses how the internal assessment of uncertainty is a desirable capability for thermal-hydraulic system codes. This consists of the possibility of obtaining proper uncertainty bands each time a nuclear plant transient scenario is calculated. A methodology suitable for introducing such a capability into a system code is discussed. At the basis of the derivation of the code with (the capability of) internal assessment of uncertainty (CIAU), there is the uncertainty methodology based on the accuracy extrapolation (UMAE), previously proposed by the University of Pisa, although other uncertainty methodologies can be used for the same purpose. The idea of the CIAU is the identification and the characterization of standard plant statuses and the association of uncertainty to each status. One hypercube and one time interval identify the plant status. Quantity and time uncertainties are combined for each plant status. The RELAP5/MOD3.2 system code has been used inside the CIAU to show the applicability of the proposed method. The derivation of the methodology is discussed, and reference results of pressurized water reactor plant transients are shown bounded by the CIAU calculated uncertainty bands. Recently, a new activity has been started with the aim to extend the CIAU to the 3D neutronics/thermal-hydraulics coupled codes. (authors)
APC: A New Code for Atmospheric Polarization Computations
Korkin, Sergey V.; Lyapustin, Alexei I.; Rozanov, Vladimir V.
2014-01-01
A new polarized radiative transfer code Atmospheric Polarization Computations (APC) is described. The code is based on separation of the diffuse light field into anisotropic and smooth (regular) parts. The anisotropic part is computed analytically. The smooth regular part is computed numerically using the discrete ordinates method. Vertical stratification of the atmosphere, common types of bidirectional surface reflection and scattering by spherical particles or spheroids are included. A particular consideration is given to computation of the bidirectional polarization distribution function (BPDF) of the waved ocean surface.
INTERFROST: a benchmark of Thermo-Hydraulic codes for cold regions hydrology
Grenier, Christophe; Roux, Nicolas; Costard, François; Pessel, Marc
2014-05-01
Large focus was put recently on the impact of climate changes in boreal regions due to the large temperature amplitudes expected. Large portions of these regions, corresponding to permafrost areas, are covered by water bodies (lakes, rivers) with very specific evolution and water budget. These water bodies generate taliks (unfrozen zones below) that may play a key role in the context of climate change. Recent studies and modeling exercises showed that a fully coupled 2D or 3D Thermo-Hydraulic (TH) approach is a minimal requirement to model and understand the evolution of the river and lake - soil continuum in a changing climate (e.g. Mc Kenzie et al., 2007; Bense et al 2009, Rowland et al 2011; Painter 2011; Grenier et al 2012; Painter et al 2012 and others from the 2012 special issue Hydrogeology Journal: "Hydrogeology of cold regions"). However, 3D studies are still scarce while numerical approaches can only be validated against analytical solutions for the purely thermal equation with conduction and 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 different codes on provided test cases and/or to have controlled experiments for validation. We propose here to join the INTERFROST benchmark exercise addressing these issues. We give an overview of some of its test cases (phase I) as well as provide the present stand of the exercise and invite other research groups to join. This initial phase of the benchmark consists of some test cases inspired by existing literature (e.g. Mc Kenzie et al., 2007) as well as new ones. Some experimental cases in cold room complement the validation approach. In view of a Phase II, the project is open as well to other test cases reflecting a numerical or a process oriented interest or answering a more general concern among the cold region community. A further purpose of the benchmark exercise is to propel discussions for the
Neutron noise computation using panda deterministic code
Energy Technology Data Exchange (ETDEWEB)
Humbert, Ph. [CEA Bruyeres le Chatel (France)
2003-07-01
PANDA is a general purpose discrete ordinates neutron transport code with deterministic and non deterministic applications. In this paper we consider the adaptation of PANDA to stochastic neutron counting problems. More specifically we consider the first two moments of the count number probability distribution. In a first part we will recall the equations for the single neutron and source induced count number moments with the corresponding expression for the excess of relative variance or Feynman function. In a second part we discuss the numerical solution of these inhomogeneous adjoint time dependent transport coupled equations with discrete ordinate methods. Finally, numerical applications are presented in the third part. (author)
HT2DINV: A 2D forward and inverse code for steady-state and transient hydraulic tomography problems
Soueid Ahmed, A.; Jardani, A.; Revil, A.; Dupont, J. P.
2015-12-01
Hydraulic tomography is a technique used to characterize the spatial heterogeneities of storativity and transmissivity fields. The responses of an aquifer to a source of hydraulic stimulations are used to recover the features of the estimated fields using inverse techniques. We developed a 2D free source Matlab package for performing hydraulic tomography analysis in steady state and transient regimes. The package uses the finite elements method to solve the ground water flow equation for simple or complex geometries accounting for the anisotropy of the material properties. The inverse problem is based on implementing the geostatistical quasi-linear approach of Kitanidis combined with the adjoint-state method to compute the required sensitivity matrices. For undetermined inverse problems, the adjoint-state method provides a faster and more accurate approach for the evaluation of sensitivity matrices compared with the finite differences method. Our methodology is organized in a way that permits the end-user to activate parallel computing in order to reduce the computational burden. Three case studies are investigated demonstrating the robustness and efficiency of our approach for inverting hydraulic parameters.
Detailed thermal-hydraulic computation into a containment building
Energy Technology Data Exchange (ETDEWEB)
Caruso. A.; Flour, I.; Simonin, O. [EDF/LNH, Chatou (France); Cherbonnel, C [EDF/SEPTEN, Villeurbanne (France)
1995-09-01
This paper deals with numerical predictions of the influence of water sprays upon stratifications into a containment building using a two-dimensional two-phase flow code. Basic equations and closure assumptions are briefly presented. A test case in a situation involving spray evaporation is first detailed to illustrate the validation step. Then results are presented for a compressible recirculating flow into a containment building with condensation phenomena.
Kruyt, N.P.; Esch, van B.P.M.; Jonker, J.B.
1999-01-01
A numerical method is presented for the computation of unsteady, three-dimensional potential flows in hydraulic pumps and turbines. The superelement method has been extended in order to eliminate slave degrees of freedom not only from the governing Laplace equation, but also from the Kutta condition
Directory of Open Access Journals (Sweden)
2016-01-01
Full Text Available To determine the operating fields of the tolerances of hydraulic systems parameters for various conditions of work and phases of flight given mathematical relationships and the results obtained in Mathcad in analytical form for the board computer system.
Computer code for intraply hybrid composite design
Chamis, C. C.; Sinclair, J. H.
1981-01-01
A computer program has been developed and is described herein for intraply hybrid composite design (INHYD). The program includes several composite micromechanics theories, intraply hybrid composite theories and a hygrothermomechanical theory. These theories provide INHYD with considerable flexibility and capability which the user can exercise through several available options. Key features and capabilities of INHYD are illustrated through selected samples.
Computer codes used during upgrading activities at MINT TRIGA reactor
Energy Technology Data Exchange (ETDEWEB)
Mohammad Suhaimi Kassim; Adnan Bokhari; Mohd. Idris Taib [Malaysian Institute for Nuclear Technology Research, Kajang (Malaysia)
1999-10-01
MINT TRIGA Reactor is a 1-MW swimming pool nuclear research reactor commissioned in 1982. In 1993, a project was initiated to upgrade the thermal power to 2 MW. The IAEA assistance was sought to assist the various activities relevant to an upgrading exercise. For neutronics calculations, the IAEA has provided expert assistance to introduce the WIMS code, TRIGAP, and EXTERMINATOR2. For thermal-hydraulics calculations, PARET and RELAP5 were introduced. Shielding codes include ANISN and MERCURE. However, in the middle of 1997, MINT has decided to change the scope of the project to safety upgrading of the MINT Reactor. This paper describes some of the activities carried out during the upgrading process. (author)
Energy Technology Data Exchange (ETDEWEB)
Park, Youngjae; Kim, Iljin; Kim, Hyungdae [Kyung Hee University, Yongin (Korea, Republic of)
2015-10-15
Diverse integral/small-modular reactors (SMRs) have been developed. Once-through steam generator (OTSG) which generates superheated steam without steam separator and dryer was used in the SMRs to reduce volume of steam generator. It would be possible to design a new steam generator with best estimate thermal-hydraulic codes such as RELAP and MARS. However, it is not convenience to use the general purpose thermal-hydraulic analysis code to design a specific component of nuclear power plants. A widely used simulation tool for thermal-hydraulic analysis of drum-type steam generators is ATHOS, which allows 3D analysis. On the other hand, a simple 1D thermal-hydraulic analysis code might be accurate enough for the conceptual design of OTSG. In this study, thermal-hydraulic analysis code for conceptual design of OTSG was developed using 1D homogeneous equilibrium model (HEM). A benchmark calculation was also conducted to verify and validate the prediction accuracy of the developed code by comparing with the analysis results with MARS. Finally, conceptual design of OTSG was conducted by the developed code. A simple 1D thermal-hydraulic analysis code was developed for the purpose of conceptual design OTSG for SMRs. A set of benchmark calculations was conducted to verify and validate the analysis accuracy of the developed code by comparing results obtained with a best-estimated thermal-hydraulic analysis code, MARS. Finally, analysis of two different OTSG design concepts with superheating and recirculation was demonstrated using the developed code.
Computer vision cracks the leaf code.
Wilf, Peter; Zhang, Shengping; Chikkerur, Sharat; Little, Stefan A; Wing, Scott L; Serre, Thomas
2016-03-22
Understanding the extremely variable, complex shape and venation characters of angiosperm leaves is one of the most challenging problems in botany. Machine learning offers opportunities to analyze large numbers of specimens, to discover novel leaf features of angiosperm clades that may have phylogenetic significance, and to use those characters to classify unknowns. Previous computer vision approaches have primarily focused on leaf identification at the species level. It remains an open question whether learning and classification are possible among major evolutionary groups such as families and orders, which usually contain hundreds to thousands of species each and exhibit many times the foliar variation of individual species. Here, we tested whether a computer vision algorithm could use a database of 7,597 leaf images from 2,001 genera to learn features of botanical families and orders, then classify novel images. The images are of cleared leaves, specimens that are chemically bleached, then stained to reveal venation. Machine learning was used to learn a codebook of visual elements representing leaf shape and venation patterns. The resulting automated system learned to classify images into families and orders with a success rate many times greater than chance. Of direct botanical interest, the responses of diagnostic features can be visualized on leaf images as heat maps, which are likely to prompt recognition and evolutionary interpretation of a wealth of novel morphological characters. With assistance from computer vision, leaves are poised to make numerous new contributions to systematic and paleobotanical studies.
DEFF Research Database (Denmark)
Dam Jensen, Mette; Ingildsen, Pernille; Rasmussen, Michael R.;
2005-01-01
shown to be more effective than others. To improve the design of less effective plants Computational Fluid Dynamics (CFD) modelling of hydraulics and sedimentation has been applied. The paper discusses the results at one particular plant experiencing problems with partly short-circuiting of the inlet...... been suggested and tested by means of computational fluid dynamics modelling. The most promissing design change have been found and reported....
Validation of CESAR Thermal-hydraulic Module of ASTEC V1.2 Code on BETHSY Experiments
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.
Assessment of computer codes for VVER-440/213-type nuclear power plants
Energy Technology Data Exchange (ETDEWEB)
Szabados, L.; Ezsol, Gy.; Perneczky [Atomic Energy Research Institute, Budapest (Hungary)
1995-09-01
Nuclear power plant of VVER-440/213 designed by the former USSR have a number of special features. As a consequence of these features the transient behaviour of such a reactor system should be different from the PWR system behaviour. To study the transient behaviour of the Hungarian Paks Nuclear Power Plant of VVER-440/213-type both analytical and experimental activities have been performed. The experimental basis of the research in the PMK-2 integral-type test facility , which is a scaled down model of the plant. Experiments performed on this facility have been used to assess thermal-hydraulic system codes. Four tests were selected for {open_quotes}Standard Problem Exercises{close_quotes} of the International Atomic Energy Agency. Results of the 4th Exercise, of high international interest, are presented in the paper, focusing on the essential findings of the assessment of computer codes.
HUDU: The Hanford Unified Dose Utility computer code
Energy Technology Data Exchange (ETDEWEB)
Scherpelz, R.I.
1991-02-01
The Hanford Unified Dose Utility (HUDU) computer program was developed to provide rapid initial assessment of radiological emergency situations. The HUDU code uses a straight-line Gaussian atmospheric dispersion model to estimate the transport of radionuclides released from an accident site. For dose points on the plume centerline, it calculates internal doses due to inhalation and external doses due to exposure to the plume. The program incorporates a number of features unique to the Hanford Site (operated by the US Department of Energy), including a library of source terms derived from various facilities' safety analysis reports. The HUDU code was designed to run on an IBM-PC or compatible personal computer. The user interface was designed for fast and easy operation with minimal user training. The theoretical basis and mathematical models used in the HUDU computer code are described, as are the computer code itself and the data libraries used. Detailed instructions for operating the code are also included. Appendices to the report contain descriptions of the program modules, listings of HUDU's data library, and descriptions of the verification tests that were run as part of the code development. 14 refs., 19 figs., 2 tabs.
Computer code applicability assessment for the advanced Candu reactor
Energy Technology Data Exchange (ETDEWEB)
Wren, D.J.; Langman, V.J.; Popov, N.; Snell, V.G. [Atomic Energy of Canada Ltd (Canada)
2004-07-01
AECL Technologies, the 100%-owned US subsidiary of Atomic Energy of Canada Ltd. (AECL), is currently the proponents of a pre-licensing review of the Advanced Candu Reactor (ACR) with the United States Nuclear Regulatory Commission (NRC). A key focus topic for this pre-application review is the NRC acceptance of the computer codes used in the safety analysis of the ACR. These codes have been developed and their predictions compared against experimental results over extended periods of time in Canada. These codes have also undergone formal validation in the 1990's. In support of this formal validation effort AECL has developed, implemented and currently maintains a Software Quality Assurance program (SQA) to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. This paper discusses the SQA program used to develop, qualify and maintain the computer codes used in ACR safety analysis, including the current program underway to confirm the applicability of these computer codes for use in ACR safety analyses. (authors)
Energy Technology Data Exchange (ETDEWEB)
Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Scari, Maria E., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: melizabethscari@yahoo.com [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Belo Horizonte (Brazil); Miro, Rafael; Verdu, Gumersindo, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidad Politecnica de Valencia (Spain). Departamento de Ingenieria Quimica y Nuclear
2015-07-01
Simulations and analyses of nuclear reactors have been improved by utilization of coupled thermal-hydraulic (TH) and neutron kinetics (NK) system codes especially to simulate transients that involve strong feedback effects between NK and TH. The TH-NK coupling technique was initially developed and used to simulate the behavior of power reactors; however, several coupling methodologies are now being applied for research reactors. This work presents the coupling methodology application between RELAP5 and PARCS codes using as a model the TRIGA IPR-R1 research reactor. Analyses of steady state and transient conditions and comparisons with results from simulations using only the RELAP5 code are being presented in this paper. (author)
Energy Technology Data Exchange (ETDEWEB)
KALINICHENKO,S.D.KROSHILIN,A.E.KROSHILIN,V.E.SMIRNOV,A.V.KOHUT,P.
2004-03-15
In this paper we present verification results of the BAGIRA code that was performed using data from integral thermal-hydraulic experimental test facilities as well as data obtained from operating nuclear power plants. BAGIRA is a three-dimensional numerical best-estimate code that includes non-homogeneous modeling. Special consideration was given to the recently completed experimental data from the PSB-VVER integral test facility (EREC, Electrogorsk, Russia)--a new Russian large-scale four-loop unit, which has been designed to model the primary circuits of VVER-1000 type reactors. It is demonstrated that the code BAGIRA can be used to analyze nuclear reactor behavior under normal and accident conditions.
Experimental methodology for computational fluid dynamics code validation
Energy Technology Data Exchange (ETDEWEB)
Aeschliman, D.P.; Oberkampf, W.L.
1997-09-01
Validation of Computational Fluid Dynamics (CFD) codes is an essential element of the code development process. Typically, CFD code validation is accomplished through comparison of computed results to previously published experimental data that were obtained for some other purpose, unrelated to code validation. As a result, it is a near certainty that not all of the information required by the code, particularly the boundary conditions, will be available. The common approach is therefore unsatisfactory, and a different method is required. This paper describes a methodology developed specifically for experimental validation of CFD codes. The methodology requires teamwork and cooperation between code developers and experimentalists throughout the validation process, and takes advantage of certain synergisms between CFD and experiment. The methodology employs a novel uncertainty analysis technique which helps to define the experimental plan for code validation wind tunnel experiments, and to distinguish between and quantify various types of experimental error. The methodology is demonstrated with an example of surface pressure measurements over a model of varying geometrical complexity in laminar, hypersonic, near perfect gas, 3-dimensional flow.
Energy Technology Data Exchange (ETDEWEB)
Zemulis, G.; Jasiulevicius, A. [Kaunas University of Technology, Dept. of Thermal and Nuclear Energy, Kaunas, (Lithuania)
2001-07-01
Reactor safety is the most important issue in nuclear engineering. It concerns the capability of the nuclear object to withhold the main safety and reliability criterion within specified range during both normal operation and transient conditions. Three types of assessment are to be performed in order to establish the nuclear power plant safety level: neutronic calculations; thermal hydraulic calculations; mechanical design calculations. Calculations of the thermal hydraulic parameters of the RBMK-1500 reactor main circulation circuit (MCC) are presented in this paper. The aim of this work was to test the capability of the APROS code to simulate the behavior of the RBMK-1500 type reactor main circulation circuit during normal operation and transients. (author)
Computer Security: better code, fewer problems
Stefan Lueders, Computer Security Team
2016-01-01
The origin of many security incidents is negligence or unintentional mistakes made by web developers or programmers. In the rush to complete the work, due to skewed priorities, or just to ignorance, basic security principles can be omitted or forgotten. The resulting vulnerabilities lie dormant until the evil side spots them and decides to hit hard. Computer security incidents in the past have put CERN’s reputation at risk due to websites being defaced with negative messages about the Organization, hash files of passwords being extracted, restricted data exposed… And it all started with a little bit of negligence! If you check out the Top 10 web development blunders, you will see that the most prevalent mistakes are: Not filtering input, e.g. accepting “<“ or “>” in input fields even if only a number is expected. Not validating that input: you expect a birth date? So why accept letters? &...
Rejwer, Ewa
2015-01-01
Strong interaction of closely located, nearly parallel hydraulic fractures and its influence on their propagation are studied. Both computational and physical aspects of the problem are considered. It is shown that from the computational point of view, when a distance between cracks is small as compared with their sizes, the system becomes ill-conditioned and numerical results deteriorate. The physical consequence of the interaction consists in decreasing of the crack opening and even greater decrease of conductivity. Then the resistance to fluid flow grows what results in the propagation of only those fractures, the distance between which is large enough. The research aims to suggests a means to overcome the computational difficulty and to improve numerical simulation of hydraulic fractures in shales. Numerical experiments are carried out for a 2D problem by using the complex variable hypersingular boundary element method of higher order accuracy. The condition number of the main matrix of a system, the open...
A three-dimensional magnetostatics computer code for insertion devices.
Chubar, O; Elleaume, P; Chavanne, J
1998-05-01
RADIA is a three-dimensional magnetostatics computer code optimized for the design of undulators and wigglers. It solves boundary magnetostatics problems with magnetized and current-carrying volumes using the boundary integral approach. The magnetized volumes can be arbitrary polyhedrons with non-linear (iron) or linear anisotropic (permanent magnet) characteristics. The current-carrying elements can be straight or curved blocks with rectangular cross sections. Boundary conditions are simulated by the technique of mirroring. Analytical formulae used for the computation of the field produced by a magnetized volume of a polyhedron shape are detailed. The RADIA code is written in object-oriented C++ and interfaced to Mathematica [Mathematica is a registered trademark of Wolfram Research, Inc.]. The code outperforms currently available finite-element packages with respect to the CPU time of the solver and accuracy of the field integral estimations. An application of the code to the case of a wedge-pole undulator is presented.
Low Computational Complexity Network Coding For Mobile Networks
DEFF Research Database (Denmark)
Heide, Janus
2012-01-01
Network Coding (NC) is a technique that can provide benefits in many types of networks, some examples from wireless networks are: In relay networks, either the physical or the data link layer, to reduce the number of transmissions. In reliable multicast, to reduce the amount of signaling and enable...... cooperation among receivers. In meshed networks, to simplify routing schemes and to increase robustness toward node failures. This thesis deals with implementation issues of one NC technique namely Random Linear Network Coding (RLNC) which can be described as a highly decentralized non-deterministic intra......-flow coding technique. One of the key challenges of this technique is its inherent computational complexity which can lead to high computational load and energy consumption in particular on the mobile platforms that are the target platform in this work. To increase the coding throughput several...
Recent applications of the transonic wing analysis computer code, TWING
Subramanian, N. R.; Holst, T. L.; Thomas, S. D.
1982-01-01
An evaluation of the transonic-wing-analysis computer code TWING is given. TWING utilizes a fully implicit approximate factorization iteration scheme to solve the full potential equation in conservative form. A numerical elliptic-solver grid-generation scheme is used to generate the required finite-difference mesh. Several wing configurations were analyzed, and the limits of applicability of this code was evaluated. Comparisons of computed results were made with available experimental data. Results indicate that the code is robust, accurate (when significant viscous effects are not present), and efficient. TWING generally produces solutions an order of magnitude faster than other conservative full potential codes using successive-line overrelaxation. The present method is applicable to a wide range of isolated wing configurations including high-aspect-ratio transport wings and low-aspect-ratio, high-sweep, fighter configurations.
V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009
Energy Technology Data Exchange (ETDEWEB)
Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.
2010-07-15
V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)
Directory of Open Access Journals (Sweden)
Surian Pinem
2014-01-01
Full Text Available A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the few-group neutron diffusion equation in 3-dimensional geometry for typical PWR static and transient analyses. The spatial variables are treated by using a polynomial nodal method while for the neutron dynamic solver the adiabatic and improved quasistatic methods are adopted. In this paper we report the benchmark calculation results of the code against the OECD/NEA CRP PWR rod ejection cases. The objective of this work is to determine the accuracy of NODAL3 code in analysing the reactivity initiated accident due to the control rod ejection. The NEACRP PWR rod ejection cases are chosen since many organizations participated in the NEA project using various methods as well as approximations, so that, in addition to the reference solutions, the calculation results of NODAL3 code can also be compared to other codes’ results. The transient parameters to be verified are time of power peak, power peak, final power, final average Doppler temperature, maximum fuel temperature, and final coolant temperature. The results of NODAL3 code agree well with the PHANTHER reference solutions in 1993 and 1997 (revised. Comparison with other validated codes, DYN3D/R and ANCK, shows also a satisfactory agreement.
FLASH: A finite element computer code for variably saturated flow
Energy Technology Data Exchange (ETDEWEB)
Baca, R.G.; Magnuson, S.O.
1992-05-01
A numerical model was developed for use in performance assessment studies at the INEL. The numerical model, referred to as the FLASH computer code, is designed to simulate two-dimensional fluid flow in fractured-porous media. The code is specifically designed to model variably saturated flow in an arid site vadose zone and saturated flow in an unconfined aquifer. In addition, the code also has the capability to simulate heat conduction in the vadose zone. This report presents the following: description of the conceptual frame-work and mathematical theory; derivations of the finite element techniques and algorithms; computational examples that illustrate the capability of the code; and input instructions for the general use of the code. The FLASH computer code is aimed at providing environmental scientists at the INEL with a predictive tool for the subsurface water pathway. This numerical model is expected to be widely used in performance assessments for: (1) the Remedial Investigation/Feasibility Study process and (2) compliance studies required by the US Department of Energy Order 5820.2A.
Energy Technology Data Exchange (ETDEWEB)
Mann, F.M.
1998-01-26
The Tank Waste Remediation System (TWRS) is responsible for the safe storage, retrieval, and disposal of waste currently being held in 177 underground tanks at the Hanford Site. In order to successfully carry out its mission, TWRS must perform environmental analyses describing the consequences of tank contents leaking from tanks and associated facilities during the storage, retrieval, or closure periods and immobilized low-activity tank waste contaminants leaving disposal facilities. Because of the large size of the facilities and the great depth of the dry zone (known as the vadose zone) underneath the facilities, sophisticated computer codes are needed to model the transport of the tank contents or contaminants. This document presents the code selection criteria for those vadose zone analyses (a subset of the above analyses) where the hydraulic properties of the vadose zone are constant in time the geochemical behavior of the contaminant-soil interaction can be described by simple models, and the geologic or engineered structures are complicated enough to require a two-or three dimensional model. Thus, simple analyses would not need to use the fairly sophisticated codes which would meet the selection criteria in this document. Similarly, those analyses which involve complex chemical modeling (such as those analyses involving large tank leaks or those analyses involving the modeling of contaminant release from glass waste forms) are excluded. The analyses covered here are those where the movement of contaminants can be relatively simply calculated from the moisture flow. These code selection criteria are based on the information from the low-level waste programs of the US Department of Energy (DOE) and of the US Nuclear Regulatory Commission as well as experience gained in the DOE Complex in applying these criteria. Appendix table A-1 provides a comparison between the criteria in these documents and those used here. This document does not define the models (that
MMA, A Computer Code for Multi-Model Analysis
Poeter, Eileen P.; Hill, Mary C.
2007-01-01
This report documents the Multi-Model Analysis (MMA) computer code. MMA can be used to evaluate results from alternative models of a single system using the same set of observations for all models. As long as the observations, the observation weighting, and system being represented are the same, the models can differ in nearly any way imaginable. For example, they may include different processes, different simulation software, different temporal definitions (for example, steady-state and transient models could be considered), and so on. The multiple models need to be calibrated by nonlinear regression. Calibration of the individual models needs to be completed before application of MMA. MMA can be used to rank models and calculate posterior model probabilities. These can be used to (1) determine the relative importance of the characteristics embodied in the alternative models, (2) calculate model-averaged parameter estimates and predictions, and (3) quantify the uncertainty of parameter estimates and predictions in a way that integrates the variations represented by the alternative models. There is a lack of consensus on what model analysis methods are best, so MMA provides four default methods. Two are based on Kullback-Leibler information, and use the AIC (Akaike Information Criterion) or AICc (second-order-bias-corrected AIC) model discrimination criteria. The other two default methods are the BIC (Bayesian Information Criterion) and the KIC (Kashyap Information Criterion) model discrimination criteria. Use of the KIC criterion is equivalent to using the maximum-likelihood Bayesian model averaging (MLBMA) method. AIC, AICc, and BIC can be derived from Frequentist or Bayesian arguments. The default methods based on Kullback-Leibler information have a number of theoretical advantages, including that they tend to favor more complicated models as more data become available than do the other methods, which makes sense in many situations. Many applications of MMA will
Auder, Benjamin; Iooss, Bertrand; Marques, Michel
2010-01-01
To perform uncertainty, sensitivity or optimization analysis on scalar variables calculated by a cpu time expensive computer code, a widely accepted methodology consists in first identifying the most influential uncertain inputs (by screening techniques), and then in replacing the cpu time expensive model by a cpu inexpensive mathematical function, called a metamodel. This paper extends this methodology to the functional output case, for instance when the model output variables are curves. The screening approach is based on the analysis of variance and principal component analysis of output curves. The functional metamodeling consists in a curve classification step, a dimension reduction step, then a classical metamodeling step. An industrial nuclear reactor application (dealing with uncertainties in the pressurized thermal shock analysis) illustrates all these steps.
Highly Optimized Code Generation for Stencil Codes with Computation Reuse for GPUs
Institute of Scientific and Technical Information of China (English)
Wen-Jing Ma; Kan Gao; Guo-Ping Long
2016-01-01
Computation reuse is known as an effective optimization technique. However, due to the complexity of modern GPU architectures, there is yet not enough understanding regarding the intriguing implications of the interplay of compu-tation reuse and hardware specifics on application performance. In this paper, we propose an automatic code generator for a class of stencil codes with inherent computation reuse on GPUs. For such applications, the proper reuse of intermediate results, combined with careful register and on-chip local memory usage, has profound implications on performance. Current state of the art does not address this problem in depth, partially due to the lack of a good program representation that can expose all potential computation reuse. In this paper, we leverage the computation overlap graph (COG), a simple representation of data dependence and data reuse with “element view”, to expose potential reuse opportunities. Using COG, we propose a portable code generation and tuning framework for GPUs. Compared with current state-of-the-art code generators, our experimental results show up to 56.7%performance improvement on modern GPUs such as NVIDIA C2050.
Parallelization of Finite Element Analysis Codes Using Heterogeneous Distributed Computing
Ozguner, Fusun
1996-01-01
Performance gains in computer design are quickly consumed as users seek to analyze larger problems to a higher degree of accuracy. Innovative computational methods, such as parallel and distributed computing, seek to multiply the power of existing hardware technology to satisfy the computational demands of large applications. In the early stages of this project, experiments were performed using two large, coarse-grained applications, CSTEM and METCAN. These applications were parallelized on an Intel iPSC/860 hypercube. It was found that the overall speedup was very low, due to large, inherently sequential code segments present in the applications. The overall execution time T(sub par), of the application is dependent on these sequential segments. If these segments make up a significant fraction of the overall code, the application will have a poor speedup measure.
Institute of Scientific and Technical Information of China (English)
Feng Yi; Li Li; Tian Shujun
2003-01-01
Optimization design of hydraulic manifold blocks (HMB) is studied as a complex solid spatial layout problem. Based on comprehensive research into structure features and design rules of HMB, an optimal mathematical model for this problem is presented. Using human-computer cooperative genetic algorithm (GA) and its hybrid optimization strategies, integrated layout and connection design schemes of HMB can be automatically optimized. An example is given to testify it.
Prodeto, a computer code for probabilistic fatigue design
Energy Technology Data Exchange (ETDEWEB)
Braam, H. [ECN-Solar and Wind Energy, Petten (Netherlands); Christensen, C.J.; Thoegersen, M.L. [Risoe National Lab., Roskilde (Denmark); Ronold, K.O. [Det Norske Veritas, Hoevik (Norway)
1999-03-01
A computer code for structural relibility analyses of wind turbine rotor blades subjected to fatigue loading is presented. With pre-processors that can transform measured and theoretically predicted load series to load range distributions by rain-flow counting and with a family of generic distribution models for parametric representation of these distribution this computer program is available for carying through probabilistic fatigue analyses of rotor blades. (au)
Institute of Scientific and Technical Information of China (English)
无
2010-01-01
Previously it was assumed that the pressure within the cavity or on the cavity surface remained constant and the vapor pressure of clean water at 20°C and 0 m altitude was utilized as the computational boundary for cavitating flows in hydraulic turbomachinery. Cavitation was confused with vaporization, and the effect of water quality on cavitation pressure characteristics was not taken into account. In recent years, lots of experiments of cavitation pressure characteristics of different water qualities including different sand concentrations of sand water and different altitudes of clean water have been performed by the authors, and the important influences of water quality on cavitation pressure characteristic have been validated. Thus the water quality should be involved in the cavitating flows computation. In the present paper, the effect of water quality on the cavitation pressure characteristic is analyzed and the computational method and theory of cavitating flows for hydraulic turbomachinery that considers the influence of water quality are proposed. The theory is suitable for both the potential flow method and the two-phase flow method for cavitating flows simulation. Finally, the validation results for cavitating flows in a hydraulic tur- bine indicate the significant influences of water quality on the cavitating flow performance.
Methods and computer codes for nuclear systems calculations
Indian Academy of Sciences (India)
B P Kochurov; A P Knyazev; A Yu Kwaretzkheli
2007-02-01
Some numerical methods for reactor cell, sub-critical systems and 3D models of nuclear reactors are presented. The methods are developed for steady states and space–time calculations. Computer code TRIFON solves space-energy problem in (, ) systems of finite height and calculates heterogeneous few-group matrix parameters of reactor cells. These parameters are used as input data in the computer code SHERHAN solving the 3D heterogeneous reactor equation for steady states and 3D space–time neutron processes simulation. Modification of TRIFON was developed for the simulation of space–time processes in sub-critical systems with external sources. An option of SHERHAN code for the system with external sources is under development.
Computer code for double beta decay QRPA based calculations
Energy Technology Data Exchange (ETDEWEB)
Barbero, C. A.; Mariano, A. [Departamento de Física, Facultad de Ciencias Exactas, Universidad Nacional de La Plata, La Plata, Argentina and Instituto de Física La Plata, CONICET, La Plata (Argentina); Krmpotić, F. [Instituto de Física La Plata, CONICET, La Plata, Argentina and Instituto de Física Teórica, Universidade Estadual Paulista, São Paulo (Brazil); Samana, A. R.; Ferreira, V. dos Santos [Departamento de Ciências Exatas e Tecnológicas, Universidade Estadual de Santa Cruz, BA (Brazil); Bertulani, C. A. [Department of Physics, Texas A and M University-Commerce, Commerce, TX (United States)
2014-11-11
The computer code developed by our group some years ago for the evaluation of nuclear matrix elements, within the QRPA and PQRPA nuclear structure models, involved in neutrino-nucleus reactions, muon capture and β{sup ±} processes, is extended to include also the nuclear double beta decay.
Plagiarism Detection Algorithm for Source Code in Computer Science Education
Liu, Xin; Xu, Chan; Ouyang, Boyu
2015-01-01
Nowadays, computer programming is getting more necessary in the course of program design in college education. However, the trick of plagiarizing plus a little modification exists among some students' home works. It's not easy for teachers to judge if there's plagiarizing in source code or not. Traditional detection algorithms cannot fit this…
Connecting Neural Coding to Number Cognition: A Computational Account
Prather, Richard W.
2012-01-01
The current study presents a series of computational simulations that demonstrate how the neural coding of numerical magnitude may influence number cognition and development. This includes behavioral phenomena cataloged in cognitive literature such as the development of numerical estimation and operational momentum. Though neural research has…
General review of the MOSTAS computer code for wind turbines
Dungundji, J.; Wendell, J. H.
1981-01-01
The MOSTAS computer code for wind turbine analysis is reviewed, and techniques and methods used in its analyses are described. Impressions of its strengths and weakness, and recommendations for its application, modification, and further development are made. Basic techniques used in wind turbine stability and response analyses for systems with constant and periodic coefficients are reviewed.
Energy Technology Data Exchange (ETDEWEB)
Alva N, J.
2010-07-01
In this thesis, some fundamental knowledge is presented about uncertainty analysis and about diverse methodologies applied in the study of nuclear power plant transient event analysis, particularly related to thermal hydraulics phenomena. These concepts and methodologies mentioned in this work come from a wide bibliographical research in the nuclear power subject. Methodologies for uncertainty analysis have been developed by quite diverse institutions, and they have been widely used worldwide for application to results from best-estimate-type computer codes in nuclear reactor thermal hydraulics and safety analysis. Also, the main uncertainty sources, types of uncertainties, and aspects related to best estimate modeling and methods are introduced. Once the main bases of uncertainty analysis have been set, and some of the known methodologies have been introduced, it is presented in detail the CSAU methodology, which will be applied in the analyses. The main objective of this thesis is to compare the results of an uncertainty and sensibility analysis by using the Response Surface Technique to the application of W ilks formula, apply through a loss coolant experiment and an event of rise in a BWR. Both techniques are options in the part of uncertainty and sensibility analysis of the CSAU methodology, which was developed for the analysis of transients and accidents at nuclear power plants, and it is the base of most of the methodologies used in licensing of nuclear power plants practically everywhere. Finally, the results of applying both techniques are compared and discussed. (Author)
Energy Technology Data Exchange (ETDEWEB)
Berna, G. A; Bohn, M. P.; Rausch, W. N.; Williford, R. E.; Lanning, D. D.
1981-01-01
FRAPCON-2 is a FORTRAN IV computer code that calculates the steady state response of light Mater reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, deformation, and tai lure histories of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (a) heat conduction through the fuel and cladding, (b) cladding elastic and plastic deformation, (c) fuel-cladding mechanical interaction, (d) fission gas release, (e} fuel rod internal gas pressure, (f) heat transfer between fuel and cladding, (g) cladding oxidation, and (h) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat transfer correlations. FRAPCON-2 is programmed for use on the CDC Cyber 175 and 176 computers. The FRAPCON-2 code Is designed to generate initial conditions for transient fuel rod analysis by either the FRAP-T6 computer code or the thermal-hydraulic code, RELAP4/MOD7 Version 2.
TAPINS: A THERMAL-HYDRAULIC SYSTEM CODE FOR TRANSIENT ANALYSIS OF A FULLY-PASSIVE INTEGRAL PWR
Directory of Open Access Journals (Sweden)
YEON-GUN LEE
2013-08-01
Full Text Available REX-10 is a fully-passive small modular reactor in which the coolant flow is driven by natural circulation, the RCS is pressurized by a steam-gas pressurizer, and the decay heat is removed by the PRHRS. To confirm design decisions and analyze the transient responses of an integral PWR such as REX-10, a thermal-hydraulic system code named TAPINS (Thermal-hydraulic Analysis Program for INtegral reactor System is developed in this study. Based on a one-dimensional four-equation drift-flux model, TAPINS incorporates mathematical models for the core, the helical-coil steam generator, and the steam-gas pressurizer. The system of difference equations derived from the semi-implicit finite-difference scheme is numerically solved by the Newton Block Gauss Seidel (NBGS method. TAPINS is characterized by applicability to transients with non-equilibrium effects, better prediction of the transient behavior of a pressurizer containing non-condensable gas, and code assessment by using the experimental data from the autonomous integral effect tests in the RTF (REX-10 Test Facility. Details on the hydrodynamic models as well as a part of validation results that reveal the features of TAPINS are presented in this paper.
Energy Technology Data Exchange (ETDEWEB)
Di Maio, P.A. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128, Palermo (Italy); Merola, M.; Mitteau, R.; Raffray, R. [ITER Organization, Route de Vinon sur Verdon, 13067 Saint Paul, Lez Durance (France); Vallone, E., E-mail: eug.vallone@gmail.com [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128, Palermo (Italy)
2016-11-01
Highlights: • A benchmarking activity has been carried out focusing the attention on the cooling circuits of ITER Shield Blocks #08 and #14. • A theoretical-computational fluid-dynamic approach based on the Finite Volume Method has been followed, adopting a commercial code. • Hydraulic characteristic functions and spatial distributions of coolant mass flow rate, velocity and pressure drop have been assessed. • Results obtained have allowed code benchmarking for Blanket modules and the numerical predictions have been found to be generally lower than but quite close to the experimental results (lower than 10%). - Abstract: As a consequence of its position and functions, the ITER blanket system will be subjected to significant heat loads under nominal reference conditions. Therefore, the design of its cooling system is particularly demanding. Coolant water is distributed individually to the 440 blanket modules (BMs) through manifold piping, which makes it a highly parallelized system. The mass flow rate distribution is finely tuned to meet all operation constraints: adequate margin to burn out in the plasma facing components, even distribution of water flow among the so-called plasma-facing “fingers” of the Blanket First Wall panels, high enough water flow rate to avoid excessive water temperature in the outlet pipes, maximum allowable water velocity lower than 7 m/s in manifold pipes. Furthermore the overall pressure drop and flow rate in each BM shall be within the fixed specified design limit to avoid an unduly unbalance of cooling among the 440 modules. Analyses have to be carried out following a computational fluid-dynamic (CFD) approach based on the finite volume method and adopting a CFD commercial code to assess the thermal-hydraulic behaviour of each single circuit of the ITER blanket cooling system. This paper describes the code benchmarking needed to determine the best method to get reliable and timely results. Since experimental tests are
Research on the improvement of nuclear safety -Improvement of level 1 PSA computer code package-
Energy Technology Data Exchange (ETDEWEB)
Park, Chang Kyoo; Kim, Tae Woon; Kim, Kil Yoo; Han, Sang Hoon; Jung, Won Dae; Jang, Seung Chul; Yang, Joon Un; Choi, Yung; Sung, Tae Yong; Son, Yung Suk; Park, Won Suk; Jung, Kwang Sub; Kang Dae Il; Park, Jin Heui; Hwang, Mi Jung; Hah, Jae Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1995-07-01
This year is the third year of the Government-sponsored mid- and long-term nuclear power technology development project. The scope of this sub project titled on `The improvement of level-1 PSA computer codes` is divided into three main activities : (1) Methodology development on the underdeveloped fields such as risk assessment technology for plant shutdown and low power situations, (2) Computer code package development for level-1 PSA, (3) Applications of new technologies to reactor safety assessment. At first, in this area of shutdown risk assessment technology development, plant outage experiences of domestic plants are reviewed and plant operating states (POS) are decided. A sample core damage frequency is estimated for over draining event in RCS low water inventory i.e. mid-loop operation. Human reliability analysis and thermal hydraulic support analysis are identified to be needed to reduce uncertainty. Two design improvement alternatives are evaluated using PSA technique for mid-loop operation situation: one is use of containment spray system as backup of shutdown cooling system and the other is installation of two independent level indication system. Procedure change is identified more preferable option to hardware modification in the core damage frequency point of view. Next, level-1 PSA code KIRAP is converted to PC-windows environment. For the improvement of efficiency in performing PSA, the fast cutest generation algorithm and an analytical technique for handling logical loop in fault tree modeling are developed. 48 figs, 15 tabs, 59 refs. (Author).
Application of the thermal-hydraulic codes in VVER-440 steam generators modelling
Energy Technology Data Exchange (ETDEWEB)
Matejovic, P.; Vranca, L.; Vaclav, E. [Nuclear Power Plant Research Inst. VUJE (Slovakia)
1995-12-31
Performances with the CATHARE2 V1.3U and RELAP5/MOD3.0 application to the VVER-440 SG modelling during normal conditions and during transient with secondary water lowering are described. Similar recirculation model was chosen for both codes. In the CATHARE calculation, no special measures were taken with the aim to optimize artificially flow rate distribution coefficients for the junction between SG riser and steam dome. Contrary to RELAP code, the CATHARE code is able to predict reasonable the secondary swell level in nominal conditions. Both codes are able to model properly natural phase separation on the SG water level. 6 refs.
Computed radiography simulation using the Monte Carlo code MCNPX
Energy Technology Data Exchange (ETDEWEB)
Correa, S.C.A. [Programa de Engenharia Nuclear/COPPE, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil); Centro Universitario Estadual da Zona Oeste (CCMAT)/UEZO, Av. Manuel Caldeira de Alvarenga, 1203, Campo Grande, 23070-200, Rio de Janeiro, RJ (Brazil); Souza, E.M. [Programa de Engenharia Nuclear/COPPE, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil); Silva, A.X., E-mail: ademir@con.ufrj.b [PEN/COPPE-DNC/Poli CT, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil); Cassiano, D.H. [Instituto de Radioprotecao e Dosimetria/CNEN Av. Salvador Allende, s/n, Recreio, 22780-160, Rio de Janeiro, RJ (Brazil); Lopes, R.T. [Programa de Engenharia Nuclear/COPPE, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil)
2010-09-15
Simulating X-ray images has been of great interest in recent years as it makes possible an analysis of how X-ray images are affected owing to relevant operating parameters. In this paper, a procedure for simulating computed radiographic images using the Monte Carlo code MCNPX is proposed. The sensitivity curve of the BaFBr image plate detector as well as the characteristic noise of a 16-bit computed radiography system were considered during the methodology's development. The results obtained confirm that the proposed procedure for simulating computed radiographic images is satisfactory, as it allows obtaining results comparable with experimental data.
Fault-tolerant quantum computing with color codes
Landahl, Andrew J; Rice, Patrick R
2011-01-01
We present and analyze protocols for fault-tolerant quantum computing using color codes. We present circuit-level schemes for extracting the error syndrome of these codes fault-tolerantly. We further present an integer-program-based decoding algorithm for identifying the most likely error given the syndrome. We simulated our syndrome extraction and decoding algorithms against three physically-motivated noise models using Monte Carlo methods, and used the simulations to estimate the corresponding accuracy thresholds for fault-tolerant quantum error correction. We also used a self-avoiding walk analysis to lower-bound the accuracy threshold for two of these noise models. We present and analyze two architectures for fault-tolerantly computing with these codes: one with 2D arrays of qubits are stacked atop each other and one in a single 2D substrate. Our analysis demonstrates that color codes perform slightly better than Kitaev's surface codes when circuit details are ignored. When these details are considered, w...
New Parallel computing framework for radiation transport codes
Energy Technology Data Exchange (ETDEWEB)
Kostin, M.A.; /Michigan State U., NSCL; Mokhov, N.V.; /Fermilab; Niita, K.; /JAERI, Tokai
2010-09-01
A new parallel computing framework has been developed to use with general-purpose radiation transport codes. The framework was implemented as a C++ module that uses MPI for message passing. The module is significantly independent of radiation transport codes it can be used with, and is connected to the codes by means of a number of interface functions. The framework was integrated with the MARS15 code, and an effort is under way to deploy it in PHITS. Besides the parallel computing functionality, the framework offers a checkpoint facility that allows restarting calculations with a saved checkpoint file. The checkpoint facility can be used in single process calculations as well as in the parallel regime. Several checkpoint files can be merged into one thus combining results of several calculations. The framework also corrects some of the known problems with the scheduling and load balancing found in the original implementations of the parallel computing functionality in MARS15 and PHITS. The framework can be used efficiently on homogeneous systems and networks of workstations, where the interference from the other users is possible.
New Parallel computing framework for radiation transport codes
Kostin, M A; Niita, K
2012-01-01
A new parallel computing framework has been developed to use with general-purpose radiation transport codes. The framework was implemented as a C++ module that uses MPI for message passing. The module is significantly independent of radiation transport codes it can be used with, and is connected to the codes by means of a number of interface functions. The frame work was integrated with the MARS15 code, and an effort is under way to deploy it in PHITS. Besides the parallel computing functionality, the framework offers a checkpoint facility that allows restarting calculations with a saved checkpoint file. The checkpoint facility can be used in single process calculations as well as in the parallel regime. Several checkpoint files can be merged into one thus combining results of several calculations. The framework also corrects some of the known problems with the sch eduling and load balancing found in the original implementations of the parallel computing functionality in MARS15 and PHITS. The framework can be...
Development of a computer code for dynamic analysis of the primary circuit of advanced reactors
Energy Technology Data Exchange (ETDEWEB)
Rocha, Jussie Soares da; Lira, Carlos A.B.O.; Magalhaes, Mardson A. de Sa, E-mail: cabol@ufpe.b [Universidade Federal de Pernambuco (DEN/UFPE), Recife, PE (Brazil). Dept. de Energia Nuclear
2011-07-01
Currently, advanced reactors are being developed, seeking for enhanced safety, better performance and low environmental impacts. Reactor designs must follow several steps and numerous tests before a conceptual project could be certified. In this sense, computational tools become indispensable in the preparation of such projects. Thus, this study aimed at the development of a computational tool for thermal-hydraulic analysis by coupling two computer codes to evaluate the influence of transients caused by pressure variations and flow surges in the region of the primary circuit of IRIS reactor between the core and the pressurizer. For the simulation, it was used a situation of 'insurge', characterized by the entry of water in the pressurizer, due to the expansion of the refrigerant in the primary circuit. This expansion was represented by a pressure disturbance in step form, through the block 'step' of SIMULINK, thus enabling the transient startup. The results showed that the dynamic tool, obtained through the coupling of the codes, generated very satisfactory responses within model limitations, preserving the most important phenomena in the process. (author)
LMFBR models for the ORIGEN2 computer code
Energy Technology Data Exchange (ETDEWEB)
Croff, A.G.; McAdoo, J.W.; Bjerke, M.A.
1983-06-01
Reactor physics calculations have led to the development of nine liquid-metal fast breeder reactor (LMFBR) models for the ORIGEN2 computer code. Four of the models are based on the U-Pu fuel cycle, two are based on the Th-U-Pu fuel cycle, and three are based on the Th-/sup 233/U fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST are given.
LMFBR models for the ORIGEN2 computer code
Energy Technology Data Exchange (ETDEWEB)
Croff, A.G.; McAdoo, J.W.; Bjerke, M.A.
1981-10-01
Reactor physics calculations have led to the development of nine liquid-metal fast breeder reactor (LMFBR) models for the ORIGEN2 computer code. Four of the models are based on the U-Pu fuel cycle, two are based on the Th-U-Pu fuel cycle, and three are based on the Th-/sup 238/U fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST are given.
User's manual for HDR3 computer code
Energy Technology Data Exchange (ETDEWEB)
Arundale, C.J.
1982-10-01
A description of the HDR3 computer code and instructions for its use are provided. HDR3 calculates space heating costs for a hot dry rock (HDR) geothermal space heating system. The code also compares these costs to those of a specific oil heating system in use at the National Aeronautics and Space Administration Flight Center at Wallops Island, Virginia. HDR3 allows many HDR system parameters to be varied so that the user may examine various reservoir management schemes and may optimize reservoir design to suit a particular set of geophysical and economic parameters.
Energy Technology Data Exchange (ETDEWEB)
Petruzzi, Alessandro; D' Auria, Francesco [University of Pisa, San Piero a Grado (Italy). Nuclear Research Group San Piero a Grado (GRNSPG); Galetti, Regina, E-mail: regina@cnen.gov.b [National Commission for Nuclear Energy (CNEN), Rio de Janeiro, RJ (Brazil); Bajs, Tomislav [University of Zagreb (Croatia). Fac. of Electrical Engineering and Computing. Dept. of Power Systems; Reventos, Francesc [Technical University of Catalonia, Barcelona (Spain). Dept. of Physics and Nuclear Engineering
2011-07-01
Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers, vendors, and research organizations. Computer code user represents a source of uncertainty that may significantly affect the results of system code calculations. Code user training and qualification represent an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes the experience in applying a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In addition, this paper presents the organization and the main features of the 3D S.UN.COP (scaling, uncertainty, and 3D coupled code calculations) seminars during which particular emphasis is given to practical applications in connection with the licensing process of best estimate plus uncertainty methodologies, showing the designer, utility and regulatory approaches. (author)
Computer codes for evaluation of control room habitability (HABIT)
Energy Technology Data Exchange (ETDEWEB)
Stage, S.A. [Pacific Northwest Lab., Richland, WA (United States)
1996-06-01
This report describes the Computer Codes for Evaluation of Control Room Habitability (HABIT). HABIT is a package of computer codes designed to be used for the evaluation of control room habitability in the event of an accidental release of toxic chemicals or radioactive materials. Given information about the design of a nuclear power plant, a scenario for the release of toxic chemicals or radionuclides, and information about the air flows and protection systems of the control room, HABIT can be used to estimate the chemical exposure or radiological dose to control room personnel. HABIT is an integrated package of several programs that previously needed to be run separately and required considerable user intervention. This report discusses the theoretical basis and physical assumptions made by each of the modules in HABIT and gives detailed information about the data entry windows. Sample runs are given for each of the modules. A brief section of programming notes is included. A set of computer disks will accompany this report if the report is ordered from the Energy Science and Technology Software Center. The disks contain the files needed to run HABIT on a personal computer running DOS. Source codes for the various HABIT routines are on the disks. Also included are input and output files for three demonstration runs.
War of ontology worlds: mathematics, computer code, or Esperanto?
Directory of Open Access Journals (Sweden)
Andrey Rzhetsky
2011-09-01
Full Text Available The use of structured knowledge representations-ontologies and terminologies-has become standard in biomedicine. Definitions of ontologies vary widely, as do the values and philosophies that underlie them. In seeking to make these views explicit, we conducted and summarized interviews with a dozen leading ontologists. Their views clustered into three broad perspectives that we summarize as mathematics, computer code, and Esperanto. Ontology as mathematics puts the ultimate premium on rigor and logic, symmetry and consistency of representation across scientific subfields, and the inclusion of only established, non-contradictory knowledge. Ontology as computer code focuses on utility and cultivates diversity, fitting ontologies to their purpose. Like computer languages C++, Prolog, and HTML, the code perspective holds that diverse applications warrant custom designed ontologies. Ontology as Esperanto focuses on facilitating cross-disciplinary communication, knowledge cross-referencing, and computation across datasets from diverse communities. We show how these views align with classical divides in science and suggest how a synthesis of their concerns could strengthen the next generation of biomedical ontologies.
War of Ontology Worlds: Mathematics, Computer Code, or Esperanto?
Rzhetsky, Andrey; Evans, James A.
2011-01-01
The use of structured knowledge representations—ontologies and terminologies—has become standard in biomedicine. Definitions of ontologies vary widely, as do the values and philosophies that underlie them. In seeking to make these views explicit, we conducted and summarized interviews with a dozen leading ontologists. Their views clustered into three broad perspectives that we summarize as mathematics, computer code, and Esperanto. Ontology as mathematics puts the ultimate premium on rigor and logic, symmetry and consistency of representation across scientific subfields, and the inclusion of only established, non-contradictory knowledge. Ontology as computer code focuses on utility and cultivates diversity, fitting ontologies to their purpose. Like computer languages C++, Prolog, and HTML, the code perspective holds that diverse applications warrant custom designed ontologies. Ontology as Esperanto focuses on facilitating cross-disciplinary communication, knowledge cross-referencing, and computation across datasets from diverse communities. We show how these views align with classical divides in science and suggest how a synthesis of their concerns could strengthen the next generation of biomedical ontologies. PMID:21980276
Benchmarking of computer codes and approaches for modeling exposure scenarios
Energy Technology Data Exchange (ETDEWEB)
Seitz, R.R. [EG and G Idaho, Inc., Idaho Falls, ID (United States); Rittmann, P.D.; Wood, M.I. [Westinghouse Hanford Co., Richland, WA (United States); Cook, J.R. [Westinghouse Savannah River Co., Aiken, SC (United States)
1994-08-01
The US Department of Energy Headquarters established a performance assessment task team (PATT) to integrate the activities of DOE sites that are preparing performance assessments for the disposal of newly generated low-level waste. The PATT chartered a subteam with the task of comparing computer codes and exposure scenarios used for dose calculations in performance assessments. This report documents the efforts of the subteam. Computer codes considered in the comparison include GENII, PATHRAE-EPA, MICROSHIELD, and ISOSHLD. Calculations were also conducted using spreadsheets to provide a comparison at the most fundamental level. Calculations and modeling approaches are compared for unit radionuclide concentrations in water and soil for the ingestion, inhalation, and external dose pathways. Over 30 tables comparing inputs and results are provided.
Computational Complexity of Decoding Orthogonal Space-Time Block Codes
Ayanoglu, Ender; Karipidis, Eleftherios
2009-01-01
The computational complexity of optimum decoding for an orthogonal space-time block code G satisfying the orthogonality property that the Hermitian transpose of G multiplied by G is equal to a constant c times the sum of the squared symbols of the code times an identity matrix, where c is a positive integer is quantified. Four equivalent techniques of optimum decoding which have the same computational complexity are specified. Modifications to the basic formulation in special cases are calculated and illustrated by means of examples. This paper corrects and extends [1],[2], and unifies them with the results from the literature. In addition, a number of results from the literature are extended to the case c > 1.
Bragg optics computer codes for neutron scattering instrument design
Energy Technology Data Exchange (ETDEWEB)
Popovici, M.; Yelon, W.B.; Berliner, R.R. [Missouri Univ. Research Reactor, Columbia, MO (United States); Stoica, A.D. [Institute of Physics and Technology of Materials, Bucharest (Romania)
1997-09-01
Computer codes for neutron crystal spectrometer design, optimization and experiment planning are described. Phase space distributions, linewidths and absolute intensities are calculated by matrix methods in an extension of the Cooper-Nathans resolution function formalism. For modeling the Bragg reflection on bent crystals the lamellar approximation is used. Optimization is done by satisfying conditions of focusing in scattering and in real space, and by numerically maximizing figures of merit. Examples for three-axis and two-axis spectrometers are given.
General review of the MOSTAS computer code for wind turbines
Energy Technology Data Exchange (ETDEWEB)
Dugundji, J.; Wendell, J.H.
1981-06-01
The MOSTAS computer code for wind turbine analysis is reviewed, and the techniques and methods used in its analyses are described in some detail. Some impressions of its strengths and weaknesses, and some recommendations for its application, modification, and further development are made. Additionally, some basic techniques used in wind turbine stability and response analyses for systems with constant and periodic coefficients are reviewed in the Appendices.
Refactoring Android Java Code for On-Demand Computation Offloading
Zhang, Ying; Huang, Gang; Liu, Xuanzhe; Zhang, Wei; Zhang, Wei; Mei, Hong; Yang, Shunxiang
2012-01-01
International audience; Computation offloading is a promising way to improve the performance as well as reduce the battery energy consumption of a smartphone application by executing some part of the application on a remote server. Supporting such capability is not easy to smartphone app developers for 1) correctness: some codes, e.g. those for GPS, gravity and other sensors, can only run on the smartphone so that the developers have to identify which part of the application cannot be offload...
Energy Technology Data Exchange (ETDEWEB)
Kadi, Rabah, E-mail: kadi.rkhaled@hotmail.com [Laboratory for Thermal-Hydraulics, Nuclear Research Center of Birine (Algeria); Aissani, Slimane [Hydrocarbons and Chemistry Faculty, University of Boumerdes (Algeria); Bouam, Abdellah [Laboratory for Thermal-Hydraulics, Nuclear Research Center of Birine (Algeria)
2015-11-15
Highlights: • TransAT CMFD code application to DCC phenomenon. • LEIS methodology to predict the condensing steam flow rate. • Validation of interfacial phase-change heat transfer and turbulence models. • Correction of damping function at the free surface region. • Numerical validation of previous models using LIM and KAERI & KAIST test facilities. - Abstract: The use of CFD for the industrial studies related to PTS, including DCC is already possible; improvements of the two-phase modeling capabilities have to be undertaken to qualify the codes for the simulation of such flows. The DCC in horizontally stratified flow regime constitutes very considerable challenge exercises for a computational fluid dynamics (CFD) simulation of the thermal hydraulics PTS phenomenon because the interplay between turbulence and interfacial heat and mass transfer problem. The main purpose of our study is to investigate numerically the DCC in horizontally stratified steam water flow in a 2D and 3D channel using TransAT CMFD code. The new methodology known as Large-Eddy & Interface (LEIS) have been implemented for treatment of turbulence combined with interface tracking ITM (level set approach). Among of the so-called ‘coarse-grained’ ITM's models, the modified original surface divergence has been chosen as well as the treatment of the turbulence by URANS and VLES. This contribution addressed on the validation of interfacial phase-change heat transfer and turbulence models with special correction of the damping function at the free surface for single and combined-effect thermal hydraulic studies for LIM and KAERI & KAIST test facilities. The LIES methodology was found to apply successfully to predict the condensing steam flow rate in the all cases of the LIM test case involving a Smooth to Wavy turbulent, concurrent stratified steam-water flow in a 2D channel. The CMFD TransAT code predicting capability is analyzed, comparing the liquid temperature and to much the
Methodology for computational fluid dynamics code verification/validation
Energy Technology Data Exchange (ETDEWEB)
Oberkampf, W.L.; Blottner, F.G.; Aeschliman, D.P.
1995-07-01
The issues of verification, calibration, and validation of computational fluid dynamics (CFD) codes has been receiving increasing levels of attention in the research literature and in engineering technology. Both CFD researchers and users of CFD codes are asking more critical and detailed questions concerning the accuracy, range of applicability, reliability and robustness of CFD codes and their predictions. This is a welcomed trend because it demonstrates that CFD is maturing from a research tool to the world of impacting engineering hardware and system design. In this environment, the broad issue of code quality assurance becomes paramount. However, the philosophy and methodology of building confidence in CFD code predictions has proven to be more difficult than many expected. A wide variety of physical modeling errors and discretization errors are discussed. Here, discretization errors refer to all errors caused by conversion of the original partial differential equations to algebraic equations, and their solution. Boundary conditions for both the partial differential equations and the discretized equations will be discussed. Contrasts are drawn between the assumptions and actual use of numerical method consistency and stability. Comments are also made concerning the existence and uniqueness of solutions for both the partial differential equations and the discrete equations. Various techniques are suggested for the detection and estimation of errors caused by physical modeling and discretization of the partial differential equations.
Validation and Application of the Thermal Hydraulic System Code TRACE for Analysis of BWR Transients
Directory of Open Access Journals (Sweden)
V. H. Sánchez
2012-01-01
Full Text Available The Karlsruhe Institute of Technology (KIT is participating on (Code Applications and Maintenance Program CAMP of the US Nuclear Regulatory Commission (NRC to validate TRACE code for LWR transient analysis. The application of TRACE for the safety assessment of BWR requires a throughout verification and validation using experimental data from separate effect and integral tests but also using plant data. The validation process is normally focused on safety-relevant phenomena for example, pressure drop, void fraction, heat transfer, and critical power models. The purpose of this paper is to validate selected BWR-relevant TRACE-models using both data of bundle tests such as the (Boiling Water Reactor Full-Size Fine-Mesh Bundle Test BFBT and plant data recorded during a turbine trip event (TUSA occurred in a Type-72 German BWR plant. For the validation, TRACE models of the BFBT bundle and of the BWR plant were developed. The performed investigations have shown that the TRACE code is appropriate to describe main BWR-safety-relevant phenomena (pressure drop, void fraction, and critical power with acceptable accuracy. The comparison of the predicted global BWR plant parameters for the TUSA event with the measured plant data indicates that the code predictions are following the main trends of the measured parameters such as dome pressure and reactor power.
Energy Technology Data Exchange (ETDEWEB)
Peterson, Per
2012-10-30
The objective of the 3-year project was to collect integral effects test (IET) data to validate the RELAP5-3D code and other thermal hydraulics codes for use in predicting the transient thermal hydraulics response of liquid salt cooled reactor systems, including integral transient response for forced and natural circulation operation. The reference system for the project is a modular, 900-MWth Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a specific type of Fluoride salt-cooled High temperature Reactor (FHR). Two experimental facilities were developed for thermal-hydraulic integral effects tests (IETs) and separate effects tests (SETs). The facilities use simulant fluids for the liquid fluoride salts, with very little distortion to the heat transfer and fluid dynamics behavior. The CIET Test Bay facility was designed, built, and operated. IET data for steady state and transient natural circulation was collected. SET data for convective heat transfer in pebble beds and straight channel geometries was collected. The facility continues to be operational and will be used for future experiments, and for component development. The CIET 2 facility is larger in scope, and its construction and operation has a longer timeline than the duration of this grant. The design for the CIET 2 facility has drawn heavily on the experience and data collected on the CIET Test Bay, and it was completed in parallel with operation of the CIET Test Bay. CIET 2 will demonstrate start-up and shut-down transients and control logic, in addition to LOFC and LOHS transients, and buoyant shut down rod operation during transients. Design of the CIET 2 Facility is complete, and engineering drawings have been submitted to an external vendor for outsourced quality controlled construction. CIET 2 construction and operation continue under another NEUP grant. IET data from both CIET facilities is to be used for validation of system codes used for FHR modeling, such as RELAP5-3D. A set of
MELCOR computer code manuals: Primer and user`s guides, Version 1.8.3 September 1994. Volume 1
Energy Technology Data Exchange (ETDEWEB)
Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.; Stuart, D.S.; Thompson, S.L. [Sandia National Labs., Albuquerque, NM (United States); Hodge, S.A.; Hyman, C.R.; Sanders, R.L. [Oak Ridge National Lab., TN (United States)
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, and 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.
Energy Technology Data Exchange (ETDEWEB)
Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Scherer, W.
1998-04-01
V.S.O.P. (97) is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies processing of cross sections, the setup of the reactor and of the fuel element, repeated neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to HTRs and to two spatial dimensions. V.S.O.P. (97) can simulate the reactor operation from the initial core towards the equilibrium core. V.S.O.P. (97) - on the basis of V.S.O.P. (94) - has been improved with regard to a more detailed treatment of the build-up and the depletion of the heavy metal isotopes. Their chains now include the minor actinides. Resonance cross sections of the lumped resonance absorbers are evaluated burnup-dependent. Beyond this, the code has been reviewed in many details, aiming at an improved precision in the computer simulation of the features of the reactors and of their fuel cycle. The code consists of about 65000 FORTRAN statements. A memory of 32 MB should be available for its use. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Holt, L., E-mail: lars.holt@tuev-sued.de [TÜV SÜD Energietechnik GmbH Baden-Württemberg, Gottlieb-Daimler-Str. 7, 70794 Filderstadt (Germany); Technical University München, Department of Nuclear Engineering, Boltzmannstr. 15, D-85748 Garching bei München (Germany); Rohde, U.; Kliem, S.; Baier, S. [Helmholtz-Zentrum Dresden—Rossendorf, Reactor Safety Division, PO Box 510119, D-01314 Dresden (Germany); Seidl, M. [E.ON Kernkraft GmbH, Tresckowstr. 5, D-30457 Hannover (Germany); Van Uffelen, P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Macián-Juan, R. [Technical University München, Department of Nuclear Engineering, Boltzmannstr. 15, D-85748 Garching bei München (Germany)
2016-02-15
Highlights: • General coupling interface was developed for the fuel performance code TRANSURANUS. • With this new tool simplified fuel behavior models in codes can be replaced. • The reactor dynamics code DYN3D was coupled to TRANSURANUS at assembly level. • The feedback from detailed online fuel behavior modeling is analyzed for reactivity initiated accident (RIA). • The thermal hydraulics can be affected strongly even in fresh fuel assemblies. - Abstract: Recently the reactor dynamics code DYN3D (including an internal fuel behavior model) was coupled to the fuel performance code TRANSURANUS at assembly level. The coupled code system applies the new general TRANSURANUS coupling interface, hence it can be used for one-way or two-way coupling. In the coupling, DYN3D provides process time, time-dependent rod power and thermal hydraulics conditions to TRANSURANUS, which in case of the two-way coupling approach replaces completely the internal DYN3D fuel behavior model and transfers parameters like radial fuel temperature distribution and cladding temperature back to DYN3D. For the first time results of the coupled code system are presented for a post-critical-heat-flux heat transfer. The corresponding heat transfer regime is mostly film boiling, where the cladding temperature can rise several hundreds of degrees. The simulated boron dilution transient assumed an injection of a 36 m{sup 3} slug of under-borated coolant into a German pressurized water reactor (PWR) core initiated from a sub-critical reactor state (extreme reactivity initiated accident (RIA) conditions). The feedback from detailed fuel behavior modeling was found negligible on the neutron kinetics and thermal hydraulics during the first power rise. In a later phase of the transient, the node injected energy can differ 25 J/g, even still around 20 J/g for nodes without film boiling. Furthermore, the thermal hydraulics can be affected strongly even in fresh fuel assemblies, where film boiling
Improvement of level-1 PSA computer code package
Energy Technology Data Exchange (ETDEWEB)
Kim, Tae Woon; Park, C. K.; Kim, K. Y.; Han, S. H.; Jung, W. D.; Chang, S. C.; Yang, J. E.; Sung, T. Y.; Kang, D. I.; Park, J. H.; Lee, Y. H.; Kim, S. H.; Hwang, M. J.; Choi, S. Y.
1997-07-01
This year the fifth (final) year of the phase-I of the Government-sponsored Mid- and Long-term Nuclear Power Technology Development Project. The scope of this subproject titled on `The improvement of level-1 PSA Computer Codes` is divided into two main activities : (1) improvement of level-1 PSA methodology, (2) development of applications methodology of PSA techniques to operations and maintenance of nuclear power plant. Level-1 PSA code KIRAP is converted to PC-Windows environment. For the improvement of efficiency in performing PSA, the fast cutset generation algorithm and an analytical technique for handling logical loop in fault tree modeling are developed. Using about 30 foreign generic data sources, generic component reliability database (GDB) are developed considering dependency among source data. A computer program which handles dependency among data sources are also developed based on three stage bayesian updating technique. Common cause failure (CCF) analysis methods are reviewed and CCF database are established. Impact vectors can be estimated from this CCF database. A computer code, called MPRIDP, which handles CCF database are also developed. A CCF analysis reflecting plant-specific defensive strategy against CCF event is also performed. A risk monitor computer program, called Risk Monster, are being developed for the application to the operation and maintenance of nuclear power plant. The PSA application technique is applied to review the feasibility study of on-line maintenance and to the prioritization of in-service test (IST) of motor-operated valves (MOV). Finally, the root cause analysis (RCA) and reliability-centered maintenance (RCM) technologies are adopted and applied to the improvement of reliability of emergency diesel generators (EDG) of nuclear power plant. To help RCA and RCM analyses, two software programs are developed, which are EPIS and RAM Pro. (author). 129 refs., 20 tabs., 60 figs.
Computationally efficient sub-band coding of ECG signals.
Husøy, J H; Gjerde, T
1996-03-01
A data compression technique is presented for the compression of discrete time electrocardiogram (ECG) signals. The compression system is based on sub-band coding, a technique traditionally used for compressing speech and images. The sub-band coder employs quadrature mirror filter banks (QMF) with up to 32 critically sampled sub-bands. Both finite impulse response (FIR) and the more computationally efficient infinite impulse response (IIR) filter banks are considered as candidates in a complete ECG coding system. The sub-bands are threshold, quantized using uniform quantizers and run-length coded. The output of the run-length coder is further compressed by a Huffman coder. Extensive simulations indicate that 16 sub-bands are a suitable choice for this application. Furthermore, IIR filter banks are preferable due to their superiority in terms of computational efficiency. We conclude that the present scheme, which is suitable for real time implementation on a PC, can provide compression ratios between 5 and 15 without loss of clinical information.
Codes for Computationally Simple Channels: Explicit Constructions with Optimal Rate
Guruswami, Venkatesan
2010-01-01
In this paper, we consider coding schemes for computationally bounded channels, which can introduce an arbitrary set of errors as long as (a) the fraction of errors is bounded with high probability by a parameter p and (b) the process which adds the errors can be described by a sufficiently "simple" circuit. For three classes of channels, we provide explicit, efficiently encodable/decodable codes of optimal rate where only inefficiently decodable codes were previously known. In each case, we provide one encoder/decoder that works for every channel in the class. (1) Unique decoding for additive errors: We give the first construction of poly-time encodable/decodable codes for additive (a.k.a. oblivious) channels that achieve the Shannon capacity 1-H(p). Such channels capture binary symmetric errors and burst errors as special cases. (2) List-decoding for log-space channels: A space-S(n) channel reads and modifies the transmitted codeword as a stream, using at most S(n) bits of workspace on transmissions of n bi...
Compressing industrial computed tomography images by means of contour coding
Jiang, Haina; Zeng, Li
2013-10-01
An improved method for compressing industrial computed tomography (CT) images is presented. To have higher resolution and precision, the amount of industrial CT data has become larger and larger. Considering that industrial CT images are approximately piece-wise constant, we develop a compression method based on contour coding. The traditional contour-based method for compressing gray images usually needs two steps. The first is contour extraction and then compression, which is negative for compression efficiency. So we merge the Freeman encoding idea into an improved method for two-dimensional contours extraction (2-D-IMCE) to improve the compression efficiency. By exploiting the continuity and logical linking, preliminary contour codes are directly obtained simultaneously with the contour extraction. By that, the two steps of the traditional contour-based compression method are simplified into only one. Finally, Huffman coding is employed to further losslessly compress preliminary contour codes. Experimental results show that this method can obtain a good compression ratio as well as keeping satisfactory quality of compressed images.
Energy Technology Data Exchange (ETDEWEB)
Queral, C.; Coriso, M.; Garcia Sedano, P. J.; Ruiz, J. A.; Posada, J. M.; Jimenez Varas, G.; Sol, I.; Herranz, L. E.
2011-07-01
Use of thermal-hydraulic codes is extended all over many different aspects of nuclear engineering. This article groups and briefly describes the main features of some of the well known codes as an introduction to their recent applications in the Spain nuclear sector. the broad range and quality of applications highlight the maturity achieved both in industry and research organizations and universities within the Spanish nuclear sector. (Author)
Energy Technology Data Exchange (ETDEWEB)
Lizorkin, M.; Nikonov, S. [Kurchatov Institute for Atomic Energy, Moscow (Russian Federation); Langenbuch, S.; Velkov, K. [Gesellschaft fur Anlagen- und Reaktorsicherheit (GRS) mbH, Garching (Germany)
2006-07-01
The coupled thermal-hydraulics and neutron-kinetics code ATHLET/BIPR-VVER was developed within a co-operation between the RRC Kurchatov Institute (KI) and GRS. The modeling capability of this coupled code as well as the status of validation by benchmark activities and comparison with plant measurements are described. The paper is focused on the modeling of flow mixing in the reactor pressure vessel including its validation and the application for the safety justification of VVER plants. (authors)
PLUTO code for computational Astrophysics: News and Developments
Tzeferacos, P.; Mignone, A.
2012-01-01
We present an overview on recent developments and functionalities available with the PLUTO code for astrophysical fluid dynamics. The recent extension of the code to a conservative finite difference formulation and high order spatial discretization of the compressible equations of magneto-hydrodynamics (MHD), complementary to its finite volume approach, allows for a highly accurate treatment of smooth flows, while avoiding loss of accuracy near smooth extrema and providing sharp non-oscillatory transitions at discontinuities. Among the novel features, we present alternative, fully explicit treatments to include non-ideal dissipative processes (namely viscosity, resistivity and anisotropic thermal conduction), that do not suffer from the usual timestep limitation of explicit time stepping. These methods, offsprings of the multistep Runge-Kutta family that use a Chebyshev polynomial recursion, are competitive substitutes of computationally expensive implicit schemes that involve sparse matrix inversion. Several multi-dimensional benchmarks and appli-cations assess the potential of PLUTO to efficiently handle many astrophysical problems.
Atucha II NPP full scope simulator modelling with the thermal hydraulic code TRAC{sub R}T
Energy Technology Data Exchange (ETDEWEB)
Alonso, Pablo Rey; Ruiz, Jose Antonio; Rivero, Norberto, E-mail: prey@tecnatom.e, E-mail: jaruiz@tecnatom.e, E-mail: nrivero@tecnatom.e [Tecnatom S.A., Madrid (Spain)
2011-07-01
In February 2010 NA-SA (Nucleoelectrica Argentina S.A.) awarded Tecnatom the Atucha II full scope simulator project. NA-SA is a public company owner of the Argentinean nuclear power plants. Atucha II is due to enter in operation shortly. Atucha II NPP is a PHWR type plant cooled by the water of the Parana River and has the same design as the Atucha I unit, doubling its power capacity. Atucha II will produce 745 MWe utilizing heavy water as coolant and moderator, and natural uranium as fuel. A plant singular feature is the permanent core refueling. TRAC{sub R}T is the first real time thermal hydraulic six-equations code used in the training simulation industry for NSSS modeling. It is the result from adapting to real time the best estimate code TRACG. TRAC{sub R}T is based on first principle conservation equations for mass, energy and momentum for liquid and steam phases, with two phase flows under non homogeneous and non equilibrium conditions. At present, it has been successfully implemented in twelve full scope replica simulators in different training centers throughout the world. To ease the modeling task, TRAC{sub R}T includes a graphical pre-processing tool designed to optimize this process and alleviate the burden of entering alpha numerical data in an input file. (author)
Directory of Open Access Journals (Sweden)
Viet-Anh Phung
2015-01-01
Full Text Available In earlier study we have demonstrated that RELAP5 can predict flow instability parameters (flow rate, oscillation period, temperature, and pressure in single channel tests in CIRCUS-IV facility. The main goals of this work are to (i validate RELAP5 and TRACE capabilities in prediction of two-phase flow instability and flow regimes and (ii assess the effect of improvement in flow regime identification on code predictions. Most of the results of RELAP5 and TRACE calculation are in reasonable agreement with experimental data from CIRCUS-IV. However, both codes misidentified instantaneous flow regimes which were observed in the test with high speed camera. One of the reasons for the incorrect identification of the flow regimes is the small tube flow regime transition model in RELAP5 and the combined bubbly-slug flow regime in TRACE. We found that calculation results are sensitive to flow regime boundaries of RELAP5 which were modified in order to match the experimental data on flow regimes. Although the flow regime became closer to the experimental one, other predicted thermal hydraulic parameters showed larger discrepancy with the experimental data than with the base case calculations where flow regimes were misidentified.
Parallel Computing Characteristics of CUPID code under MPI and Hybrid environment
Energy Technology Data Exchange (ETDEWEB)
Lee, Jae Ryong; Yoon, Han Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Jeon, Byoung Jin; Choi, Hyoung Gwon [Seoul National Univ. of Science and Technology, Seoul (Korea, Republic of)
2014-05-15
In this paper, a characteristic of parallel algorithm is presented for solving an elliptic type equation of CUPID via domain decomposition method using the MPI and the parallel performance is estimated in terms of a scalability which shows the speedup ratio. In addition, the time-consuming pattern of major subroutines is studied. Two different grid systems are taken into account: 40,000 meshes for coarse system and 320,000 meshes for fine system. Since the matrix of the CUPID code differs according to whether the flow is single-phase or two-phase, the effect of matrix shape is evaluated. Finally, the effect of the preconditioner for matrix solver is also investigated. Finally, the hybrid (OpenMP+MPI) parallel algorithm is introduced and discussed in detail for solving pressure solver. Component-scale thermal-hydraulics code, CUPID has been developed for two-phase flow analysis, which adopts a three-dimensional, transient, three-field model, and parallelized to fulfill a recent demand for long-transient and highly resolved multi-phase flow behavior. In this study, the parallel performance of the CUPID code was investigated in terms of scalability. The CUPID code was parallelized with domain decomposition method. The MPI library was adopted to communicate the information at the neighboring domain. For managing the sparse matrix effectively, the CSR storage format is used. To take into account the characteristics of the pressure matrix which turns to be asymmetric for two-phase flow, both single-phase and two-phase calculations were run. In addition, the effect of the matrix size and preconditioning was also investigated. The fine mesh calculation shows better scalability than the coarse mesh because the number of coarse mesh does not need to decompose the computational domain excessively. The fine mesh can be present good scalability when dividing geometry with considering the ratio between computation and communication time. For a given mesh, single-phase flow
Multicode comparison of selected source-term computer codes
Energy Technology Data Exchange (ETDEWEB)
Hermann, O.W.; Parks, C.V.; Renier, J.P.; Roddy, J.W.; Ashline, R.C.; Wilson, W.B.; LaBauve, R.J.
1989-04-01
This report summarizes the results of a study to assess the predictive capabilities of three radionuclide inventory/depletion computer codes, ORIGEN2, ORIGEN-S, and CINDER-2. The task was accomplished through a series of comparisons of their output for several light-water reactor (LWR) models (i.e., verification). Of the five cases chosen, two modeled typical boiling-water reactors (BWR) at burnups of 27.5 and 40 GWd/MTU and two represented typical pressurized-water reactors (PWR) at burnups of 33 and 50 GWd/MTU. In the fifth case, identical input data were used for each of the codes to examine the results of decay only and to show differences in nuclear decay constants and decay heat rates. Comparisons were made for several different characteristics (mass, radioactivity, and decay heat rate) for 52 radionuclides and for nine decay periods ranging from 30 d to 10,000 years. Only fission products and actinides were considered. The results are presented in comparative-ratio tables for each of the characteristics, decay periods, and cases. A brief summary description of each of the codes has been included. Of the more than 21,000 individual comparisons made for the three codes (taken two at a time), nearly half (45%) agreed to within 1%, and an additional 17% fell within the range of 1 to 5%. Approximately 8% of the comparison results disagreed by more than 30%. However, relatively good agreement was obtained for most of the radionuclides that are expected to contribute the greatest impact to waste disposal. Even though some defects have been noted, each of the codes in the comparison appears to produce respectable results. 12 figs., 12 tabs.
Code Verification of the HIGRAD Computational Fluid Dynamics Solver
Energy Technology Data Exchange (ETDEWEB)
Van Buren, Kendra L. [Los Alamos National Laboratory; Canfield, Jesse M. [Los Alamos National Laboratory; Hemez, Francois M. [Los Alamos National Laboratory; Sauer, Jeremy A. [Los Alamos National Laboratory
2012-05-04
The purpose of this report is to outline code and solution verification activities applied to HIGRAD, a Computational Fluid Dynamics (CFD) solver of the compressible Navier-Stokes equations developed at the Los Alamos National Laboratory, and used to simulate various phenomena such as the propagation of wildfires and atmospheric hydrodynamics. Code verification efforts, as described in this report, are an important first step to establish the credibility of numerical simulations. They provide evidence that the mathematical formulation is properly implemented without significant mistakes that would adversely impact the application of interest. Highly accurate analytical solutions are derived for four code verification test problems that exercise different aspects of the code. These test problems are referred to as: (i) the quiet start, (ii) the passive advection, (iii) the passive diffusion, and (iv) the piston-like problem. These problems are simulated using HIGRAD with different levels of mesh discretization and the numerical solutions are compared to their analytical counterparts. In addition, the rates of convergence are estimated to verify the numerical performance of the solver. The first three test problems produce numerical approximations as expected. The fourth test problem (piston-like) indicates the extent to which the code is able to simulate a 'mild' discontinuity, which is a condition that would typically be better handled by a Lagrangian formulation. The current investigation concludes that the numerical implementation of the solver performs as expected. The quality of solutions is sufficient to provide credible simulations of fluid flows around wind turbines. The main caveat associated to these findings is the low coverage provided by these four problems, and somewhat limited verification activities. A more comprehensive evaluation of HIGRAD may be beneficial for future studies.
Knowlton, Marie; Wetzel, Robin
2006-01-01
This study compared the length of text in English Braille American Edition, the Nemeth code, and the computer braille code with the Unified English Braille Code (UEBC)--also known as Unified English Braille (UEB). The findings indicate that differences in the length of text are dependent on the type of material that is transcribed and the grade…
pyro: A teaching code for computational astrophysical hydrodynamics
Zingale, Michael
2013-01-01
We describe pyro: a simple, freely-available code to aid students in learning the computational hydrodynamics methods widely used in astrophysics. pyro is written with simplicity and learning in mind and intended to allow students to experiment with various methods popular in the field, including those for advection, compressible and incompressible hydrodynamics, multigrid, and diffusion in a finite-volume framework. We show some of the test problems from pyro, describe its design philosophy, and suggest extensions for students to build their understanding of these methods.
pyro: A teaching code for computational astrophysical hydrodynamics
Zingale, M.
2014-10-01
We describe pyro: a simple, freely-available code to aid students in learning the computational hydrodynamics methods widely used in astrophysics. pyro is written with simplicity and learning in mind and intended to allow students to experiment with various methods popular in the field, including those for advection, compressible and incompressible hydrodynamics, multigrid, and diffusion in a finite-volume framework. We show some of the test problems from pyro, describe its design philosophy, and suggest extensions for students to build their understanding of these methods.
Preliminary study of CANDU moderator thermal hydraulics using the CUPID code
Energy Technology Data Exchange (ETDEWEB)
Park, Sang Gi; Jeong Jae Jun [Pusan National Univ., Busan (Korea, Republic of); Lee, Jae Ryong; Kim, Hyoung Tae [KAERI, Daejeon (Korea, Republic of)
2012-10-15
When the moderator cooling system fails, moderator may act as to remove decay heat which occurs in fuel. During loss of coolant accident (LOCA), the film boiling occurs in the Calandria tube (CT) because the hot pressure tube would deform into contacting with the calandria tube. And lower subcooling would decrease the margin of the CT to dryout. So, it is important to estimate a local subcooling of the moderator inside the Calandria vessel. However, in order to predict the internal temperature the study of empirical experiments and calculations are needed because only the inlet/outlet temperature can be measured in real reactor. In this study, the internal flow of the moderator was predicted by using the CUPID code, which has been developed in KAERI. The CUPID adopts three dimensional, transient, two phase and three field model, and includes various physical models and correlations of the interfacial mass, momentum and energy transfer for the closure relations of the two fluid model. The CUPID code shows single phase and two phase flow through two phase flow calculations of virtual can be applied.
Ichikawa, Ryoko; Masuhara, Yasuhiro; Kasahara, Fumio
The Best Estimate Plus Uncertainty (BEPU) method has been prepared for the regulatory cross-check analysis at Japan Nuclear Energy Safety Organization (JNES) on base of the three-dimensional neutron-kinetics/thermal- hydraulics coupled code SKETCH-INS/TRACE5.0. In the preparation, TRACE5.0 is verified against the large-scale thermal-hydraulic tests carried out with NUPEC facility. These tests were focused on the pressure drop of steam-liquid two phase flow and void fraction distribution. From the comparison of the experimental data with other codes (RELAP5/MOD3.3 and TRAC-BF1), TRACE5.0 was judged better than other codes. It was confirmed that TRACE5.0 has high reliability for thermal hydraulics behavior and are used as a best-estimate code for the statistical safety evaluation. Next, the coupled code SKETCH-INS/TRACE5.0 was applied to turbine trip tests performed at the Peach Bottom-2 BWR4 Plant. The turbine trip event shows the rapid power peak due to the voids collapse with the pressure increase. The analyzed peak value of core power is better simulated than the previous version SKETCH-INS/TRAC-BF1. And the statistical safety evaluation using SKETCH-INS/TRACE5.0 was applied to the loss of load transient for examining the influence of the choice of sampling method.
Geometric plane shapes for computer-generated holographic engraving codes
Augier, Ángel G.; Rabal, Héctor; Sánchez, Raúl B.
2017-04-01
We report a new theoretical and experimental study on hologravures, as holographic computer-generated laser-engravings. A geometric theory of images based on the general principles of light ray behaviour is shown. The models used are also applicable for similar engravings obtained by any non-laser method, and the solutions allow for the analysis of particular situations, not only in the case of light reflection mode, but also in transmission mode geometry. This approach is a novel perspective allowing the three-dimensional (3D) design of engraved images for specific ends. We prove theoretically that plane curves of very general geometric shapes can be used to encode image information onto a two-dimensional (2D) engraving, showing notable influence on the behaviour of reconstructed images that appears as an exciting investigation topic, extending its applications. Several cases of code using particular curvilinear shapes are experimentally studied. The computer-generated objects are coded by using the chosen curve type, and engraved by a laser on a plane surface of suitable material. All images are recovered optically by adequate illumination. The pseudoscopic or orthoscopic character of these images is considered, and an appropriate interpretation is presented.
Smith, P D
1982-01-01
BASIC Hydraulics aims to help students both to become proficient in the BASIC programming language by actually using the language in an important field of engineering and to use computing as a means of mastering the subject of hydraulics. The book begins with a summary of the technique of computing in BASIC together with comments and listing of the main commands and statements. Subsequent chapters introduce the fundamental concepts and appropriate governing equations. Topics covered include principles of fluid mechanics; flow in pipes, pipe networks and open channels; hydraulic machinery;
Evaluation of detonation energy from EXPLO5 computer code results
Energy Technology Data Exchange (ETDEWEB)
Suceska, M. [Brodarski Institute, Zagreb (Croatia). Marine Research and Special Technologies
1999-10-01
The detonation energies of several high explosives are evaluated from the results of chemical-equilibrium computer code named EXPLO5. Two methods of the evaluation of detonation energy are applied: (a) Direct evaluation from the internal energy of detonation products at the CJ point and the energy of shock compression of the detonation products, i.e. by equating the detonation energy and the heat of detonation, and (b) evaluation from the expansion isentrope of detonation products, applying the JWL model. These energies are compared to the energies computed from cylinder test derived JWL coefficients. It is found out that the detonation energies obtained directly from the energy of detonation products at the CJ point are uniformly to high (0.9445{+-}0.577 kJ/cm{sup 3}) while the detonation energies evaluated from the expansion isentrope, are in a considerable agreement (0.2072{+-}0.396 kJ/cm{sup 3}) with the energies calculated from cylinder test derived JWL coefficients. (orig.) [German] Die Detonationsenergien verschiedener Hochleistungssprengstoffe werden bewertet aus den Ergebnissen des Computer Codes fuer chemische Gleichgewichte genannt EXPLO5. Zwei Methoden wurden angewendet: (a) Direkte Bewertung aus der inneren Energie der Detonationsprodukte am CJ-Punkt und aus der Energie der Stosskompression der Detonationsprodukte, d.h. durch Gleichsetzung von Detonationsenergie und Detonationswaerme, (b) Auswertung durch die Expansions-Isentrope der Detonationsprodukte unter Anwendung des JWL-Modells. Diese Energien werden verglichen mit den berechneten Energien mit aus dem Zylindertest abgeleiteten JWL-Koeffizienten. Es wird gefunden, dass die Detonationsenergien, die direkt aus der Energie der Detonationsprodukte beim CJ-Punkt erhalten wurden, einheitlich zu hoch sind (0,9445{+-}0,577 kJ/cm{sup 3}), waehrend die aus der Expansions-Isentrope erhaltenen in guter Uebereinstimmung sind (0,2072{+-}0,396 kJ/cm{sup 3}) mit den berechneten Energien mit aus dem Zylindertest
Good, Jonathon; Keenan, Sarah; Mishra, Punya
2016-01-01
The popular press is rife with examples of how students in the United States and around the globe are learning to program, make, and tinker. The Hour of Code, maker-education, and similar efforts are advocating that more students be exposed to principles found within computer science. We propose an expansion beyond simply teaching computational…
DEFF Research Database (Denmark)
Jensen, M.D.; Ingildsen, P.; Rasmussen, Michael R.;
2006-01-01
Aeration tank settling is a control method allowing settling in the process tank during highhydraulic load. The control method is patented. Aeration tank settling has been applied in several wastewater treatment plants using the present design of the process tanks. Some process tank designs...... haveshown to be more effective than others. To improve the design of less effective plants, computational fluiddynamics (CFD) modelling of hydraulics and sedimentation has been applied. This paper discusses theresults at one particular plant experiencing problems with partly short-circuiting of the inlet...... and outletcausing a disruption of the sludge blanket at the outlet and thereby reducing the retention of sludge in theprocess tank. The model has allowed us to establish a clear picture of the problems arising at the plantduring aeration tank settling. Secondly, several process tank design changes have been...
A computer code to simulate X-ray imaging techniques
Energy Technology Data Exchange (ETDEWEB)
Duvauchelle, Philippe E-mail: philippe.duvauchelle@insa-lyon.fr; Freud, Nicolas; Kaftandjian, Valerie; Babot, Daniel
2000-09-01
A computer code was developed to simulate the operation of radiographic, radioscopic or tomographic devices. The simulation is based on ray-tracing techniques and on the X-ray attenuation law. The use of computer-aided drawing (CAD) models enables simulations to be carried out with complex three-dimensional (3D) objects and the geometry of every component of the imaging chain, from the source to the detector, can be defined. Geometric unsharpness, for example, can be easily taken into account, even in complex configurations. Automatic translations or rotations of the object can be performed to simulate radioscopic or tomographic image acquisition. Simulations can be carried out with monochromatic or polychromatic beam spectra. This feature enables, for example, the beam hardening phenomenon to be dealt with or dual energy imaging techniques to be studied. The simulation principle is completely deterministic and consequently the computed images present no photon noise. Nevertheless, the variance of the signal associated with each pixel of the detector can be determined, which enables contrast-to-noise ratio (CNR) maps to be computed, in order to predict quantitatively the detectability of defects in the inspected object. The CNR is a relevant indicator for optimizing the experimental parameters. This paper provides several examples of simulated images that illustrate some of the rich possibilities offered by our software. Depending on the simulation type, the computation time order of magnitude can vary from 0.1 s (simple radiographic projection) up to several hours (3D tomography) on a PC, with a 400 MHz microprocessor. Our simulation tool proves to be useful in developing new specific applications, in choosing the most suitable components when designing a new testing chain, and in saving time by reducing the number of experimental tests.
Energy Technology Data Exchange (ETDEWEB)
Chaparro V, F. J.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, UP - Adolfo Lopez Mateos, Edif. 9, 07738 Mexico D. F. (Mexico); Rodriguez H, A.; Gomez T, A. M. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Sanchez E, V. H.; Jager, W., E-mail: evalle@esfm.ipn.mx [Karlsruhe Institute of Technology, Hermann-von-Helmholtz Platz I, D-76344 Eggenstein - Leopoldshafen (Germany)
2014-10-15
In this article the results of the design of a pressure vessel of a BWR/5 similar to the type of Laguna Verde NPP are presented, using the Trace code. A thermo hydraulics Vessel component capable of simulating the behavior of fluids and heat transfer that occurs within the reactor vessel was created. The Vessel component consists of a three-dimensional cylinder divided into 19 axial sections, 4 azimuthal sections and two concentric radial rings. The inner ring is used to contain the core and the central part of the reactor, while the outer ring is used as a down comer. Axial an azimuthal divisions were made with the intention that the dimensions of the internal components, heights and orientation of the external connections match the reference values of a reactor BWR/5 type. In the model internal components as, fuel assemblies, steam separators, jet pumps, guide tubes, etc. are included and main external connections as, steam lines, feed-water or penetrations of the recirculation system. The model presents significant simplifications because the object is to keep symmetry between each azimuthal section of the vessel. In most internal components lack a detailed description of the geometry and initial values of temperature, pressure, fluid velocity, etc. given that it only considered the most representative data, however with these simulations are obtained acceptable results in important parameters such as the total flow through the core, the pressure in the vessel, percentage of vacuums fraction, pressure drop in the core and the steam separators. (Author)
Reasoning with Computer Code: a new Mathematical Logic
Pissanetzky, Sergio
2013-01-01
A logic is a mathematical model of knowledge used to study how we reason, how we describe the world, and how we infer the conclusions that determine our behavior. The logic presented here is natural. It has been experimentally observed, not designed. It represents knowledge as a causal set, includes a new type of inference based on the minimization of an action functional, and generates its own semantics, making it unnecessary to prescribe one. This logic is suitable for high-level reasoning with computer code, including tasks such as self-programming, objectoriented analysis, refactoring, systems integration, code reuse, and automated programming from sensor-acquired data. A strong theoretical foundation exists for the new logic. The inference derives laws of conservation from the permutation symmetry of the causal set, and calculates the corresponding conserved quantities. The association between symmetries and conservation laws is a fundamental and well-known law of nature and a general principle in modern theoretical Physics. The conserved quantities take the form of a nested hierarchy of invariant partitions of the given set. The logic associates elements of the set and binds them together to form the levels of the hierarchy. It is conjectured that the hierarchy corresponds to the invariant representations that the brain is known to generate. The hierarchies also represent fully object-oriented, self-generated code, that can be directly compiled and executed (when a compiler becomes available), or translated to a suitable programming language. The approach is constructivist because all entities are constructed bottom-up, with the fundamental principles of nature being at the bottom, and their existence is proved by construction. The new logic is mathematically introduced and later discussed in the context of transformations of algorithms and computer programs. We discuss what a full self-programming capability would really mean. We argue that self
Energy Technology Data Exchange (ETDEWEB)
Cho, H. K.; Lee, S. J.; Kang, K. H.; Yoon, H. Y. [Korea Atomic Energy Research Inst., 1045 Daeduk-daero, Daejeon (Korea, Republic of)
2012-07-01
For the analysis of transient two-phase flows in nuclear reactor components, a three-dimensional thermal hydraulics code, named CUPID, has been being developed. In the present study, the CUPID code was applied for the simulation of the PASCAL (PAFS Condensing Heat Removal Assessment Loop) test facility constructed with an aim of validating the cooling and operational performance of the PAFS (Passive Auxiliary Feedwater System). The PAFS is one of the advanced safety features adopted in the APR+ (Advanced Power Reactor +), which is intended to completely replace the conventional active auxiliary feedwater system. This paper presents the preliminary simulation results of the PASCAL facility performed with the CUPID code in order to verify its applicability to the thermal-hydraulic phenomena inside the system. A standalone calculation for the passive condensation cooling tank was performed by imposing a heat source boundary condition and the transient thermal-hydraulic behaviors inside the system, such as the water level, temperature and velocity, were qualitatively investigated. The simulation results verified that the natural circulation and boiling phenomena in the water pool can be well reproduced by the CUPID code. (authors)
Interface design of VSOP'94 computer code for safety analysis
Energy Technology Data Exchange (ETDEWEB)
Natsir, Khairina, E-mail: yenny@batan.go.id; Andiwijayakusuma, D.; Wahanani, Nursinta Adi [Center for Development of Nuclear Informatics - National Nuclear Energy Agency, PUSPIPTEK, Serpong, Tangerang, Banten (Indonesia); Yazid, Putranto Ilham [Center for Nuclear Technology, Material and Radiometry- National Nuclear Energy Agency, Jl. Tamansari No.71, Bandung 40132 (Indonesia)
2014-09-30
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.
Compute-and-Forward: Harnessing Interference through Structured Codes
Nazer, Bobak
2009-01-01
Interference is usually viewed as an obstacle to communication in wireless networks. This paper proposes a new strategy, compute-and-forward, that exploits interference to obtain significantly higher rates between users in a network. The key idea is that relays should decode linear functions of transmitted messages according to their observed channel coefficients rather than ignoring the interference as noise. After decoding these linear equations, the relays simply send them towards the destinations, which given enough equations, can recover their desired messages. The underlying codes are based on nested lattices whose algebraic structure ensures that integer combinations of codewords can be decoded reliably. Encoders map messages from a finite field to a lattice and decoders recover equations of lattice points which are then mapped back to equations over the finite field. This scheme is applicable even if the transmitters lack channel state information. Its potential is demonstrated through examples drawn ...
Benchmark Solutions for Computational Aeroacoustics (CAA) Code Validation
Scott, James R.
2004-01-01
NASA has conducted a series of Computational Aeroacoustics (CAA) Workshops on Benchmark Problems to develop a set of realistic CAA problems that can be used for code validation. In the Third (1999) and Fourth (2003) Workshops, the single airfoil gust response problem, with real geometry effects, was included as one of the benchmark problems. Respondents were asked to calculate the airfoil RMS pressure and far-field acoustic intensity for different airfoil geometries and a wide range of gust frequencies. This paper presents the validated that have been obtained to the benchmark problem, and in addition, compares them with classical flat plate results. It is seen that airfoil geometry has a strong effect on the airfoil unsteady pressure, and a significant effect on the far-field acoustic intensity. Those parts of the benchmark problem that have not yet been adequately solved are identified and presented as a challenge to the CAA research community.
Fire aerosol experiment and comparisons with computer code predictions
Energy Technology Data Exchange (ETDEWEB)
Gregory, W.S.; Nichols, B.D.; White, B.W.; Smith, P.R.; Leslie, I.H.; Corkran, J.R.
1988-01-01
Los Alamos National Laboratory, in cooperation with New Mexico State University, has carried on a series of tests to provide experimental data on fire-generated aerosol transport. These data will be used to verify the aerosol transport capabilities of the FIRAC computer code. FIRAC was developed by Los Alamos for the US Nuclear Regulatory Commission. It is intended to be used by safety analysts to evaluate the effects of hypothetical fires on nuclear plants. One of the most significant aspects of this analysis deals with smoke and radioactive material movement throughout the plant. The tests have been carried out using an industrial furnace that can generate gas temperatures to 300/degree/C. To date, we have used quartz aerosol with a median diameter of about 10 ..mu..m as the fire aerosol simulant. We also plan to use fire-generated aerosols of polystyrene and polymethyl methacrylate (PMMA). The test variables include two nominal gas flow rates (150 and 300 ft/sup 3//min) and three nominal gas temperatures (ambient, 150/degree/C, and 300/degree/C). The test results are presented in the form of plots of aerosol deposition vs length of duct. In addition, the mass of aerosol caught in a high-efficiency particulate air (HEPA) filter during the tests is reported. The tests are simulated with the FIRAC code, and the results are compared with the experimental data. 3 refs., 10 figs., 1 tab.
Application of the RESRAD computer code to VAMP scenario S
Energy Technology Data Exchange (ETDEWEB)
Gnanapragasam, E.K.; Yu, C.
1997-03-01
The RESRAD computer code developed at Argonne National Laboratory was among 11 models from 11 countries participating in the international Scenario S validation of radiological assessment models with Chernobyl fallout data from southern Finland. The validation test was conducted by the Multiple Pathways Assessment Working Group of the Validation of Environmental Model Predictions (VAMP) program coordinated by the International Atomic Energy Agency. RESRAD was enhanced to provide an output of contaminant concentrations in environmental media and in food products to compare with measured data from southern Finland. Probability distributions for inputs that were judged to be most uncertain were obtained from the literature and from information provided in the scenario description prepared by the Finnish Centre for Radiation and Nuclear Safety. The deterministic version of RESRAD was run repeatedly to generate probability distributions for the required predictions. These predictions were used later to verify the probabilistic RESRAD code. The RESRAD predictions of radionuclide concentrations are compared with measured concentrations in selected food products. The radiological doses predicted by RESRAD are also compared with those estimated by the Finnish Centre for Radiation and Nuclear Safety.
Computer Tensor Codes to Design the War Drive
Maccone, C.
To address problems in Breakthrough Propulsion Physics (BPP) and design the Warp Drive one needs sheer computing capabilities. This is because General Relativity (GR) and Quantum Field Theory (QFT) are so mathematically sophisticated that the amount of analytical calculations is prohibitive and one can hardly do all of them by hand. In this paper we make a comparative review of the main tensor calculus capabilities of the three most advanced and commercially available “symbolic manipulator” codes. We also point out that currently one faces such a variety of different conventions in tensor calculus that it is difficult or impossible to compare results obtained by different scholars in GR and QFT. Mathematical physicists, experimental physicists and engineers have each their own way of customizing tensors, especially by using different metric signatures, different metric determinant signs, different definitions of the basic Riemann and Ricci tensors, and by adopting different systems of physical units. This chaos greatly hampers progress toward the design of the Warp Drive. It is thus suggested that NASA would be a suitable organization to establish standards in symbolic tensor calculus and anyone working in BPP should adopt these standards. Alternatively other institutions, like CERN in Europe, might consider the challenge of starting the preliminary implementation of a Universal Tensor Code to design the Warp Drive.
Computer code for the atomistic simulation of lattice defects and dynamics. [COMENT code
Energy Technology Data Exchange (ETDEWEB)
Schiffgens, J.O.; Graves, N.J.; Oster, C.A.
1980-04-01
This document has been prepared to satisfy the need for a detailed, up-to-date description of a computer code that can be used to simulate phenomena on an atomistic level. COMENT was written in FORTRAN IV and COMPASS (CDC assembly language) to solve the classical equations of motion for a large number of atoms interacting according to a given force law, and to perform the desired ancillary analysis of the resulting data. COMENT is a dual-purpose intended to describe static defect configurations as well as the detailed motion of atoms in a crystal lattice. It can be used to simulate the effect of temperature, impurities, and pre-existing defects on radiation-induced defect production mechanisms, defect migration, and defect stability.
Computational fluid dynamics simulation and geometric design of hydraulic turbine draft tube
Directory of Open Access Journals (Sweden)
JB Sosa
2015-10-01
Full Text Available Any hydraulic reaction turbine is installed with a draft tube that impacts widely the entire turbine performance, on which its functions are as follows: drive the flux in appropriate manner after it releases its energy to the runner; recover the suction head by a suction effect; and improve the dynamic energy in the runner outlet. All these functions are strongly linked to the geometric definition of the draft tube. This article proposes a geometric parametrization and analysis of a Francis turbine draft tube. Based on the parametric definition, geometric changes in the draft tube are proposed and the turbine performance is modeled by computational fluid dynamics; the boundary conditions are set by measurements performed in a hydroelectric power plant. This modeling allows us to see the influence of the draft tube shape on the entire turbine performance. The numerical analysis is based on the steady-state solution of the turbine component flows for different guide vanes opening and multiple modified draft tubes. The computational fluid dynamics predictions are validated using hydroelectric plant measurements. The prediction of the turbine performance is successful and it is linked to the draft tube geometric features; therefore, it is possible to obtain a draft tube parameter value that results in a desired turbine performance.
Energy Technology Data Exchange (ETDEWEB)
Kotsarev, Alexander; Lizorkin, Mikhail [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation); Bencik, Marek; Hadek, Jan [UJV Rez, a.s., Rez (Czech Republic); Kozmenkov, Yaroslav; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany)
2016-09-15
The 7th AER dynamic benchmark is a continuation of the efforts to validate the codes systematically for the estimation of the transient behavior of VVER type nuclear power plants. The main part of the benchmark is the simulation of the re-connection of an isolated circulation loop with low temperature in a VVER-440 plant. This benchmark was calculated by the National Research Centre ''Kurchatov Institute'' (with the code ATHLET/BIPR-VVER), UJV Rez (with the code RELAP5-3D {sup copyright}) and HZDR (with the code DYN3D/ATHLET). The paper gives an overview of the behavior of the main thermal hydraulic and neutron kinetic parameters in the provided solutions.
Computation and analysis of cavitating flow in Francis-class hydraulic turbines
Leonard, Daniel J.
Hydropower is the most proven renewable energy technology, supplying the world with 16% of its electricity. Conventional hydropower generates a vast majority of that percentage. Although a mature technology, hydroelectric generation shows great promise for expansion through new dams and plants in developing hydro countries. Moreover, in developed hydro countries, such as the United States, installing generating units in existing dams and the modern refurbishment of existing plants can greatly expand generating capabilities with little to no further impact on the environment. In addition, modern computational technology and fluid dynamics expertise has led to substantial improvements in modern turbine design and performance. Cavitation has always presented a problem in hydroturbines, causing performance breakdown, erosion, damage, vibration, and noise. While modern turbines are usually designed to be cavitation-free at their best efficiency point, due to the variable demand of the energy market it is fairly common to operate at off-design conditions. Here, cavitation and its deleterious effects are unavoidable, and hence, cavitation is a limiting factor on the design and operation of these turbines. Multiphase Computational Fluid Dynamics (CFD) has been used in recent years to model cavitating flow for a large range of problems, including turbomachinery. However, CFD of cavitating flow in hydroturbines is still in its infancy. This dissertation presents steady-periodic Reynolds-averaged Navier-Stokes simulations of a cavitating Francis-class hydroturbine at model and prototype scales. Computational results of the reduced-scale model and full-scale prototype, undergoing performance breakdown, are compared with empirical model data and prototype performance estimations based on standard industry scalings from the model data. Mesh convergence of the simulations is also displayed. Comparisons are made between the scales to display that cavitation performance breakdown
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
Implementation of a 3D mixing layer code on parallel computers
Energy Technology Data Exchange (ETDEWEB)
Roe, K.; Thakur, R.; Dang, T.; Bogucz, E. [Syracuse Univ., NY (United States)
1995-09-01
This paper summarizes our progress and experience in the development of a Computational-Fluid-Dynamics code on parallel computers to simulate three-dimensional spatially-developing mixing layers. In this initial study, the three-dimensional time-dependent Euler equations are solved using a finite-volume explicit time-marching algorithm. The code was first programmed in Fortran 77 for sequential computers. The code was then converted for use on parallel computers using the conventional message-passing technique, while we have not been able to compile the code with the present version of HPF compilers.
Energy Technology Data Exchange (ETDEWEB)
Kim, Geon-Woo; Lee, Jeong-Hun; Park, Goon-Cherl; Cho, Hyoung-Kyu [Seoul National University, Seoul (Korea, Republic of); Im, Kihak [National Fusion Research Institute, Daejeon (Korea, Republic of)
2015-10-15
A preliminary concept for the Korean fusion demonstration reactor (K-DEMO) has been studied by the National Fusion Research Institute (NFRI) based on the National Fusion Roadmap of Korea. The feasibility studies have been performed in order to establish the conceptual design guidelines of the breeding blanket. As a part of the NFRI research, Seoul National University (SNU) is conducting thermal design, evaluation and validation of the water-cooled breeding blanket for the K-DEMO reactor. The purpose of this study is to extend the capability of MARS-KS to the overall blanket system analysis which includes 736 blanket modules in total. The strategy for the multi-module blanket system analysis using MARS-KS is introduced and the analysis result of the 46 blanket modules of single sector was summarized. A thermal-hydraulic analysis code for a nuclear reactor safety, MARS-KS, was applied for thermal analysis of the conceptual design of the K-DEMO breeding blanket. Then, a methodology to simulate multiple blanket modules was proposed, which uses a supervisor program to handle each blanket module individually at first and then distribute the flow rate considering the pressure drop that occurs in each module. For a feasibility test of the proposed methodology, 46 blankets in a sector, which are connected with each other through the common headers for the sector inlet and outlet, were simulated. The calculation results of flow rates, pressure drops, and temperatures showed the validity of the calculation. Because of parallelization using the MPI system, the computational time could be reduced significantly. In future, this methodology will be extended to an efficient simulation of multiple sectors, and further validation for transient simulation will be carried out for more practical applications.
Grenier, Christophe; Rühaak, Wolfram
2016-04-01
Climate change impacts in permafrost regions have received considerable attention recently due to the pronounced warming trends experienced in recent decades and which have been projected into the future. Large portions of these permafrost regions are characterized by surface water bodies (lakes, rivers) that interact with the surrounding permafrost often generating taliks (unfrozen zones) within the permafrost that allow for hydrologic interactions between the surface water bodies and underlying aquifers and thus influence the hydrologic response of a landscape to 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 past and future evolution such units (Kurylyk et al. 2014). However, there is presently a paucity of 3D numerical studies of permafrost thaw and associated hydrological changes, which can be partly attributed to the difficulty in verifying multi-dimensional results produced by numerical models. A benchmark exercise was initialized at the end of 2014. Participants convened from USA, Canada, Europe, representing 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 (Kurylyk et al. 2014; Grenier et al. in prep.; Rühaak et al. 2015). They range from simpler, purely thermal 1D cases to more complex, coupled 2D TH cases (benchmarks TH1, TH2, and TH3). Some experimental cases conducted in a 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 case databases 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 TH2 & TH3. Both cases
Caviedes-Voullième, Daniel; Kesserwani, Georges
2015-12-01
Numerical modelling of wide ranges of different physical scales, which are involved in Shallow Water (SW) problems, has been a key challenge in computational hydraulics. Adaptive meshing techniques have been commonly coupled with numerical methods in an attempt to address this challenge. The combination of MultiWavelets (MW) with the Runge-Kutta Discontinuous Galerkin (RKDG) method offers a new philosophy to readily achieve mesh adaptivity driven by the local variability of the numerical solution, and without requiring more than one threshold value set by the user. However, the practical merits and implications of the MWRKDG, in terms of how far it contributes to address the key challenge above, are yet to be explored. This work systematically explores this, through the verification and validation of the MWRKDG for selected steady and transient benchmark tests, which involves the features of real SW problems. Our findings reveal a practical promise of the SW-MWRKDG solver, in terms of efficient and accurate mesh-adaptivity, but also suggest further improvement in the SW-RKDG reference scheme to better intertwine with, and harness the prowess of, the MW-based adaptivity.
Energy Technology Data Exchange (ETDEWEB)
Vasiliev, A., E-mail: vasil@ibrae.ac.ru [Nuclear Safety Institute (IBRAE), B. Tulskaya 52, 115191 Moscow (Russian Federation); Stuckert, J., E-mail: juri.stuckert@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)
2015-03-15
Highlights: • Modeling of processes in porous debris regions. • Analysis of coolability of massive debris bed. • Complexity of simulation of flow regime near boiling curve. - Abstract: The thermal hydraulic and SFD (Severe Fuel Damage) best estimate computer modeling code SOCRAT/V3 was used for the post-test analysis of the QUENCH-17 experiment performed at KIT on January 2013. The objective of this test was to examine the formation of a debris bed inside the completely oxidized region of the bundle without melt formation and to investigate the coolability behavior during the reflood. The test bundle for QUENCH-17 test was intentionally changed in comparison to basic QUENCH bundles (usually 21 heated rod simulators) with the emphasis to investigate debris behavior phenomena. Only 12 periphery fuel rod simulators were heated by centerline tungsten heaters. 9 unheated fuel rod simulators were located in the inner part of the test bundle. This is why the massive porous debris formation in the inner part of the bundle was not influenced by the presence of tungsten heaters. The QUENCH-17 test conditions simulated a hypothetical scenario of nuclear power plant severe accident sequence with debris bed formation in which the overheated up to 1800 K core would be flooded from the bottom by ECCS (Emergency Core Cooling System). The QUENCH-17 test included the following phases: (1) heat-up phase (heat-up rate up to 0.25 K/s); (2) oxidation phase (the cladding temperature about 1800 K in hottest region, steam mass flow rate 2 g/s); (3) bottom flood phase (characteristic cooling time about 600 s, water mass flow rate 10 g/s). SOCRAT/V3 computer modeling code was used for calculation of basic thermal hydraulic, oxidation and thermal mechanical behavior during all phases of the experiment. The calculated results are in a good agreement with experimental data which justifies the adequacy of modeling capabilities of SOCRAT code system.
Superimposed Code Theoretic Analysis of Deoxyribonucleic Acid (DNA) Codes and DNA Computing
2010-01-01
DNA Codes Based on Fibonacci Ensembles of DNA Sequences ”, 2008 IEEE Proceedings of International Symposium on Information Theory, pp. 2292 – 2296...2008, pp. 525-34. 28. A. Macula, et al., “Random Coding Bounds for DNA Codes Based on Fibonacci Ensembles of DNA Sequences ”, 2008 IEEE...component of this innovation is the combinatorial method of bio-memory design and detection that encodes item or process information as numerical sequences
Directory of Open Access Journals (Sweden)
Itamar Iliuk
2016-01-01
Full Text Available Thermal-hydraulic analysis of plate-type fuel has great importance to the establishment of safety criteria, also to the licensing of the future nuclear reactor with the objective of propelling the Brazilian nuclear submarine. In this work, an analysis of a single plate-type fuel surrounding by two water channels was performed using the RELAP5 thermal-hydraulic code. To realize the simulations, a plate-type fuel with the meat of uranium dioxide sandwiched between two Zircaloy-4 plates was proposed. A partial loss of flow accident was simulated to show the behavior of the model under this type of accident. The results show that the critical heat flux was detected in the central region along the axial direction of the plate when the right water channel was blocked.
V. S. O. P. ('94) Computer Code System for Reactor Physics and Fuel Cycle Simulation
Teuchert, E.; Haas, K. A.; Rütten, H. J.; Brockmann, Hans; Gerwin, Helmut; Ohlig, U.; Scherer, Winfried
1994-01-01
V. S. O. P. ('Very Superior Old Programs) is a system of codes lurked together for the simulationof reactor life histories and temporary in-depth research. In comprises neutron cross sectionlibraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculationwith depletion and shut-down features, in-core and out-of--pile fuel management, fuel cyclecost analysis, and thermal hydraulics (at present restricted to 's). Various techniques havebeen employed to accelerat...
V. S. O. P. - Computer Code System for Reactor Physics and Fuel Cycle Simulation
Teuchert, E.; Hansen, U.; Haas, K. A.
1980-01-01
V .S .O .P . (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprisesneutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based onneutron flux synthesis with depletion and shut-down features, incore and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employe...
Energy Technology Data Exchange (ETDEWEB)
Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.
1998-03-01
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 SPARC 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.
User manual for PACTOLUS: a code for computing power costs.
Energy Technology Data Exchange (ETDEWEB)
Huber, H.D.; Bloomster, C.H.
1979-02-01
PACTOLUS is a computer code for calculating the cost of generating electricity. Through appropriate definition of the input data, PACTOLUS can calculate the cost of generating electricity from a wide variety of power plants, including nuclear, fossil, geothermal, solar, and other types of advanced energy systems. The purpose of PACTOLUS is to develop cash flows and calculate the unit busbar power cost (mills/kWh) over the entire life of a power plant. The cash flow information is calculated by two principal models: the Fuel Model and the Discounted Cash Flow Model. The Fuel Model is an engineering cost model which calculates the cash flow for the fuel cycle costs over the project lifetime based on input data defining the fuel material requirements, the unit costs of fuel materials and processes, the process lead and lag times, and the schedule of the capacity factor for the plant. For nuclear plants, the Fuel Model calculates the cash flow for the entire nuclear fuel cycle. For fossil plants, the Fuel Model calculates the cash flow for the fossil fuel purchases. The Discounted Cash Flow Model combines the fuel costs generated by the Fuel Model with input data on the capital costs, capital structure, licensing time, construction time, rates of return on capital, tax rates, operating costs, and depreciation method of the plant to calculate the cash flow for the entire lifetime of the project. The financial and tax structure for both investor-owned utilities and municipal utilities can be simulated through varying the rates of return on equity and debt, the debt-equity ratios, and tax rates. The Discounted Cash Flow Model uses the principal that the present worth of the revenues will be equal to the present worth of the expenses including the return on investment over the economic life of the project. This manual explains how to prepare the input data, execute cases, and interpret the output results. (RWR)
MMA, A Computer Code for Multi-Model Analysis
Energy Technology Data Exchange (ETDEWEB)
Eileen P. Poeter and Mary C. Hill
2007-08-20
This report documents the Multi-Model Analysis (MMA) computer code. MMA can be used to evaluate results from alternative models of a single system using the same set of observations for all models. As long as the observations, the observation weighting, and system being represented are the same, the models can differ in nearly any way imaginable. For example, they may include different processes, different simulation software, different temporal definitions (for example, steady-state and transient models could be considered), and so on. The multiple models need to be calibrated by nonlinear regression. Calibration of the individual models needs to be completed before application of MMA. MMA can be used to rank models and calculate posterior model probabilities. These can be used to (1) determine the relative importance of the characteristics embodied in the alternative models, (2) calculate model-averaged parameter estimates and predictions, and (3) quantify the uncertainty of parameter estimates and predictions in a way that integrates the variations represented by the alternative models. There is a lack of consensus on what model analysis methods are best, so MMA provides four default methods. Two are based on Kullback-Leibler information, and use the AIC (Akaike Information Criterion) or AICc (second-order-bias-corrected AIC) model discrimination criteria. The other two default methods are the BIC (Bayesian Information Criterion) and the KIC (Kashyap Information Criterion) model discrimination criteria. Use of the KIC criterion is equivalent to using the maximum-likelihood Bayesian model averaging (MLBMA) method. AIC, AICc, and BIC can be derived from Frequentist or Bayesian arguments. The default methods based on Kullback-Leibler information have a number of theoretical advantages, including that they tend to favor more complicated models as more data become available than do the other methods, which makes sense in many situations.
Reducing Computational Overhead of Network Coding with Intrinsic Information Conveying
DEFF Research Database (Denmark)
Heide, Janus; Zhang, Qi; Pedersen, Morten V.
This paper investigated the possibility of intrinsic information conveying in network coding systems. The information is embedded into the coding vector by constructing the vector based on a set of predefined rules. This information can subsequently be retrieved by any receiver. The starting point...... to the overall energy consumption, which is particular problematic for mobile battery-driven devices. In RLNC coding is performed over a FF (Finite Field). We propose to divide this field into sub fields, and let each sub field signify some information or state. In order to embed the information correctly...... the coding operations must be performed in a particular way, which we introduce. Finally we evaluate the suggested system and find that the amount of coding can be significantly reduced both at nodes that recode and decode....
3-D field computation: The near-triumph of commerical codes
Energy Technology Data Exchange (ETDEWEB)
Turner, L.R.
1995-07-01
In recent years, more and more of those who design and analyze magnets and other devices are using commercial codes rather than developing their own. This paper considers the commercial codes and the features available with them. Other recent trends with 3-D field computation include parallel computation and visualization methods such as virtual reality systems.
Directory of Open Access Journals (Sweden)
Taewan Kim
2012-01-01
Full Text Available In order to assess the accuracy and validity of subchannel, system, and computational fluid dynamics codes, the Paul Scherrer Institut has participated in the OECD/NRC PSBT benchmark with the thermal-hydraulic system code TRACE5.0 developed by US NRC, the subchannel code FLICA4 developed by CEA, and the computational fluid dynamic code STAR-CD developed by CD-adapco. The PSBT benchmark consists of a series of void distribution exercises and departure from nucleate boiling exercises. The results reveal that the prediction by the subchannel code FLICA4 agrees with the experimental data reasonably well in both steady-state and transient conditions. The analyses of single-subchannel experiments by means of the computational fluid dynamic code STAR-CD with the CD-adapco boiling model indicate that the prediction of the void fraction has no significant discrepancy from the experiments. The analyses with TRACE point out the necessity to perform additional assessment of the subcooled boiling model and bulk condensation model of TRACE.
ORNL ALICE: a statistical model computer code including fission competition. [In FORTRAN
Energy Technology Data Exchange (ETDEWEB)
Plasil, F.
1977-11-01
A listing of the computer code ORNL ALICE is given. This code is a modified version of computer codes ALICE and OVERLAID ALICE. It allows for higher excitation energies and for a greater number of evaporated particles than the earlier versions. The angular momentum removal option was made more general and more internally consistent. Certain roundoff errors are avoided by keeping a strict accounting of partial probabilities. Several output options were added.
Comparison of computer codes for estimates of the symmetric coupled bunch instabilities growth times
Angal-Kalinin, Deepa
2002-01-01
The standard computer codes used for estimating the growth times of the symmetric coupled bunch instabilities are ZAP and BBI.The code Vlasov was earlier used for the LHC for the estimates of the coupled bunch instabilities growth time[1]. The results obtained by these three codes have been compared and the options under which their results can be compared are discussed. The differences in the input and the output for these three codes are given for a typical case.
Quantum error correcting codes and one-way quantum computing: Towards a quantum memory
Schlingemann, D
2003-01-01
For realizing a quantum memory we suggest to first encode quantum information via a quantum error correcting code and then concatenate combined decoding and re-encoding operations. This requires that the encoding and the decoding operation can be performed faster than the typical decoherence time of the underlying system. The computational model underlying the one-way quantum computer, which has been introduced by Hans Briegel and Robert Raussendorf, provides a suitable concept for a fast implementation of quantum error correcting codes. It is shown explicitly in this article is how encoding and decoding operations for stabilizer codes can be realized on a one-way quantum computer. This is based on the graph code representation for stabilizer codes, on the one hand, and the relation between cluster states and graph codes, on the other hand.
Energy Technology Data Exchange (ETDEWEB)
VOOGD, J.A.
1999-04-19
An analysis of three software proposals is performed to recommend a computer code for immobilized low activity waste flow and transport modeling. The document uses criteria restablished in HNF-1839, ''Computer Code Selection Criteria for Flow and Transport Codes to be Used in Undisturbed Vadose Zone Calculation for TWRS Environmental Analyses'' as the basis for this analysis.
Energy Technology Data Exchange (ETDEWEB)
Cho, Hyoung Kyu; Lee, Seung Jun; Yoon, Han Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2012-05-15
The need for a multi-dimensional analysis of transient thermal hydraulic phenomena in a component of a nuclear reactor is increasing with the advanced design features, such as a direct vessel injection system, a gravity-driven safety injection system, and a passive cooling system. Motivated by this, the development of a new thermal-hydraulic analysis code, named CUPID, is in progress at KAERI (Korea Atomic Energy Research Institute). Its numerical solver and two-phase flow models have been verified against standard conceptual problems of single and two-phase flows and validated for thermal-hydraulic experiments in our previous studies. The simulation of the passive secondary cooling system, PAFS (Passive Auxiliary Feedwater System) has been considered as one of the practical applications of CUPID. In the present study, the PCCT (Passive Condensation Cooling Tank) of the PASCAL test facility was analyzed with CUPID prior to simulating the prototype PAFS system. The objectives of the PASCAL simulation were to validate physical models of CUPID and its applicability to the PAFS analysis. This paper presents the two-dimensional transient calculation results and the comparisons with the experimental data
Quantum computation with topological codes from qubit to topological fault-tolerance
Fujii, Keisuke
2015-01-01
This book presents a self-consistent review of quantum computation with topological quantum codes. The book covers everything required to understand topological fault-tolerant quantum computation, ranging from the definition of the surface code to topological quantum error correction and topological fault-tolerant operations. The underlying basic concepts and powerful tools, such as universal quantum computation, quantum algorithms, stabilizer formalism, and measurement-based quantum computation, are also introduced in a self-consistent way. The interdisciplinary fields between quantum information and other fields of physics such as condensed matter physics and statistical physics are also explored in terms of the topological quantum codes. This book thus provides the first comprehensive description of the whole picture of topological quantum codes and quantum computation with them.
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.
Two-Phase Flow in Geothermal Wells: Development and Uses of a Good Computer Code
Energy Technology Data Exchange (ETDEWEB)
Ortiz-Ramirez, Jaime
1983-06-01
A computer code is developed for vertical two-phase flow in geothermal wellbores. The two-phase correlations used were developed by Orkiszewski (1967) and others and are widely applicable in the oil and gas industry. The computer code is compared to the flowing survey measurements from wells in the East Mesa, Cerro Prieto, and Roosevelt Hot Springs geothermal fields with success. Well data from the Svartsengi field in Iceland are also used. Several applications of the computer code are considered. They range from reservoir analysis to wellbore deposition studies. It is considered that accurate and workable wellbore simulators have an important role to play in geothermal reservoir engineering.
Efficient Quantification of Uncertainties in Complex Computer Code Results Project
National Aeronautics and Space Administration — Propagation of parameter uncertainties through large computer models can be very resource intensive. Frameworks and tools for uncertainty quantification are...
Second Generation Integrated Composite Analyzer (ICAN) Computer Code
Murthy, Pappu L. N.; Ginty, Carol A.; Sanfeliz, Jose G.
1993-01-01
This manual updates the original 1986 NASA TP-2515, Integrated Composite Analyzer (ICAN) Users and Programmers Manual. The various enhancements and newly added features are described to enable the user to prepare the appropriate input data to run this updated version of the ICAN code. For reference, the micromechanics equations are provided in an appendix and should be compared to those in the original manual for modifications. A complete output for a sample case is also provided in a separate appendix. The input to the code includes constituent material properties, factors reflecting the fabrication process, and laminate configuration. The code performs micromechanics, macromechanics, and laminate analyses, including the hygrothermal response of polymer-matrix-based fiber composites. The output includes the various ply and composite properties, the composite structural response, and the composite stress analysis results with details on failure. The code is written in FORTRAN 77 and can be used efficiently as a self-contained package (or as a module) in complex structural analysis programs. The input-output format has changed considerably from the original version of ICAN and is described extensively through the use of a sample problem.
Code of Ethical Conduct for Computer-Using Educators: An ICCE Policy Statement.
Computing Teacher, 1987
1987-01-01
Prepared by the International Council for Computers in Education's Ethics and Equity Committee, this code of ethics for educators using computers covers nine main areas: curriculum issues, issues relating to computer access, privacy/confidentiality issues, teacher-related issues, student issues, the community, school organizational issues,…
Computational Participation: Understanding Coding as an Extension of Literacy Instruction
Burke, Quinn; O'Byrne, W. Ian; Kafai, Yasmin B.
2016-01-01
Understanding the computational concepts on which countless digital applications run offers learners the opportunity to no longer simply read such media but also become more discerning end users and potentially innovative "writers" of new media themselves. To think computationally--to solve problems, to design systems, and to process and…
A computational method for oleo-acoustics, application to hydraulic shock absorbers
B. Koren (Barry); P.F.M. Michielsen (Paul); J.-W. Kars; P. Wesseling
1995-01-01
textabstractTo predict high-frequency oil-flow phenomena in hydraulic-shock-absorber designs, a mathematical-physical model is proposed. The model consists of the 2-D unsteady Euler equations in axial-symmetric coordinates and an appropriate equation of state for oil. The main topic of the paper is
Selection of a computer code for Hanford low-level waste engineered-system performance assessment
Energy Technology Data Exchange (ETDEWEB)
McGrail, B.P.; Mahoney, L.A.
1995-10-01
Planned performance assessments for the proposed disposal of low-level waste (LLW) glass produced from remediation of wastes stored in underground tanks at Hanford, Washington will require calculations of radionuclide release rates from the subsurface disposal facility. These calculations will be done with the aid of computer codes. Currently available computer codes were ranked in terms of the feature sets implemented in the code that match a set of physical, chemical, numerical, and functional capabilities needed to assess release rates from the engineered system. The needed capabilities were identified from an analysis of the important physical and chemical process expected to affect LLW glass corrosion and the mobility of radionuclides. The highest ranked computer code was found to be the ARES-CT code developed at PNL for the US Department of Energy for evaluation of and land disposal sites.
Validation of the NCC Code for Staged Transverse Injection and Computations for a RBCC Combustor
Ajmani, Kumud; Liu, Nan-Suey
2005-01-01
The NCC code was validated for a case involving staged transverse injection into Mach 2 flow behind a rearward facing step. Comparisons with experimental data and with solutions from the FPVortex code was then used to perform computations to study fuel-air mixing for the combustor of a candidate rocket based combined cycle engine geometry. Comparisons with a one-dimensional analysis and a three-dimensional code (VULCAN) were performed to assess the qualitative and quantitative performance of the NCC solver.
Challenges of Computational Processing of Code-Switching
Çetinoğlu, Özlem; Schulz, Sarah; Vu, Ngoc Thang
2016-01-01
This paper addresses challenges of Natural Language Processing (NLP) on non-canonical multilingual data in which two or more languages are mixed. It refers to code-switching which has become more popular in our daily life and therefore obtains an increasing amount of attention from the research community. We report our experience that cov- ers not only core NLP tasks such as normalisation, language identification, language modelling, part-of-speech tagging and dependency parsing but also more...
POTRE: A computer code for the assessment of dose from ingestion
Energy Technology Data Exchange (ETDEWEB)
Hanusik, V.; Mitro, A.; Niedel, S.; Grosikova, B.; Uvirova, E.; Stranai, I. (Institute of Radioecology and Applied Nuclear Techniques, Kosice (Czechoslovakia))
1991-01-01
The paper describes the computer code PORET and the auxiliary database system which allow to assess the radiation exposure from ingestion of foodstuffs contaminated by radionuclides released from a nuclear facility during normal operation into the atmosphere. (orig.).
Energy Technology Data Exchange (ETDEWEB)
Kalimullah, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Olson, Arne P. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States); Hanan, N. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States)
2015-04-07
The document compiles in a single volume several verification and validation works done for the PLTEMP/ANL code during the years of its development and improvement. Some works that are available in the open literature are simply referenced at the outset, and are not included in the document. PLTEMP has been used in conversion safety analysis reports of several US and foreign research reactors that have been licensed and converted. A list of such reactors is given. Each chapter of the document deals with the verification or validation of a specific model. The model verification is usually done by comparing the code with hand calculation, Microsoft spreadsheet calculation, or Mathematica calculation. The model validation is done by comparing the code with experimental data or a more validated code like the RELAP5 code.
Speeding-up MADYMO 3D on serial and parallel computers using a portable coding environment
Tsiandikos, T.; Rooijackers, H.F.L.; Asperen, F.G.J. van; Lupker, H.A.
1996-01-01
This paper outlines the strategy and methodology used to create a portable coding environment for the commercial package MADYMO. The objective is to design a global data structure that efficiently utilises the memory and cache of computers, so that one source code can be used for serial, vector and
Moral, Cristian; de Antonio, Angelica; Ferre, Xavier; Lara, Graciela
2015-01-01
Introduction: In this article we propose a qualitative analysis tool--a coding system--that can support the formalisation of the information-seeking process in a specific field: research in computer science. Method: In order to elaborate the coding system, we have conducted a set of qualitative studies, more specifically a focus group and some…
Holbrook, M. Cay; MacCuspie, P. Ann
2010-01-01
Braille-reading mathematicians, scientists, and computer scientists were asked to examine the usability of the Unified English Braille Code (UEB) for technical materials. They had little knowledge of the code prior to the study. The research included two reading tasks, a short tutorial about UEB, and a focus group. The results indicated that the…
Metropol, a computer code for the simulation of transport of contaminants with groundwater
Sauter FJ; Hassanizadeh SM; Leijnse A; Glasbergen P; Slot AFM
1990-01-01
In this report a description is given of the computer code METROPOL. This code simulates the three dimensional flow of groundwater with varying density and the simultaneous transport of contaminants in low concentration and is based on the finite element method. The basic equations for groundwater
Comparison of different computer platforms for running the Versatile Advection Code
Toth, G.; Keppens, R.; Sloot, P.; Bubak, M.; Hertzberger, B.
1998-01-01
The Versatile Advection Code is a general tool for solving hydrodynamical and magnetohydrodynamical problems arising in astrophysics. We compare the performance of the code on different computer platforms, including work stations and vector and parallel supercomputers. Good parallel scaling can be a
Code and papers: computing publication patterns in the LHC era
CERN. Geneva
2012-01-01
Publications in scholarly journals establish the body of knowledge deriving from scientific research; they also play a fundamental role in the career path of scientists and in the evaluation criteria of funding agencies. This presentation reviews the evolution of computing-oriented publications in HEP following the start of operation of LHC. Quantitative analyses are illustrated, which document the production of scholarly papers on computing-related topics by HEP experiments and core tools projects (including distributed computing R&D), and the citations they receive. Several scientometric indicators are analyzed to characterize the role of computing in HEP literature. Distinctive features of scholarly publication production in the software-oriented and hardware-oriented experimental HEP communities are highlighted. Current patterns and trends are compared to the situation in previous generations' HEP experiments at LEP, Tevatron and B-factories. The results of this scientometric analysis document objec...
Proposed standards for peer-reviewed publication of computer code
Computer simulation models are mathematical abstractions of physical systems. In the area of natural resources and agriculture, these physical systems encompass selected interacting processes in plants, soils, animals, or watersheds. These models are scientific products and have become important i...
Energy Technology Data Exchange (ETDEWEB)
Joshua J. Cogliati; Abderrafi M. Ougouag
2006-10-01
A comprehensive, high fidelity model for pebble flow has been developed and embodied in the PEBBLES computer code. In this paper, a description of the physical artifacts included in the model is presented and some results from using the computer code for predicting the features of pebble flow and packing in a realistic pebble bed reactor design are shown. The sensitivity of models to various physical parameters is also discussed.
Computation of Grobner basis for systematic encoding of generalized quasi-cyclic codes
Van, Vo Tam; Mita, Seiichi
2008-01-01
Generalized quasi-cyclic (GQC) codes form a wide and useful class of linear codes that includes thoroughly quasi-cyclic codes, finite geometry (FG) low density parity check (LDPC) codes, and Hermitian codes. Although it is known that the systematic encoding of GQC codes is equivalent to the division algorithm in the theory of Grobner basis of modules, there has been no algorithm that computes Grobner basis for all types of GQC codes. In this paper, we propose two algorithms to compute Grobner basis for GQC codes from their parity check matrices: echelon canonical form algorithm and transpose algorithm. Both algorithms require sufficiently small number of finite-field operations with the order of the third power of code-length. Each algorithm has its own characteristic; the first algorithm is composed of elementary methods, and the second algorithm is based on a novel formula and is faster than the first one for high-rate codes. Moreover, we show that a serial-in serial-out encoder architecture for FG LDPC cod...
Directory of Open Access Journals (Sweden)
Nakhaei Mohammad
2014-03-01
Full Text Available Knowledge of soil hydraulic and thermal properties is essential for studies involving the combined effects of soil temperature and water input on water flow and redistribution processes under field conditions. The objective of this study was to estimate the parameters characterizing these properties from a transient water flow and heat transport field experiment. Real-time sensors built by the authors were used to monitor soil temperatures at depths of 40, 80, 120, and 160 cm during a 10-hour long ring infiltration experiment. Water temperatures and cumulative infiltration from a single infiltration ring were monitored simultaneously. The soil hydraulic parameters (the saturated water content θ s, empirical shape parameters α and n, and the saturated hydraulic conductivity Ks and soil thermal conductivity parameters (coefficients b1, b2, and b3 in the thermal conductivity function were estimated from cumulative infiltration and temperature measurements by inversely solving a two-dimensional water flow and heat transport using HYDRUS-2D. Three scenarios with a different, sequentially decreasing number of optimized parameters were considered. In scenario 1, seven parameters (θ s, Ks, α, n, b1, b2, and b3 were included in the inverse problem. The results indicated that this scenario does not provide a unique solution. In scenario 2, six parameters (Ks, α, n, b1, b2, and b3 were included in the inverse problem. The results showed that this scenario also results in a non-unique solution. Only scenario 3, in which five parameters (α, n, b1, b2, and b3 were included in the inverse problem, provided a unique solution. The simulated soil temperatures and cumulative infiltration during the ring infiltration experiment compared reasonably well with their corresponding observed values.
Windtalking Computers: Frequency Normalization, Binary Coding Systems and Encryption
Zirkind, Givon
2009-01-01
The goal of this paper is to discuss the application of known techniques, knowledge and technology in a novel way, to encrypt computer and non-computer data. To-date most computers use base 2 and most encryption systems use ciphering and/or an encryption algorithm, to convert data into a secret message. The method of having the computer "speak another secret language" as used in human military secret communications has never been imitated. The author presents the theory and several possible implementations of a method for computers for secret communications analogous to human beings using a secret language or; speaking multiple languages. The kind of encryption scheme proposed significantly increases the complexity of and the effort needed for, decryption. As every methodology has its drawbacks, so too, the data of the proposed system has its drawbacks. It is not as compressed as base 2 would be. However, this is manageable and acceptable, if the goal is very strong encryption: At least two methods and their ...
SENDIN and SENTINEL: two computer codes to assess the effects of nuclear data changes
Energy Technology Data Exchange (ETDEWEB)
Marable, J. H.; Drischler, J. D.; Weisbin, C. R.
1977-07-01
A description is given of the computer code SENTINEL, which provides a simple means for finding the effects on calculated reactor and shielding performance parameters due to proposed changes in the cross section data base. This code uses predetermined detailed sensitivity coefficients in SENPRO format, which is described in Appendix A. Knowledge of details of the particular reactor and/or shielding assemblies is not required of the user. Also described is the computer code SENDIN, which converts unformatted (binary) sensitivity files to card image form and vice versa. This is useful for transferring sensitivity files from one installation to another.
TPASS: a gamma-ray spectrum analysis and isotope identification computer code
Energy Technology Data Exchange (ETDEWEB)
Dickens, J.K.
1981-03-01
The gamma-ray spectral data-reduction and analysis computer code TPASS is described. This computer code is used to analyze complex Ge(Li) gamma-ray spectra to obtain peak areas corrected for detector efficiencies, from which are determined gamma-ray yields. These yields are compared with an isotope gamma-ray data file to determine the contributions to the observed spectrum from decay of specific radionuclides. A complete FORTRAN listing of the code and a complex test case are given.
Development of a system of computer codes for severe accident analyses and its applications
Energy Technology Data Exchange (ETDEWEB)
Chang, Soon Hong; Cheon, Moon Heon; Cho, Nam jin; No, Hui Cheon; Chang, Hyeon Seop; Moon, Sang Kee; Park, Seok Jeong; Chung, Jee Hwan [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)
1991-12-15
The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in Nuclear Power Plants. This system of codes is necessary to conduct individual plant examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident resistance. The scope and contents of this study are as follows : development of a system of computer codes for severe accident analyses, development of severe accident management strategy.
Energy Technology Data Exchange (ETDEWEB)
Lottes, S.A.; Bojanowski, C.; Shen, J.; Xie, Z.; Zhai, Y. (Energy Systems); (Turner-Fairbank Highway Research Center)
2012-04-09
The computational fluid dynamics (CFD) and computational structural mechanics (CSM) focus areas at Argonne's Transportation Research and Analysis Computing Center (TRACC) initiated a project to support and compliment the experimental programs at the Turner-Fairbank Highway Research Center (TFHRC) with high performance computing based analysis capabilities in August 2010. The project was established with a new interagency agreement between the Department of Energy and the Department of Transportation to provide collaborative research, development, and benchmarking of advanced three-dimensional computational mechanics analysis methods to the aerodynamics and hydraulics laboratories at TFHRC for a period of five years, beginning in October 2010. The analysis methods employ well-benchmarked and supported commercial computational mechanics software. Computational mechanics encompasses the areas of Computational Fluid Dynamics (CFD), Computational Wind Engineering (CWE), Computational Structural Mechanics (CSM), and Computational Multiphysics Mechanics (CMM) applied in Fluid-Structure Interaction (FSI) problems. The major areas of focus of the project are wind and water effects on bridges - superstructure, deck, cables, and substructure (including soil), primarily during storms and flood events - and the risks that these loads pose to structural failure. For flood events at bridges, another major focus of the work is assessment of the risk to bridges caused by scour of stream and riverbed material away from the foundations of a bridge. Other areas of current research include modeling of flow through culverts to improve design allowing for fish passage, modeling of the salt spray transport into bridge girders to address suitability of using weathering steel in bridges, CFD analysis of the operation of the wind tunnel in the TFHRC wind engineering laboratory. This quarterly report documents technical progress on the project tasks for the period of October through
Energy Technology Data Exchange (ETDEWEB)
Lottes, S.A.; Bojanowski, C.; Shen, J.; Xie, Z.; Zhai, Y. (Energy Systems); (Turner-Fairbank Highway Research Center)
2012-06-28
The computational fluid dynamics (CFD) and computational structural mechanics (CSM) focus areas at Argonne's Transportation Research and Analysis Computing Center (TRACC) initiated a project to support and compliment the experimental programs at the Turner-Fairbank Highway Research Center (TFHRC) with high performance computing based analysis capabilities in August 2010. The project was established with a new interagency agreement between the Department of Energy and the Department of Transportation to provide collaborative research, development, and benchmarking of advanced three-dimensional computational mechanics analysis methods to the aerodynamics and hydraulics laboratories at TFHRC for a period of five years, beginning in October 2010. The analysis methods employ well benchmarked and supported commercial computational mechanics software. Computational mechanics encompasses the areas of Computational Fluid Dynamics (CFD), Computational Wind Engineering (CWE), Computational Structural Mechanics (CSM), and Computational Multiphysics Mechanics (CMM) applied in Fluid-Structure Interaction (FSI) problems. The major areas of focus of the project are wind and water effects on bridges - superstructure, deck, cables, and substructure (including soil), primarily during storms and flood events - and the risks that these loads pose to structural failure. For flood events at bridges, another major focus of the work is assessment of the risk to bridges caused by scour of stream and riverbed material away from the foundations of a bridge. Other areas of current research include modeling of flow through culverts to improve design allowing for fish passage, modeling of the salt spray transport into bridge girders to address suitability of using weathering steel in bridges, CFD analysis of the operation of the wind tunnel in the TFHRC wind engineering laboratory. This quarterly report documents technical progress on the project tasks for the period of January through
Energy Technology Data Exchange (ETDEWEB)
Lottes, S.A.; Kulak, R.F.; Bojanowski, C. (Energy Systems)
2011-12-09
The computational fluid dynamics (CFD) and computational structural mechanics (CSM) focus areas at Argonne's Transportation Research and Analysis Computing Center (TRACC) initiated a project to support and compliment the experimental programs at the Turner-Fairbank Highway Research Center (TFHRC) with high performance computing based analysis capabilities in August 2010. The project was established with a new interagency agreement between the Department of Energy and the Department of Transportation to provide collaborative research, development, and benchmarking of advanced three-dimensional computational mechanics analysis methods to the aerodynamics and hydraulics laboratories at TFHRC for a period of five years, beginning in October 2010. The analysis methods employ well-benchmarked and supported commercial computational mechanics software. Computational mechanics encompasses the areas of Computational Fluid Dynamics (CFD), Computational Wind Engineering (CWE), Computational Structural Mechanics (CSM), and Computational Multiphysics Mechanics (CMM) applied in Fluid-Structure Interaction (FSI) problems. The major areas of focus of the project are wind and water effects on bridges - superstructure, deck, cables, and substructure (including soil), primarily during storms and flood events - and the risks that these loads pose to structural failure. For flood events at bridges, another major focus of the work is assessment of the risk to bridges caused by scour of stream and riverbed material away from the foundations of a bridge. Other areas of current research include modeling of flow through culverts to assess them for fish passage, modeling of the salt spray transport into bridge girders to address suitability of using weathering steel in bridges, CFD analysis of the operation of the wind tunnel in the TFCHR wind engineering laboratory, vehicle stability under high wind loading, and the use of electromagnetic shock absorbers to improve vehicle stability
Energy Technology Data Exchange (ETDEWEB)
Lottes, S.A.; Kulak, R.F.; Bojanowski, C. (Energy Systems)
2011-08-26
The computational fluid dynamics (CFD) and computational structural mechanics (CSM) focus areas at Argonne's Transportation Research and Analysis Computing Center (TRACC) initiated a project to support and compliment the experimental programs at the Turner-Fairbank Highway Research Center (TFHRC) with high performance computing based analysis capabilities in August 2010. The project was established with a new interagency agreement between the Department of Energy and the Department of Transportation to provide collaborative research, development, and benchmarking of advanced three-dimensional computational mechanics analysis methods to the aerodynamics and hydraulics laboratories at TFHRC for a period of five years, beginning in October 2010. The analysis methods employ well-benchmarked and supported commercial computational mechanics software. Computational mechanics encompasses the areas of Computational Fluid Dynamics (CFD), Computational Wind Engineering (CWE), Computational Structural Mechanics (CSM), and Computational Multiphysics Mechanics (CMM) applied in Fluid-Structure Interaction (FSI) problems. The major areas of focus of the project are wind and water loads on bridges - superstructure, deck, cables, and substructure (including soil), primarily during storms and flood events - and the risks that these loads pose to structural failure. For flood events at bridges, another major focus of the work is assessment of the risk to bridges caused by scour of stream and riverbed material away from the foundations of a bridge. Other areas of current research include modeling of flow through culverts to assess them for fish passage, modeling of the salt spray transport into bridge girders to address suitability of using weathering steel in bridges, vehicle stability under high wind loading, and the use of electromagnetic shock absorbers to improve vehicle stability under high wind conditions. This quarterly report documents technical progress on the project
Energy Technology Data Exchange (ETDEWEB)
Studer, E.; Dabbene, F.; Magnaud, J. P.; Blumenfeld, L.; Quillico, J. J.; Paillere, H. [CEA Saclay, Gif-sur-Yvette (France)
2003-07-01
Twenty four years after the Three Mile Island Accident, Hydrogen risk remains a safety issue for current and future Pressurized Water Reactors (PWR). The formation of a combustible gas mixture in the complex geometry of a reactor containment depends on the understanding of hydrogen production, complex 3D flow due to gas/steam injection, natural convection, heat transfer by condensation on walls and effect of mitigation devices. Lumped parameter safety codes mainly developed for full containment analysis are not able to accurately predict the local gas mixing within the containment. 3D CFD codes are required but a thorough validation process on well-instrumented experimental data is necessary before they can be used with a high degree of confidence. The MISTRA coupled effect test facility has been recently built at CEA to fulfill these objectives: numerous measurement points in the gaseous volume (temperature and gas concentration) and the use of Laser technology (L.D.V. and P.I.V.) provide suitable experimental data for code validation. The in-house CEA-IRSN CAST3M/TONUS code is developed and validated against experimental data provided by this facility. Some of these tests have been proposed to the international community for code benchmarking (MICOCO benchmark and OECD/ISP47 exercise). Finally, extrapolation to global containment scale requires the validation of the code on more complex flow patterns and a detailed investigation of scaling effects. These two items will be the guidelines of future MISTRA tests.
Proceedings of the conference on computer codes and the linear accelerator community
Energy Technology Data Exchange (ETDEWEB)
Cooper, R.K. (comp.)
1990-07-01
The conference whose proceedings you are reading was envisioned as the second in a series, the first having been held in San Diego in January 1988. The intended participants were those people who are actively involved in writing and applying computer codes for the solution of problems related to the design and construction of linear accelerators. The first conference reviewed many of the codes both extant and under development. This second conference provided an opportunity to update the status of those codes, and to provide a forum in which emerging new 3D codes could be described and discussed. The afternoon poster session on the second day of the conference provided an opportunity for extended discussion. All in all, this conference was felt to be quite a useful interchange of ideas and developments in the field of 3D calculations, parallel computation, higher-order optics calculations, and code documentation and maintenance for the linear accelerator community. A third conference is planned.
Exact Gap Computation for Code Coverage Metrics in ISO-C
Richter, Dirk; 10.4204/EPTCS.80.4
2012-01-01
Test generation and test data selection are difficult tasks for model based testing. Tests for a program can be meld to a test suite. A lot of research is done to quantify the quality and improve a test suite. Code coverage metrics estimate the quality of a test suite. This quality is fine, if the code coverage value is high or 100%. Unfortunately it might be impossible to achieve 100% code coverage because of dead code for example. There is a gap between the feasible and theoretical maximal possible code coverage value. Our review of the research indicates, none of current research is concerned with exact gap computation. This paper presents a framework to compute such gaps exactly in an ISO-C compatible semantic and similar languages. We describe an efficient approximation of the gap in all the other cases. Thus, a tester can decide if more tests might be able or necessary to achieve better coverage.
Energy Technology Data Exchange (ETDEWEB)
Liu Ping, E-mail: ping.liu@areva.co [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), P.O. Box 3640, D-76021 Karlsruhe (Germany); Gabrielli, Fabrizio; Rineiski, Andrei; Maschek, Werner [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), P.O. Box 3640, D-76021 Karlsruhe (Germany); Bruna, Giovanni B. [Reactor Safety Division, French Institute for Radioprotection and Nuclear Safety (IRSN), B.P. 17, 92262 Fontenay aux Roses Cedex (France)
2010-10-15
SIMMER-III, a neutronics and thermal-hydraulics coupled code, was originally developed for core disruptive accident analyses of liquid metal cooled fast reactors. Due to its versatility in investigating scenarios of core disruption, the code has also been extended to the simulation of transients in thermal neutron systems such as the criticality accident at the JCO fuel fabrication plant, and, in recent years, applied to water-moderated thermal research reactor transient studies, too. Originally, SIMMER considered only cylindrical fuel pin geometry. Therefore, implementation of a plate-type fuel model to the SIMMER-III code is of importance for the analysis of research reactors adopting this kind of fuel. Furthermore, validation of the SIMMER-III modeling of light water-cooled thermal reactor reactivity initiated transients is of necessity. This paper presents the work carried out on the SIMMER-III code in the framework of a KIT and IRSN joint activity aimed at providing the code with experimental reactor transient study capabilities. The first step of the job was the implementation of a new fuel model in SIMMER-III. Verification on this new model indicates that it can well simulate the steady-state temperature profile in the fuel. Secondly, three cases with the shortest reactor periods of 5.0 ms, 4.6 ms and 3.2 ms among the Special Power Excursion Reactor Tests (SPERT) performed in the SPERT I D-12/25 facility have been simulated. Comparison of the results between the SIMMER-III simulation and the reported SPERT results indicates that although there is space for further improvement on the modeling of negative feedback mechanisms, with this plate-type fuel model SIMMER-III can well represent the transient phenomena of reactivity driven power excursion.
Visualization of elastic wavefields computed with a finite difference code
Energy Technology Data Exchange (ETDEWEB)
Larsen, S. [Lawrence Livermore National Lab., CA (United States); Harris, D.
1994-11-15
The authors have developed a finite difference elastic propagation model to simulate seismic wave propagation through geophysically complex regions. To facilitate debugging and to assist seismologists in interpreting the seismograms generated by the code, they have developed an X Windows interface that permits viewing of successive temporal snapshots of the (2D) wavefield as they are calculated. The authors present a brief video displaying the generation of seismic waves by an explosive source on a continent, which propagate to the edge of the continent then convert to two types of acoustic waves. This sample calculation was part of an effort to study the potential of offshore hydroacoustic systems to monitor seismic events occurring onshore.
Energy Technology Data Exchange (ETDEWEB)
Mejia S, D. M. [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col. Narvarte, 03020 Ciudad de Mexico (Mexico); Del Valle G, E., E-mail: dulcemaria.mejia@cnsns.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, 07738 Ciudad de Mexico (Mexico)
2015-09-15
The developed models for Parcs and Trace codes corresponding for the cycle 15 of the Unit 1 of the Laguna Verde nuclear power plant are described. The first focused to the neutronic simulation and the second to thermal hydraulics. The model developed for Parcs consists of a core of 444 fuel assemblies wrapped in a radial reflective layer and two layers, a superior and another inferior, of axial reflector. The core consists of 27 total axial planes. The model for Trace includes the vessel and its internal components as well as various safety systems. The coupling between the two codes is through two maps that allow its intercommunication. Both codes are used in coupled form performing a dynamic simulation that allows obtaining acceptably a stable state from which is carried out the closure of all the main steam isolation valves (MSIVs) followed by the performance of safety relief valves (SRVs) and ECCS. The results for the power and reactivities introduced by the moderator density, the fuel temperature and total temperature are shown. Data are also provided like: the behavior of the pressure in the steam dome, the water level in the downcomer, the flow through the MSIVs and SRVs. The results are explained for the power, the pressure in the steam dome and the water level in the downcomer which show agreement with the actions of the MSIVs, SRVs and ECCS. (Author)
Jin, Xuezhou; R, Meyder
2005-04-01
The HCPB concept has been a European DEMO reference concept for nearly one decade. Detailed thermal-hydraulic study on the control behavior of the whole system is one of the important parts of this development. The thermal-hydraulic effect of the TBM-combined cooling circuit during a cyclic operation in ITER has been studied using the system code RELAP5. The RELAP5 is based on an one-dimensional, transient two-fluid model for the flow of a two-phase steam-water mixture that can contain noncondensable components like Helium. The RELAP5-models are modified to take the cyclic operation of the circulator, heat exchanger, bypass, valves etc in to account. A sequence of operational phases is investigated, starting from the cold state through the heating phase that brings the system to a stand-by condition, followed by typical power cycles applied in ITER. The results show that the implemented control mechanisms keep the inlet temperature to the TBM and the total mass flow rate at the required values through all phases.
Institute of Scientific and Technical Information of China (English)
Jin Xuezhou; R. Meyder
2005-01-01
The HCPB concept has been a European DEMO reference concept for nearly one decade. Detailed thermal-hydraulic study on the control behavior of the whole system is one of the important parts of this development. The thermal-hydraulic effect of the TBM-combined cooling circuit during a cyclic operation in ITER has been studied using the system code RELAP5. The RELAP5 is based on an one-dimensional, transient two-fluid model for the flow of a two-phase steam-water mixture that can contain noncondensable components like Helium. The RELAP5-models are modified to take the cyclic operation of the circulator, heat exchanger, bypass, valves etc in to account. A sequence of operational phases is investigated, starting from the cold state through the heating phase that brings the system to a stand-by condition, followed by typical power cycles applied in ITER. The results show that the implemented control mechanisms keep the inlet temperature to the TBM and the total mass flow rate at the required values through all phases.
Introduction to error correcting codes in quantum computers
Salas, P J
2006-01-01
The goal of this paper is to review the theoretical basis for achieving a faithful quantum information transmission and processing in the presence of noise. Initially encoding and decoding, implementing gates and quantum error correction will be considered error free. Finally we will relax this non realistic assumption, introducing the quantum fault-tolerant concept. The existence of an error threshold permits to conclude that there is no physical law preventing a quantum computer from being built. An error model based on the depolarizing channel will be able to provide a simple estimation of the storage or memory computation error threshold: < 5.2 10-5. The encoding is made by means of the [[7,1,3
High-Performance Java Codes for Computational Fluid Dynamics
Riley, Christopher; Chatterjee, Siddhartha; Biswas, Rupak; Biegel, Bryan (Technical Monitor)
2001-01-01
The computational science community is reluctant to write large-scale computationally -intensive applications in Java due to concerns over Java's poor performance, despite the claimed software engineering advantages of its object-oriented features. Naive Java implementations of numerical algorithms can perform poorly compared to corresponding Fortran or C implementations. To achieve high performance, Java applications must be designed with good performance as a primary goal. This paper presents the object-oriented design and implementation of two real-world applications from the field of Computational Fluid Dynamics (CFD): a finite-volume fluid flow solver (LAURA, from NASA Langley Research Center), and an unstructured mesh adaptation algorithm (2D_TAG, from NASA Ames Research Center). This work builds on our previous experience with the design of high-performance numerical libraries in Java. We examine the performance of the applications using the currently available Java infrastructure and show that the Java version of the flow solver LAURA performs almost within a factor of 2 of the original procedural version. Our Java version of the mesh adaptation algorithm 2D_TAG performs within a factor of 1.5 of its original procedural version on certain platforms. Our results demonstrate that object-oriented software design principles are not necessarily inimical to high performance.
Compendium of computer codes for the safety analysis of fast breeder reactors
Energy Technology Data Exchange (ETDEWEB)
1977-10-01
The objective of the compendium is to provide the reader with a guide which briefly describes many of the computer codes used for liquid metal fast breeder reactor safety analyses, since it is for this system that most of the codes have been developed. The compendium is designed to address the following frequently asked questions from individuals in licensing and research and development activities: (1) What does the code do. (2) To what safety problems has it been applied. (3) What are the code's limitations. (4) What is being done to remove these limitations. (5) How does the code compare with experimental observations and other code predictions. (6) What reference documents are available.
Performance evaluation of moment-method codes on an Intel iPSC/860 hypercube computer
Energy Technology Data Exchange (ETDEWEB)
Klimkowski, K.; Ling, H. (Texas Univ., Austin (United States))
1993-09-01
An analytical evaluation is conducted of the performance of a moment-method code on a parallel computer, treating algorithmic complexity costs within the framework of matrix size and the 'subblock-size' matrix-partitioning parameter. A scaled-efficiencies analysis is conducted for the measured computation times of the matrix-fill operation and LU decomposition. 6 refs.
Computing the Feng-Rao distances for codes from order domains
DEFF Research Database (Denmark)
Ruano Benito, Diego
2007-01-01
We compute the Feng–Rao distance of a code coming from an order domain with a simplicial value semigroup. The main tool is the Apéry set of a semigroup that can be computed using a Gröbner basis....
Development of MCNPX-ESUT computer code for simulation of neutron/gamma pulse height distribution
Abolfazl Hosseini, Seyed; Vosoughi, Naser; Zangian, Mehdi
2015-05-01
In this paper, the development of the MCNPX-ESUT (MCNPX-Energy Engineering of Sharif University of Technology) computer code for simulation of neutron/gamma pulse height distribution is reported. Since liquid organic scintillators like NE-213 are well suited and routinely used for spectrometry in mixed neutron/gamma fields, this type of detectors is selected for simulation in the present study. The proposed algorithm for simulation includes four main steps. The first step is the modeling of the neutron/gamma particle transport and their interactions with the materials in the environment and detector volume. In the second step, the number of scintillation photons due to charged particles such as electrons, alphas, protons and carbon nuclei in the scintillator material is calculated. In the third step, the transport of scintillation photons in the scintillator and lightguide is simulated. Finally, the resolution corresponding to the experiment is considered in the last step of the simulation. Unlike the similar computer codes like SCINFUL, NRESP7 and PHRESP, the developed computer code is applicable to both neutron and gamma sources. Hence, the discrimination of neutron and gamma in the mixed fields may be performed using the MCNPX-ESUT computer code. The main feature of MCNPX-ESUT computer code is that the neutron/gamma pulse height simulation may be performed without needing any sort of post processing. In the present study, the pulse height distributions due to a monoenergetic neutron/gamma source in NE-213 detector using MCNPX-ESUT computer code is simulated. The simulated neutron pulse height distributions are validated through comparing with experimental data (Gohil et al. Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 664 (2012) 304-309.) and the results obtained from similar computer codes like SCINFUL, NRESP7 and Geant4. The simulated gamma pulse height distribution for a 137Cs
Robust Coding for Lossy Computing with Observation Costs
Ahmadi, Behzad
2011-01-01
An encoder wishes to minimize the bit rate necessary to guarantee that a decoder is able to calculate a symbol-wise function of a sequence available only at the encoder and a sequence that can be measured only at the decoder. This classical problem, first studied by Yamamoto, is addressed here by including two new aspects: (i) The decoder obtains noisy measurements of its sequence, where the quality of such measurements can be controlled via a cost-constrained "action" sequence, which is taken at the decoder or at the encoder; (ii) Measurement at the decoder may fail in a way that is unpredictable to the encoder, thus requiring robust encoding. The considered scenario generalizes known settings such as the Heegard-Berger-Kaspi and the "source coding with a vending machine" problems. The rate-distortion-cost function is derived in relevant special cases, along with general upper and lower bounds. Numerical examples are also worked out to obtain further insight into the optimal system design.
Automatic Parallelization Tool: Classification of Program Code for Parallel Computing
Directory of Open Access Journals (Sweden)
Mustafa Basthikodi
2016-04-01
Full Text Available Performance growth of single-core processors has come to a halt in the past decade, but was re-enabled by the introduction of parallelism in processors. Multicore frameworks along with Graphical Processing Units empowered to enhance parallelism broadly. Couples of compilers are updated to developing challenges forsynchronization and threading issues. Appropriate program and algorithm classifications will have advantage to a great extent to the group of software engineers to get opportunities for effective parallelization. In present work we investigated current species for classification of algorithms, in that related work on classification is discussed along with the comparison of issues that challenges the classification. The set of algorithms are chosen which matches the structure with different issues and perform given task. We have tested these algorithms utilizing existing automatic species extraction toolsalong with Bones compiler. We have added functionalities to existing tool, providing a more detailed characterization. The contributions of our work include support for pointer arithmetic, conditional and incremental statements, user defined types, constants and mathematical functions. With this, we can retain significant data which is not captured by original speciesof algorithms. We executed new theories into the device, empowering automatic characterization of program code.
The Uncertainty Test for the MAAP Computer Code
Energy Technology Data Exchange (ETDEWEB)
Park, S. H.; Song, Y. M.; Park, S. Y.; Ahn, K. I.; Kim, K. R.; Lee, Y. J. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2008-10-15
After the Three Mile Island Unit 2 (TMI-2) and Chernobyl accidents, safety issues for a severe accident are treated in various aspects. Major issues in our research part include a level 2 PSA. The difficulty in expanding the level 2 PSA as a risk information activity is the uncertainty. In former days, it attached a weight to improve the quality in a internal accident PSA, but the effort is insufficient for decrease the phenomenon uncertainty in the level 2 PSA. In our country, the uncertainty degree is high in the case of a level 2 PSA model, and it is necessary to secure a model to decrease the uncertainty. We have not yet experienced the uncertainty assessment technology, the assessment system itself depends on advanced nations. In advanced nations, the severe accident simulator is implemented in the hardware level. But in our case, basic function in a software level can be implemented. In these circumstance at home and abroad, similar instances are surveyed such as UQM and MELCOR. Referred to these instances, SAUNA (Severe Accident UNcertainty Analysis) system is being developed in our project to assess and decrease the uncertainty in a level 2 PSA. It selects the MAAP code to analyze the uncertainty in a severe accident.
Skála, J.; Baruffa, F.; Büchner, J.; Rampp, M.
2015-08-01
Context. The numerical simulation of turbulence and flows in almost ideal astrophysical plasmas with large Reynolds numbers motivates the implementation of magnetohydrodynamical (MHD) computer codes with low resistivity. They need to be computationally efficient and scale well with large numbers of CPU cores, allow obtaining a high grid resolution over large simulation domains, and be easily and modularly extensible, for instance, to new initial and boundary conditions. Aims: Our aims are the implementation, optimization, and verification of a computationally efficient, highly scalable, and easily extensible low-dissipative MHD simulation code for the numerical investigation of the dynamics of astrophysical plasmas with large Reynolds numbers in three dimensions (3D). Methods: The new GOEMHD3 code discretizes the ideal part of the MHD equations using a fast and efficient leap-frog scheme that is second-order accurate in space and time and whose initial and boundary conditions can easily be modified. For the investigation of diffusive and dissipative processes the corresponding terms are discretized by a DuFort-Frankel scheme. To always fulfill the Courant-Friedrichs-Lewy stability criterion, the time step of the code is adapted dynamically. Numerically induced local oscillations are suppressed by explicit, externally controlled diffusion terms. Non-equidistant grids are implemented, which enhance the spatial resolution, where needed. GOEMHD3 is parallelized based on the hybrid MPI-OpenMP programing paradigm, adopting a standard two-dimensional domain-decomposition approach. Results: The ideal part of the equation solver is verified by performing numerical tests of the evolution of the well-understood Kelvin-Helmholtz instability and of Orszag-Tang vortices. The accuracy of solving the (resistive) induction equation is tested by simulating the decay of a cylindrical current column. Furthermore, we show that the computational performance of the code scales very
Development of a model and computer code to describe solar grade silicon production processes
Gould, R. K.; Srivastava, R.
1979-01-01
Two computer codes were developed for describing flow reactors in which high purity, solar grade silicon is produced via reduction of gaseous silicon halides. The first is the CHEMPART code, an axisymmetric, marching code which treats two phase flows with models describing detailed gas-phase chemical kinetics, particle formation, and particle growth. It can be used to described flow reactors in which reactants, mix, react, and form a particulate phase. Detailed radial gas-phase composition, temperature, velocity, and particle size distribution profiles are computed. Also, deposition of heat, momentum, and mass (either particulate or vapor) on reactor walls is described. The second code is a modified version of the GENMIX boundary layer code which is used to compute rates of heat, momentum, and mass transfer to the reactor walls. This code lacks the detailed chemical kinetics and particle handling features of the CHEMPART code but has the virtue of running much more rapidly than CHEMPART, while treating the phenomena occurring in the boundary layer in more detail.
Issues in computational fluid dynamics code verification and validation
Energy Technology Data Exchange (ETDEWEB)
Oberkampf, W.L.; Blottner, F.G.
1997-09-01
A broad range of mathematical modeling errors of fluid flow physics and numerical approximation errors are addressed in computational fluid dynamics (CFD). It is strongly believed that if CFD is to have a major impact on the design of engineering hardware and flight systems, the level of confidence in complex simulations must substantially improve. To better understand the present limitations of CFD simulations, a wide variety of physical modeling, discretization, and solution errors are identified and discussed. Here, discretization and solution errors refer to all errors caused by conversion of the original partial differential, or integral, conservation equations representing the physical process, to algebraic equations and their solution on a computer. The impact of boundary conditions on the solution of the partial differential equations and their discrete representation will also be discussed. Throughout the article, clear distinctions are made between the analytical mathematical models of fluid dynamics and the numerical models. Lax`s Equivalence Theorem and its frailties in practical CFD solutions are pointed out. Distinctions are also made between the existence and uniqueness of solutions to the partial differential equations as opposed to the discrete equations. Two techniques are briefly discussed for the detection and quantification of certain types of discretization and grid resolution errors.
Computer code simulations of explosions in flow networks and comparison with experiments
Gregory, W. S.; Nichols, B. D.; Moore, J. A.; Smith, P. R.; Steinke, R. G.; Idzorek, R. D.
1987-10-01
A program of experimental testing and computer code development for predicting the effects of explosions in air-cleaning systems is being carried out for the Department of Energy. This work is a combined effort by the Los Alamos National Laboratory and New Mexico State University (NMSU). Los Alamos has the lead responsibility in the project and develops the computer codes; NMSU performs the experimental testing. The emphasis in the program is on obtaining experimental data to verify the analytical work. The primary benefit of this work will be the development of a verified computer code that safety analysts can use to analyze the effects of hypothetical explosions in nuclear plant air cleaning systems. The experimental data show the combined effects of explosions in air-cleaning systems that contain all of the important air-cleaning elements (blowers, dampers, filters, ductwork, and cells). A small experimental set-up consisting of multiple rooms, ductwork, a damper, a filter, and a blower was constructed. Explosions were simulated with a shock tube, hydrogen/air-filled gas balloons, and blasting caps. Analytical predictions were made using the EVENT84 and NF85 computer codes. The EVENT84 code predictions were in good agreement with the effects of the hydrogen/air explosions, but they did not model the blasting cap explosions adequately. NF85 predicted shock entrance to and within the experimental set-up very well. The NF85 code was not used to model the hydrogen/air or blasting cap explosions.
Algorithms and computer codes for atomic and molecular quantum scattering theory
Energy Technology Data Exchange (ETDEWEB)
Thomas, L. (ed.)
1979-01-01
This workshop has succeeded in bringing up 11 different coupled equation codes on the NRCC computer, testing them against a set of 24 different test problems and making them available to the user community. These codes span a wide variety of methodologies, and factors of up to 300 were observed in the spread of computer times on specific problems. A very effective method was devised for examining the performance of the individual codes in the different regions of the integration range. Many of the strengths and weaknesses of the codes have been identified. Based on these observations, a hybrid code has been developed which is significantly superior to any single code tested. Thus, not only have the original goals been fully met, the workshop has resulted directly in an advancement of the field. All of the computer programs except VIVS are available upon request from the NRCC. Since an improved version of VIVS is contained in the hybrid program, VIVAS, it was not made available for distribution. The individual program LOGD is, however, available. In addition, programs which compute the potential energy matrices of the test problems are also available. The software library names for Tests 1, 2 and 4 are HEH2, LICO, and EN2, respectively.
Energy Technology Data Exchange (ETDEWEB)
Carbajo, Juan (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin, Madison, WI); Schmidt, Rodney Cannon; Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Ludewig, Hans (Brookhaven National Laboratory, Upton, NY); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache %3CU%2B2013%3E CEA, France)
2011-06-01
This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the
Institute of Scientific and Technical Information of China (English)
无
2010-01-01
The eukaryotic genome contains varying numbers of non-coding RNA(ncRNA) genes."Computational RNomics" takes a multidisciplinary approach,like information science,to resolve the structure and function of ncRNAs.Here,we review the main issues in "Computational RNomics" of data storage and management,ncRNA gene identification and characterization,ncRNA target identification and functional prediction,and we summarize the main methods and current content of "computational RNomics".
[Vascular assessment in stroke codes: role of computed tomography angiography].
Mendigaña Ramos, M; Cabada Giadas, T
2015-01-01
Advances in imaging studies for acute ischemic stroke are largely due to the development of new efficacious treatments carried out in the acute phase. Together with computed tomography (CT) perfusion studies, CT angiography facilitates the selection of patients who are likely to benefit from appropriate early treatment. CT angiography plays an important role in the workup for acute ischemic stroke because it makes it possible to confirm vascular occlusion, assess the collateral circulation, and obtain an arterial map that is very useful for planning endovascular treatment. In this review about CT angiography, we discuss the main technical characteristics, emphasizing the usefulness of the technique in making the right diagnosis and improving treatment strategies. Copyright © 2012 SERAM. Published by Elsevier España, S.L.U. All rights reserved.
Symbolic coding for noninvertible systems: uniform approximation and numerical computation
Beyn, Wolf-Jürgen; Hüls, Thorsten; Schenke, Andre
2016-11-01
It is well known that the homoclinic theorem, which conjugates a map near a transversal homoclinic orbit to a Bernoulli subshift, extends from invertible to specific noninvertible dynamical systems. In this paper, we provide a unifying approach that combines such a result with a fully discrete analog of the conjugacy for finite but sufficiently long orbit segments. The underlying idea is to solve appropriate discrete boundary value problems in both cases, and to use the theory of exponential dichotomies to control the errors. This leads to a numerical approach that allows us to compute the conjugacy to any prescribed accuracy. The method is demonstrated for several examples where invertibility of the map fails in different ways.
Benchmark Problems Used to Assess Computational Aeroacoustics Codes
Dahl, Milo D.; Envia, Edmane
2005-01-01
The field of computational aeroacoustics (CAA) encompasses numerical techniques for calculating all aspects of sound generation and propagation in air directly from fundamental governing equations. Aeroacoustic problems typically involve flow-generated noise, with and without the presence of a solid surface, and the propagation of the sound to a receiver far away from the noise source. It is a challenge to obtain accurate numerical solutions to these problems. The NASA Glenn Research Center has been at the forefront in developing and promoting the development of CAA techniques and methodologies for computing the noise generated by aircraft propulsion systems. To assess the technological advancement of CAA, Glenn, in cooperation with the Ohio Aerospace Institute and the AeroAcoustics Research Consortium, organized and hosted the Fourth CAA Workshop on Benchmark Problems. Participants from industry and academia from both the United States and abroad joined to present and discuss solutions to benchmark problems. These demonstrated technical progress ranging from the basic challenges to accurate CAA calculations to the solution of CAA problems of increasing complexity and difficulty. The results are documented in the proceedings of the workshop. Problems were solved in five categories. In three of the five categories, exact solutions were available for comparison with CAA results. A fourth category of problems representing sound generation from either a single airfoil or a blade row interacting with a gust (i.e., problems relevant to fan noise) had approximate analytical or completely numerical solutions. The fifth category of problems involved sound generation in a viscous flow. In this case, the CAA results were compared with experimental data.
Energy Technology Data Exchange (ETDEWEB)
Bordy, J.M.; Kodeli, I.; Menard, St.; Bouchet, J.L.; Renard, F.; Martin, E.; Blazy, L.; Voros, S.; Bochud, F.; Laedermann, J.P.; Beaugelin, K.; Makovicka, L.; Quiot, A.; Vermeersch, F.; Roche, H.; Perrin, M.C.; Laye, F.; Bardies, M.; Struelens, L.; Vanhavere, F.; Gschwind, R.; Fernandez, F.; Quesne, B.; Fritsch, P.; Lamart, St.; Crovisier, Ph.; Leservot, A.; Antoni, R.; Huet, Ch.; Thiam, Ch.; Donadille, L.; Monfort, M.; Diop, Ch.; Ricard, M
2006-07-01
The purpose of this conference was to describe the present state of computer codes dedicated to radiation transport or radiation source assessment or dosimetry. The presentations have been parted into 2 sessions: 1) methodology and 2) uses in industrial or medical or research domains. It appears that 2 different calculation strategies are prevailing, both are based on preliminary Monte-Carlo calculations with data storage. First, quick simulations made from a database of particle histories built though a previous Monte-Carlo simulation and secondly, a neuronal approach involving a learning platform generated through a previous Monte-Carlo simulation. This document gathers the slides of the presentations.
Energy Technology Data Exchange (ETDEWEB)
Gomez Garcia-Torano, I.; Jimenez, G.
2013-07-01
The Thermo-hydraulic code GOTHIC is often used in the nuclear industry for licensing transient analysis inside containment of generation II (PWR, BWR) plants as Gen III and III + (AP1000, ESBWR, APWR). After entering the mass and energy released to the containment, previously calculated by other codes (basis, TRACE), GOTHIC allows to calculate in detail the evolution of basic parameters in the containment.
Energy Technology Data Exchange (ETDEWEB)
Hoffman, F. O.; Miller, C. W.; Shaeffer, D. L.; Garten, Jr., C. T.; Shor, R. W.; Ensminger, J. T.
1977-04-01
The objective of this paper is to present a compilation of computer codes for the assessment of accidental or routine releases of radioactivity to the environment from nuclear power facilities. The capabilities of 83 computer codes in the areas of environmental transport and radiation dosimetry are summarized in tabular form. This preliminary analysis clearly indicates that the initial efforts in assessment methodology development have concentrated on atmospheric dispersion, external dosimetry, and internal dosimetry via inhalation. The incorporation of terrestrial and aquatic food chain pathways has been a more recent development and reflects the current requirements of environmental legislation and the needs of regulatory agencies. The characteristics of the conceptual models employed by these codes are reviewed. The appendixes include abstracts of the codes and indexes by author, key words, publication description, and title.
Energy Technology Data Exchange (ETDEWEB)
Veloso, Marcelo Antonio
2003-07-01
PANTERA-2 (from Programa para Analise Termo-hidraulica de Reatores a Agua - Program for Thermal-hydraulic Analysis of Water Reactors, Version 2), whose fundamentals are described in this work, is intended to carry out rod bundle subchannel analysis in conjunction with multiloop simulation. It solves simultaneously the conservation equations of mass, axial and lateral momentum, and energy for subchannel geometry coupled with the balance equations that describe the fluid flows in any number of coolant loops connected to a pressure vessel containing the rod bundle. As far as subchannel analysis is concerned, the basic computational strategy of PANTERA-2 comes from COBRA codes, but an alternative implicit solution method oriented to the pressure field has been used to solve the finite difference approximations for the balance laws. The results provided by the subchannel model comprise the fluid density, enthalpy, flow rate, and pressure fields in the subchannels. The loop model predicts the individual loop flows, total flow through the pressure vessel, and pump rotational speeds as a function of time subsequent to the failure of any number of the coolant pumps. The flow transients in the loops may initiated by partial, total or sequential loss of electric power to the operating pumps. Transient events caused by either shaft break or rotor locking may also be simulated. The changes in rotational speed of the pumps as a function of rime are determined from a torque balance. Pump dynamic head and hydraulic torque are calculated as a function of rotational speed and volumetric flow from two polar homologous curves supplied to the code in the tabular form. In order to illustrate the analytical capability of PANTERA-2, three sample problems are presented and discussed. Comparisons between calculated and measured results indicate that the program reproduces with a good accuracy experimental data for subchannel exit temperatures and critical heat fluxes in 5x5 rod bundles. It
Compendium of computer codes for the researcher in magnetic fusion energy
Energy Technology Data Exchange (ETDEWEB)
Porter, G.D. (ed.)
1989-03-10
This is a compendium of computer codes, which are available to the fusion researcher. It is intended to be a document that permits a quick evaluation of the tools available to the experimenter who wants to both analyze his data, and compare the results of his analysis with the predictions of available theories. This document will be updated frequently to maintain its usefulness. I would appreciate receiving further information about codes not included here from anyone who has used them. The information required includes a brief description of the code (including any special features), a bibliography of the documentation available for the code and/or the underlying physics, a list of people to contact for help in running the code, instructions on how to access the code, and a description of the output from the code. Wherever possible, the code contacts should include people from each of the fusion facilities so that the novice can talk to someone ''down the hall'' when he first tries to use a code. I would also appreciate any comments about possible additions and improvements in the index. I encourage any additional criticism of this document. 137 refs.
Yang, Xiao-Chen; Zhang, Yan; Gui, Xing-Min; Hu, Sheng-Shou
2011-10-01
The advent of various technologies has allowed mechanical blood pumps to become more reliable and versatile in recent decades. In our study group, a novel structure of axial flow blood pump was developed for assisting the left ventricle. The design point of the left ventricular assist blood pump 25 (LAP-25) was chosen at 4 Lpm with 100 mm Hg according to our clinical practice. Computational fluid dynamics was used to design and analyze the performance of the LAP-25. In order to obtain a required hydraulic performance and a satisfactory hemolytic property in the LAP-25 of a smaller size, a novel structure was developed including an integrated shroud impeller, a streamlined impeller hub, and main impeller blades with splitter blades; furthermore, tandem cascades were introduced in designing the diffuser. The results of numerical simulation show the LAP-25 can generate flow rates of 3-5 Lpm at rotational speeds of 8500-10,500 rpm, producing pressure rises of 27.5-148.3 mm Hg with hydraulic efficiency points ranging from 13.4 to 27.5%. Moreover, the fluid field and the hemolytic property of the LAP-25 were estimated, and the mean hemolysis index of the pump was 0.0895% with Heuser's estimated model. In conclusion, the design of the LAP-25 shows an acceptable result.
Arvand, Arash; Hahn, Nicole; Hormes, Marcus; Akdis, Mustafa; Martin, Michael; Reul, Helmut
2004-10-01
A mixed-flow blood pump for long-term applications has been developed at the Helmholtz-Institute in Aachen, Germany. Central features of this implantable pump are a centrally integrated motor, a blood-immersed mechanical bearing, magnetic coupling of the impeller, and a shrouded impeller, which allows a relatively wide clearance. The aim of the study was a numerical analysis of hydraulic and hemolytic properties of different impeller design configurations. In vitro testing and numerical simulation techniques (computational fluid dynamics [CFD]) were applied to achieve a comprehensive overview. Pressure-flow charts were experimentally measured in a mock loop in order to validate the CFD data. In vitro hemolysis tests were performed at the main operating point of each impeller design. General flow patterns, pressure-flow charts, secondary flow rates, torque, and axial forces on the impeller were calculated by means of CFD. Furthermore, based on streak line techniques, shear stress (stress loading), exposure times, and volume percentage with critical stress loading have been determined. Comparison of CFD data with pressure head measurements showed excel-lent agreement. Also, impressive trend conformity was observed between in-vitro hemolysis results and numerical data. Comparison of design variations yielded clear trends and results. Design C revealed the best hydraulic and hemolytic properties and was chosen as the final design for the mixed-flow rotary blood pump.
Energy Technology Data Exchange (ETDEWEB)
Chung, B. D.; Bae, S. W.; Jeong, J. J.; Lee, S. M
2005-04-15
A new multi-dimensional component has been developed to allow for more flexible 3D capabilities in the system code, MARS. This component can be applied in the Cartesian and cylindrical coordinates. For the development of this model, the 3D convection and diffusion terms are implemented in the momentum and energy equation. And a simple Prandtl's mixing length model is applied for the turbulent viscosity. The developed multi-dimensional component was assessed against five conceptual problems with analytic solution. And some SETs are calculated and compared with experimental data. With this newly developed multi-dimensional flow module, the MARS code can realistic calculate the flow fields in pools such as those occurring in the core, steam generators and IRWST.
Institute of Scientific and Technical Information of China (English)
无
2006-01-01
The technology of passive safety is the current trend among safety systems in nuclear power plant. Passive residual heat removal system (PRHRS), a major part of passive safety systems of Chinese advanced PWR, is a novel design with three-fold natural circulation. On the basis of reasonable physics and mathematics models, MITAP-PRHRS code was developed to analyze steady and transient characteristics of the PRHRS. The calculation and analysis show that the code simulates steady characteristics of the PRHRS very well, and it is able to simulate transient characteristics of all startup modes of the PRHRS. However, the quantitative description is poor during the initial stages of the transition process when water hammer occurs.
POPCYCLE: a computer code for calculating nuclear and fossil plant levelized life-cycle power costs
Energy Technology Data Exchange (ETDEWEB)
Hardie, R.W.
1982-02-01
POPCYCLE, a computer code designed to calculate levelized life-cycle power costs for nuclear and fossil electrical generating plants is described. Included are (1) derivations of the equations and a discussion of the methodology used by POPCYCLE, (2) a description of the input required by the code, (3) a listing of the input for a sample case, and (4) the output for a sample case.
Fault-tolerant quantum computation with asymmetric Bacon-Shor codes
Brooks, Peter; Preskill, John
2013-03-01
We develop a scheme for fault-tolerant quantum computation based on asymmetric Bacon-Shor codes, which works effectively against highly biased noise dominated by dephasing. We find the optimal Bacon-Shor block size as a function of the noise strength and the noise bias, and estimate the logical error rate and overhead cost achieved by this optimal code. Our fault-tolerant gadgets, based on gate teleportation, are well suited for hardware platforms with geometrically local gates in two dimensions.
HIFI: a computer code for projectile fragmentation accompanied by incomplete fusion
Energy Technology Data Exchange (ETDEWEB)
Wu, J.R.
1980-07-01
A brief summary of a model proposed to describe projectile fragmentation accompanied by incomplete fusion and the instructions for the use of the computer code HIFI are given. The code HIFI calculates single inclusive spectra, coincident spectra and excitation functions resulting from particle-induced reactions. It is a multipurpose program which can calculate any type of coincident spectra as long as the reaction is assumed to take place in two steps.
SAMDIST: A Computer Code for Calculating Statistical Distributions for R-Matrix Resonance Parameters
Energy Technology Data Exchange (ETDEWEB)
Leal, L.C.
1995-01-01
The: SAMDIST computer code has been developed to calculate distribution of resonance parameters of the Reich-Moore R-matrix type. The program assumes the parameters are in the format compatible with that of the multilevel R-matrix code SAMMY. SAMDIST calculates the energy-level spacing distribution, the resonance width distribution, and the long-range correlation of the energy levels. Results of these calculations are presented in both graphic and tabular forms.
SAMDIST A Computer Code for Calculating Statistical Distributions for R-Matrix Resonance Parameters
Leal, L C
1995-01-01
The: SAMDIST computer code has been developed to calculate distribution of resonance parameters of the Reich-Moore R-matrix type. The program assumes the parameters are in the format compatible with that of the multilevel R-matrix code SAMMY. SAMDIST calculates the energy-level spacing distribution, the resonance width distribution, and the long-range correlation of the energy levels. Results of these calculations are presented in both graphic and tabular forms.
The development of an intelligent interface to a computational fluid dynamics flow-solver code
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.
ANL/HTP: a computer code for the simulation of heat pipe operation
Energy Technology Data Exchange (ETDEWEB)
McLennan, G.A.
1983-11-01
ANL/HTP is a computer code for the simulation of heat pipe operation, to predict heat pipe performance and temperature distributions during steady state operation. Source and sink temperatures and heat transfer coefficients can be set as input boundary conditions, and varied for parametric studies. Five code options are included to calculate performance for fixed operating conditions, or to vary any one of the four boundary conditions to determine the heat pipe limited performance. The performance limits included are viscous, sonic, entrainment capillary, and boiling, using the best available theories to model these effects. The code has built-in models for a number of wick configurations - open grooves, screen-covered grooves, screen-wrap, and arteries, with provision for expansion. The current version of the code includes the thermophysical properties of sodium as the working fluid in an expandable subroutine. The code-calculated performance agrees quite well with measured experiment data.
LEADS-DC: A computer code for intense dc beam nonlinear transport simulation
Institute of Scientific and Technical Information of China (English)
无
2011-01-01
An intense dc beam nonlinear transport code has been developed. The code is written in Visual FORTRAN 6.6 and has ~13000 lines. The particle distribution in the transverse cross section is uniform or Gaussian. The space charge forces are calculated by the PIC (particle in cell) scheme, and the effects of the applied fields on the particle motion are calculated with the Lie algebraic method through the third order approximation. Obviously,the solutions to the equations of particle motion are self-consistent. The results obtained from the theoretical analysis have been put in the computer code. Many optical beam elements are contained in the code. So, the code can simulate the intense dc particle motions in the beam transport lines, high voltage dc accelerators and ion implanters.
Energy Technology Data Exchange (ETDEWEB)
Williams, P. T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dickson, T. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yin, S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
2007-12-01
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 the 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.
SCALE: A modular code system for performing standardized computer analyses for licensing evaluation
Energy Technology Data Exchange (ETDEWEB)
NONE
1997-03-01
This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U.S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This manual covers an array of modules written for the SCALE package, consisting of drivers, system libraries, cross section and materials properties libraries, input/output routines, storage modules, and help files.
ASHMET: a computer code for estimating insolation incident on tilted surfaces
Energy Technology Data Exchange (ETDEWEB)
Elkin, R.F.; Toelle, R.G.
1980-05-01
A computer code, ASHMET, has been developed by MSFC to estimate the amount of solar insolation incident on the surfaces of solar collectors. Both tracking and fixed-position collectors have been included. Climatological data for 248 US locations are built into the code. This report describes the methodology of the code, and its input and output. The basic methodology used by ASHMET is the ASHRAE clear-day insolation relationships modified by a clearness index derived from SOLMET-measured solar radiation data to a horizontal surface.
Solution of 3-dimensional time-dependent viscous flows. Part 2: Development of the computer code
Weinberg, B. C.; Mcdonald, H.
1980-01-01
There is considerable interest in developing a numerical scheme for solving the time dependent viscous compressible three dimensional flow equations to aid in the design of helicopter rotors. The development of a computer code to solve a three dimensional unsteady approximate form of the Navier-Stokes equations employing a linearized block emplicit technique in conjunction with a QR operator scheme is described. Results of calculations of several Cartesian test cases are presented. The computer code can be applied to more complex flow fields such as these encountered on rotating airfoils.
Directory of Open Access Journals (Sweden)
J. A. P. Pollacco
2017-06-01
Full Text Available Descriptions of soil hydraulic properties, such as the soil moisture retention curve, θ(h, and saturated hydraulic conductivities, Ks, are a prerequisite for hydrological models. Since the measurement of Ks is expensive, it is frequently derived from statistical pedotransfer functions (PTFs. Because it is usually more difficult to describe Ks than θ(h from pedotransfer functions, Pollacco et al. (2013 developed a physical unimodal model to compute Ks solely from hydraulic parameters derived from the Kosugi θ(h. This unimodal Ks model, which is based on a unimodal Kosugi soil pore-size distribution, was developed by combining the approach of Hagen–Poiseuille with Darcy's law and by introducing three tortuosity parameters. We report here on (1 the suitability of the Pollacco unimodal Ks model to predict Ks for a range of New Zealand soils from the New Zealand soil database (S-map and (2 further adaptations to this model to adapt it to dual-porosity structured soils by computing the soil water flux through a continuous function of an improved bimodal pore-size distribution. The improved bimodal Ks model was tested with a New Zealand data set derived from historical measurements of Ks and θ(h for a range of soils derived from sandstone and siltstone. The Ks data were collected using a small core size of 10 cm diameter, causing large uncertainty in replicate measurements. Predictions of Ks were further improved by distinguishing topsoils from subsoil. Nevertheless, as expected, stratifying the data with soil texture only slightly improved the predictions of the physical Ks models because the Ks model is based on pore-size distribution and the calibrated parameters were obtained within the physically feasible range. The improvements made to the unimodal Ks model by using the new bimodal Ks model are modest when compared to the unimodal model, which is explained by the poor accuracy of measured total porosity. Nevertheless, the new bimodal
Pollacco, Joseph Alexander Paul; Webb, Trevor; McNeill, Stephen; Hu, Wei; Carrick, Sam; Hewitt, Allan; Lilburne, Linda
2017-06-01
Descriptions of soil hydraulic properties, such as the soil moisture retention curve, θ(h), and saturated hydraulic conductivities, Ks, are a prerequisite for hydrological models. Since the measurement of Ks is expensive, it is frequently derived from statistical pedotransfer functions (PTFs). Because it is usually more difficult to describe Ks than θ(h) from pedotransfer functions, Pollacco et al. (2013) developed a physical unimodal model to compute Ks solely from hydraulic parameters derived from the Kosugi θ(h). This unimodal Ks model, which is based on a unimodal Kosugi soil pore-size distribution, was developed by combining the approach of Hagen-Poiseuille with Darcy's law and by introducing three tortuosity parameters. We report here on (1) the suitability of the Pollacco unimodal Ks model to predict Ks for a range of New Zealand soils from the New Zealand soil database (S-map) and (2) further adaptations to this model to adapt it to dual-porosity structured soils by computing the soil water flux through a continuous function of an improved bimodal pore-size distribution. The improved bimodal Ks model was tested with a New Zealand data set derived from historical measurements of Ks and θ(h) for a range of soils derived from sandstone and siltstone. The Ks data were collected using a small core size of 10 cm diameter, causing large uncertainty in replicate measurements. Predictions of Ks were further improved by distinguishing topsoils from subsoil. Nevertheless, as expected, stratifying the data with soil texture only slightly improved the predictions of the physical Ks models because the Ks model is based on pore-size distribution and the calibrated parameters were obtained within the physically feasible range. The improvements made to the unimodal Ks model by using the new bimodal Ks model are modest when compared to the unimodal model, which is explained by the poor accuracy of measured total porosity. Nevertheless, the new bimodal model provides an
Aeschliman, D. P.; Oberkampf, W. L.; Blottner, F. G.
Verification, calibration, and validation (VCV) of Computational Fluid Dynamics (CFD) codes is an essential element of the code development process. The exact manner in which code VCV activities are planned and conducted, however, is critically important. It is suggested that the way in which code validation, in particular, is often conducted--by comparison to published experimental data obtained for other purposes--is in general difficult and unsatisfactory, and that a different approach is required. This paper describes a proposed methodology for CFD code VCV that meets the technical requirements and is philosophically consistent with code development needs. The proposed methodology stresses teamwork and cooperation between code developers and experimentalists throughout the VCV process, and takes advantage of certain synergisms between CFD and experiment. A novel approach to uncertainty analysis is described which can both distinguish between and quantify various types of experimental error, and whose attributes are used to help define an appropriate experimental design for code VCV experiments. The methodology is demonstrated with an example of laminar, hypersonic, near perfect gas, 3-dimensional flow over a sliced sphere/cone of varying geometrical complexity.
NADAC and MERGE: computer codes for processing neutron activation analysis data
Energy Technology Data Exchange (ETDEWEB)
Heft, R.E.; Martin, W.E.
1977-05-19
Absolute disintegration rates of specific radioactive products induced by neutron irradition of a sample are determined by spectrometric analysis of gamma-ray emissions. Nuclide identification and quantification is carried out by a complex computer code GAMANAL (described elsewhere). The output of GAMANAL is processed by NADAC, a computer code that converts the data on observed distintegration rates to data on the elemental composition of the original sample. Computations by NADAC are on an absolute basis in that stored nuclear parameters are used rather than the difference between the observed disintegration rate and the rate obtained by concurrent irradiation of elemental standards. The NADAC code provides for the computation of complex cases including those involving interrupted irradiations, parent and daughter decay situations where the daughter may also be produced independently, nuclides with very short half-lives compared to counting interval, and those involving interference by competing neutron-induced reactions. The NADAC output consists of a printed report, which summarizes analytical results, and a card-image file, which can be used as input to another computer code MERGE. The purpose of MERGE is to combine the results of multiple analyses and produce a single final answer, based on all available information, for each element found.
Energy Technology Data Exchange (ETDEWEB)
Baes, C.F. III; Sharp, R.D.; Sjoreen, A.L.; Hermann, O.W.
1984-11-01
TERRA is a computer code which calculates concentrations of radionuclides and ingrowing daughters in surface and root-zone soil, produce and feed, beef, and milk from a given deposition rate at any location in the conterminous United States. The code is fully integrated with seven other computer codes which together comprise a Computerized Radiological Risk Investigation System, CRRIS. Output from either the long range (> 100 km) atmospheric dispersion code RETADD-II or the short range (<80 km) atmospheric dispersion code ANEMOS, in the form of radionuclide air concentrations and ground deposition rates by downwind location, serves as input to TERRA. User-defined deposition rates and air concentrations may also be provided as input to TERRA through use of the PRIMUS computer code. The environmental concentrations of radionuclides predicted by TERRA serve as input to the ANDROS computer code which calculates population and individual intakes, exposures, doses, and risks. TERRA incorporates models to calculate uptake from soil and atmospheric deposition on four groups of produce for human consumption and four groups of livestock feeds. During the environmental transport simulation, intermediate calculations of interception fraction for leafy vegetables, produce directly exposed to atmospherically depositing material, pasture, hay, and silage are made based on location-specific estimates of standing crop biomass. Pasture productivity is estimated by a model which considers the number and types of cattle and sheep, pasture area, and annual production of other forages (hay and silage) at a given location. Calculations are made of the fraction of grain imported from outside the assessment area. TERRA output includes the above calculations and estimated radionuclide concentrations in plant produce, milk, and a beef composite by location.
Directory of Open Access Journals (Sweden)
Siniša Šadek
2010-01-01
Full Text Available RELAP5/SCDAPSIM and MAAP4 are two widely used severe accident computer codes for the integral analysis of the core and the reactor pressure vessel behaviour following the core degradation. The objective of the paper is the comparison of code results obtained by application of different modelling options and the evaluation of influence of thermal hydraulic behaviour of the plant on core damage progression. The analysed transient was postulated station blackout in NPP Krško with a leakage from reactor coolant pump seals. Two groups of calculations were performed where each group had a different break area and, thus, a different leakage rate. Analyses have shown that MAAP4 results were more sensitive to varying thermal hydraulic conditions in the primary system. User-defined parameters had to be carefully selected when the MAAP4 model was developed, in contrast to the RELAP5/SCDAPSIM model where those parameters did not have any significant impact on final results.
Energy Technology Data Exchange (ETDEWEB)
Kroeger, P.G.; Kennett, R.J.; Colman, J.; Ginsberg, T. (Brookhaven National Lab., Upton, NY (United States))
1991-10-01
This report documents the THATCH code, which can be used to model general thermal and flow networks of solids and coolant channels in two-dimensional r-z geometries. The main application of THATCH is to model reactor thermo-hydraulic transients in High-Temperature Gas-Cooled Reactors (HTGRs). The available modules simulate pressurized or depressurized core heatup transients, heat transfer to general exterior sinks or to specific passive Reactor Cavity Cooling Systems, which can be air or water-cooled. Graphite oxidation during air or water ingress can be modelled, including the effects of added combustion products to the gas flow and the additional chemical energy release. A point kinetics model is available for analyzing reactivity excursions; for instance due to water ingress, and also for hypothetical no-scram scenarios. For most HTGR transients, which generally range over hours, a user-selected nodalization of the core in r-z geometry is used. However, a separate model of heat transfer in the symmetry element of each fuel element is also available for very rapid transients. This model can be applied coupled to the traditional coarser r-z nodalization. This report described the mathematical models used in the code and the method of solution. It describes the code and its various sub-elements. Details of the input data and file usage, with file formats, is given for the code, as well as for several preprocessing and postprocessing options. The THATCH model of the currently applicable 350 MW{sub th} reactor is described. Input data for four sample cases are given with output available in fiche form. Installation requirements and code limitations, as well as the most common error indications are listed. 31 refs., 23 figs., 32 tabs.
Lilley, D. G.; Rhode, D. L.
1982-01-01
A primitive pressure-velocity variable finite difference computer code was developed to predict swirling recirculating inert turbulent flows in axisymmetric combustors in general, and for application to a specific idealized combustion chamber with sudden or gradual expansion. The technique involves a staggered grid system for axial and radial velocities, a line relaxation procedure for efficient solution of the equations, a two-equation k-epsilon turbulence model, a stairstep boundary representation of the expansion flow, and realistic accommodation of swirl effects. A user's manual, dealing with the computational problem, showing how the mathematical basis and computational scheme may be translated into a computer program is presented. A flow chart, FORTRAN IV listing, notes about various subroutines and a user's guide are supplied as an aid to prospective users of the code.
Just-in-Time Compilation-Inspired Methodology for Parallelization of Compute Intensive Java Code
Directory of Open Access Journals (Sweden)
GHULAM MUSTAFA
2017-01-01
Full Text Available Compute intensive programs generally consume significant fraction of execution time in a small amount of repetitive code. Such repetitive code is commonly known as hotspot code. We observed that compute intensive hotspots often possess exploitable loop level parallelism. A JIT (Just-in-Time compiler profiles a running program to identify its hotspots. Hotspots are then translated into native code, for efficient execution. Using similar approach, we propose a methodology to identify hotspots and exploit their parallelization potential on multicore systems. Proposed methodology selects and parallelizes each DOALL loop that is either contained in a hotspot method or calls a hotspot method. The methodology could be integrated in front-end of a JIT compiler to parallelize sequential code, just before native translation. However, compilation to native code is out of scope of this work. As a case study, we analyze eighteen JGF (Java Grande Forum benchmarks to determine parallelization potential of hotspots. Eight benchmarks demonstrate a speedup of up to 7.6x on an 8-core system
Computational simulation of thermal hydraulic processes in the model LMFBR fuel assembly
Bayaskhalanov, M. V.; Merinov, I. G.; Korsun, A. S.; Vlasov, M. N.
2017-01-01
The aim of this study was to verify a developed software module on the experimental fuel assembly with partial blockage of the flow section. The developed software module for simulation of thermal hydraulic processes in liquid metal coolant is based on theory of anisotropic porous media with specially developed integral turbulence model for coefficients determination. The finite element method is used for numerical solution. Experimental data for hexahedral assembly with electrically heated smooth cylindrical rods cooled by liquid sodium are considered. The results of calculation obtained with developed software module for a case of corner blockade are presented. The calculated distribution of coolant velocities showed the presence of the vortex flow behind the blockade. Features vortex region are in a good quantitative and qualitative agreement with experimental data. This demonstrates the efficiency of the hydrodynamic unit for developed software module. But obtained radial coolant temperature profiles differ significantly from the experimental in the vortex flow region. The possible reasons for this discrepancy were analyzed.
PIC codes for plasma accelerators on emerging computer architectures (GPUS, Multicore/Manycore CPUS)
Vincenti, Henri
2016-03-01
The advent of exascale computers will enable 3D simulations of a new laser-plasma interaction regimes that were previously out of reach of current Petasale computers. However, the paradigm used to write current PIC codes will have to change in order to fully exploit the potentialities of these new computing architectures. Indeed, achieving Exascale computing facilities in the next decade will be a great challenge in terms of energy consumption and will imply hardware developments directly impacting our way of implementing PIC codes. As data movement (from die to network) is by far the most energy consuming part of an algorithm future computers will tend to increase memory locality at the hardware level and reduce energy consumption related to data movement by using more and more cores on each compute nodes (''fat nodes'') that will have a reduced clock speed to allow for efficient cooling. To compensate for frequency decrease, CPU machine vendors are making use of long SIMD instruction registers that are able to process multiple data with one arithmetic operator in one clock cycle. SIMD register length is expected to double every four years. GPU's also have a reduced clock speed per core and can process Multiple Instructions on Multiple Datas (MIMD). At the software level Particle-In-Cell (PIC) codes will thus have to achieve both good memory locality and vectorization (for Multicore/Manycore CPU) to fully take advantage of these upcoming architectures. In this talk, we present the portable solutions we implemented in our high performance skeleton PIC code PICSAR to both achieve good memory locality and cache reuse as well as good vectorization on SIMD architectures. We also present the portable solutions used to parallelize the Pseudo-sepctral quasi-cylindrical code FBPIC on GPUs using the Numba python compiler.
Klipa, Vladimir; Zumr, David; Snehota, Michal; Dohnal, Michal
2016-04-01
relationship between the soil bulk density and hydraulic conductivity was detected - in general unsaturated soil hydraulic conductivity was higher when the soil bulk density was high. Differences in trends of unsaturated hydraulic conductivities in the years 2012 - 2014 and year 2015 were probably caused by different agricultural management. In years 2012 - 2014, ploughing and sowing (2012 - winter barley, 2013 - oat, 2014 - winter wheat) were carried out in autumn whereas in year 2015 they were done in spring (white mustard). The impact of individual agricultural procedures was not fully apparent in the dataset. New useful information on underlying changes of pore geometry that affected the hydraulic conductivity should be obtained from detailed analysis of X-ray computed tomography images that is currently being performed.
Energy Technology Data Exchange (ETDEWEB)
TP Clement
1999-06-24
RT3DV1 (Reactive Transport in 3-Dimensions) is computer code that solves the coupled partial differential equations that describe reactive-flow and transport of multiple mobile and/or immobile species in three-dimensional saturated groundwater systems. RT3D is a generalized multi-species version of the US Environmental Protection Agency (EPA) transport code, MT3D (Zheng, 1990). The current version of RT3D uses the advection and dispersion solvers from the DOD-1.5 (1997) version of MT3D. As with MT3D, RT3D also requires the groundwater flow code MODFLOW for computing spatial and temporal variations in groundwater head distribution. The RT3D code was originally developed to support the contaminant transport modeling efforts at natural attenuation demonstration sites. As a research tool, RT3D has also been used to model several laboratory and pilot-scale active bioremediation experiments. The performance of RT3D has been validated by comparing the code results against various numerical and analytical solutions. The code is currently being used to model field-scale natural attenuation at multiple sites. The RT3D code is unique in that it includes an implicit reaction solver that makes the code sufficiently flexible for simulating various types of chemical and microbial reaction kinetics. RT3D V1.0 supports seven pre-programmed reaction modules that can be used to simulate different types of reactive contaminants including benzene-toluene-xylene mixtures (BTEX), and chlorinated solvents such as tetrachloroethene (PCE) and trichloroethene (TCE). In addition, RT3D has a user-defined reaction option that can be used to simulate any other types of user-specified reactive transport systems. This report describes the mathematical details of the RT3D computer code and its input/output data structure. It is assumed that the user is familiar with the basics of groundwater flow and contaminant transport mechanics. In addition, RT3D users are expected to have some experience in
Physical implementation of a Majorana fermion surface code for fault-tolerant quantum computation
Vijay, Sagar; Fu, Liang
2016-12-01
We propose a physical realization of a commuting Hamiltonian of interacting Majorana fermions realizing Z 2 topological order, using an array of Josephson-coupled topological superconductor islands. The required multi-body interaction Hamiltonian is naturally generated by a combination of charging energy induced quantum phase-slips on the superconducting islands and electron tunneling between islands. Our setup improves on a recent proposal for implementing a Majorana fermion surface code (Vijay et al 2015 Phys. Rev. X 5 041038), a ‘hybrid’ approach to fault-tolerant quantum computation that combines (1) the engineering of a stabilizer Hamiltonian with a topologically ordered ground state with (2) projective stabilizer measurements to implement error correction and a universal set of logical gates. Our hybrid strategy has advantages over the traditional surface code architecture in error suppression and single-step stabilizer measurements, and is widely applicable to implementing stabilizer codes for quantum computation.
Modeling of BWR core meltdown accidents - for application in the MELRPI. MOD2 computer code
Energy Technology Data Exchange (ETDEWEB)
Koh, B R; Kim, S H; Taleyarkhan, R P; Podowski, M Z; Lahey, Jr, R T
1985-04-01
This report summarizes improvements and modifications made in the MELRPI computer code. A major difference between this new, updated version of the code, called MELRPI.MOD2, and the one reported previously, concerns the inclusion of a model for the BWR emergency core cooling systems (ECCS). This model and its computer implementation, the ECCRPI subroutine, account for various emergency injection modes, for both intact and rubblized geometries. Other changes to MELRPI deal with an improved model for canister wall oxidation, rubble bed modeling, and numerical integration of system equations. A complete documentation of the entire MELRPI.MOD2 code is also given, including an input guide, list of subroutines, sample input/output and program listing.
Computational approaches towards understanding human long non-coding RNA biology.
Jalali, Saakshi; Kapoor, Shruti; Sivadas, Ambily; Bhartiya, Deeksha; Scaria, Vinod
2015-07-15
Long non-coding RNAs (lncRNAs) form the largest class of non-protein coding genes in the human genome. While a small subset of well-characterized lncRNAs has demonstrated their significant role in diverse biological functions like chromatin modifications, post-transcriptional regulation, imprinting etc., the functional significance of a vast majority of them still remains an enigma. Increasing evidence of the implications of lncRNAs in various diseases including cancer and major developmental processes has further enhanced the need to gain mechanistic insights into the lncRNA functions. Here, we present a comprehensive review of the various computational approaches and tools available for the identification and annotation of long non-coding RNAs. We also discuss a conceptual roadmap to systematically explore the functional properties of the lncRNAs using computational approaches.
Algorithms and computer codes for atomic and molecular quantum scattering theory. Volume I
Energy Technology Data Exchange (ETDEWEB)
Thomas, L. (ed.)
1979-01-01
The goals of this workshop are to identify which of the existing computer codes for solving the coupled equations of quantum molecular scattering theory perform most efficiently on a variety of test problems, and to make tested versions of those codes available to the chemistry community through the NRCC software library. To this end, many of the most active developers and users of these codes have been invited to discuss the methods and to solve a set of test problems using the LBL computers. The first volume of this workshop report is a collection of the manuscripts of the talks that were presented at the first meeting held at the Argonne National Laboratory, Argonne, Illinois June 25-27, 1979. It is hoped that this will serve as an up-to-date reference to the most popular methods with their latest refinements and implementations.
Tight bounds on computing error-correcting codes by bounded-depth circuits with arbitrary gates
DEFF Research Database (Denmark)
Gál, Anna; Hansen, Kristoffer Arnsfelt; Koucký, Michal;
2011-01-01
We bound the minimum number w of wires needed to compute any (asymptotically good) error-correcting code C:01(n)01n with minimum distance (n), using unbounded fan-in circuits of depth d with arbitrary gates. Our main results are: (1) If d=2 then w=(n(lognloglogn)2) . (2) If d=3 then w=(nlglgn). (3...
Application of Multiple Description Coding for Adaptive QoS Mechanism for Mobile Cloud Computing
Directory of Open Access Journals (Sweden)
Ilan Sadeh
2014-02-01
Full Text Available Multimedia transmission over cloud infrastructure is a hot research topic worldwide. It is very strongly related to video streaming, VoIP, mobile networks, and computer networks. The goal is a reliable integration of telephony, video and audio transmission, computing and broadband transmission based on cloud computing. One right approach to pave the way for mobile multimedia and cloud computing is Multiple Description Coding (MDC, i.e. the solution would be: TCP/IP and similar protocols to be used for transmission of text files, and Multiple Description Coding “Send and Forget” algorithm to be used as transmission method for Multimedia over the cloud. Multiple Description Coding would improve the Quality of Service and would provide new service of rate adaptive streaming. This paper presents a new approach for improving the quality of multimedia and other services in the cloud, by using Multiple Description Coding (MDC. Firsty MDC Send and Forget Algorithm is compared with the existing protocols such as TCP/IP, UDP, RTP, etc. Then the Achievable Rate Region for MDC system is evaluated. Finally, a new subset of Quality of Service that considers the blocking in multi-terminal multimedia network and fidelity losses is considered.
Ivanov, Anisoara; Neacsu, Andrei
2011-01-01
This study describes the possibility and advantages of utilizing simple computer codes to complement the teaching techniques for high school physics. The authors have begun working on a collection of open source programs which allow students to compare the results and graphics from classroom exercises with the correct solutions and further more to…
Methods, algorithms and computer codes for calculation of electron-impact excitation parameters
Bogdanovich, P; Stonys, D
2015-01-01
We describe the computer codes, developed at Vilnius University, for the calculation of electron-impact excitation cross sections, collision strengths, and excitation rates in the plane-wave Born approximation. These codes utilize the multireference atomic wavefunctions which are also adopted to calculate radiative transition parameters of complex many-electron ions. This leads to consistent data sets suitable in plasma modelling codes. Two versions of electron scattering codes are considered in the present work, both of them employing configuration interaction method for inclusion of correlation effects and Breit-Pauli approximation to account for relativistic effects. These versions differ only by one-electron radial orbitals, where the first one employs the non-relativistic numerical radial orbitals, while another version uses the quasirelativistic radial orbitals. The accuracy of produced results is assessed by comparing radiative transition and electron-impact excitation data for neutral hydrogen, helium...
Computer code to interchange CDS and wave-drag geometry formats
Johnson, V. S.; Turnock, D. L.
1986-01-01
A computer program has been developed on the PRIME minicomputer to provide an interface for the passage of aircraft configuration geometry data between the Rockwell Configuration Development System (CDS) and a wireframe geometry format used by aerodynamic design and analysis codes. The interface program allows aircraft geometry which has been developed in CDS to be directly converted to the wireframe geometry format for analysis. Geometry which has been modified in the analysis codes can be transformed back to a CDS geometry file and examined for physical viability. Previously created wireframe geometry files may also be converted into CDS geometry files. The program provides a useful link between a geometry creation and manipulation code and analysis codes by providing rapid and accurate geometry conversion.
Users manual for CAFE-3D : a computational fluid dynamics fire code.
Energy Technology Data Exchange (ETDEWEB)
Khalil, Imane; Lopez, Carlos; Suo-Anttila, Ahti Jorma (Alion Science and Technology, Albuquerque, NM)
2005-03-01
The Container Analysis Fire Environment (CAFE) computer code has been developed to model all relevant fire physics for predicting the thermal response of massive objects engulfed in large fires. It provides realistic fire thermal boundary conditions for use in design of radioactive material packages and in risk-based transportation studies. The CAFE code can be coupled to commercial finite-element codes such as MSC PATRAN/THERMAL and ANSYS. This coupled system of codes can be used to determine the internal thermal response of finite element models of packages to a range of fire environments. This document is a user manual describing how to use the three-dimensional version of CAFE, as well as a description of CAFE input and output parameters. Since this is a user manual, only a brief theoretical description of the equations and physical models is included.
TEMP: a computer code to calculate fuel pin temperatures during a transient. [LMFBR
Energy Technology Data Exchange (ETDEWEB)
Bard, F E; Christensen, B Y; Gneiting, B C
1980-04-01
The computer code TEMP calculates fuel pin temperatures during a transient. It was developed to accommodate temperature calculations in any system of axi-symmetric concentric cylinders. When used to calculate fuel pin temperatures, the code will handle a fuel pin as simple as a solid cylinder or as complex as a central void surrounded by fuel that is broken into three regions by two circumferential cracks. Any fuel situation between these two extremes can be analyzed along with additional cladding, heat sink, coolant or capsule regions surrounding the fuel. The one-region version of the code accurately calculates the solution to two problems having closed-form solutions. The code uses an implicit method, an explicit method and a Crank-Nicolson (implicit-explicit) method.
NASCRAC - A computer code for fracture mechanics analysis of crack growth
Harris, D. O.; Eason, E. D.; Thomas, J. M.; Bianca, C. J.; Salter, L. D.
1987-01-01
NASCRAC - a computer code for fracture mechanics analysis of crack growth - is described in this paper. The need for such a code is increasing as requirements grow for high reliability and low weight in aerospace components. The code is comprehensive and versatile, as well as user friendly. The major purpose of the code is calculation of fatigue, corrosion fatigue, or stress corrosion crack growth, and a variety of crack growth relations can be selected by the user. Additionally, crack retardation models are included. A very wide variety of stress intensity factor solutions are contained in the code, and extensive use is made of influence functions. This allows complex stress gradients in three-dimensional crack problems to be treated easily and economically. In cases where previous stress intensity factor solutions are not adequate, new influence functions can be calculated by the code. Additional features include incorporation of J-integral solutions from the literature and a capability for estimating elastic-plastic stress redistribution from the results of a corresponding elastic analysis. An example problem is presented which shows typical outputs from the code.
A proposed framework for computational fluid dynamics code calibration/validation
Energy Technology Data Exchange (ETDEWEB)
Oberkampf, W.L.
1993-12-31
The paper reviews the terminology and methodology that have been introduced during the last several years for building confidence n the predictions from Computational Fluid Dynamics (CID) codes. Code validation terminology developed for nuclear reactor analyses and aerospace applications is reviewed and evaluated. Currently used terminology such as ``calibrated code,`` ``validated code,`` and a ``validation experiment`` is discussed along with the shortcomings and criticisms of these terms. A new framework is proposed for building confidence in CFD code predictions that overcomes some of the difficulties of past procedures and delineates the causes of uncertainty in CFD predictions. Building on previous work, new definitions of code verification and calibration are proposed. These definitions provide more specific requirements for the knowledge level of the flow physics involved and the solution accuracy of the given partial differential equations. As part of the proposed framework, categories are also proposed for flow physics research, flow modeling research, and the application of numerical predictions. The contributions of physical experiments, analytical solutions, and other numerical solutions are discussed, showing that each should be designed to achieve a distinctively separate purpose in building confidence in accuracy of CFD predictions. A number of examples are given for each approach to suggest methods for obtaining the highest value for CFD code quality assurance.
An Object-Oriented Computer Code for Aircraft Engine Weight Estimation
Tong, Michael T.; Naylor, Bret A.
2009-01-01
Reliable engine-weight estimation at the conceptual design stage is critical to the development of new aircraft engines. It helps to identify the best engine concept amongst several candidates. At NASA Glenn Research Center (GRC), the Weight Analysis of Turbine Engines (WATE) computer code, originally developed by Boeing Aircraft, has been used to estimate the engine weight of various conceptual engine designs. The code, written in FORTRAN, was originally developed for NASA in 1979. Since then, substantial improvements have been made to the code to improve the weight calculations for most of the engine components. Most recently, to improve the maintainability and extensibility of WATE, the FORTRAN code has been converted into an object-oriented version. The conversion was done within the NASA's NPSS (Numerical Propulsion System Simulation) framework. This enables WATE to interact seamlessly with the thermodynamic cycle model which provides component flow data such as airflows, temperatures, and pressures, etc., that are required for sizing the components and weight calculations. The tighter integration between the NPSS and WATE would greatly enhance system-level analysis and optimization capabilities. It also would facilitate the enhancement of the WATE code for next-generation aircraft and space propulsion systems. In this paper, the architecture of the object-oriented WATE code (or WATE++) is described. Both the FORTRAN and object-oriented versions of the code are employed to compute the dimensions and weight of a 300-passenger aircraft engine (GE90 class). Both versions of the code produce essentially identical results as should be the case.
Multiple frequencies sequential coding for SSVEP-based brain-computer interface.
Directory of Open Access Journals (Sweden)
Yangsong Zhang
Full Text Available BACKGROUND: Steady-state visual evoked potential (SSVEP-based brain-computer interface (BCI has become one of the most promising modalities for a practical noninvasive BCI system. Owing to both the limitation of refresh rate of liquid crystal display (LCD or cathode ray tube (CRT monitor, and the specific physiological response property that only a very small number of stimuli at certain frequencies could evoke strong SSVEPs, the available frequencies for SSVEP stimuli are limited. Therefore, it may not be enough to code multiple targets with the traditional frequencies coding protocols, which poses a big challenge for the design of a practical SSVEP-based BCI. This study aimed to provide an innovative coding method to tackle this problem. METHODOLOGY/PRINCIPAL FINDINGS: In this study, we present a novel protocol termed multiple frequencies sequential coding (MFSC for SSVEP-based BCI. In MFSC, multiple frequencies are sequentially used in each cycle to code the targets. To fulfill the sequential coding, each cycle is divided into several coding epochs, and during each epoch, certain frequency is used. Obviously, different frequencies or the same frequency can be presented in the coding epochs, and the different epoch sequence corresponds to the different targets. To show the feasibility of MFSC, we used two frequencies to realize four targets and carried on an offline experiment. The current study shows that: 1 MFSC is feasible and efficient; 2 the performance of SSVEP-based BCI based on MFSC can be comparable to some existed systems. CONCLUSIONS/SIGNIFICANCE: The proposed protocol could potentially implement much more targets with the limited available frequencies compared with the traditional frequencies coding protocol. The efficiency of the new protocol was confirmed by real data experiment. We propose that the SSVEP-based BCI under MFSC might be a promising choice in the future.
Energy Technology Data Exchange (ETDEWEB)
Dunn, Floyd E. [Argonne National Lab. (ANL), Argonne, IL (United States); Hu, Lin-wen [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States). Nuclear Reactor Lab.; Wilson, Erik [Argonne National Lab. (ANL), Argonne, IL (United States)
2016-12-01
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 on 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.
Dodge, W. G.
1968-01-01
Computer program determines the forced vibration in three dimensional space of a multiple degree of freedom beam type structural system. Provision is made for the longitudinal axis of the analytical model to change orientation at any point along its length. This program is used by industries in which structural design dynamic analyses are performed.
Development and Verification of Thermal-hydraulic Analysis Code for Annular Fuel%环形燃料热工水力性能分析程序开发及验证
Institute of Scientific and Technical Information of China (English)
刁均辉; 季松涛; 张应超
2015-01-01
本工作开发了环形燃料子通道分析程序SAAF。采用SAAF计算了西屋公司四环路压水堆所用环形燃料组件的热工水力性能，并与VIPRE‐01的计算结果进行比较。结果表明，SAAF与VIPRE‐01的计算结果符合较好，SAAF可用于环形燃料热工水力设计分析。%A sub‐channel thermal‐hydraulic analysis code named SAAF (sub‐channel analyzer for annular fuel) for annular fuel was developed .The thermal‐hydraulic prop‐erties of annular fuel pins for Westinghouse 4‐loop PWR were calculated by SAAF code , and the calculating results of SAAF and VIPRE‐01 codes were compared .The results show that the SAAF code can be used to determine the thermal‐hydraulic properties of the annular fuel .
Zhao, Shengmei; Wang, Le; Liang, Wenqiang; Cheng, Weiwen; Gong, Longyan
2015-10-01
In this paper, we propose a high performance optical encryption (OE) scheme based on computational ghost imaging (GI) with QR code and compressive sensing (CS) technique, named QR-CGI-OE scheme. N random phase screens, generated by Alice, is a secret key and be shared with its authorized user, Bob. The information is first encoded by Alice with QR code, and the QR-coded image is then encrypted with the aid of computational ghost imaging optical system. Here, measurement results from the GI optical system's bucket detector are the encrypted information and be transmitted to Bob. With the key, Bob decrypts the encrypted information to obtain the QR-coded image with GI and CS techniques, and further recovers the information by QR decoding. The experimental and numerical simulated results show that the authorized users can recover completely the original image, whereas the eavesdroppers can not acquire any information about the image even the eavesdropping ratio (ER) is up to 60% at the given measurement times. For the proposed scheme, the number of bits sent from Alice to Bob are reduced considerably and the robustness is enhanced significantly. Meantime, the measurement times in GI system is reduced and the quality of the reconstructed QR-coded image is improved.
MOLOCH computer code for molecular-dynamics simulation of processes in condensed matter
Directory of Open Access Journals (Sweden)
Derbenev I.V.
2011-01-01
Full Text Available Theoretical and experimental investigation into properties of condensed matter is one of the mainstreams in RFNC-VNIITF scientific activity. The method of molecular dynamics (MD is an innovative method of theoretical materials science. Modern supercomputers allow the direct simulation of collective effects in multibillion atom sample, making it possible to model physical processes on the atomistic level, including material response to dynamic load, radiation damage, influence of defects and alloying additions upon material mechanical properties, or aging of actinides. During past ten years, the computer code MOLOCH has been developed at RFNC-VNIITF. It is a parallel code suitable for massive parallel computing. Modern programming techniques were used to make the code almost 100% efficient. Practically all instruments required for modelling were implemented in the code: a potential builder for different materials, simulation of physical processes in arbitrary 3D geometry, and calculated data processing. A set of tests was developed to analyse algorithms efficiency. It can be used to compare codes with different MD implementation between each other.
Computational hydraulics of a cascade of experimental-scale landside dam failures
Wright, N.; Guan, M.
2015-12-01
Abstract: Landslide dams typically comprise unconsolidated and poorly sorted material, and are vulnerable to rapid failure and breaching, particularly in mountainous areas during high intense rainfalls. A large flash flood with high-concentrated sediment can be formed in a short period, and the magnitude is likely to be amplified along the flow direction due to the inclusion of a large amount of sediment. This can result in significant and sudden flood risk downstream for human life and property. Numerous field evidence has indicated the various risks of landslide dam failures. In general, cascading landslide dams can be formed along the sloping channel due to the randomness and unpredictability of landslides, which complexes the hydraulics of landslide dam failures. The failure process of a single dam and subsequent floods has attracted attention in multidisciplinary studies. However, the dynamic failure process of cascading landslide dams has been poorly understood. From a viewpoint of simulation, this study evaluates the formation and development of rapid sediment-charged floods due to cascading failure of landslide dams through detailed hydro-morphodynamic modelling. The model used is based on shallow water theory and it has been successful in predicting the flow and morphological process during sudden dam-break, as well as full and partial dyke-breach. Various experimental-scale scenarios are modelled, including: (1) failure of a single full dam in a sloping channel, (2) failure of two dams in a sloping channel, (3) failure of multiple landslide dams (four) in a sloping channel. For each scenario, different failure modes (sudden/gradual) and bed boundary (fixed /mobile) are assumed and simulated. The study systematically explores the tempo-spatial evolution of landslide-induced floods (discharge, flow velocity, and flow concentration) and geomorphic properties along the sloping channel. The effects of in-channel erosion and flow-driven sediment from dams on
Directory of Open Access Journals (Sweden)
Daniel Litinski
2017-09-01
Full Text Available We present a scalable architecture for fault-tolerant topological quantum computation using networks of voltage-controlled Majorana Cooper pair boxes and topological color codes for error correction. Color codes have a set of transversal gates which coincides with the set of topologically protected gates in Majorana-based systems, namely, the Clifford gates. In this way, we establish color codes as providing a natural setting in which advantages offered by topological hardware can be combined with those arising from topological error-correcting software for full-fledged fault-tolerant quantum computing. We provide a complete description of our architecture, including the underlying physical ingredients. We start by showing that in topological superconductor networks, hexagonal cells can be employed to serve as physical qubits for universal quantum computation, and we present protocols for realizing topologically protected Clifford gates. These hexagonal-cell qubits allow for a direct implementation of open-boundary color codes with ancilla-free syndrome read-out and logical T gates via magic-state distillation. For concreteness, we describe how the necessary operations can be implemented using networks of Majorana Cooper pair boxes, and we give a feasibility estimate for error correction in this architecture. Our approach is motivated by nanowire-based networks of topological superconductors, but it could also be realized in alternative settings such as quantum-Hall–superconductor hybrids.
Once-through CANDU reactor models for the ORIGEN2 computer code
Energy Technology Data Exchange (ETDEWEB)
Croff, A.G.; Bjerke, M.A.
1980-11-01
Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % /sup 235/U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given.
Adaptive Mesh Computations with the PLUTO Code for Astrophysical Fluid Dynamics
Mignone, A.; Zanni, C.
2012-07-01
We present an overview of the current version of the PLUTO code for numerical simulations of astrophysical fluid flows over block-structured adaptive grids. The code preserves its modular framework for the solution of the classical or relativistic magnetohydrodynamics (MHD) equations while exploiting the distributed infrastructure of the Chombo library for multidimensional adaptive mesh refinement (AMR) parallel computations. Equations are evolved in time using an explicit second-order, dimensionally unsplit time stepping scheme based on a cell-centered discretization of the flow variables. Efficiency and robustness are shown through multidimensional benchmarks and applications to problems of astrophysical relevance.
Experimental assessment of computer codes used for safety analysis of integral reactors
Energy Technology Data Exchange (ETDEWEB)
Falkov, A.A.; Kuul, V.S.; Samoilov, O.B. [OKB Mechanical Engineering, Nizhny Novgorod (Russian Federation)
1995-09-01
Peculiarities of integral reactor thermohydraulics in accidents are associated with presence of noncondensable gas in built-in pressurizer, absence of pumped ECCS, use of guard vessel for LOCAs localisation and passive RHRS through in-reactor HX`s. These features defined the main trends in experimental investigations and verification efforts for computer codes applied. The paper reviews briefly the performed experimental investigation of thermohydraulics of AST-500, VPBER600-type integral reactors. The characteristic of UROVEN/MB-3 code for LOCAs analysis in integral reactors and results of its verification are given. The assessment of RELAP5/mod3 applicability for accident analysis in integral reactor is presented.
The MELTSPREAD-1 computer code for the analysis of transient spreading in containments
Energy Technology Data Exchange (ETDEWEB)
Farmer, M.T.; Sienicki, J.J.; Spencer, B.W.
1990-01-01
A one-dimensional, multicell, Eulerian finite difference computer code (MELTSPREAD-1) has been developed to provide an improved prediction of the gravity driven spreading and thermal interactions of molten corium flowing over a concrete or steel surface. In this paper, the modeling incorporated into the code is described and the spreading models are benchmarked against a simple dam break'' problem as well as water simulant spreading data obtained in a scaled apparatus of the Mk I containment. Results are also presented for a scoping calculation of the spreading behavior and shell thermal response in the full scale Mk I system following vessel meltthrough. 24 refs., 15 figs.
Energy Technology Data Exchange (ETDEWEB)
Strenge, D.L.; Peloquin, R.A.
1981-04-01
The computer code HADOC (Hanford Acute Dose Calculations) is described and instructions for its use are presented. The code calculates external dose from air submersion and inhalation doses following acute radionuclide releases. Atmospheric dispersion is calculated using the Hanford model with options to determine maximum conditions. Building wake effects and terrain variation may also be considered. Doses are calculated using dose conversion factor supplied in a data library. Doses are reported for one and fifty year dose commitment periods for the maximum individual and the regional population (within 50 miles). The fractional contribution to dose by radionuclide and exposure mode are also printed if requested.
Extensive use of computational fluid dynamics in the upgrading of hydraulic turbines
Energy Technology Data Exchange (ETDEWEB)
Sabourin, M.; De Henau, V. [GEC Alsthom Electromechanical Inc., Tracy, PQ (Canada); Eremeef, R. [GEC Alsthom Neyrpic, Grenoble (France)
1995-12-31
The use of computational fluid flow dynamics (CFD) and the Navier Stokes equations by GEC Alsthom for turbine rehabilitation were discussed. The process of runner rehabilitation was discussed from a fluid flow perspective, which accounts for the spiral case-distributor set and draft tube. The Kootenay turbine rehabilitation was described with regard to it spiral case and stay vane. The numerical analysis used to model upstream components was explained. The influence of draft tube effects was emphasized as an important efficiency factor. The differences between draft tubes at Sir Adam Beck 2 and La Grande 2 were discussed. Computational fluid flow modelling was claimed to have produced global performance enhancements in a reasonably short time, and at a reasonable cost. 6 refs., 6 figs., 4 tabs.
Energy Technology Data Exchange (ETDEWEB)
Sano, M. [Hiroshima City University, Hiroshima (Japan)
2000-03-15
Described herein is education of hydraulics and pneumatics in Hiroshima City University. Department of Computer Science is responsible for the education, covering a wide educational range from basics of information processing methodology to application of mathematical procedures. This university provides no subject directly related to hydraulics and pneumatics, which, however, can be studied by the courses of control engineering or modern control theories. These themes are taken up for graduation theses for bachelors and masters; 2 for dynamic characteristics of pneumatic cylinders, and one for pneumatic circuit simulation. Images of the terms hydraulics and pneumatics are outdated for students of information-related departments. Hydraulics and pneumatics are being forced to rapidly change, like other branches of science, and it may be time to make a drastic change from hardware to software, because their developments have been excessively oriented to hardware. It is needless to say that they are based on hardware, but it may be worthy of drastically changing these branches of science by establishing virtual fluid power systems. It is also proposed to introduce the modern multi-media techniques into the education of hydraulics and pneumatics. (NEDO)
1977-02-01
temperatures are initialized, the external structure temperature is changed from degrees Farenheit to Rankine and raised to the fourth power, and the...Section 1000 - The fluid and wall temperatures are initialized, the external structure temperature is changed from degrecs Farenheit to Rankine and...Computational Methods SECTION 1000 The fluid and wall temperatures are initialized, the external structure temperature is changed from degrees Farenheit to
Rutishauser, David
2006-01-01
The motivation for this work comes from an observation that amidst the push for Massively Parallel (MP) solutions to high-end computing problems such as numerical physical simulations, large amounts of legacy code exist that are highly optimized for vector supercomputers. Because re-hosting legacy code often requires a complete re-write of the original code, which can be a very long and expensive effort, this work examines the potential to exploit reconfigurable computing machines in place of a vector supercomputer to implement an essentially unmodified legacy source code. Custom and reconfigurable computing resources could be used to emulate an original application's target platform to the extent required to achieve high performance. To arrive at an architecture that delivers the desired performance subject to limited resources involves solving a multi-variable optimization problem with constraints. Prior research in the area of reconfigurable computing has demonstrated that designing an optimum hardware implementation of a given application under hardware resource constraints is an NP-complete problem. The premise of the approach is that the general issue of applying reconfigurable computing resources to the implementation of an application, maximizing the performance of the computation subject to physical resource constraints, can be made a tractable problem by assuming a computational paradigm, such as vector processing. This research contributes a formulation of the problem and a methodology to design a reconfigurable vector processing implementation of a given application that satisfies a performance metric. A generic, parametric, architectural framework for vector processing implemented in reconfigurable logic is developed as a target for a scheduling/mapping algorithm that maps an input computation to a given instance of the architecture. This algorithm is integrated with an optimization framework to arrive at a specification of the architecture parameters
Directory of Open Access Journals (Sweden)
Abdellah Ait moussa
2014-08-01
Full Text Available The design and optimization of turbo machine impellers such as those in pumps and turbines is a highly complicated task due to the complex three-dimensional shape of the impeller blades and surrounding devices. Small differences in geometry can lead to significant changes in the performance of these machines. We report here an efficient numerical technique that automatically optimizes the geometry of these blades for maximum performance. The technique combines, mathematical modeling of the impeller blades using non-uniform rational B-spline (NURBS, Computational fluid dynamics (CFD with Geometry Parameterizations in turbulent flow simulation and the Globalized and bounded Nelder-Mead (GBNM algorithm in geometry optimization.
Energy Technology Data Exchange (ETDEWEB)
Reginatto, M.; Goldhagen, P.
1998-06-01
The problem of analyzing data from a multisphere neutron spectrometer to infer the energy spectrum of the incident neutrons is discussed. The main features of the code MAXED, a computer program developed to apply the maximum entropy principle to the deconvolution (unfolding) of multisphere neutron spectrometer data, are described, and the use of the code is illustrated with an example. A user`s guide for the code MAXED is included in an appendix. The code is available from the authors upon request.
The MELTSPREAD-1 computer code for the analysis of transient spreading in containments
Energy Technology Data Exchange (ETDEWEB)
Farmer, M.T.; Sienicki, J.J.; Spencer, B.W.
1990-01-01
Transient spreading of molten core materials is important in the assessment of severe-accident sequences for Mk-I boiling water reactors (BWRs). Of interest is whether core materials are able to spread over the pedestal and drywell floors to contact the containment shell and cause thermally induced shell failure, or whether heat transfer to underlying concrete and overlying water will freeze the melt short of the shell. The development of a computational capability for the assessment of this problem was initiated by Sienicki et al. in the form of MELTSPREAD-O code. Development is continuing in the form of the MELTSPREAD-1 code, which contains new models for phenomena that were ignored in the earlier code. This paper summarizes these new models, provides benchmarking calculations of the relocation model against an analytical solution as well as simulant spreading data, and summarizes the results of a scoping calculation for the full Mk-I system.
Computer code simulations of the formation of Meteor Crater, Arizona - Calculations MC-1 and MC-2
Roddy, D. J.; Schuster, S. H.; Kreyenhagen, K. N.; Orphal, D. L.
1980-01-01
It has been widely accepted that hypervelocity impact processes play a major role in the evolution of the terrestrial planets and satellites. In connection with the development of quantitative methods for the description of impact cratering, it was found that the results provided by two-dimensional finite difference, computer codes is greatly improved when initial impact conditions can be defined and when the numerical results can be tested against field and laboratory data. In order to address this problem, a numerical code study of the formation of Meteor (Barringer) Crater, Arizona, has been undertaken. A description is presented of the major results from the first two code calculations, MC-1 and MC-2, that have been completed for Meteor Crater. Both calculations used an iron meteorite with a kinetic energy of 3.8 Megatons. Calculation MC-1 had an impact velocity of 25 km/sec and MC-2 had an impact velocity of 15 km/sec.
WOLF: a computer code package for the calculation of ion beam trajectories
Energy Technology Data Exchange (ETDEWEB)
Vogel, D.L.
1985-10-01
The WOLF code solves POISSON'S equation within a user-defined problem boundary of arbitrary shape. The code is compatible with ANSI FORTRAN and uses a two-dimensional Cartesian coordinate geometry represented on a triangular lattice. The vacuum electric fields and equipotential lines are calculated for the input problem. The use may then introduce a series of emitters from which particles of different charge-to-mass ratios and initial energies can originate. These non-relativistic particles will then be traced by WOLF through the user-defined region. Effects of ion and electron space charge are included in the calculation. A subprogram PISA forms part of this code and enables optimization of various aspects of the problem. The WOLF package also allows detailed graphics analysis of the computed results to be performed.
HYDRA-II: A hydrothermal analysis computer code: Volume 3, Verification/validation assessments
Energy Technology Data Exchange (ETDEWEB)
McCann, R.A.; Lowery, P.S.
1987-10-01
HYDRA-II is a hydrothermal computer code capable of three-dimensional analysis of coupled conduction, convection, and thermal radiation problems. This code is especially appropriate for simulating the steady-state performance of spent fuel storage systems. The code has been evaluated for this application for the US Department of Energy's Commercial Spent Fuel Management Program. HYDRA-II provides a finite difference solution in cartesian coordinates to the equations governing the conservation of mass, momentum, and energy. A cylindrical coordinate system may also be used to enclose the cartesian coordinate system. This exterior coordinate system is useful for modeling cylindrical cask bodies. The difference equations for conservation of momentum are enhanced by the incorporation of directional porosities and permeabilities that aid in modeling solid structures whose dimensions may be smaller than the computational mesh. The equation for conservation of energy permits modeling of orthotropic physical properties and film resistances. Several automated procedures are available to model radiation transfer within enclosures and from fuel rod to fuel rod. The documentation of HYDRA-II is presented in three separate volumes. Volume I - Equations and Numerics describes the basic differential equations, illustrates how the difference equations are formulated, and gives the solution procedures employed. Volume II - User's Manual contains code flow charts, discusses the code structure, provides detailed instructions for preparing an input file, and illustrates the operation of the code by means of a model problem. This volume, Volume III - Verification/Validation Assessments, provides a comparison between the analytical solution and the numerical simulation for problems with a known solution. This volume also documents comparisons between the results of simulations of single- and multiassembly storage systems and actual experimental data. 11 refs., 55 figs., 13 tabs.
HYDRA-II: A hydrothermal analysis computer code: Volume 2, User's manual
Energy Technology Data Exchange (ETDEWEB)
McCann, R.A.; Lowery, P.S.; Lessor, D.L.
1987-09-01
HYDRA-II is a hydrothermal computer code capable of three-dimensional analysis of coupled conduction, convection, and thermal radiation problems. This code is especially appropriate for simulating the steady-state performance of spent fuel storage systems. The code has been evaluated for this application for the US Department of Energy's Commercial Spent Fuel Management Program. HYDRA-II provides a finite-difference solution in cartesian coordinates to the equations governing the conservation of mass, momentum, and energy. A cylindrical coordinate system may also be used to enclose the cartesian coordinate system. This exterior coordinate system is useful for modeling cylindrical cask bodies. The difference equations for conservation of momentum incorporate directional porosities and permeabilities that are available to model solid structures whose dimensions may be smaller than the computational mesh. The equation for conservation of energy permits modeling of orthotropic physical properties and film resistances. Several automated methods are available to model radiation transfer within enclosures and from fuel rod to fuel rod. The documentation of HYDRA-II is presented in three separate volumes. Volume 1 - Equations and Numerics describes the basic differential equations, illustrates how the difference equations are formulated, and gives the solution procedures employed. This volume, Volume 2 - User's Manual, contains code flow charts, discusses the code structure, provides detailed instructions for preparing an input file, and illustrates the operation of the code by means of a sample problem. The final volume, Volume 3 - Verification/Validation Assessments, provides a comparison between the analytical solution and the numerical simulation for problems with a known solution. 6 refs.
Zhang, Shuai; Morita, Koji; Shirakawa, Noriyuki; Yamamoto, Yuichi
The COMPASS code is designed based on the moving particle semi-implicit method to simulate various complex mesoscale phenomena relevant to core disruptive accidents of sodium-cooled fast reactors. In this study, a computational framework for fluid-solid mixture flow simulations was developed for the COMPASS code. The passively moving solid model was used to simulate hydrodynamic interactions between fluid and solids. Mechanical interactions between solids were modeled by the distinct element method. A multi-time-step algorithm was introduced to couple these two calculations. The proposed computational framework for fluid-solid mixture flow simulations was verified by the comparison between experimental and numerical studies on the water-dam break with multiple solid rods.
Automatic code generation in SPARK: Applications of computer algebra and compiler-compilers
Energy Technology Data Exchange (ETDEWEB)
Nataf, J.M.; Winkelmann, F.
1992-09-01
We show how computer algebra and compiler-compilers are used for automatic code generation in the Simulation Problem Analysis and Research Kernel (SPARK), an object oriented environment for modeling complex physical systems that can be described by differential-algebraic equations. After a brief overview of SPARK, we describe the use of computer algebra in SPARK`s symbolic interface, which generates solution code for equations that are entered in symbolic form. We also describe how the Lex/Yacc compiler-compiler is used to achieve important extensions to the SPARK simulation language, including parametrized macro objects and steady-state resetting of a dynamic simulation. The application of these methods to solving the partial differential equations for two-dimensional heat flow is illustrated.
Automatic code generation in SPARK: Applications of computer algebra and compiler-compilers
Energy Technology Data Exchange (ETDEWEB)
Nataf, J.M.; Winkelmann, F.
1992-09-01
We show how computer algebra and compiler-compilers are used for automatic code generation in the Simulation Problem Analysis and Research Kernel (SPARK), an object oriented environment for modeling complex physical systems that can be described by differential-algebraic equations. After a brief overview of SPARK, we describe the use of computer algebra in SPARK's symbolic interface, which generates solution code for equations that are entered in symbolic form. We also describe how the Lex/Yacc compiler-compiler is used to achieve important extensions to the SPARK simulation language, including parametrized macro objects and steady-state resetting of a dynamic simulation. The application of these methods to solving the partial differential equations for two-dimensional heat flow is illustrated.
Computing element evolution towards Exascale and its impact on legacy simulation codes
Energy Technology Data Exchange (ETDEWEB)
Colin de Verdiere, Guillaume J.L. [CEA, DAM, DIF, Arpajon (France)
2015-12-15
In the light of the current race towards the Exascale, this article highlights the main features of the forthcoming computing elements that will be at the core of next generations of supercomputers. The market analysis, underlying this work, shows that computers are facing a major evolution in terms of architecture. As a consequence, it is important to understand the impacts of those evolutions on legacy codes or programming methods. The problems of dissipated power and memory access are discussed and will lead to a vision of what should be an exascale system. To survive, programming languages had to respond to the hardware evolutions either by evolving or with the creation of new ones. From the previous elements, we elaborate why vectorization, multithreading, data locality awareness and hybrid programming will be the key to reach the exascale, implying that it is time to start rewriting codes. (orig.)
Chen, Y. S.
1986-03-01
In this report, a numerical method for solving the equations of motion of three-dimensional incompressible flows in nonorthogonal body-fitted coordinate (BFC) systems has been developed. The equations of motion are transformed to a generalized curvilinear coordinate system from which the transformed equations are discretized using finite difference approximations in the transformed domain. The hybrid scheme is used to approximate the convection terms in the governing equations. Solutions of the finite difference equations are obtained iteratively by using a pressure-velocity correction algorithm (SIMPLE-C). Numerical examples of two- and three-dimensional, laminar and turbulent flow problems are employed to evaluate the accuracy and efficiency of the present computer code. The user's guide and computer program listing of the present code are also included.
Abstracts of digital computer code packages assembled by the Radiation Shielding Information Center
Energy Technology Data Exchange (ETDEWEB)
Carter, B.J.; Maskewitz, B.F.
1985-04-01
This publication, ORNL/RSIC-13, Volumes I to III Revised, has resulted from an internal audit of the first 168 packages of computing technology in the Computer Codes Collection (CCC) of the Radiation Shielding Information Center (RSIC). It replaces the earlier three documents published as single volumes between 1966 to 1972. A significant number of the early code packages were considered to be obsolete and were removed from the collection in the audit process and the CCC numbers were not reassigned. Others not currently being used by the nuclear R and D community were retained in the collection to preserve technology not replaced by newer methods, or were considered of potential value for reference purposes. Much of the early technology, however, has improved through developer/RSIC/user interaction and continues at the forefront of the advancing state-of-the-art.
Improvement of Level-1 PSA computer code package -A study for nuclear safety improvement-
Energy Technology Data Exchange (ETDEWEB)
Park, Chang Kyu; Kim, Tae Woon; Ha, Jae Joo; Han, Sang Hoon; Cho, Yeong Kyun; Jeong, Won Dae; Jang, Seung Cheol; Choi, Young; Seong, Tae Yong; Kang, Dae Il; Hwang, Mi Jeong; Choi, Seon Yeong; An, Kwang Il [Korea Atomic Energy Res. Inst., Taejon (Korea, Republic of)
1994-07-01
This year is the second year of the Government-sponsored Mid- and Long-Term Nuclear Power Technology Development Project. The scope of this subproject titled on `The Improvement of Level-1 PSA Computer Codes` is divided into three main activities : (1) Methodology development on the under-developed fields such as risk assessment technology for plant shutdown and external events, (2) Computer code package development for Level-1 PSA, (3) Applications of new technologies to reactor safety assessment. At first, in the area of PSA methodology development, foreign PSA reports on shutdown and external events have been reviewed and various PSA methodologies have been compared. Level-1 PSA code KIRAP and CCF analysis code COCOA are converted from KOS to Windows. Human reliability database has been also established in this year. In the area of new technology applications, fuzzy set theory and entropy theory are used to estimate component life and to develop a new measure of uncertainty importance. Finally, in the field of application study of PSA technique to reactor regulation, a strategic study to develop a dynamic risk management tool PEPSI and the determination of inspection and test priority of motor operated valves based on risk importance worths have been studied. (Author).
[Series: Medical Applications of the PHITS Code (2): Acceleration by Parallel Computing].
Furuta, Takuya; Sato, Tatsuhiko
2015-01-01
Time-consuming Monte Carlo dose calculation becomes feasible owing to the development of computer technology. However, the recent development is due to emergence of the multi-core high performance computers. Therefore, parallel computing becomes a key to achieve good performance of software programs. A Monte Carlo simulation code PHITS contains two parallel computing functions, the distributed-memory parallelization using protocols of message passing interface (MPI) and the shared-memory parallelization using open multi-processing (OpenMP) directives. Users can choose the two functions according to their needs. This paper gives the explanation of the two functions with their advantages and disadvantages. Some test applications are also provided to show their performance using a typical multi-core high performance workstation.
PREMOR: a point reactor exposure model computer code for survey analysis of power plant performance
Energy Technology Data Exchange (ETDEWEB)
Vondy, D.R.
1979-10-01
The PREMOR computer code was written to exploit a simple, two-group point nuclear reactor power plant model for survey analysis. Up to thirteen actinides, fourteen fission products, and one lumped absorber nuclide density are followed over a reactor history. Successive feed batches are accounted for with provision for from one to twenty batches resident. The effect of exposure of each of the batches to the same neutron flux is determined.
Tight bounds on computing error-correcting codes by bounded-depth circuits with arbitrary gates
DEFF Research Database (Denmark)
Gal, A.; Hansen, Kristoffer Arnsfelt; Koucky, Michal
2013-01-01
We bound the minimum number w of wires needed to compute any (asymptotically good) error-correcting code C:{0,1}Ω(n)→{0,1}n with minimum distance Ω(n), using unbounded fan-in circuits of depth d with arbitrary gates. Our main results are: 1) if d=2, then w=Θ(n (lgn/lglgn)2); 2) if d=3, then w...
Chia-Chang Hu
2005-01-01
A novel space-time adaptive near-far robust code-synchronization array detector for asynchronous DS-CDMA systems is developed in this paper. There are the same basic requirements that are needed by the conventional matched filter of an asynchronous DS-CDMA system. For the real-time applicability, a computationally efficient architecture of the proposed detector is developed that is based on the concept of the multistage Wiener filter (MWF) of Goldstein and Reed. This multistage technique resu...
Method for computing self-consistent solution in a gun code
Nelson, Eric M
2014-09-23
Complex gun code computations can be made to converge more quickly based on a selection of one or more relaxation parameters. An eigenvalue analysis is applied to error residuals to identify two error eigenvalues that are associated with respective error residuals. Relaxation values can be selected based on these eigenvalues so that error residuals associated with each can be alternately reduced in successive iterations. In some examples, relaxation values that would be unstable if used alone can be used.
Tight bounds on computing error-correcting codes by bounded-depth circuits with arbitrary gates
DEFF Research Database (Denmark)
Gál, Anna; Hansen, Kristoffer Arnsfelt; Koucký, Michal;
2012-01-01
We bound the minimum number w of wires needed to compute any (asymptotically good) error-correcting code C:{0,1}Ω(n) -> {0,1}n with minimum distance Ω(n), using unbounded fan-in circuits of depth d with arbitrary gates. Our main results are: (1) If d=2 then w = Θ(n ({log n/ log log n})2). (2) If d...
DEFF Research Database (Denmark)
Johansen, Peter Meincke
1996-01-01
New uniform closed-form expressions for physical theory of diffraction equivalent edge currents are derived for truncated incremental wedge strips. In contrast to previously reported expressions, the new expressions are well-behaved for all directions of incidence and observation and take a finit...... value for zero strip length. Consequently, the new equivalent edge currents are, to the knowledge of the author, the first that are well-suited for implementation in general computer codes...
Walowit, Jed A.
1994-01-01
A viewgraph presentation is made showing the capabilities of the computer code SPIRALI. Overall capabilities of SPIRALI include: computes rotor dynamic coefficients, flow, and power loss for cylindrical and face seals; treats turbulent, laminar, Couette, and Poiseuille dominated flows; fluid inertia effects are included; rotor dynamic coefficients in three (face) or four (cylindrical) degrees of freedom; includes effects of spiral grooves; user definable transverse film geometry including circular steps and grooves; independent user definable friction factor models for rotor and stator; and user definable loss coefficients for sudden expansions and contractions.
Multilevel Coding Schemes for Compute-and-Forward with Flexible Decoding
Hern, Brett
2011-01-01
We consider the design of coding schemes for the wireless two-way relaying channel when there is no channel state information at the transmitter. In the spirit of the compute and forward paradigm, we present a multilevel coding scheme that permits computation (or, decoding) of a class of functions at the relay. The function to be computed (or, decoded) is then chosen depending on the channel realization. We define such a class of functions which can be decoded at the relay using the proposed coding scheme and derive rates that are universally achievable over a set of channel gains when this class of functions is used at the relay. We develop our framework with general modulation formats in mind, but numerical results are presented for the case where each node transmits using the QPSK constellation. Numerical results with QPSK show that the flexibility afforded by our proposed scheme results in substantially higher rates than those achievable by always using a fixed function or by adapting the function at the ...
Energy Technology Data Exchange (ETDEWEB)
Kasinathan, N.; Rajakumar, A.; Vaidyanathan, G.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)
1995-09-01
Post shutdown decay heat removal is an important safety requirement in any nuclear system. In order to improve the reliability of this function, Liquid metal (sodium) cooled fast breeder reactors (LMFBR) are equipped with redundant hot pool dipped immersion coolers connected to natural draught air cooled heat exchangers through intermediate sodium circuits. During decay heat removal, flow through the core, immersion cooler primary side and in the intermediate sodium circuits are also through natural convection. In order to establish the viability and validate computer codes used in making predictions, a 1:20 scale experimental model called RAMONA with water as coolant has been built and experimental simulation of decay heat removal situation has been performed at KfK Karlsruhe. Results of two such experiments have been compiled and published as benchmarks. This paper brings out the results of the numerical simulation of one of the benchmark case through a 1D/2D coupled code system, DHDYN-1D/THYC-2D and the salient features of the comparisons. Brief description of the formulations of the codes are also included.
Computer Simulation of Turbulent Flow through a Hydraulic Turbine Draft Tube
Institute of Scientific and Technical Information of China (English)
HU Ying; CHENG Heming; WANG Quanlong; YU Zhikun
2006-01-01
Based on the Navier-Stokes equations and the standard k-ε turbulence model, this paper presents the derivation of the governing equations for the turbulent flow field in a draft tube. The mathematical model for the turbulent flow through a draft tube is set up when the boundary conditions, including the inlet boundary conditions, the outlet boundary conditions and the wall boundary conditions, have been implemented. The governing equations are formulated in a discrete form on a staggered grid system by the finite volume method. The second-order central difference approximation and hybrid scheme are used for discretization. The computation and analysis on internal flow through a draft tube have been carried out by using the simplec algorithm and cfx-tasc flow software so as to obtain the simulated flow fields. The calculation results at the design operating condition for the draft tube are presented in this paper. Thereby, an effective method for simulating the internal flow field in a draft tube has been explored.
Energy Technology Data Exchange (ETDEWEB)
Kim, Jong Bum; Jeong, Ji Young; Lee, Tae Ho; Kim, Sung Kyun; Euh, Dong Jin; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2016-10-15
The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V and V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V and V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.
Energy Technology Data Exchange (ETDEWEB)
Morales S, J. B.; Sanchez J, J. [UNAM, Facultad de Ingenieria, Circuito Interior s/n, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Espinosa P, G., E-mail: jaimebmoraless@gmail.co [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)
2010-10-15
pools. This work presents Pars and steam injectors models, simulations and results of implementations in a best estimated thermal-hydraulics code, which are to be used for evaluations of these systems as long-term supports to ECCS. Lumped parameter models based on heat and mass balances have also been developed to test required additional best estimated code routines needed for the representation of Pars. (Author)
Multiphase integral reacting flow computer code (ICOMFLO): User`s guide
Energy Technology Data Exchange (ETDEWEB)
Chang, S.L.; Lottes, S.A.; Petrick, M.
1997-11-01
A copyrighted computational fluid dynamics computer code, ICOMFLO, has been developed for the simulation of multiphase reacting flows. The code solves conservation equations for gaseous species and droplets (or solid particles) of various sizes. General conservation laws, expressed by elliptic type partial differential equations, are used in conjunction with rate equations governing the mass, momentum, enthalpy, species, turbulent kinetic energy, and turbulent dissipation. Associated phenomenological submodels of the code include integral combustion, two parameter turbulence, particle evaporation, and interfacial submodels. A newly developed integral combustion submodel replacing an Arrhenius type differential reaction submodel has been implemented to improve numerical convergence and enhance numerical stability. A two parameter turbulence submodel is modified for both gas and solid phases. An evaporation submodel treats not only droplet evaporation but size dispersion. Interfacial submodels use correlations to model interfacial momentum and energy transfer. The ICOMFLO code solves the governing equations in three steps. First, a staggered grid system is constructed in the flow domain. The staggered grid system defines gas velocity components on the surfaces of a control volume, while the other flow properties are defined at the volume center. A blocked cell technique is used to handle complex geometry. Then, the partial differential equations are integrated over each control volume and transformed into discrete difference equations. Finally, the difference equations are solved iteratively by using a modified SIMPLER algorithm. The results of the solution include gas flow properties (pressure, temperature, density, species concentration, velocity, and turbulence parameters) and particle flow properties (number density, temperature, velocity, and void fraction). The code has been used in many engineering applications, such as coal-fired combustors, air
Agarwal, Sapan; Quach, Tu-Thach; Parekh, Ojas; Hsia, Alexander H.; DeBenedictis, Erik P.; James, Conrad D.; Marinella, Matthew J.; Aimone, James B.
2016-01-01
The exponential increase in data over the last decade presents a significant challenge to analytics efforts that seek to process and interpret such data for various applications. Neural-inspired computing approaches are being developed in order to leverage the computational properties of the analog, low-power data processing observed in biological systems. Analog resistive memory crossbars can perform a parallel read or a vector-matrix multiplication as well as a parallel write or a rank-1 update with high computational efficiency. For an N × N crossbar, these two kernels can be O(N) more energy efficient than a conventional digital memory-based architecture. If the read operation is noise limited, the energy to read a column can be independent of the crossbar size (O(1)). These two kernels form the basis of many neuromorphic algorithms such as image, text, and speech recognition. For instance, these kernels can be applied to a neural sparse coding algorithm to give an O(N) reduction in energy for the entire algorithm when run with finite precision. Sparse coding is a rich problem with a host of applications including computer vision, object tracking, and more generally unsupervised learning. PMID:26778946
Directory of Open Access Journals (Sweden)
Sapan eAgarwal
2016-01-01
Full Text Available The exponential increase in data over the last decade presents a significant challenge to analytics efforts that seek to process and interpret such data for various applications. Neural-inspired computing approaches are being developed in order to leverage the computational advantages of the analog, low-power data processing observed in biological systems. Analog resistive memory crossbars can perform a parallel read or a vector-matrix multiplication as well as a parallel write or a rank-1 update with high computational efficiency. For an NxN crossbar, these two kernels are at a minimum O(N more energy efficient than a digital memory-based architecture. If the read operation is noise limited, the energy to read a column can be independent of the crossbar size (O(1. These two kernels form the basis of many neuromorphic algorithms such as image, text, and speech recognition. For instance, these kernels can be applied to a neural sparse coding algorithm to give an O(N reduction in energy for the entire algorithm. Sparse coding is a rich problem with a host of applications including computer vision, object tracking, and more generally unsupervised learning.
Error threshold in topological quantum-computing models with color codes
Katzgraber, Helmut; Bombin, Hector; Martin-Delgado, Miguel A.
2009-03-01
Dealing with errors in quantum computing systems is possibly one of the hardest tasks when attempting to realize physical devices. By encoding the qubits in topological properties of a system, an inherent protection of the quantum states can be achieved. Traditional topologically-protected approaches are based on the braiding of quasiparticles. Recently, a braid-less implementation using brane-net condensates in 3-colexes has been proposed. In 2D it allows the transversal implementation of the whole Clifford group of quantum gates. In this work, we compute the error threshold for this topologically-protected quantum computing system in 2D, by means of mapping its error correction process onto a random 3-body Ising model on a triangular lattice. Errors manifest themselves as random perturbation of the plaquette interaction terms thus introducing frustration. Our results from Monte Carlo simulations suggest that these topological color codes are similarly robust to perturbations as the toric codes. Furthermore, they provide more computational capabilities and the possibility of having more qubits encoded in the quantum memory.
Chooi, K Y; Comerford, A; Sherwin, S J; Weinberg, P D
2016-06-01
The hydraulic resistances of the intima and media determine water flux and the advection of macromolecules into and across the arterial wall. Despite several experimental and computational studies, these transport processes and their dependence on transmural pressure remain incompletely understood. Here, we use a combination of experimental and computational methods to ascertain how the hydraulic permeability of the rat abdominal aorta depends on these two layers and how it is affected by structural rearrangement of the media under pressure. Ex vivo experiments determined the conductance of the whole wall, the thickness of the media and the geometry of medial smooth muscle cells (SMCs) and extracellular matrix (ECM). Numerical methods were used to compute water flux through the media. Intimal values were obtained by subtraction. A mechanism was identified that modulates pressure-induced changes in medial transport properties: compaction of the ECM leading to spatial reorganization of SMCs. This is summarized in an empirical constitutive law for permeability and volumetric strain. It led to the physiologically interesting observation that, as a consequence of the changes in medial microstructure, the relative contributions of the intima and media to the hydraulic resistance of the wall depend on the applied pressure; medial resistance dominated at pressures above approximately 93 mmHg in this vessel.
Energy Technology Data Exchange (ETDEWEB)
Farmer, M.T.; Sienicki, J.J.; Spencer, B.W.; Chu, C.C.
1992-01-01
A transient, one dimensional, finite difference computer code (MELTSPREAD-1) has been developed to predict spreading behavior of high temperature melts flowing over concrete and/or steel surfaces submerged in water, or without the effects of water if the surface is initially dry. This paper provides a summary overview of models and correlations currently implemented in the code, code validation activities completed thus far, LWR spreading-related safety issues for which the code has been applied, and the status of documentation for the code.
Energy Technology Data Exchange (ETDEWEB)
Farmer, M.T.; Sienicki, J.J.; Spencer, B.W.; Chu, C.C.
1992-04-01
A transient, one dimensional, finite difference computer code (MELTSPREAD-1) has been developed to predict spreading behavior of high temperature melts flowing over concrete and/or steel surfaces submerged in water, or without the effects of water if the surface is initially dry. This paper provides a summary overview of models and correlations currently implemented in the code, code validation activities completed thus far, LWR spreading-related safety issues for which the code has been applied, and the status of documentation for the code.
V.S.O.P. (99/09) Computer Code System for Reactor Physics and Fuel Cycle Simulation; Version 2009
Rütten, H.-J.; Haas, K. A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.
2010-01-01
V.S.O.P.(99/ 09) represents the further development of V.S.O.P.(99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of...
Automatic Generation of OpenMP Directives and Its Application to Computational Fluid Dynamics Codes
Yan, Jerry; Jin, Haoqiang; Frumkin, Michael; Yan, Jerry (Technical Monitor)
2000-01-01
The shared-memory programming model is a very effective way to achieve parallelism on shared memory parallel computers. As great progress was made in hardware and software technologies, performance of parallel programs with compiler directives has demonstrated large improvement. The introduction of OpenMP directives, the industrial standard for shared-memory programming, has minimized the issue of portability. In this study, we have extended CAPTools, a computer-aided parallelization toolkit, to automatically generate OpenMP-based parallel programs with nominal user assistance. We outline techniques used in the implementation of the tool and discuss the application of this tool on the NAS Parallel Benchmarks and several computational fluid dynamics codes. This work demonstrates the great potential of using the tool to quickly port parallel programs and also achieve good performance that exceeds some of the commercial tools.
Bolève, A.; Vandemeulebrouck, J.; Grangeon, J.
2012-11-01
In the present study, we propose the combination of two geophysical techniques, which we have applied to a dyke located in southeastern France that has a visible downstream flood area: the self-potential (SP) and hydro-acoustic methods. These methods are sensitive to two different types of signals: electric signals and water-soil pressure disturbances, respectively. The advantages of the SP technique lie in the high rate of data acquisition, which allows assessment of long dykes, and direct diagnosis in terms of leakage area delimitation and quantification. Coupled with punctual hydro-acoustic cartography, a leakage position can be precisely located, therefore allowing specific remediation decisions with regard to the results of the geophysical investigation. Here, the precise localization of leakage from an earth dyke has been identified using SP and hydro-acoustic signals, with the permeability of the preferential fluid flow area estimated by forward SP modeling. Moreover, we propose a general 'abacus' diagram for the estimation of hydraulic permeability of dyke leakage according to the magnitude of over water SP anomalies and the associated uncertainty.
On the Computational Complexity of Sphere Decoder for Lattice Space-Time Coded MIMO Channel
Abediseid, Walid
2011-01-01
The exact complexity analysis of the basic sphere decoder for general space-time codes applied to multi-input multi-output (MIMO) wireless channel is known to be difficult. In this work, we shed the light on the computational complexity of sphere decoding for the quasi-static, LAttice Space-Time (LAST) coded MIMO channel. Specifically, we derive the asymptotic tail distribution of the decoder's computational complexity in the high signal-to-noise ratio (SNR) regime. For the uncoded $M\\times N$ MIMO channel (e.g., V-BLAST), the analysis in [6] revealed that the tail distribution of such a decoder is of a Pareto-type with tail exponent that is equivalent to $N-M+1$. In our analysis, we show that the tail exponent of the sphere decoder's complexity distribution is equivalent to the diversity-multiplexing tradeoff achieved by LAST coding and lattice decoding schemes. This leads to extend the channel's tradeoff to include the decoding complexity. Moreover, we show analytically how minimum-mean square-error decisio...
Energy Technology Data Exchange (ETDEWEB)
Chung, Chang Hyun; You, Young Woo; Huh, Chang Wook; Kim, Ju Yeul; Kim Do Hyung; Kim, Yoon Ik; Yang, Hui Chang [Seoul National University, Seoul (Korea, Republic of); Jae, Moo Sung [Hansung University, Seoul (Korea, Republic of)
1997-07-01
The objective of this study is to develop the appropriate procedure that can evaluate the human error in LP/S(lower power/shutdown) and the computer code that calculate the human error probabilities(HEPs) using this framework. The assessment of applicability of the typical HRA methodologies to LP/S is conducted and a new HRA procedure, SEPLOT (Systematic Evaluation Procedure for LP/S Operation Tasks) which presents the characteristics of LP/S is developed by selection and categorization of human actions by reviewing present studies. This procedure is applied to evaluate the LOOP(Loss of Off-site Power) sequence and the HEPs obtained by using SEPLOT are used to quantitative evaluation of the core uncovery frequency. In this evaluation one of the dynamic reliability computer codes, DYLAM-3 which has the advantages against the ET/FT is used. The SEPLOT developed in this study can give the basis and arrangement as to the human error evaluation technique. And this procedure can make it possible to assess the dynamic aspects of accidents leading to core uncovery applying the HEPs obtained by using the SEPLOT as input data to DYLAM-3 code, Eventually, it is expected that the results of this study will contribute to improve safety in LP/S and reduce uncertainties in risk. 57 refs. 17 tabs., 33 figs. (author)
Calculations of reactor-accident consequences, Version 2. CRAC2: computer code user's guide
Energy Technology Data Exchange (ETDEWEB)
Ritchie, L.T.; Johnson, J.D.; Blond, R.M.
1983-02-01
The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems.
Validation of Advanced Computer Codes for VVER Technology: LB-LOCA Transient in PSB-VVER Facility
Directory of Open Access Journals (Sweden)
A. Del Nevo
2012-01-01
Full Text Available The OECD/NEA PSB-VVER project provided unique and useful experimental data for code validation from PSB-VVER test facility. This facility represents the scaled-down layout of the Russian-designed pressurized water reactor, namely, VVER-1000. Five experiments were executed, dealing with loss of coolant scenarios (small, intermediate, and large break loss of coolant accidents, a primary-to-secondary leak, and a parametric study (natural circulation test aimed at characterizing the VVER system at reduced mass inventory conditions. The comparative analysis, presented in the paper, regards the large break loss of coolant accident experiment. Four participants from three different institutions were involved in the benchmark and applied their own models and set up for four different thermal-hydraulic system codes. The benchmark demonstrated the performances of such codes in predicting phenomena relevant for safety on the basis of fixed criteria.
SEACC: the systems engineering and analysis computer code for small wind systems
Energy Technology Data Exchange (ETDEWEB)
Tu, P.K.C.; Kertesz, V.
1983-03-01
The systems engineering and analysis (SEA) computer program (code) evaluates complete horizontal-axis SWECS performance. Rotor power output as a function of wind speed and energy production at various wind regions are predicted by the code. Efficiencies of components such as gearbox, electric generators, rectifiers, electronic inverters, and batteries can be included in the evaluation process to reflect the complete system performance. Parametric studies can be carried out for blade design characteristics such as airfoil series, taper rate, twist degrees and pitch setting; and for geometry such as rotor radius, hub radius, number of blades, coning angle, rotor rpm, etc. Design tradeoffs can also be performed to optimize system configurations for constant rpm, constant tip speed ratio and rpm-specific rotors. SWECS energy supply as compared to the load demand for each hour of the day and during each session of the year can be assessed by the code if the diurnal wind and load distributions are known. Also available during each run of the code is blade aerodynamic loading information.
A fully parallel, high precision, N-body code running on hybrid computing platforms
Capuzzo-Dolcetta, R; Punzo, D
2012-01-01
We present a new implementation of the numerical integration of the classical, gravitational, N-body problem based on a high order Hermite's integration scheme with block time steps, with a direct evaluation of the particle-particle forces. The main innovation of this code (called HiGPUs) is its full parallelization, exploiting both OpenMP and MPI in the use of the multicore Central Processing Units as well as either Compute Unified Device Architecture (CUDA) or OpenCL for the hosted Graphic Processing Units. We tested both performance and accuracy of the code using up to 256 GPUs in the supercomputer IBM iDataPlex DX360M3 Linux Infiniband Cluster provided by the italian supercomputing consortium CINECA, for values of N up to 8 millions. We were able to follow the evolution of a system of 8 million bodies for few crossing times, task previously unreached by direct summation codes. The code is freely available to the scientific community.
Development of a space radiation Monte Carlo computer simulation based on the FLUKA and ROOT codes
Pinsky, L; Ferrari, A; Sala, P; Carminati, F; Brun, R
2001-01-01
This NASA funded project is proceeding to develop a Monte Carlo-based computer simulation of the radiation environment in space. With actual funding only initially in place at the end of May 2000, the study is still in the early stage of development. The general tasks have been identified and personnel have been selected. The code to be assembled will be based upon two major existing software packages. The radiation transport simulation will be accomplished by updating the FLUKA Monte Carlo program, and the user interface will employ the ROOT software being developed at CERN. The end-product will be a Monte Carlo-based code which will complement the existing analytic codes such as BRYNTRN/HZETRN presently used by NASA to evaluate the effects of radiation shielding in space. The planned code will possess the ability to evaluate the radiation environment for spacecraft and habitats in Earth orbit, in interplanetary space, on the lunar surface, or on a planetary surface such as Mars. Furthermore, it will be usef...
The Proteus Navier-Stokes code. [two and three dimensional computational fluid dynamics
Towne, Charles E.; Schwab, John R.
1992-01-01
An effort is currently underway at NASA Lewis to develop two and three dimensional Navier-Stokes codes, called Proteus, for aerospace propulsion applications. Proteus solves the Reynolds-averaged, unsteady, compressible Navier-Stokes equations in strong conservation law form. Turbulence is modeled using a Baldwin-Lomax based algebraic eddy viscosity model. In addition, options are available to solve thin layer or Euler equations, and to eliminate the energy equation by assuming constant stagnation enthalpy. An extensive series of validation cases have been run, primarily using the two dimensional planar/axisymmetric version of the code. Several flows were computed that have exact solution such as: fully developed channel and pipe flow; Couette flow with and without pressure gradients; unsteady Couette flow formation; flow near a suddenly accelerated flat plate; flow between concentric rotating cylinders; and flow near a rotating disk. The two dimensional version of the Proteus code has been released, and the three dimensional code is scheduled for release in late 1991.
Institute of Scientific and Technical Information of China (English)
刘志弢; 秦本科; 解衡; 王炳华
2009-01-01
比较分析了目前世界上典型的压水堆核电站热工水力系统程序的研发历程、发展现状、应用范围,着重指出了最佳估算、程序耦合、程序评估在热工水力系统程序研发中的重要作用,阐述了各国热工水力系统程序研发模式对我国自主创新的借鉴意义.%Research and development of thermal-hydraulic system codes for nuclear power plants with pressurized water reactors were analyzed on their history, status and application ranges. The important roles of best-estimate methodology, codes coupling and codes qualification were pointed out. The development models of thermal-hydraulic system codes around the world provide references to China's self-innovation.
Energy Technology Data Exchange (ETDEWEB)
Wulff, W; Cheng, H S; Diamond, D J; Khatib-Rahbar, M
1984-01-01
This report documents the physical models and the numerical methods employed in the BWR systems code RAMONA-3B. The RAMONA-3B code simulates three-dimensional neutron kinetics and multichannel core hydraulics of nonhomogeneous, nonequilibrium two-phase flows. RAMONA-3B is programmed to calculate the steady and transient conditions in the main steam supply system for normal and abnormal operational transients, including the performances of plant control and protection systems. Presented are code capabilities and limitations, models and solution techniques, the results of development code assessment and suggestions for improving the code in the future.
Revised uranium--plutonium cycle PWR and BWR models for the ORIGEN computer code
Energy Technology Data Exchange (ETDEWEB)
Croff, A. G.; Bjerke, M. A.; Morrison, G. W.; Petrie, L. M.
1978-09-01
Reactor physics calculations and literature searches have been conducted, leading to the creation of revised enriched-uranium and enriched-uranium/mixed-oxide-fueled PWR and BWR reactor models for the ORIGEN computer code. These ORIGEN reactor models are based on cross sections that have been taken directly from the reactor physics codes and eliminate the need to make adjustments in uncorrected cross sections in order to obtain correct depletion results. Revised values of the ORIGEN flux parameters THERM, RES, and FAST were calculated along with new parameters related to the activation of fuel-assembly structural materials not located in the active fuel zone. Recommended fuel and structural material masses and compositions are presented. A summary of the new ORIGEN reactor models is given.
A general panel sizing computer code and its application to composite structural panels
Anderson, M. S.; Stroud, W. J.
1978-01-01
A computer code for obtaining the dimensions of optimum (least mass) stiffened composite structural panels is described. The procedure, which is based on nonlinear mathematical programming and a rigorous buckling analysis, is applicable to general cross sections under general loading conditions causing buckling. A simplified method of accounting for bow-type imperfections is also included. Design studies in the form of structural efficiency charts for axial compression loading are made with the code for blade and hat stiffened panels. The effects on panel mass of imperfections, material strength limitations, and panel stiffness requirements are also examined. Comparisons with previously published experimental data show that accounting for imperfections improves correlation between theory and experiment.
Development of system of computer codes for severe accident analysis and its applications
Energy Technology Data Exchange (ETDEWEB)
Jang, H. S.; Jeon, M. H.; Cho, N. J. and others [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)
1992-01-15
The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in nuclear power plants. This system of codes is necessary to conduct Individual Plant Examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident-resistance. Severe accident can be mitigated by the proper accident management strategies. Some operator action for mitigation can lead to more disastrous result and thus uncertain severe accident phenomena must be well recognized. There must be further research for development of severe accident management strategies utilizing existing plant resources as well as new design concepts.
Institute of Scientific and Technical Information of China (English)
彭小东; 杨朝晖; 刘善均; 鞠小明
2014-01-01
For the metal spiral casing of water turbines, a new equivalent pipe algorithm is developed based on the idea of equiangu-lar spiral. Prototype tests and computations are carried out to investigate the hydraulic transient characteristics. The computation re-sults by using the new model are in a good agreement with the prototype test data with respect to the maximum speed of the tur-bine-generator unit, the maximum water hammer pressure in the spiral casing and the maximum vacuum in the draft tube. The propo-sed method is a significant improvement over the conventional algorithm with the accuracy increased and the error reduced by about 3%.
Energy Technology Data Exchange (ETDEWEB)
Müller, C.; Hughes, E. D.; Niederauer, G. F.; Wilkening, H.; Travis, J. R.; Spore, J. W.; Royl, P.; Baumann, W.
1998-10-01
Los Alamos National Laboratory (LANL) and Forschungszentrum Karlsruhe (FzK) are developing GASFLOW, a three-dimensional (3D) fluid dynamics field code as a best- estimate tool to characterize local phenomena within a flow field. Examples of 3D phenomena include circulation patterns; flow stratification; hydrogen distribution mixing and stratification; combustion and flame propagation; effects of noncondensable gas distribution on local condensation and evaporation; and aerosol entrainment, transport, and deposition. An analysis with GASFLOW will result in a prediction of the gas composition and discrete particle distribution in space and time throughout the facility and the resulting pressure and temperature loadings on the walls and internal structures with or without combustion. A major application of GASFLOW is for predicting the transport, mixing, and combustion of hydrogen and other gases in nuclear reactor containment and other facilities. It has been applied to situations involving transporting and distributing combustible gas mixtures. It has been used to study gas dynamic behavior in low-speed, buoyancy-driven flows, as well as sonic flows or diffusion dominated flows; and during chemically reacting flows, including deflagrations. The effects of controlling such mixtures by safety systems can be analyzed. The code version described in this manual is designated GASFLOW 2.1, which combines previous versions of the United States Nuclear Regulatory Commission code HMS (for Hydrogen Mixing Studies) and the Department of Energy and FzK versions of GASFLOW. The code was written in standard Fortran 90. This manual comprises three volumes. Volume I describes the governing physical equations and computational model. Volume II describes how to use the code to set up a model geometry, specify gas species and material properties, define initial and boundary conditions, and specify different outputs, especially graphical displays. Sample problems are included. Volume
Horizontal steam generator PGV-1000 thermal-hydraulic analysis
Energy Technology Data Exchange (ETDEWEB)
Ubra, O. [Skoda Company, Prague (Switzerland); Doubek, M. [Czech Technical Univ., Prague (Switzerland)
1995-12-31
A computer program for the steady state thermal-hydraulic analysis of horizontal steam generator PGV-1000 is presented. The program provides the capability to analyze steam generator PGV-1000 primary side flow and temperature distribution, primary side pressure drops, heat transfer between the primary and secondary sides and multidimensional heat flux distribution. A special attention is paid to the thermal-hydraulics of the secondary side. The code predicts 3-D distribution of the void fraction at the secondary side, mass redistribution under the submerged perforated sheet and the steam generator level profile. By means of developed computer program a detailed thermal-hydraulic study of the PGV-1000 has been carried out. A wide range of calculations has been performed and a set of important steam generator characteristics has been obtained. Some of them are presented in the paper. (orig.). 5 refs.
DEFF Research Database (Denmark)
Mohebbi, Ali; Engelsholm, Signe K.D.; Puthusserypady, Sadasivan
2015-01-01
In this pilot study, a novel and minimalistic Brain Computer Interface (BCI) based wheelchair control application was developed. The system was based on pseudorandom code modulated Visual Evoked Potentials (c-VEPs). The visual stimuli in the scheme were generated based on the Gold code...
DEFF Research Database (Denmark)
Sessarego, Matias; Ramos García, Néstor; Sørensen, Jens Nørkær
2017-01-01
Aerodynamic and structural dynamic performance analysis of modern wind turbines are routinely estimated in the wind energy field using computational tools known as aeroelastic codes. Most aeroelastic codes use the blade element momentum (BEM) technique to model the rotor aerodynamics and a modal...
Prediction of detonation and JWL eos parameters of energetic materials using EXPLO5 computer code
CSIR Research Space (South Africa)
Peter, Xolani
2016-09-01
Full Text Available (Cowperthwaite and Zwisler, 1976), CHEETAH (Fried, 1996), EXPLO5(Sućeska , 2001), BARUT-X (Cengiz et al., 2007). These computer codes describe the detonation on the basis of the solution of Euler’s hydrodynamic equation based on the description of an equation... of detonation products equation of state from cylinder test: Analytical model and numerical analysis. Thermal Science, 19(1), pp. 35-48. Fried, L.E., 1996. CHEETAH 1.39 user’s manual. Lawrence Livermore National Laboratory. Göbel, M., 2009. Energetic...
On the application of computational fluid dynamics codes for liquefied natural gas dispersion.
Luketa-Hanlin, Anay; Koopman, Ronald P; Ermak, Donald L
2007-02-20
Computational fluid dynamics (CFD) codes are increasingly being used in the liquefied natural gas (LNG) industry to predict natural gas dispersion distances. This paper addresses several issues regarding the use of CFD for LNG dispersion such as specification of the domain, grid, boundary and initial conditions. A description of the k-epsilon model is presented, along with modifications required for atmospheric flows. Validation issues pertaining to the experimental data from the Burro, Coyote, and Falcon series of LNG dispersion experiments are also discussed. A description of the atmosphere is provided as well as discussion on the inclusion of the Coriolis force to model very large LNG spills.
Discrete logarithm computations over finite fields using Reed-Solomon codes
Augot, Daniel; Morain, François
2012-01-01
Cheng and Wan have related the decoding of Reed-Solomon codes to the computation of discrete logarithms over finite fields, with the aim of proving the hardness of their decoding. In this work, we experiment with solving the discrete logarithm over GF(q^h) using Reed-Solomon decoding. For fixed h and q going to infinity, we introduce an algorithm (RSDL) needing O~(h! q^2) operations over GF(q), operating on a q x q matrix with (h+2) q non-zero coefficients. We give faster variants including a...
Resin Matrix/Fiber Reinforced Composite Material, Ⅱ: Method of Solution and Computer Code
Institute of Scientific and Technical Information of China (English)
Li Chensha(李辰砂); Jiao Caishan; Liu Ying; Wang Zhengping; Wang Hongjie; Cao Maosheng
2003-01-01
According to a mathematical model which describes the curing process of composites constructed from continuous fiber-reinforced, thermosetting resin matrix prepreg materials, and the consolidation of the composites, the solution method to the model is made and a computer code is developed, which for flat-plate composites cured by a specified cure cycle, provides the variation of temperature distribution, the cure reaction process in the resin, the resin flow and fibers stress inside the composite, the void variation and the residual stress distribution.
Fuel burnup analysis for Thai research reactor by using MCNPX computer code
Sangkaew, S.; Angwongtrakool, T.; Srimok, B.
2017-06-01
This paper presents the fuel burnup analysis of the Thai research reactor (TRR-1/M1), TRIGA Mark-III, operated by Thailand Institute of Nuclear Technology (TINT) in Bangkok, Thailand. The modelling software used in this analysis is MCNPX (MCNP eXtended) version 2.6.0, a Fortran90 Monte Carlo radiation transport computer code. The analysis results will cover the core excess reactivity, neutron fluxes at the irradiation positions and neutron detector tubes, power distribution, fuel burnup, and fission products based on fuel cycle of first reactor core arrangement.
Reznik, A. L.; Tuzikov, A. V.; Solov'ev, A. A.; Torgov, A. V.
2016-11-01
Original codes and combinatorial-geometrical computational schemes are presented, which are developed and applied for finding exact analytical formulas that describe the probability of errorless readout of random point images recorded by a scanning aperture with a limited number of threshold levels. Combinatorial problems encountered in the course of the study and associated with the new generalization of Catalan numbers are formulated and solved. An attempt is made to find the explicit analytical form of these numbers, which is, on the one hand, a necessary stage of solving the basic research problem and, on the other hand, an independent self-consistent problem.
Apparatus, Method, and Computer Program for a Resolution-Enhanced Pseudo-Noise Code Technique
Li, Steven X. (Inventor)
2015-01-01
An apparatus, method, and computer program for a resolution enhanced pseudo-noise coding technique for 3D imaging is provided. In one embodiment, a pattern generator may generate a plurality of unique patterns for a return to zero signal. A plurality of laser diodes may be configured such that each laser diode transmits the return to zero signal to an object. Each of the return to zero signal includes one unique pattern from the plurality of unique patterns to distinguish each of the transmitted return to zero signals from one another.
Institute of Scientific and Technical Information of China (English)
无
2001-01-01
The fundamental algorithm of light beam propagation in high powerlaser system is investigated and the corresponding computational codes are given. It is shown that the number of modulation ring due to the diffraction is related to the size of the pinhole in spatial filter (in terms of the times of diffraction limitation, i.e. TDL) and the Fresnel number of the laser system; for the complex laser system with multi-spatial filters and free space, the system can be investigated by the reciprocal rule of operators.
Modeling of field lysimeter release data using the computer code dust
Energy Technology Data Exchange (ETDEWEB)
Sullivan, T.M.; Fitzgerald, I.T. (Brookhaven National Lab., Upton, NY (United States)); McConnell, J.W.; Rogers, R.D. (Idaho National Engineering Lab., Idaho Falls, ID (United States))
1993-01-01
In this study, it was attempted to match the experimentally measured mass release data collected over a period of seven years by investigators from Idaho National Engineering Laboratory from the lysimeters at Oak Ridge National Laboratory and Argonne National Laboratory using the computer code DUST. The influence of the dispersion coefficient and distribution coefficient on mass release was investigated. Both were found to significantly influence mass release over the seven year period. It is recommended that these parameters be measured on a site specific basis to enhance the understanding of the system.
Modeling of field lysimeter release data using the computer code dust
Energy Technology Data Exchange (ETDEWEB)
Sullivan, T.M.; Fitzgerald, I.T. [Brookhaven National Lab., Upton, NY (United States); McConnell, J.W.; Rogers, R.D. [Idaho National Engineering Lab., Idaho Falls, ID (United States)
1993-03-01
In this study, it was attempted to match the experimentally measured mass release data collected over a period of seven years by investigators from Idaho National Engineering Laboratory from the lysimeters at Oak Ridge National Laboratory and Argonne National Laboratory using the computer code DUST. The influence of the dispersion coefficient and distribution coefficient on mass release was investigated. Both were found to significantly influence mass release over the seven year period. It is recommended that these parameters be measured on a site specific basis to enhance the understanding of the system.
Issues and future direction of thermal-hydraulics research and development in nuclear power reactors
Energy Technology Data Exchange (ETDEWEB)
Saha, P., E-mail: pradip.saha@ge.com [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Aksan, N. [GRNSPG Group, University of Pisa (Italy); Andersen, J. [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Yan, J. [Westinghouse Electric Co., Columbia, SC (United States); Simoneau, J.P. [AREVA, Lyon (France); Leung, L. [Atomic Energy of Canada Ltd., Chalk River, Ontario (Canada); Bertrand, F. [CEA, DEN, DER, F-13108 Saint-Paul-Lez-Durance (France); Aoto, K.; Kamide, H. [Japan Atomic Energy Agency, Chiyoda-ku, Tokyo (Japan)
2013-11-15
The paper archives the proceedings of an expert panel discussion on the issues and future direction of thermal-hydraulic research and development in nuclear power reactors held at the NURETH-14 conference in Toronto, Canada, in September 2011. Thermal-hydraulic issues related to both operating and advanced reactors are presented. Advances in thermal-hydraulics have significantly improved the performance of operating reactors. Further thermal-hydraulics research and development is continuing in both experimental and computational areas for operating reactors, reactors under construction or ready for near-term deployment, and advanced Generation-IV reactors. As the computing power increases, the fine-scale multi-physics computational models, coupled with the systems analysis code, are expected to provide answers to many challenging problems in both operating and advanced reactor designs.
Energy Technology Data Exchange (ETDEWEB)
Nichols, B.D.; Mueller, C.; Necker, G.A.; Travis, J.R.; Spore, J.W.; Lam, K.L.; Royl, P.; Redlinger, R.; Wilson, T.L.
1998-10-01
Los Alamos National Laboratory (LANL) and Forschungszentrum Karlsruhe (FzK) are developing GASFLOW, a three-dimensional (3D) fluid dynamics field code as a best-estimate tool to characterize local phenomena within a flow field. Examples of 3D phenomena include circulation patterns; flow stratification; hydrogen distribution mixing and stratification; combustion and flame propagation; effects of noncondensable gas distribution on local condensation and evaporation; and aerosol entrainment, transport, and deposition. An analysis with GASFLOW will result in a prediction of the gas composition and discrete particle distribution in space and time throughout the facility and the resulting pressure and temperature loadings on the walls and internal structures with or without combustion. A major application of GASFLOW is for predicting the transport, mixing, and combustion of hydrogen and other gases in nuclear reactor containments and other facilities. It has been applied to situations involving transporting and distributing combustible gas mixtures. It has been used to study gas dynamic behavior (1) in low-speed, buoyancy-driven flows, as well as sonic flows or diffusion dominated flows; and (2) during chemically reacting flows, including deflagrations. The effects of controlling such mixtures by safety systems can be analyzed. The code version described in this manual is designated GASFLOW 2.1, which combines previous versions of the United States Nuclear Regulatory Commission code HMS (for Hydrogen Mixing Studies) and the Department of Energy and FzK versions of GASFLOW. The code was written in standard Fortran 90. This manual comprises three volumes. Volume I describes the governing physical equations and computational model. Volume II describes how to use the code to set up a model geometry, specify gas species and material properties, define initial and boundary conditions, and specify different outputs, especially graphical displays. Sample problems are included
Digital Poetry: A Narrow Relation between Poetics and the Codes of the Computational Logic
Laurentiz, Silvia
The project "Percorrendo Escrituras" (Walking Through Writings Project) has been developed at ECA-USP Fine Arts Department. Summarizing, it intends to study different structures of digital information that share the same universe and are generators of a new aesthetics condition. The aim is to search which are the expressive possibilities of the computer among the algorithm functions and other of its specific properties. It is a practical, theoretical and interdisciplinary project where the study of programming evolutionary language, logic and mathematics take us to poetic experimentations. The focus of this research is the digital poetry, and it comes from poetics of permutation combinations and culminates with dynamic and complex systems, autonomous, multi-user and interactive, through agents generation derivations, filtration and emergent standards. This lecture will present artworks that use some mechanisms introduced by cybernetics and the notion of system in digital poetry that demonstrate the narrow relationship between poetics and the codes of computational logic.
1975-01-01
Major developments are examined which have taken place to date in the analysis of the power and energy demands on the APU/Hydraulic/Actuator Subsystem for space shuttle during the entry-to-touchdown (not including rollout) flight regime. These developments are given in the form of two subroutines which were written for use with the Space Shuttle Functional Simulator. The first subroutine calculates the power and energy demand on each of the three hydraulic systems due to control surface (inboard/outboard elevons, rudder, speedbrake, and body flap) activity. The second subroutine incorporates the R. I. priority rate limiting logic which limits control surface deflection rates as a function of the number of failed hydraulic. Typical results of this analysis are included, and listings of the subroutines are presented in appendicies.
Chen, Sheng-Hong
2015-01-01
This book discusses in detail the planning, design, construction and management of hydraulic structures, covering dams, spillways, tunnels, cut slopes, sluices, water intake and measuring works, ship locks and lifts, as well as fish ways. Particular attention is paid to considerations concerning the environment, hydrology, geology and materials etc. in the planning and design of hydraulic projects. It also considers the type selection, profile configuration, stress/stability calibration and engineering countermeasures, flood releasing arrangements and scouring protection, operation and maintenance etc. for a variety of specific hydraulic structures. The book is primarily intended for engineers, undergraduate and graduate students in the field of civil and hydraulic engineering who are faced with the challenges of extending our understanding of hydraulic structures ranging from traditional to groundbreaking, as well as designing, constructing and managing safe, durable hydraulic structures that are economical ...
Determination of thermal-hydraulic loads on reactor internals in a DBA-situation
Energy Technology Data Exchange (ETDEWEB)
Ville Lestinen; Timo Toppila [POB 10, 00048 FORTUM (Finland)
2005-07-01
Full text of publication follows: According to Finnish regulatory requirements, reactor internals have to stay intact in a design basis accident (DBA) situation, so that control rods can still penetrate into the core. To fulfill this demand some criteria must be followed in periodical in-service inspections. This is the motivation for studying and developing more detailed methods for analysis of thermal-hydraulic loads on reactor internals during the DBA-situation for the Loviisa NPP in Finland. The objective of this research program is to connect thermal-hydraulic and mechanical analysis methods with the goal to produce a reliable method for determination of thermal-hydraulic and mechanical loads on reactor internals in the accident situation. The tools studied are thermal-hydraulic system codes, computational fluid dynamics (CFD) codes and finite element analysis (FEA) codes. This paper concentrates mainly on thermal-hydraulic part of the research, but also the mechanical aspects are discussed. Firstly, the paper includes a short literary review of the available methods to analyse the described problem including both thermal-hydraulic and structural analysis parts. Secondly, different possibilities to carry out thermal-hydraulic analyses have been studied. The DBA-case includes complex physical phenomena and therefore modelling is difficult. The accident situation can be for example LLOCA. When the pipe has broken, the pressure decreases and water starts to evaporate, which consumes energy and that way limits the pressure decrease. After some period of time, the system reaches a new equilibrium state. To perform exact thermal-hydraulic analysis also two phase phenomena must be included. Therefore CFD codes are not capable of modelling the DBA situation very well, but the use of CFD codes requires that the effect of two phase flow must be added somehow. One method to calculate two phase phenomena with CFD codes is to use thermal-hydraulic system codes to calculate
An implementation of a tree code on a SIMD, parallel computer
Olson, Kevin M.; Dorband, John E.
1994-01-01
We describe a fast tree algorithm for gravitational N-body simulation on SIMD parallel computers. The tree construction uses fast, parallel sorts. The sorted lists are recursively divided along their x, y and z coordinates. This data structure is a completely balanced tree (i.e., each particle is paired with exactly one other particle) and maintains good spatial locality. An implementation of this tree-building algorithm on a 16k processor Maspar MP-1 performs well and constitutes only a small fraction (approximately 15%) of the entire cycle of finding the accelerations. Each node in the tree is treated as a monopole. The tree search and the summation of accelerations also perform well. During the tree search, node data that is needed from another processor is simply fetched. Roughly 55% of the tree search time is spent in communications between processors. We apply the code to two problems of astrophysical interest. The first is a simulation of the close passage of two gravitationally, interacting, disk galaxies using 65,636 particles. We also simulate the formation of structure in an expanding, model universe using 1,048,576 particles. Our code attains speeds comparable to one head of a Cray Y-MP, so single instruction, multiple data (SIMD) type computers can be used for these simulations. The cost/performance ratio for SIMD machines like the Maspar MP-1 make them an extremely attractive alternative to either vector processors or large multiple instruction, multiple data (MIMD) type parallel computers. With further optimizations (e.g., more careful load balancing), speeds in excess of today's vector processing computers should be possible.
Wavelet subband coding of computer simulation output using the A++ array class library
Energy Technology Data Exchange (ETDEWEB)
Bradley, J.N.; Brislawn, C.M.; Quinlan, D.J.; Zhang, H.D. [Los Alamos National Lab., NM (United States); Nuri, V. [Washington State Univ., Pullman, WA (United States). School of EECS
1995-07-01
The goal of the project is to produce utility software for off-line compression of existing data and library code that can be called from a simulation program for on-line compression of data dumps as the simulation proceeds. Naturally, we would like the amount of CPU time required by the compression algorithm to be small in comparison to the requirements of typical simulation codes. We also want the algorithm to accomodate a wide variety of smooth, multidimensional data types. For these reasons, the subband vector quantization (VQ) approach employed in has been replaced by a scalar quantization (SQ) strategy using a bank of almost-uniform scalar subband quantizers in a scheme similar to that used in the FBI fingerprint image compression standard. This eliminates the considerable computational burdens of training VQ codebooks for each new type of data and performing nearest-vector searches to encode the data. The comparison of subband VQ and SQ algorithms in indicated that, in practice, there is relatively little additional gain from using vector as opposed to scalar quantization on DWT subbands, even when the source imagery is from a very homogeneous population, and our subjective experience with synthetic computer-generated data supports this stance. It appears that a careful study is needed of the tradeoffs involved in selecting scalar vs. vector subband quantization, but such an analysis is beyond the scope of this paper. Our present work is focused on the problem of generating wavelet transform/scalar quantization (WSQ) implementations that can be ported easily between different hardware environments. This is an extremely important consideration given the great profusion of different high-performance computing architectures available, the high cost associated with learning how to map algorithms effectively onto a new architecture, and the rapid rate of evolution in the world of high-performance computing.
Computer simulation of Angra-2 PWR nuclear reactor core using MCNPX code
Energy Technology Data Exchange (ETDEWEB)
Medeiros, Marcos P.C. de; Rebello, Wilson F., E-mail: eng.cavaliere@ime.eb.br, E-mail: rebello@ime.eb.br [Instituto Militar de Engenharia - Secao de Engenharia Nuclear, Rio de Janeiro, RJ (Brazil); Oliveira, Claudio L. [Universidade Gama Filho, Departamento de Matematica, Rio de Janeiro, RJ (Brazil); Vellozo, Sergio O., E-mail: vellozo@cbpf.br [Centro Tecnologico do Exercito. Divisao de Defesa Quimica, Biologica e Nuclear, Rio de Janeiro, RJ (Brazil); Silva, Ademir X. da, E-mail: ademir@nuclear.ufrj.br [Coordenacao dos Programas de Pos Gaduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil)
2011-07-01
In this work the MCNPX (Monte Carlo N-Particle Transport Code) code was used to develop a computerized model of the core of Angra 2 PWR (Pressurized Water Reactor) nuclear reactor. The model was created without any kind of homogenization, but using real geometric information and material composition of that reactor, obtained from the FSAR (Final Safety Analysis Report). The model is still being improved and the version presented in this work is validated by comparing values calculated by MCNPX with results calculated by others means and presented on FSAR. This paper shows the results already obtained to K{sub eff} and K{infinity}, general parameters of the core, considering the reactor operating under stationary conditions of initial testing and operation. Other stationary operation conditions have been simulated and, in all tested cases, there was a close agreement between values calculated computationally through this model and data presented on the FSAR, which were obtained by other codes. This model is expected to become a valuable tool for many future applications. (author)
Development of computer code models for analysis of subassembly voiding in the LMFBR
Energy Technology Data Exchange (ETDEWEB)
Hinkle, W [ed.
1979-12-01
The research program discussed in this report was started in FY1979 under the combined sponsorship of the US Department of Energy (DOE), General Electric (GE) and Hanford Engineering Development Laboratory (HEDL). The objective of the program is to develop multi-dimensional computer codes which can be used for the analysis of subassembly voiding incoherence under postulated accident conditions in the LMFBR. Two codes are being developed in parallel. The first will use a two fluid (6 equation) model which is more difficult to develop but has the potential for providing a code with the utmost in flexibility and physical consistency for use in the long term. The other will use a mixture (< 6 equation) model which is less general but may be more amenable to interpretation and use of experimental data and therefore, easier to develop for use in the near term. To assure that the models developed are not design dependent, geometries and transient conditions typical of both foreign and US designs are being considered.
Sodium combustion computer code ASSCOPS version 2.0 user`s manual
Energy Technology Data Exchange (ETDEWEB)
Ishikawa, Hiroyasu; Futagami, Satoshi; Ohno, Shuji; Seino, Hiroshi; Miyake, Osamu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center
1997-12-01
ASSCOPS (Analysis of Simultaneous Sodium Combustion in Pool and Spray) has been developed for analyses of thermal consequences of sodium leak and fire accidents in LMFBRs. This report presents a description of the computational models, input, and output as the user`s manual of ASSCOPS version 2.0. ASSCOPS is an integrated code based on the sodium pool fire code SOFIRE II developed by the Atomics International Division of Rockwell International, and the sodium spray fire code SPRAY developed by the Hanford Engineering Development Laboratory in the U.S. The experimental studies conducted at PNC have been reflected in the ASSCOPS improvement. The users of ASSCOPS need to specify the sodium leak conditions (leak flow rate and temperature, etc.), the cell geometries (volume and structure surface area and thickness, etc.), and the atmospheric initial conditions, such as gas temperature, pressure, and gas composition. ASSCOPS calculates the time histories of atmospheric pressure and temperature changes along with those of the structural temperatures. (author)
Hierarchical surface code for network quantum computing with modules of arbitrary size
Li, Ying; Benjamin, Simon C.
2016-10-01
The network paradigm for quantum computing involves interconnecting many modules to form a scalable machine. Typically it is assumed that the links between modules are prone to noise while operations within modules have a significantly higher fidelity. To optimize fault tolerance in such architectures we introduce a hierarchical generalization of the surface code: a small "patch" of the code exists within each module and constitutes a single effective qubit of the logic-level surface code. Errors primarily occur in a two-dimensional subspace, i.e., patch perimeters extruded over time, and the resulting noise threshold for intermodule links can exceed ˜10 % even in the absence of purification. Increasing the number of qubits within each module decreases the number of qubits necessary for encoding a logical qubit. But this advantage is relatively modest, and broadly speaking, a "fine-grained" network of small modules containing only about eight qubits is competitive in total qubit count versus a "course" network with modules containing many hundreds of qubits.
Wiegand, D.E.
1962-05-01
A hydraulic servo is designed in which a small pressure difference produced at two orifices by an electrically operated flapper arm in a constantly flowing hydraulic loop is hydraulically amplified by two constant flow pumps, two additional orifices, and three unconnected ball pistons. Two of the pistons are of one size and operate against the additional orifices, and the third piston is of a different size and operates between and against the first two pistons. (AEC)
Institute of Scientific and Technical Information of China (English)
刘伟; 朱元兵; 白宁; 单建强; 张博; 苟军利; 厉井钢
2014-01-01
GE3×3 test bundle experiments were simulated with sub-channel analysis code ATHAS.Comparisons of the obtained results by ATHAS code with the experimental measurements and other sub-channel codes show that ATHAS is capable to predict thermal-hydraulic parameters distribution in GE3 ×3 components accurately.All of this demonstrates the reasonable physical models and powerful application functions of ATHAS.The work of this thesis can be taken example by the design and development of thermal-hydraulic program of nuclear power plant in China.%利用具有自主知识产权的子通道程序 ATHAS对 GE3×3组件进行稳态计算，并将 ATHAS的预测值与实验测量值及其他子通道程序的预测值进行了对比分析，结果表明：ATHAS 能够准确预测GE3×3组件内的热工水力参数分布，展示了 ATHAS可靠的物理模型。本文对 ATHAS 进行稳态验证的思路和方法，对我国核电站热工水力软件自主化的设计开发具有借鉴意义。
Institute of Scientific and Technical Information of China (English)
M. Garbey; C. Picard
2008-01-01
The goal of this paper is to present a versatile framework for solution verification of PDE's.We first generalize the Richardson Extrapolation technique to an optimized extrapolation solution procedure that constructs the best consistent solution from a set of two or three coarse grid solution in the discrete norm of choice. This technique generalizes the Least Square Extrapolation method introduced by one of the author and W. Shyy. We second establish the conditioning number of the problem in a reduced space that approximates the main feature of the numerical solution thanks to a sensitivity analysis. Overall our method produces an a posteriori error estimation in this reduced space of approximation. The key feature of our method is that our construction does not require an internal knowledge of the software neither the source code that produces the solution to be verified. It can be applied in principle as a postprocessing procedure to off the shelf commercial code. We demonstrate the robustness of our method with two steady problems that are separately an incompressible back step flow test case and a heat transfer problem for a battery. Our error estimate might be ultimately verified with a near by manufactured solution. While our procedure is systematic and requires numerous computation of residuals, one can take advantage of distributed computing to get quickly the error estimate.
DCS Hydraulics Submittal, Butler County, Alabama, USA
Federal Emergency Management Agency, Department of Homeland Security — Hydraulics data includes spatial datasets and data tables necessary for documenting the hydraulic procedures for computing flood elevations for a flood insurance...
DCS Hydraulics Submittal, Bullock County, Alabama, USA
Federal Emergency Management Agency, Department of Homeland Security — Hydraulics data includes spatial datasets and data tables necessary for documenting the hydraulic procedures for computing flood elevations for a flood insurance...
DCS Hydraulics Submittal, Covington County, Alabama, USA
Federal Emergency Management Agency, Department of Homeland Security — Hydraulics data includes spatial datasets and data tables necessary for documenting the hydraulic procedures for computing flood elevations for a flood insurance...
Computer code to predict the heat of explosion of high energy materials
Energy Technology Data Exchange (ETDEWEB)
Muthurajan, H. [Armament Research and Development Establishment, Pashan, Pune 411021 (India)], E-mail: muthurajan_h@rediffmail.com; Sivabalan, R.; Pon Saravanan, N.; Talawar, M.B. [High Energy Materials Research Laboratory, Sutarwadi, Pune 411 021 (India)
2009-01-30
The computational approach to the thermochemical changes involved in the process of explosion of a high energy materials (HEMs) vis-a-vis its molecular structure aids a HEMs chemist/engineers to predict the important thermodynamic parameters such as heat of explosion of the HEMs. Such a computer-aided design will be useful in predicting the performance of a given HEM as well as in conceiving futuristic high energy molecules that have significant potential in the field of explosives and propellants. The software code viz., LOTUSES developed by authors predicts various characteristics of HEMs such as explosion products including balanced explosion reactions, density of HEMs, velocity of detonation, CJ pressure, etc. The new computational approach described in this paper allows the prediction of heat of explosion ({delta}H{sub e}) without any experimental data for different HEMs, which are comparable with experimental results reported in literature. The new algorithm which does not require any complex input parameter is incorporated in LOTUSES (version 1.5) and the results are presented in this paper. The linear regression analysis of all data point yields the correlation coefficient R{sup 2} = 0.9721 with a linear equation y = 0.9262x + 101.45. The correlation coefficient value 0.9721 reveals that the computed values are in good agreement with experimental values and useful for rapid hazard assessment of energetic materials.
Computer code to predict the heat of explosion of high energy materials.
Muthurajan, H; Sivabalan, R; Pon Saravanan, N; Talawar, M B
2009-01-30
The computational approach to the thermochemical changes involved in the process of explosion of a high energy materials (HEMs) vis-à-vis its molecular structure aids a HEMs chemist/engineers to predict the important thermodynamic parameters such as heat of explosion of the HEMs. Such a computer-aided design will be useful in predicting the performance of a given HEM as well as in conceiving futuristic high energy molecules that have significant potential in the field of explosives and propellants. The software code viz., LOTUSES developed by authors predicts various characteristics of HEMs such as explosion products including balanced explosion reactions, density of HEMs, velocity of detonation, CJ pressure, etc. The new computational approach described in this paper allows the prediction of heat of explosion (DeltaH(e)) without any experimental data for different HEMs, which are comparable with experimental results reported in literature. The new algorithm which does not require any complex input parameter is incorporated in LOTUSES (version 1.5) and the results are presented in this paper. The linear regression analysis of all data point yields the correlation coefficient R(2)=0.9721 with a linear equation y=0.9262x+101.45. The correlation coefficient value 0.9721 reveals that the computed values are in good agreement with experimental values and useful for rapid hazard assessment of energetic materials.
Implementation of discrete transfer radiation method into swift computational fluid dynamics code
Directory of Open Access Journals (Sweden)
Baburić Mario
2004-01-01
Full Text Available The Computational Fluid Dynamics (CFD has developed into a powerful tool widely used in science, technology and industrial design applications, when ever fluid flow, heat transfer, combustion, or other complicated physical processes, are involved. During decades of development of CFD codes scientists were writing their own codes, that had to include not only the model of processes that were of interest, but also a whole spectrum of necessary CFD procedures, numerical techniques, pre-processing and post-processing. That has arrested much of the scientist effort in work that has been copied many times over, and was not actually producing the added value. The arrival of commercial CFD codes brought relief to many engineers that could now use the user-function approach for mod el ling purposes, en trusting the application to do the rest of the work. This pa per shows the implementation of Discrete Transfer Radiation Method into AVL’s commercial CFD code SWIFT with the help of user defined functions. Few standard verification test cases were per formed first, and in order to check the implementation of the radiation method it self, where the comparisons with available analytic solution could be performed. After wards, the validation was done by simulating the combustion in the experimental furnace at IJmuiden (Netherlands, for which the experimental measurements were available. The importance of radiation prediction in such real-size furnaces is proved again to be substantial, where radiation itself takes the major fraction of over all heat transfer. The oil-combustion model used in simulations was the semi-empirical one that has been developed at the Power Engineering Department, and which is suit able for a wide range of typical oil flames.
Energy Technology Data Exchange (ETDEWEB)
NONE
1997-03-01
This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U.S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This volume consists of the section of the manual dealing with three of the functional modules in the code. Those are the Morse-SGC for the SCALE system, Heating 7.2, and KENO V.a. The manual describes the latest released versions of the codes.
Automated Development of Accurate Algorithms and Efficient Codes for Computational Aeroacoustics
Goodrich, John W.; Dyson, Rodger W.
1999-01-01
The simulation of sound generation and propagation in three space dimensions with realistic aircraft components is a very large time dependent computation with fine details. Simulations in open domains with embedded objects require accurate and robust algorithms for propagation, for artificial inflow and outflow boundaries, and for the definition of geometrically complex objects. The development, implementation, and validation of methods for solving these demanding problems is being done to support the NASA pillar goals for reducing aircraft noise levels. Our goal is to provide algorithms which are sufficiently accurate and efficient to produce usable results rapidly enough to allow design engineers to study the effects on sound levels of design changes in propulsion systems, and in the integration of propulsion systems with airframes. There is a lack of design tools for these purposes at this time. Our technical approach to this problem combines the development of new, algorithms with the use of Mathematica and Unix utilities to automate the algorithm development, code implementation, and validation. We use explicit methods to ensure effective implementation by domain decomposition for SPMD parallel computing. There are several orders of magnitude difference in the computational efficiencies of the algorithms which we have considered. We currently have new artificial inflow and outflow boundary conditions that are stable, accurate, and unobtrusive, with implementations that match the accuracy and efficiency of the propagation methods. The artificial numerical boundary treatments have been proven to have solutions which converge to the full open domain problems, so that the error from the boundary treatments can be driven as low as is required. The purpose of this paper is to briefly present a method for developing highly accurate algorithms for computational aeroacoustics, the use of computer automation in this process, and a brief survey of the algorithms that
Teaching Thermal Hydraulics & Numerical Methods: An Introductory Control Volume Primer
Energy Technology Data Exchange (ETDEWEB)
D. S. Lucas
2004-10-01
A graduate level course for Thermal Hydraulics (T/H) was taught through Idaho State University in the spring of 2004. A numerical approach was taken for the content of this course since the students were employed at the Idaho National Laboratory and had been users of T/H codes. The majority of the students had expressed an interest in learning about the Courant Limit, mass error, semi-implicit and implicit numerical integration schemes in the context of a computer code. Since no introductory text was found the author developed notes taught from his own research and courses taught for Westinghouse on the subject. The course started with a primer on control volume methods and the construction of a Homogeneous Equilibrium Model (HEM) (T/H) code. The primer was valuable for giving the students the basics behind such codes and their evolution to more complex codes for Thermal Hydraulics and Computational Fluid Dynamics (CFD). The course covered additional material including the Finite Element Method and non-equilibrium (T/H). The control volume primer and the construction of a three-equation (mass, momentum and energy) HEM code are the subject of this paper . The Fortran version of the code covered in this paper is elementary compared to its descendants. The steam tables used are less accurate than the available commercial version written in C Coupled to a Graphical User Interface (GUI). The Fortran version and input files can be downloaded at www.microfusionlab.com.
Energy Technology Data Exchange (ETDEWEB)
Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.
1998-03-01
This document describes the major modifications and improvements made to the modeling of the RAMONA-3B/MOD0 code since 1981, when the code description and assessment report was completed. The new version of the code is RAMONA-4B. RAMONA-4B is a systems transient code for application to different versions of Boiling Water Reactors (BWR) such as the current BWR, the Advanced Boiling Water Reactor (ABWR), and the Simplified Boiling Water Reactor (SBWR). This code uses a three-dimensional neutron kinetics model coupled with a multichannel, non-equilibrium, drift-flux, two-phase flow formulation of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients and instability issues. Chapter 1 is an overview of the code`s capabilities and limitations; Chapter 2 discusses the neutron kinetics modeling and the implementation of reactivity edits. Chapter 3 is an overview of the heat conduction calculations. Chapter 4 presents modifications to the thermal-hydraulics model of the vessel, recirculation loop, steam separators, boron transport, and SBWR specific components. Chapter 5 describes modeling of the plant control and safety systems. Chapter 6 presents and modeling of Balance of Plant (BOP). Chapter 7 describes the mechanistic containment model in the code. The content of this report is complementary to the RAMONA-3B code description and assessment document. 53 refs., 81 figs., 13 tabs.
Energy Technology Data Exchange (ETDEWEB)
Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.
1998-03-01
This document describes the major modifications and improvements made to the modeling of the RAMONA-3B/MOD0 code since 1981, when the code description and assessment report was completed. The new version of the code is RAMONA-4B. RAMONA-4B is a systems transient code for application to different versions of Boiling Water Reactors (BWR) such as the current BWR, the Advanced Boiling Water Reactor (ABWR), and the Simplified Boiling Water Reactor (SBWR). This code uses a three-dimensional neutron kinetics model coupled with a multichannel, non-equilibrium, drift-flux, two-phase flow formulation of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients and instability issues. Chapter 1 is an overview of the code`s capabilities and limitations; Chapter 2 discusses the neutron kinetics modeling and the implementation of reactivity edits. Chapter 3 is an overview of the heat conduction calculations. Chapter 4 presents modifications to the thermal-hydraulics model of the vessel, recirculation loop, steam separators, boron transport, and SBWR specific components. Chapter 5 describes modeling of the plant control and safety systems. Chapter 6 presents and modeling of Balance of Plant (BOP). Chapter 7 describes the mechanistic containment model in the code. The content of this report is complementary to the RAMONA-3B code description and assessment document. 53 refs., 81 figs., 13 tabs.
2010-01-01
The works aimed at the further development and validation of models for CFD codes. For this reason, the new thermal-hydraulic test facility TOPFLOW was erected and equipped with wire-mesh sensors with high spatial and time resolution. Vertical test sections with nominal diameters of DN50 and DN200 operating with air-water as well as steam-water two-phase flows provided results on the evaluation of flow patterns, on the be¬haviour of the interfacial area as well as on interfacial momentum and ...
Pon-Barry, Heather; Packard, Becky Wai-Ling; St. John, Audrey
2017-01-01
A dilemma within computer science departments is developing sustainable ways to expand capacity within introductory computer science courses while remaining committed to inclusive practices. Training near-peer mentors for peer code review is one solution. This paper describes the preparation of near-peer mentors for their role, with a focus on…
Wood, Jerry R.; Schmidt, James F.; Steinke, Ronald J.; Chima, Rodrick V.; Kunik, William G.
1987-01-01
Increased emphasis on sustained supersonic or hypersonic cruise has revived interest in the supersonic throughflow fan as a possible component in advanced propulsion systems. Use of a fan that can operate with a supersonic inlet axial Mach number is attractive from the standpoint of reducing the inlet losses incurred in diffusing the flow from a supersonic flight Mach number to a subsonic one at the fan face. The design of the experiment using advanced computational codes to calculate the components required is described. The rotor was designed using existing turbomachinery design and analysis codes modified to handle fully supersonic axial flow through the rotor. A two-dimensional axisymmetric throughflow design code plus a blade element code were used to generate fan rotor velocity diagrams and blade shapes. A quasi-three-dimensional, thin shear layer Navier-Stokes code was used to assess the performance of the fan rotor blade shapes. The final design was stacked and checked for three-dimensional effects using a three-dimensional Euler code interactively coupled with a two-dimensional boundary layer code. The nozzle design in the expansion region was analyzed with a three-dimensional parabolized viscous code which corroborated the results from the Euler code. A translating supersonic diffuser was designed using these same codes.