Science.gov

Sample records for in-vessel thermal-hydraulic phenomena

  1. 78 FR 8202 - Meeting of the Joint ACRS Subcommittees on Thermal Hydraulic Phenomena and Materials, Metallurgy...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-02-05

    ... ACRS meetings were published in the Federal Register on October 18, 2012, (77 FR 64146- 64147... Hydraulic Phenomena and Materials, Metallurgy and Reactor Fuels; Notice of Meeting The Joint ACRS Subcommittees on Thermal Hydraulic Phenomena and Materials, Metallurgy and Reactor Fuels will hold a meeting...

  2. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    SciTech Connect

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

    2007-06-30

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition from normal high

  3. Best Estimate Code System to Calculate Thermal & Hydraulic Phenomena in a Nuclear Reactor or Related System.

    1999-05-19

    Version 00 RELAP4/MOD7/101 performs best estimate analyses of nuclear reactors or related systems undergoing a transient. Transient thermal-hydraulic, two-phase phenomena are calculated from formulations of one-dimensional, homogeneous, equilibrium conservation equations for water mass, momentum, and energy. Heat structures are modeled using a transient one-dimensional heat conduction solution that is coupled to the fluid through heat transfer relations. Various explicit models are used to calculate nonhomogeneous, nonequilibrium behavior including a phase separation model, a vertical slipmore » model, and a nonequilibrium model. Other models are used to represent critical flow, reactor kinetics, pressurized water reactor reflood behavior, nuclear fuel rod swelling and blockage, and components such as pumps, valves, and accumulators.« less

  4. Study on Characteristics of Thermal-Hydraulics Phenomena in Steam Injector

    SciTech Connect

    Yuhki Takahashi; Yasuo Koizumi; Hiroyasu Ohtake; Michitsugu Mori

    2006-07-01

    Characteristics of thermal-hydraulic phenomena in the steam injector were examined. Two-types of a test section were used: a straight condensing section type and a taper condensing section type. In the straight condensing section type experiments, a water jet from a nozzle of 5 mm diameter flowed in to the test section concentrically. The inner diameter of the condensing section was 7, 10, or 20 mm and the length was 105 mm. Steam flowed into the peripheral space between the water jet and the inner wall of the condensing section. Experiments of the taper condensing section type were similar to those of the straight condensing section type. The inner diameter of the condensing section at the outlet of the water nozzle was 13.25 mm and the condensing section length was 52.9 mm. The inner diameter of the throat was 4 mm and the throat length was 5 mm. The vapor condensation rate, i.e. the condensation heat transfer rate, did not depend on the subcooling of the water jet. It was proved that the overall condensation heat transfer is controlled by the radial heat transport in the water jet since the condensation heat transfer resistance is much smaller than that of the radial heat transport in the water jet. The radial heat transport only depends on the jet flow rate. The criterion of the jet disruption was correlated by the Kelvin-Helmholtz instability wave length. The prediction provided conservative results; approximately one-tenth. When the condensing section was tapered such as the steam injector, the water jet became more stable. It was suggested that the accelerated water jet becomes more stable. (authors)

  5. In-vessel phenomena -- CORA

    SciTech Connect

    Ott, L.J.; van Rij, W.I.

    1991-01-01

    Experiment-specific models have been employed since 1986 by Oak Ridge National Laboratory (ORNL) severe accident analysis programs for the purpose of boiling water reactor experimental planning and optimum interpretation of experimental results. The large integral tests performed to date, which start from an initial undamaged core state, have involved significantly different-from-prototypic boundary and experimental conditions because of either normal facility limitations or specific experimental constraints. These experiments (ACRR: DF-4, NRU: FLHT-6, and CORA) were designed to obtain specific phenomenological information such as the degradation and interaction of prototypic components and the effects on melt progression of control-blade materials and channel boxes. Applications of ORNL models specific to the KfK CORA-16 and CORA-17 experiments are discussed and significant findings from the experimental analyses such as the following are presented: applicability of available Zircaloy oxidation kinetics correlations; influence of cladding strain on Zircaloy oxidation; influence of spacer grids on the structural heatup; and the impact of treating the gaseous coolant as a gray interacting medium. The experiment-specific models supplement and support the systems-level accident analysis codes. They allow the analyst to accurately quantify the observed experimental phenomena and to compensate for the effect of known uncertainties. They provide a basis for the efficient development of new models for phenomena that are currently not modeled (such as material interactions). They can provide validated phenomenological models (from the results of the experiments) as candidates for incorporation in the systems-level whole-core'' codes.

  6. INVESTIGATION OF FUNDAMENTAL THERMAL-HYDRAULIC PHENOMENA IN ADVANCED GAS-COOLED REACTORS

    SciTech Connect

    INVESTIGATION OF FUNDAMENTAL THERMAL-HYDRAULIC PHE

    2006-09-01

    INL LDRD funded research was conducted at MIT to experimentally characterize mixed convection heat transfer in gas-cooled fast reactor (GFR) core channels in collaboration with INL personnel. The GFR for Generation IV has generated considerable interest and is under development in the U.S., France, and Japan. One of the key candidates is a block-core configuration first proposed by MIT, has the potential to operate in Deteriorated Turbulent Heat Transfer (DTHT) regime or in the transition between the DTHT and normal forced or laminar convection regime during post-loss-of-coolant accident (LOCA) conditions. This is contrary to most industrial applications where operation is in a well-defined and well-known turbulent forced convection regime. As a result, important new need emerged to develop heat transfer correlations that make possible rigorous and accurate predictions of Decay Heat Removal (DHR) during post LOCA in these regimes. Extensive literature review on these regimes was performed and a number of the available correlations was collected in: (1) forced laminar, (2) forced turbulent, (3) mixed convection laminar, (4) buoyancy driven DTHT and (5) acceleration driven DTHT regimes. Preliminary analysis on the GFR DHR system was performed and using the literature review results and GFR conditions. It confirmed that the GFR block type core has a potential to operate in the DTHT regime. Further, a newly proposed approach proved that gas, liquid and super critical fluids all behave differently in single channel under DTHT regime conditions, thus making it questionable to extrapolate liquid or supercritical fluid data to gas flow heat transfer. Experimental data were collected with three different gases (nitrogen, helium and carbon dioxide) in various heat transfer regimes. Each gas unveiled different physical phenomena. All data basically covered the forced turbulent heat transfer regime, nitrogen data covered the acceleration driven DTHT and buoyancy driven DTHT

  7. Development of a phenomena identification and ranking table for thermal-hydraulic phenomena during a double-ended guillotine break LOCA in an SRS production reactor

    SciTech Connect

    Hanson, R.G.; Ortiz, M.G.; Bolander, M.A.; Wilson, G.E.

    1989-07-01

    A rising level of scrutiny is being directed toward the Savannah River Site (SRS) production reactors. Improved calculational capabilities are being developed to provide a best estimate analytical process to determine the safe operating margins of the reactors. The Code Scaling, Applicability, and Uncertainty (CSAU) methodology, developed by the US Nuclear Regulatory Commission to support best estimate simulations, is being applied to the best estimate limits analysis for the SRS production reactors. One of the foundational parts of the method is the identification and ranking of all the processes that occur during the specific limiting scenario. The phenomena ranking is done according to their importance to safety criteria during the transient and is used to focus the uncertainty analysis on a sufficient, yet cost effective scope of work. This report documents the thermal-hydraulic phenomena that occur during a limiting break in an SRS production reactor and their importance to the uncertainty in simulations of the reactor behavior. 9 refs., 14 figs., 10 tabs.

  8. Development of Design Technology on Thermal-Hydraulic Performance in Tight-Lattice Rod Bundles: III - Numerical Evaluation of Fluid Mixing Phenomena using Advanced Interface-Tracking Method -

    NASA Astrophysics Data System (ADS)

    Yoshida, Hiroyuki; Nagayoshi, Takuji; Takase, Kazuyuki; Akimoto, Hajime

    Thermal-hydraulic design of the current boiling water reactor (BWR) is performed by correlations with empirical results of actual-size tests. However, for the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) core, an actual size test of an embodiment of its design is required to confirm or modify such correlations. Development of a method that enables the thermal-hydraulic design of nuclear reactors without these actual size tests is desired, because these tests take a long time and entail great cost. For this reason we developed an advanced thermal-hydraulic design method for FLWRs using innovative two-phase flow simulation technology. In this study, detailed Two-Phase Flow simulation code using advanced Interface Tracking method: TPFIT is developed to calculate the detailed information of the two-phase flow. We tried to verify the TPFIT code by comparing it with the 2-channel air-water and steam-water mixing experimental results. The predicted result agrees well the observed results and bubble dynamics through the gap and cross flow behavior could be effectively predicted by the TPFIT code, and pressure difference between fluid channels is responsible for the fluid mixing.

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

    SciTech Connect

    Lindgren, Eric Richard; Durbin, Samuel G

    2007-04-01

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

  10. Thermal-Hydraulic-Analysis Program

    NASA Technical Reports Server (NTRS)

    Walton, J. T.

    1993-01-01

    ELM computer program is simple computational tool for modeling steady-state thermal hydraulics of flows of propellants through fuel-element-coolant channels in nuclear thermal rockets. Evaluates various heat-transfer-coefficient and friction-factor correlations available for turbulent pipe flow with addition of heat. Comparisons possible within one program. Machine-independent program written in FORTRAN 77.

  11. GCFR thermal-hydraulic experiments

    SciTech Connect

    Schlueter, G.; Baxi, C.B.; Dalle Donne, M.; Gat, U.; Fenech, H.; Hanson, D.; Hudina, M.

    1980-01-01

    The thermal-hydraulic experimental studies performed and planned for the Gas-Cooled Fast Reactor (GCFR) core assemblies are described. The experiments consist of basic studies performed to obtain correlations, and bundle experiments which provide input for code validation and design verification. These studies have been performed and are planned at European laboratories, US national laboratories, Universities in the US, and at General Atomic Company

  12. Helical coil thermal hydraulic model

    NASA Astrophysics Data System (ADS)

    Caramello, M.; Bertani, C.; De Salve, M.; Panella, B.

    2014-11-01

    A model has been developed in Matlab environment for the thermal hydraulic analysis of helical coil and shell steam generators. The model considers the internal flow inside one helix and its associated control volume of water on the external side, both characterized by their inlet thermodynamic conditions and the characteristic geometry data. The model evaluates the behaviour of the thermal-hydraulic parameters of the two fluids, such as temperature, pressure, heat transfer coefficients, flow quality, void fraction and heat flux. The evaluation of the heat transfer coefficients as well as the pressure drops has been performed by means of the most validated literature correlations. The model has been applied to one of the steam generators of the IRIS modular reactor and a comparison has been performed with the RELAP5/Mod.3.3 code applied to an inclined straight pipe that has the same length and the same elevation change between inlet and outlet of the real helix. The predictions of the developed model and RELAP5/Mod.3.3 code are in fairly good agreement before the dryout region, while the dryout front inside the helical pipes is predicted at a lower distance from inlet by the model.

  13. Thermal hydraulics development for CASL

    SciTech Connect

    Lowrie, Robert B

    2010-12-07

    This talk will describe the technical direction of the Thermal-Hydraulics (T-H) Project within the Consortium for Advanced Simulation of Light Water Reactors (CASL) Department of Energy Innovation Hub. CASL is focused on developing a 'virtual reactor', that will simulate the physical processes that occur within a light-water reactor. These simulations will address several challenge problems, defined by laboratory, university, and industrial partners that make up CASL. CASL's T-H efforts are encompassed in two sub-projects: (1) Computational Fluid Dynamics (CFD), (2) Interface Treatment Methods (ITM). The CFD subproject will develop non-proprietary, scalable, verified and validated macroscale CFD simulation tools. These tools typically require closures for their turbulence and boiling models, which will be provided by the ITM sub-project, via experiments and microscale (such as DNS) simulation results. The near-term milestones and longer term plans of these two sub-projects will be discussed.

  14. 77 FR 9707 - Advisory Committee on Reactor Safeguards Meeting of the ACRS Subcommittee on Thermal-Hydraulics...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-02-17

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards Meeting of the ACRS Subcommittee on Thermal-Hydraulics... for the ACRS Subcommittee meeting on Thermal-Hydraulics Phenomena scheduled to be held on February...

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

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

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

  16. INL Experimental Program Roadmap for Thermal Hydraulic Code Validation

    SciTech Connect

    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

  17. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    NASA Astrophysics Data System (ADS)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki; Hashimoto, Akihiko; Imai, Yasutomo; Ito, Masahiro

    2015-12-01

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  18. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    SciTech Connect

    Ohshima, Hiroyuki; Uwaba, Tomoyuki; Hashimoto, Akihiko; Imai, Yasutomo; Ito, Masahiro

    2015-12-31

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  19. Application of computational fluid dynamics methods to improve thermal hydraulic code analysis

    NASA Astrophysics Data System (ADS)

    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.

  20. Thermal-hydraulic modeling needs for passive reactors

    SciTech Connect

    Kelly, J.M.

    1997-07-01

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

  1. TEMPEST. Transient 3-D Thermal-Hydraulic

    SciTech Connect

    Eyler, L.L.

    1992-01-31

    TEMPEST is a transient, three-dimensional, hydrothermal program that is designed to analyze a range of coupled fluid dynamic and heat transfer systems of particular interest to the Fast Breeder Reactor (FBR) thermal-hydraulic design community. The full three-dimensional, time-dependent equations of motion, continuity, and heat transport are solved for either laminar or turbulent fluid flow, including heat diffusion and generation in both solid and liquid materials. The equations governing mass, momentum, and energy conservation for incompressible flows and small density variations (Boussinesq approximation) are solved using finite-difference techniques. Analyses may be conducted in either cylindrical or Cartesian coordinate systems. Turbulence is treated using a two-equation model. Two auxiliary plotting programs, SEQUEL and MANPLOT, for use with TEMPEST output are included. SEQUEL may be operated in batch or interactive mode; it generates data required for vector plots, contour plots of scalar quantities, line plots, grid and boundary plots, and time-history plots. MANPLOT reads the SEQUEL-generated data and creates the hardcopy plots. TEMPEST can be a valuable hydrothermal design analysis tool in areas outside the intended FBR thermal-hydraulic design community.

  2. Simulation of the PBF-Candu test with coupled thermal-hydraulic and fuel thermo-mechanical responses

    SciTech Connect

    Baschuk, J. J.

    2012-07-01

    During a large loss-of-coolant accident (LLOCA), the fuel sheath temperature is influenced by thermal-hydraulic and thermo-mechanical phenomena. The thermal-hydraulic phenomena include the heat transfer from the sheath to the coolant and surroundings. Thermo-mechanical phenomena, such as creep and thermal expansion, influence the size of the fuel-to-sheath gap, and thus the heat transfer from the fuel to the sheath. Therefore, coupling the thermal-hydraulic and thermo-mechanical analysis of an LLOCA would result in more accurate predictions of sheath temperature. This is illustrated by comparing the sheath temperature predictions from coupled and decoupled simulations of the PBF-Candu Test with experimental measurements. The codes CATHENA and ELOCA were used for the thermal-hydraulic and thermo-mechanical analysis, respectively. The predicted sheath temperatures from both the coupled and decoupled simulations were higher than the measured values. However, after the initial power pulse, when the fuel-to-sheath gap was calculated as being opened, the sheath temperatures predicted by the coupled simulation were closer to the experimental measurements. Thus, under conditions of an open fuel-to-sheath gap, a coupled thermal-hydraulic and thermo-mechanical analysis can improve predictions of sheath temperatures. (authors)

  3. Proceedings of the OECD/CSNI workshop on transient thermal-hydraulic and neutronic codes requirements

    SciTech Connect

    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.

  4. THE THREE DIMENSIONAL THERMAL HYDRAULIC CODE BAGIRA.

    SciTech Connect

    KALINICHENKO,S.D.; KOHUT,P.; KROSHILIN,A.E.; KROSHILIN,V.E.; SMIRNOV,A.V.

    2003-05-04

    BAGIRA - a thermal-hydraulic program complex was primarily developed for using it in nuclear power plant simulator models, but is also used as a best-estimate analytical tool for modeling two-phase mixture flows. The code models allow consideration of phase transients and the treatment of the hydrodynamic behavior of boiling and pressurized water reactor circuits. It provides the capability to explicitly model three-dimensional flow regimes in various regions of the primary and secondary circuits such as, the mixing regions, circular downcomer, pressurizer, reactor core, main primary loops, the steam generators, the separator-reheaters. In addition, it is coupled to a severe-accident module allowing the analysis of core degradation and fuel damage behavior. Section II will present the theoretical basis for development and selected results are presented in Section III. The primary use for the code complex is to realistically model reactor core behavior in power plant simulators providing enhanced training tools for plant operators.

  5. LMR thermal hydraulics calculations in the US

    SciTech Connect

    Dunn, F.E.; Malloy, D.J.; Mohr, D.

    1987-04-27

    A wide range of thermal hydraulics computer codes have been developed by various organizations in the US. These codes cover an extensive range of purposes from within-assembly-wise pin temperature calculations to plant wide transient analysis. The codes are used for static analysis, for analysis of protected anticipated transients, and for analysis of a wide range of unprotected transients for the more recent inherently safe LMR designs. Some of these codes are plant-specific codes with properties of a specific plant built into them. Other codes are more general and can be applied to a number of plants or designs. These codes, and the purposes for which they have been used, are described.

  6. Thermal Hydraulic Analysis of Spent Fuel Casks

    SciTech Connect

    Rector, D. R.; Cuta, J. M.; Enderlin, C. W.

    1997-10-08

    COBRA-SFS (Spent Fuel Storage) is a code for thermal-hydraulic analysis of multi-assembly spent fuel storage and transportation systems. It uses a lumped parameter finite difference approach to predict flow and temperature distributions in spent fuel storage systems and fuel assemblies, under forced and natural convection heat transfer conditions. Derived from the COBRA family of codes, which have been extensively evaluated against in-pile and out-of-pile data, COBRA-SFS retains all the important features of the COBRA codes for single phase fluid analysis, and extends the range application to include problems with two-dimensional radiative and three-dimensional conductive heat transfer. COBRA-SFS has been used to analyze various single- and multi-assembly spent fuel storage systems containing unconsolidated and consolidated fuel rods, with a variety of fill media, including air, helium and vacuum. Cycle 0 of COBRA-SFS was released in 1986. Subsequent applications of the code led to development of additional capabilities, which resulted in the release of Cycle 1 in February 1989. Since then, the code has undergone an independent technical review as part of a submittal to the Nuclear Regulatory Commission for a generic license to apply the code to spent fuel storage system analysis. Modifications and improvements to the code have been combined to form Cycle 2. Cycle 3., the newest version of COBRA-SFS, has been validated and verified for transient applications, such as a storage cask thermal response to a pool fire.

  7. An Approach of Uncertainty Evaluation for Thermal-Hydraulic Analysis

    SciTech Connect

    Katsunori Ogura; Hisashi Ninokata

    2002-07-01

    An approach to evaluate uncertainty systematically for thermal-hydraulic analysis programs is demonstrated. The approach is applied to the Peach Bottom Unit 2 Turbine Trip 2 Benchmark and is validated. (authors)

  8. Thermal Hydraulic Analysis of Spent Fuel Casks

    1997-10-08

    COBRA-SFS (Spent Fuel Storage) is a code for thermal-hydraulic analysis of multi-assembly spent fuel storage and transportation systems. It uses a lumped parameter finite difference approach to predict flow and temperature distributions in spent fuel storage systems and fuel assemblies, under forced and natural convection heat transfer conditions. Derived from the COBRA family of codes, which have been extensively evaluated against in-pile and out-of-pile data, COBRA-SFS retains all the important features of the COBRA codesmore » for single phase fluid analysis, and extends the range application to include problems with two-dimensional radiative and three-dimensional conductive heat transfer. COBRA-SFS has been used to analyze various single- and multi-assembly spent fuel storage systems containing unconsolidated and consolidated fuel rods, with a variety of fill media, including air, helium and vacuum. Cycle 0 of COBRA-SFS was released in 1986. Subsequent applications of the code led to development of additional capabilities, which resulted in the release of Cycle 1 in February 1989. Since then, the code has undergone an independent technical review as part of a submittal to the Nuclear Regulatory Commission for a generic license to apply the code to spent fuel storage system analysis. Modifications and improvements to the code have been combined to form Cycle 2. Cycle 3., the newest version of COBRA-SFS, has been validated and verified for transient applications, such as a storage cask thermal response to a pool fire.« less

  9. Process management using component thermal-hydraulic function classes

    DOEpatents

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

    1999-07-27

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

  10. Process management using component thermal-hydraulic function classes

    DOEpatents

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

    1999-01-01

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

  11. Process management using component thermal-hydraulic function classes

    SciTech Connect

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

    1997-12-01

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

  12. VIPRE-01: A thermal-hydraulic code for reactor cores:

    SciTech Connect

    Stewart, C.W.; Cuta, J.M.

    1988-03-01

    VIPRE (Versatile Internals and Component Program for Reactors;EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear reactor core safety limits including minimum departure from nucleate boiling ratio (NDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume discusses general and specific considerations in using VIPRE as a thermal-hydraulic analysis tool. Volume 1: Mathematical Modeling, explains the major thermal-hydraulic models and supporting mathematial correlations in detail. Volume 2: Users's Manual, describes the input requirements of the codes in the VIPRE code package. Volume 3: Programmer's Manual, explains the code structure and computer interface. Experimence in running VIPRE is documented in Volume 4: Applications. 25 refs., 31 figs., 7 tabs.

  13. Current and anticipated use of thermal-hydraulic codes for BWR transient and accident analyses in Japan

    SciTech Connect

    Arai, Kenji; Ebata, Shigeo

    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.

  14. Development of thermal-hydraulic analysis capabilities for Oyster creek

    SciTech Connect

    Lee, R.B.

    1987-01-01

    GPU Nuclear (GPUN) has been involved in developing analytical methodologies for Oyster Creek plant thermal-hydraulic response simulation for approx. 15 yr. Plant-system-related transient analysis is being accomplished via RETRAN02 MOD4 and loss-of-coolant accident (LOCA) analysis by SAFER-CORECOOL. This paper reviews the developmental process and lessons learned through this process.

  15. Current and anticipated uses of thermal hydraulic codes in Korea

    SciTech Connect

    Kim, Kyung-Doo; Chang, Won-Pyo

    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.

  16. Thermal-hydraulic processes involved in loss of residual heat removal during reduced inventory operation

    SciTech Connect

    Fletcher, C.D.; McHugh, P.R.; Naff, S.A.; Johnsen, G.W.

    1991-02-01

    This paper identifies the topics needed to understand pressurized water reactor response to an extended loss of residual heat removal event during refueling and maintenance outages. By identifying the possible plant conditions and cooling methods that would be used for each cooling mode, the controlling thermal-hydraulic processes and phenomena were identified. Controlling processes and phenomena include: gravity drain, core water boil-off, and reflux cooling processes. Important subcategories of the reflux cooling processes include: the initiation of reflux cooling from various plant conditions, the effects of air on reflux cooling, core level depression effects, issues regarding the steam generator secondaries, and the special case of boiler-condenser cooling with once-through steam generators. 25 refs., 6 figs., 1 tab.

  17. Overview of rod-bundle thermal-hydraulic analysis

    SciTech Connect

    Sha, W.T.

    1980-11-01

    Three methods used in rod-bundle thermal-hydraulic analysis are summarized. These methods are: (1) subchannel analysis, and its inherent assumptions are clearly stated; (2) porous medium formulation with volume porosity, surface permeability, distributed resistance and distributed heat source (sink) - the concept of surface permeability is new in porous medium formulation, and greatly facilitates modeling anisotropic effects; and (3) benchmark rod-bundle thermal-hydraulic analysis using a boundary-fitted coordinate system, and it represents the most rigorous method to date. For laminar flow, this method gives solutions without any assumptions and it requires information on rod bundle geometry and thermal physical properties of the fluid. Basic limitations and merits of each method are discussed in detail. 19 refs., 6 figs., 1 tab.

  18. Portable Life Support Subsystem Thermal Hydraulic Performance Analysis

    NASA Technical Reports Server (NTRS)

    Barnes, Bruce; Pinckney, John; Conger, Bruce

    2010-01-01

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

  19. Thermal-hydraulic interfacing code modules for CANDU reactors

    SciTech Connect

    Liu, W.S.; Gold, M.; Sills, H.

    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.

  20. Coupled Monte Carlo neutronics and thermal hydraulics for power reactors

    SciTech Connect

    Bernnat, W.; Buck, M.; Mattes, M.; Zwermann, W.; Pasichnyk, I.; Velkov, K.

    2012-07-01

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

  1. Thermal hydraulic feasibility assessment for the Spent Nuclear Fuel Project

    SciTech Connect

    Heard, F.J.; Cramer, E.R.; Beaver, T.R.; Thurgood, M.J.

    1996-01-01

    A series of scoping analyses have been completed investigating the thermal-hydraulic performance and feasibility of the Spent Nuclear Fuel Project (SNFP) Integrated Process Strategy (IPS). The SNFP was established to develop engineered solutions for the expedited removal, stabilization, and storage of spent nuclear fuel from the K Basins at the U.S. Department of Energy`s Hanford Site in Richland, Washington. The subject efforts focused on independently investigating, quantifying, and establishing the governing heat production and removal mechanisms for each of the IPS operations and configurations, obtaining preliminary results for comparison with and verification of other analyses, and providing technology-based recommendations for consideration and incorporation into the design bases for the SNFP. The goal was to develop a series fo thermal-hydraulic models that could respond to all process and safety-related issues that may arise pertaining to the SNFP. A series of sensitivity analyses were also performed to help identify those parameters that have the greatest impact on energy transfer and hence, temperature control. It is anticipated that the subject thermal-hydraulic models will form the basis for a series of advanced and more detailed models that will more accurately reflect the thermal performance of the IPS and alleviate the necessity for some of the more conservative assumptions and oversimplifications, as well as form the basis for the final process and safety analyses.

  2. Views on the future of thermal hydraulic modeling

    SciTech Connect

    Ishii, M.

    1997-07-01

    It is essential for the U.S. NRC to sustain the highest level of the thermal-hydraulics and reactor safety research expertise and continuously improve their accident analysis capability. Such expertise should span over four different areas which are strongly related to each other. These are: (1) Reactor Safety Code Development, (2) Two-phase Flow Modeling, (3) Instrumentation and Fundamental Experimental Research, and (4) Separate Effect and Integral Test. The NRC is already considering a new effort in the area of advanced thermal-hydraulics effort. Its success largely depends on the availability of a significantly improved two-phase flow formulation and constitutive relations supported by detailed experimental data. Therefore, it is recommended that the NRC start significant research efforts in the areas of two-phase flow modeling, instrumentation, basic and separate effect experiments which should be pursued systematically and with clearly defined objectives. It is desirable that some international program is developed in this area. This paper is concentrated on those items in the thermal-hydraulic area which eventually determine the quality of future accident analysis codes.

  3. Interface requirements to couple thermal-hydraulic codes to severe accident codes: ATHLET-CD

    SciTech Connect

    Trambauer, K.

    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.

  4. Modeling Reactor Coolant Systems Thermal-Hydraulic Transients

    1999-10-05

    RELAP5/MOD3.2* is used to model reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents and operational transients such as anticipated transients without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal-hydraulic systems. Control system and secondary system components are included to allow modeling of themore » plant controls, turbines, condensers, and secondary feedwater systems.« less

  5. Upgrading the HFIR Thermal-Hydraulic Legacy Code Using COMSOL

    SciTech Connect

    Bodey, Isaac T; Arimilli, Rao V; Freels, James D

    2010-01-01

    Modernization of the High Flux Isotope Reactor (HFIR) thermal-hydraulic (TH) design and safety analysis capability is an important step in preparation for the conversion of the HFIR core from a high enriched uranium (HEU) fuel to a low enriched uranium (LEU) fuel. Currently, an important part of the HFIR TH analysis is based on the legacy Steady State Heat Transfer Code (SSHTC), which adds much conservatism to the safety analysis. The multi-dimensional multi-physics capabilities of the COMSOL environment allow the analyst to relax the number and magnitude of conservatisms, imposed by the SSHTC, to present a more physical model of the TH aspect of the HFIR.

  6. Thermal hydraulics analysis of LIBRA-SP target chamber

    SciTech Connect

    Mogahed, E.A.

    1996-12-31

    LIBRA-SP is a conceptual design study of an inertially confined 1000 MWe fusion power reactor utilizing self-pinched light ion beams. There are 24 ion beams which are arranged around the reactor cavity. The reaction chamber is an upright cylinder with an inverted conical roof resembling a mushroom, and a pool floor. The vertical sides of the cylinder are occupied by a blanket zone consisting of many perforated rigid HT-9 ferritic steel tubes called PERITs (PEr-forated RIgid Tube). The breeding/cooling material, liquid lead-lithium, flows through the PERITs, providing protection to the reflector/vacuum chamber so as to make it a lifetime component. The neutronics analysis and cavity hydrodynamics calculations are performed to account for the neutron heating and also to determine the effects of vaporization/condensation processes on the surface heat flux. The steady state nuclear heating distribution at the midplane is used for thermal hydraulics calculations. The maximum surface temperature of the HT-9 is chosen to not exceed 625{degree}C to avoid drastic deterioration of the metal`s mechanical properties. This choice restricts the thermal hydraulics performance of the reaction cavity. The inlet first surface coolant bulk temperature is 370{degree}C, and the heat exchanger inlet coolant bulk temperature is 502{degree}C. 4 refs., 6 figs., 2 tabs.

  7. 3D neutronic/thermal-hydraulic coupled analysis of MYRRHA

    SciTech Connect

    Vazquez, M.; Martin-Fuertes, F.

    2012-07-01

    The current tendency in multiphysics calculations applied to reactor physics is the use of already validated computer codes, coupled by means of an iterative approach. In this paper such an approach is explained concerning neutronics and thermal-hydraulics coupled analysis with MCNPX and COBRA-IV codes using a driver program and file exchange between codes. MCNPX provides the neutronic analysis of heterogeneous nuclear systems, both in critical and subcritical states, while COBRA-IV is a subchannel code that can be used for rod bundles or core thermal-hydraulics analysis. In our model, the MCNP temperature dependence of nuclear data is handled via pseudo-material approach, mixing pre-generated cross section data set to obtain the material with the desired cross section temperature. On the other hand, COBRA-IV has been updated to allow for the simulation of liquid metal cooled reactors. The coupled computational tool can be applied to any geometry and coolant, as it is the case of single fuel assembly, at pin-by-pin level, or full core simulation with the average pin of each fuel-assembly. The coupling tool has been applied to the critical core layout of the SCK-CEN MYRRHA concept, an experimental LBE cooled fast reactor presently in engineering design stage. (authors)

  8. Teaching Thermal Hydraulics & Numerical Methods: An Introductory Control Volume Primer

    SciTech Connect

    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.

  9. Performance of a parallel thermal-hydraulics code TEMPEST

    SciTech Connect

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

    1996-11-01

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

  10. Advanced neutron source reactor thermal-hydraulic test loop facility description

    SciTech Connect

    Felde, D.K.; Farquharson, G.; Hardy, J.H.; King, J.F.; McFee, M.T.; Montgomery, B.H.; Pawel, R.E.; Power, B.H.; Shourbaji, A.A.; Siman-Tov, M.; Wood, R.J.; Yoder, G.L.

    1994-02-01

    The Thermal-Hydraulic Test Loop (THTL) is a facility for experiments constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory. The ANSR is both cooled and moderated by heavy water and uses uranium silicide fuel. The core is composed of two coaxial fuel-element annuli, each of different diameter. There are 684 parallel aluminum-clad fuel plates (252 in the inner-lower core and 432 in the outer-upper core) arranged in an involute geometry that effectively creates an array of thin rectangular flow channels. Both the fuel plates and the coolant channels are 1.27 mm thick, with a span of 87 mm (lower core), 70 mm (upper core), and 507-mm heated length. The coolant flows vertically upwards at a mass flux of 27 Mg/m{sup 2}s (inlet velocity of 25 m/s) with an inlet temperature of 45{degrees}C and inlet pressure of 3.2 MPa. The average and peak heat fluxes are approximately 6 and 12 MW/m{sup 2}, respectively. The availability of experimental data for both flow excursion (FE) and true critical heat flux (CHF) at the conditions applicable to the ANSR is very limited. The THTL was designed and built to simulate a full-length coolant subchannel of the core, allowing experimental determination of thermal limits under the expected ANSR thermal-hydraulic conditions. For these experimental studies, the involute-shaped fuel plates of the ANSR core with the narrow 1.27-mm flow gap are represented by a narrow rectangular channel. Tests in the THTL will provide both single- and two-phase thermal-hydraulic information. The specific phenomena that are to be examined are (1) single-phase heat-transfer coefficients and friction factors, (2) the point of incipient boiling, (3) nucleate boiling heat-transfer coefficients, (4) two-phase pressure-drop characteristics in the nucleate boiling regime, (5) flow instability limits, and (6) CHF limits.

  11. FY 1995 progress report on the ANS thermal-hydraulic test loop operation and results

    SciTech Connect

    Siman-Tov, M.; Felde, D.K.; Farquharson, G.; McDuffee, J.L.; McFee, M.T.; Ruggles, A.E.; Wendel, M.W.; Yoder, G.L.

    1997-07-01

    The Thermal-Hydraulic Test Loop (THTL) is an experimental facility constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory (ORNL). The THTL facility was designed and built to provide known thermal-hydraulic (T/H) conditions for a simulated full-length coolant subchannel of the ANS reactor core, thus facilitating experimental determination of FE and CHF thermal limits under expected ANSR T/H conditions. Special consideration was given to allow operation of the system in a stiff mode (constant flow) and in a soft mode (constant pressure drop) for proper implementation of true FE and DNB experiments. The facility is also designed to examine other T/H phenomena, including onset of incipient boiling (IB), single-phase heat transfer coefficients and friction factors, and two-phase heat transfer and pressure drop characteristics. Tests will also be conducted that are representative of decay heat levels at both high pressure and low pressure as well as other quasi-equilibrium conditions encountered during transient scenarios. A total of 22 FE tests and 2 CHF tests were performed during FY 1994 and FY 1995 with water flowing vertically upward. Comparison of these data as well as extensive data from other investigators led to a proposed modification to the Saha and Zuber correlation for onset of significant void (OSV), applied to FE prediction. The modification takes into account a demonstrated dependence of the OSV or FE thermal limits on subcooling levels, especially in the low subcooling regime.

  12. FY 1993 progress report on the ANS thermal-hydraulic test loop operation and results

    SciTech Connect

    Siman-Tov, M.; Felde, D.K.; Farquharson, G.

    1994-07-01

    The Thermal-Hydraulic Test Loop (THTL) is an experimental facility constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory (ORNL). Highly subcooled heavy-water coolant flows vertically upward at a very high mass flux of almost 27 MG/m{sup 2}-s. In a parallel fuel plate configuration as in the ANSR, the flow is subject to a potential excursive static-flow instability that can very rapidly lead to flow starvation and departure from nucleate boiling (DNB) in the ``hot channel``. The current correlations and experimental data bases for flow excursion (FE) and critical heat flux (CHF) seldom evaluate the specific combination of ANSR operating parameters. The THTL facility was designed and built to provide known thermal-hydraulic (T/H) conditions for a simulated full-length coolant subchannel of the ANS reactor core, thus facilitating experimental determination of FE and CHF thermal limits under expected ANSR T/H conditions. A series of FE tests with water flowing vertically upward was completed over a nominal heat flux range of 6 to 17 MW/m{sup 2}, a mass flux range of 8 to 28 Mg/m{sup 2}-s, an exit pressure range of 1.4 to 2.1 MPa, and an inlet temperature range of 40 to 50 C. FE experiments were also conducted using as ``soft`` a system as possible to secure a true FE phenomena (actual secondary burnout). True DNB experiments under similar conditions were also conducted. To the author`s knowledge, no other FE data have been reported in the literature to date that dover such a combination of conditions of high mass flux, high heat flux, and moderately high pressure.

  13. COMMIX-1B. 3-D Single-Phase Thermal Hydraulics

    SciTech Connect

    Wildman, D.J.

    1986-01-31

    COMMIX-1B is designed to perform steady-state or transient, single-phase, three-dimensional analysis of fluid flow with heat transfer in a single-component or multicomponent system. The program was developed for the analysis of heat transfer and fluid flow processes in a nuclear reactor system; however, it can easily be applied to non-nuclear systems requiring heat transfer and/or fluid flow analysis. COMMIX-1B solves the conservation equations of mass, momentum, and energy, and transport equations of turbulence parameters and provides detailed local velocity, temperature, and pressure fields for the problem under consideration. It is capable of solving thermal-hydraulic problems involving either a single component, such as a rod bundle, reactor plenum, piping system, heat exchanger, etc., or a multicomponent system that is a combination of these components.

  14. Thermal hydraulic modeling of integrated cooling water systems

    SciTech Connect

    Niyogi, K.K.; Rathi, J.S.; Phan, T.Q.; Chaudhary, A.

    1994-12-31

    Thermal hydraulic modeling of cooling water systems has been extended to multiple system configurations with heat exchangers as interface components between systems. The computer program PC-TRAX has been used as the basic tool for the system simulation. Additional heat exchanger modules have been incorporated to accurately predict the thermal performance of systems for the design as well as off-design conditions. The modeling accommodates time-dependent changes in conditions, temperature and pressure controllers, and detailed physical parameters of the heat exchangers. The modeling has been illustrated with examples from actual plant systems. An integrated system consisting of Spent Fuel Pool, Primary Component Cooling Water, and Service Water System has been successfully modeled to predict their performance under normal operations and emergency conditions. System configurations are changed from the base model by using a command module.

  15. Thermal-Hydraulic Analyses Of The LS-VHTR

    SciTech Connect

    Cliff B. Davis; Grant L. Hawkes

    2006-06-01

    Thermal-hydraulic analyses were performed to evaluate the safety characteristics of the Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR). A one-dimensional model of the LS-VHTR was developed using the RELAP5-3D computer program. The thermal calculations from the one-dimensional model of a fuel block were benchmarked against a multi-dimensional finite element model. The RELAP5-3D model was used to simulate a transient initiated by loss of forced convection in which the Reactor Vessel Auxiliary Cooling System (RVACS) passively removed decay heat. Parametric calculations were performed to investigate the effects of various parameters, including bypass flow fraction, coolant channel diameter, and the coolant outlet temperature. Additional parametric calculations investigated the effects of an enhanced RVACS design, failure to scram, and radial/axial conduction in the core.

  16. Numerical simulation of thermal-hydraulic generators running in a single regime

    NASA Astrophysics Data System (ADS)

    Chioreanu, Nicolae; Mitran, Tudor; Rus, Alexandru; Beles, Horia

    2014-06-01

    The paper presents the basis for the design of thermal-hydraulic generators running in a single regime. The thermal-hydraulic generators in a single regime running represent an absolute novelty worldwide (a pioneer invention). Based on the methodology concerning this subject, the design calculus for an experimental model was developed.

  17. Coupled thermal-hydraulic-chemical modelling of enhanced geothermal systems

    NASA Astrophysics Data System (ADS)

    Bächler, D.; Kohl, T.

    2005-05-01

    The study investigates thermal-, hydraulic- and chemically coupled processes of enhanced geothermal systems (EGS). On the basis of the two existing numerical codes, the finite element program FRACTURE and the geochemical module of CHEMTOUGH, FRACHEM was developed, to simulate coupled thermal-hydraulic-chemical (THC) processes, accounting for the Soultz specific conditions such as the high salinity of the reservoir fluid and the high temperatures. The finite element part calculates the thermal and hydraulic field and the geochemical module the chemical processes. According to the characteristics of the Soultz EGS reservoir, the geochemical module was modified. (i) The Debye-Huckel approach was replaced by the Pitzer formalism. (ii) New kinetic laws for calcite, dolomite, quartz and pyrite were implemented. (iii) The porosity-permeability relation was replaced by a new relation for fractured rock. (iv) The possibility of re-injecting the produced fluid was implemented. The sequential non-iterative approach (SNIA) was used to couple transport and reactions. Sensitivity analyses proved the proper functionality of FRACHEM, but highlighted the sensitivity of the SNIA approach to time steps. To quantify the FRACHEM results, a comparative simulation with the code SHEMAT was conducted, which validated FRACHEM. Coupled THC processes in a fractured zone in the Soultz reservoir at 3500 m (T0= 165 °C), which occur as a result of the injection of fluid (Tinj= 65 °C) at one end of the zone and the production at the other end, were modelled for 2 yr. Calcite is the most reactive mineral and therefore the porosity and permeability evolution results from the calcite reactions: near the injection point, porosity and permeability increase and near the production well they decrease. After 2 yr, the system seems to be very close to steady-state. Therefore, mineral dissolution and precipitation during the circulation of the fluid in the reservoir do not represent a limiting factor on

  18. Thermal-hydraulic processes involved in loss of residual heat removal during reduced inventory operation. Revision 1

    SciTech Connect

    Fletcher, C.D.; McHugh, P.R.; Naff, S.A.; Johnsen, G.W.

    1991-02-01

    This paper identifies the topics needed to understand pressurized water reactor response to an extended loss of residual heat removal event during refueling and maintenance outages. By identifying the possible plant conditions and cooling methods that would be used for each cooling mode, the controlling thermal-hydraulic processes and phenomena were identified. Controlling processes and phenomena include: gravity drain, core water boil-off, and reflux cooling processes. Important subcategories of the reflux cooling processes include: the initiation of reflux cooling from various plant conditions, the effects of air on reflux cooling, core level depression effects, issues regarding the steam generator secondaries, and the special case of boiler-condenser cooling with once-through steam generators. 25 refs., 6 figs., 1 tab.

  19. 75 FR 80544 - NUREG-1953, Confirmatory Thermal-Hydraulic Analysis To Support Specific Success Criteria in the...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-12-22

    ... COMMISSION NUREG-1953, Confirmatory Thermal-Hydraulic Analysis To Support Specific Success Criteria in the..., ``Confirmatory Thermal-Hydraulic Analysis to Support Specific Success Criteria in the Standardized Plant Analysis... . SUPPLEMENTARY INFORMATION: NUREG-1953, ``Confirmatory Thermal-Hydraulic Analysis to Support Specific...

  20. ORNL rod-bundle heat-transfer test data. Volume 6. Thermal-hydraulic test facility experimental data report for test 3. 05. 5B - double-ended cold-leg break simulation

    SciTech Connect

    Mullins, C.B.; Felde, D.K.; Sutton, A.G.; Gould, S.S.; Morris, D.G.; Robinson, J.J.; Schwinkendorf, K.N.

    1982-05-18

    Thermal-Hydraulic Test Facility (THTF) Test 3.05.5B was conducted by members of the ORNL PWR Blowdown Heat Transfer Separate-Effects Program on July 3, 1980. The objective of the program is to investigate heat transfer phenomena believed to occur in PWRs during accidents, including small and large break loss-of-coolant accidents. Test 3.05.5B was designed to provide transient thermal-hydraulics data in rod bundle geometry under reactor accident-type conditions. Reduced instrument responses are presented. Also included are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers.

  1. Simulation of the passive condensation cooling tank of the PASCAL test facility using the component thermal-hydraulic analysis code CUPID

    SciTech Connect

    Cho, H. K.; Lee, S. J.; Kang, K. H.; Yoon, H. Y.

    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)

  2. Engineered Barrier Systems Thermal-Hydraulic-Chemical Column Test Report

    SciTech Connect

    W.E. Lowry

    2001-12-13

    The Engineered Barrier System (EBS) Thermal-Hydraulic-Chemical (THC) Column Tests provide data needed for model validation. The EBS Degradation, Flow, and Transport Process Modeling Report (PMR) will be based on supporting models for in-drift THC coupled processes, and the in-drift physical and chemical environment. These models describe the complex chemical interaction of EBS materials, including granular materials, with the thermal and hydrologic conditions that will be present in the repository emplacement drifts. Of particular interest are the coupled processes that result in mineral and salt dissolution/precipitation in the EBS environment. Test data are needed for thermal, hydrologic, and geochemical model validation and to support selection of introduced materials (CRWMS M&O 1999c). These column tests evaluated granular crushed tuff as potential invert ballast or backfill material, under accelerated thermal and hydrologic environments. The objectives of the THC column testing are to: (1) Characterize THC coupled processes that could affect performance of EBS components, particularly the magnitude of permeability reduction (increases or decreases), the nature of minerals produced, and chemical fractionation (i.e., concentrative separation of salts and minerals due to boiling-point elevation). (2) Generate data for validating THC predictive models that will support the EBS Degradation, Flow, and Transport PMR, Rev. 01.

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

    NASA Astrophysics Data System (ADS)

    Ivanov, Aleksandar; Sanchez, Victor; Ivanov, Kostadin

    2014-06-01

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

  4. PRATHAM: Parallel Thermal Hydraulics Simulations using Advanced Mesoscopic Methods

    SciTech Connect

    Joshi, Abhijit S; Jain, Prashant K; Mudrich, Jaime A; Popov, Emilian L

    2012-01-01

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

  5. COBRA-SFS. Thermal Hydraulic Analysis of Spent Fuel Casks

    SciTech Connect

    Michener, T.E.; Rector, D.R.; Cuta, J.M.; Enderlin, C.W.

    1995-09-01

    COBRA-SFS (Spent Fuel Storage) is a code for thermal-hydraulic analysis of multi-assembly spent fuel storage and transportation systems. It uses a lumped parameter finite difference approach to predict flow and temperature distributions in spent fuel storage systems and fuel assemblies, under forced and natural convection heat transfer conditions. Derived from the COBRA family of codes, which have been extensively evaluated against in-pile and out-of-pile data, COBRA-SFS retains all the important features of the COBRA codes for single phase fluid analysis, and extends the range application to include problems with two-dimensional radiative and three-dimensional conductive heat transfer. COBRA-SFS has been used to analyze various single- and multi-assembly spent fuel storage systems containing unconsolidated and consolidated fuel rods, with a variety of fill media, including air, helium and vacuum. Cycle 0 of COBRA-SFS was released in 1986. Subsequent applications of the code led to development of additional capabilities, which resulted in the release of Cycle 1 in February 1989. Since then, the code has undergone an independent technical review as part of a submittal to the Nuclear Regulatory Commission for a generic license to apply the code to spent fuel storage system analysis. Modifications and improvements to the code have been combined to form the latest release of the code, Cycle 2.

  6. Use of laser flow visualization techniques in reactor component thermal-hydraulic studies

    SciTech Connect

    Oras, J.J.; Kasza, K.E.

    1984-01-01

    To properly design reactor components, an understanding of the various thermal hydraulic phenomena, i.e., thermal stratification flow channeling, recirculation regions, shear layers, etc., is necessary. In the liquid metal breeder reactor program, water is commonly used to replace sodium in experimental testing to facilitate the investigations, (i.e., reduce cost and allow fluid velocity measurement or flow pattern study). After water testing, limited sodium tests can be conducted to validate the extrapolation of the water results to sodium. This paper describes a novel laser flow visualization technique being utilized at ANL together with various examples of its use and plans for further development. A 3-watt argon-ion laser, in conjunction with a cylindrical opticallens, has been used to create a thin (approx. 1-mm) intense plane of laser light for the illuminiation of various flow tracers in precisely defined regions of interest within a test article having windows. Both fluorescing dyes tuned to the wavelength of the laser light (to maximize brightness and sharpness of flow image) and small (< 0.038-mm, 0.0015-in. dia.) opaque, nearly neutrally buoyant polystyrene spheres (to ensure that the particles trace out the fluid motion) have been used as flow tracers.

  7. Thermal Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Water: Performance and Stability

    NASA Astrophysics Data System (ADS)

    Lisowski, Darius D.

    This experimental study investigated the thermal hydraulic behavior and boiling mechanisms present in a scaled reactor cavity cooling system (RCCS). The experimental facility reflects a ¼ scale model of one conceptual design for decay heat removal in advanced GenIV nuclear reactors. Radiant heaters supply up to 25 kW/m2 onto a three parallel riser tube and cooling panel test section assembly, representative of a 5° sector model of the full scale concept. Derived similarity relations have preserved the thermal hydraulic flow patterns and integral system response, ensuring relevant data and similarity among scales. Attention will first be given to the characterization of design features, form and heat losses, nominal behavior, repeatability, and data uncertainty. Then, tests performed in single-phase have evaluated the steady-state behavior. Following, the transition to saturation and subsequent boiling allowed investigations onto four parametric effects at two-phase flow and will be the primary focus area of remaining analysis. Baseline conditions at two-phase flow were defined by 15.19 kW of heated power and 80% coolant inventory, and resulted in semi-periodic system oscillations by the mechanism of hydrostatic head fluctuations. Void generation was the result of adiabatic expansion of the fluid due to a reduction in hydrostatic head pressure, a phenomena similar to flashing. At higher powers of 17.84 and 20.49 kW, this effect was augmented, creating large flow excursions that followed a smooth and sinusoidal shaped path. Stabilization can occur if the steam outflow condition incorporates a nominal restriction, as it will serve to buffer the short time scale excursions of the gas space pressure and dampen oscillations. The influences of an inlet restriction, imposed by an orifice plate, introduced subcooling boiling within the heated core and resulted in chaotic interactions among the parallel risers. The penultimate parametric examined effects of boil-off and

  8. Validation and Calibration of Nuclear Thermal Hydraulics Multiscale Multiphysics Models - Subcooled Flow Boiling Study

    SciTech Connect

    Anh Bui; Nam Dinh; Brian Williams

    2013-09-01

    In addition to validation data plan, development of advanced techniques for calibration and validation of complex multiscale, multiphysics nuclear reactor simulation codes are a main objective of the CASL VUQ plan. Advanced modeling of LWR systems normally involves a range of physico-chemical models describing multiple interacting phenomena, such as thermal hydraulics, reactor physics, coolant chemistry, etc., which occur over a wide range of spatial and temporal scales. To a large extent, the accuracy of (and uncertainty in) overall model predictions is determined by the correctness of various sub-models, which are not conservation-laws based, but empirically derived from measurement data. Such sub-models normally require extensive calibration before the models can be applied to analysis of real reactor problems. This work demonstrates a case study of calibration of a common model of subcooled flow boiling, which is an important multiscale, multiphysics phenomenon in LWR thermal hydraulics. The calibration process is based on a new strategy of model-data integration, in which, all sub-models are simultaneously analyzed and calibrated using multiple sets of data of different types. Specifically, both data on large-scale distributions of void fraction and fluid temperature and data on small-scale physics of wall evaporation were simultaneously used in this work’s calibration. In a departure from traditional (or common-sense) practice of tuning/calibrating complex models, a modern calibration technique based on statistical modeling and Bayesian inference was employed, which allowed simultaneous calibration of multiple sub-models (and related parameters) using different datasets. Quality of data (relevancy, scalability, and uncertainty) could be taken into consideration in the calibration process. This work presents a step forward in the development and realization of the “CIPS Validation Data Plan” at the Consortium for Advanced Simulation of LWRs to enable

  9. Isotope Production Facility Conceptual Thermal-Hydraulic Design Review and Scoping Calculations

    SciTech Connect

    Pasamehmetoglu, K.O.; Shelton, J.D.

    1998-08-01

    The thermal-hydraulic design of the target for the Isotope Production Facility (IPF) is reviewed. In support of the technical review, scoping calculations are performed. The results of the review and scoping calculations are presented in this report.

  10. Thermal Hydraulics of the Very High Temperature Gas Cooled Reactor

    SciTech Connect

    Chang Oh; Eung Kim; Richard Schultz; Mike Patterson; Davie Petti

    2009-10-01

    The U.S Department of Energy (DOE) is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R&D) that will be critical to the success of the NGNP, primarily in the areas of: • High temperature gas reactor fuels behavior • High temperature materials qualification • Design methods development and validation • Hydrogen production technologies • Energy conversion. This paper presents current R&D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.

  11. Momentum Integral Network Method for Thermal-Hydraulic Systems Analysis.

    2000-11-20

    EPIPE is used for design or design evaluation of complex large piping systems. The piping systems can be viewed as a network of straight pipe elements (or tangents) and curved elements (pipe bends) interconnected at joints (or nodes) with intermediate supports and anchors. The system may be subject to static loads such as thermal, dead weight, internal pressure, or dynamic loads such as earthquake motions and flow-induced vibrations, or any combination of these. MINET (Momentummore » Integral NETwork) was developed for the transient analysis of intricate fluid flow and heat transfer networks, such as those found in the balance of plant in power generating facilities. It can be utilized as a stand-alone program or interfaced to another computer program for concurrent analysis. Through such coupling, a computer code limited by either the lack of required component models or large computational needs can be extended to more fully represent the thermal hydraulic system thereby reducing the need for estimating essential transient boundary conditions. The MINET representation of a system is one or more networks of volumes, segments, and boundaries linked together via heat exchangers only, i.e., heat can transfer between networks, but fluids cannot. Volumes are used to represent tanks or other volume components, as well as locations in the system where significant flow divisions or combinations occur. Segments are composed of one or more pipes, pumps, heat exchangers, turbines, and/or valves each represented by one or more nodes. Boundaries are simply points where the network interfaces with the user or another computer code. Several fluids can be simulated, including water, sodium, NaK, and air.« less

  12. Momentum Integral Network Method for Thermal-Hydraulic Systems Analysis.

    SciTech Connect

    2000-11-20

    EPIPE is used for design or design evaluation of complex large piping systems. The piping systems can be viewed as a network of straight pipe elements (or tangents) and curved elements (pipe bends) interconnected at joints (or nodes) with intermediate supports and anchors. The system may be subject to static loads such as thermal, dead weight, internal pressure, or dynamic loads such as earthquake motions and flow-induced vibrations, or any combination of these. MINET (Momentum Integral NETwork) was developed for the transient analysis of intricate fluid flow and heat transfer networks, such as those found in the balance of plant in power generating facilities. It can be utilized as a stand-alone program or interfaced to another computer program for concurrent analysis. Through such coupling, a computer code limited by either the lack of required component models or large computational needs can be extended to more fully represent the thermal hydraulic system thereby reducing the need for estimating essential transient boundary conditions. The MINET representation of a system is one or more networks of volumes, segments, and boundaries linked together via heat exchangers only, i.e., heat can transfer between networks, but fluids cannot. Volumes are used to represent tanks or other volume components, as well as locations in the system where significant flow divisions or combinations occur. Segments are composed of one or more pipes, pumps, heat exchangers, turbines, and/or valves each represented by one or more nodes. Boundaries are simply points where the network interfaces with the user or another computer code. Several fluids can be simulated, including water, sodium, NaK, and air.

  13. Current and anticipated uses of thermal-hydraulic codes in NFI

    SciTech Connect

    Tsuda, K.; Takayasu, M.

    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.

  14. Thermal hydraulics modeling of the US Geological Survey TRIGA reactor

    NASA Astrophysics Data System (ADS)

    Alkaabi, Ahmed K.

    The Geological Survey TRIGA reactor (GSTR) is a 1 MW Mark I TRIGA reactor located in Lakewood, Colorado. Single channel GSTR thermal hydraulics models built using RELAP5/MOD3.3, RELAP5-3D, TRACE, and COMSOL Multiphysics predict the fuel, outer clad, and coolant temperatures as a function of position in the core. The results from the RELAP5/MOD3.3, RELAP5-3D, and COMSOL models are similar. The TRACE model predicts significantly higher temperatures, potentially resulting from inappropriate convection correlations. To more accurately study the complex fluid flow patterns within the core, this research develops detailed RELAP5/MOD3.3 and COMSOL multichannel models of the GSTR core. The multichannel models predict lower fuel, outer clad, and coolant temperatures compared to the single channel models by up to 16.7°C, 4.8°C, and 9.6°C, respectively, as a result of the higher mass flow rates predicted by these models. The single channel models and the RELAP5/MOD3.3 multichannel model predict that the coolant temperatures in all fuel rings rise axially with core height, as the coolant in these models flows predominantly in the axial direction. The coolant temperatures predicted by the COMSOL multichannel model rise with core height in the B-, C-, and D-rings and peak and then decrease in the E-, F-, and G-rings, as the coolant tends to flow from the bottom sides of the core to the center of the core in this model. Experiments at the GSTR measured coolant temperatures in the GSTR core to validate the developed models. The axial temperature profiles measured in the GSTR show that the flow patterns predicted by the COMSOL multichannel model are consistent with the actual conditions in the core. Adjusting the RELAP5/MOD3.3 single and multichannel models by modifying the axial and cross-flow areas allow them to better predict the GSTR coolant temperatures; however, the adjusted models still fail to predict accurate axial temperature profiles in the E-, F-, and G-rings.

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

    SciTech Connect

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

    2012-07-01

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

  16. Model development and calibration for the coupled thermal, hydraulic and mechanical phenomena of the bentonite

    SciTech Connect

    Chijimatsu, M.; Borgesson, L.; Fujita, T.; Jussila, P.; Nguyen, S.; Rutqvist, J.; Jing, L.; Hernelind, J.

    2009-02-01

    In Task A of the international DECOVALEX-THMC project, five research teams study the influence of thermal-hydro-mechanical (THM) coupling on the safety of a hypothetical geological repository for spent fuel. In order to improve the analyses, the teams calibrated their bentonite models with results from laboratory experiments, including swelling pressure tests, water uptake tests, thermally gradient tests, and the CEA mock-up THM experiment. This paper describes the mathematical models used by the teams, and compares the results of their calibrations with the experimental data.

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

    SciTech Connect

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

    1996-06-01

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

  18. Thermal-hydraulic analysis for changing feedwater check valve leakage rate testing methodology

    SciTech Connect

    Fuller, R.; Harrell, J.

    1996-12-01

    The current design and testing requirements for the feedwater check valves (FWCVs) at the Grand Gulf Nuclear Station are established from original licensing requirements that necessitate extremely restrictive air testing with tight allowable leakage limits. As a direct result of these requirements, the original high endurance hard seats in the FWCVs were modified with elastomeric seals to provide a sealing surface capable of meeting the stringent air leakage limits. However, due to the relatively short functional life of the elastomeric seals compared to the hard seats, the overall reliability of the sealing function actually decreased. This degraded performance was exhibited by frequent seal failures and subsequent valve repairs. The original requirements were based on limited analysis and the belief that all of the high energy feedwater vaporized during the LOCA blowdown. These phenomena would have resulted in completely voided feedwater lines and thus a steam environment within the feedwater leak pathway. To challenge these criteria, a comprehensive design basis accident analysis was developed using the RELAP5/MOD3.1 thermal-hydraulic code. Realistic assumptions were used to more accurately model the post-accident fluid conditions within the feedwater system. The results of this analysis demonstrated that no leak path exists through the feedwater lines during the reactor blowdown phase and that sufficient subcooled water remains in various portions of the feedwater piping to form liquid water loop seals that effectively isolate this leak path. These results provided the bases for changing the leak testing requirements of the FWCVs from air to water. The analysis results also established more accurate allowable leakage limits, determined the real effective margins associated with the FWCV safety functions, and led to design changes that improved the overall functional performance of the valves.

  19. Thermal hydraulic analysis for the Oregon State TRIGA reactor using RELAP5-3D

    SciTech Connect

    Marcum, W.R.; Woods, B.G.; Hartman, M.

    2008-07-15

    Thermal hydraulic analyses have being conducted at Oregon State University (OSU) in support of the conversion of the OSU TRIGA reactor (OSTR) core from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel as part of the Reduced Enrichment for Research and Test Reactors program. The goals of the thermal hydraulic analyses were to calculate natural circulation flow rates, coolant temperatures and fuel temperatures as a function of core power for both the HEU and LEU cores; calculate peak values of fuel temperature, cladding temperature, surface heat flux as well as departure from nuclear boiling ratio (DNBR) for steady state and pulse operation; and perform accident analyses for the accident scenarios identified in the OSTR safety analysis report. RELAP5-3D Version 2.4.2 was implemented to develop a model for the thermal hydraulic study. The OSTR core conversion is planned to take place in late 2008. (author)

  20. SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

    SciTech Connect

    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.

  1. Thermal hydraulic analysis of advanced Pb-Bi cooled NPP using natural circulation

    NASA Astrophysics Data System (ADS)

    Novitrian, Su'ud, Zaki; Waris, Abdul

    2012-06-01

    We present thermal hydraulic analysis for a low power advanced nuclear reactor cooled by lead-bismuth eutectic. In this work is to study the thermal hydraulic analysis of a low power SPINNOR (Small Power Reactor, Indonesia, No On-site Refuelling) reactor with 125 MWth which a design a core with very small volume and fuel column height, resulting in a negative coolant temperature coefficient and very low channel pressure drop. And also at full power the heat can be completely removed by natural circulation in the primary circuit, thus eliminating the needs for pumps.

  2. Current and anticipated uses of thermal hydraulic codes at the Japan Atomic Energy Research Institute

    SciTech Connect

    Akimoto, Hajime; Kukita; Ohnuki, Akira

    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.

  3. Detailed thermal-hydraulic computation into a containment building

    SciTech Connect

    Caruso. A.; Flour, I.; Simonin, O.

    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.

  4. Applicability of RELAP5-3D for Thermal-Hydraulic Analyses of a Sodium-Cooled Actinide Burner Test Reactor

    SciTech Connect

    C. B. Davis

    2006-07-01

    The Actinide Burner Test Reactor (ABTR) is envisioned as a sodium-cooled, fast reactor that will burn the actinides generated in light water reactors to reduce nuclear waste and ease proliferation concerns. The RELAP5-3D computer code is being considered as the thermal-hydraulic system code to support the development of the ABTR. An evaluation was performed to determine the applicability of RELAP5-3D for the analysis of a sodium-cooled fast reactor. The applicability evaluation consisted of several steps, including identifying the important transients and phenomena expected in the ABTR, identifying the models and correlations that affect the code’s calculation of the important phenomena, and evaluating the applicability of the important models and correlations for calculating the important phenomena expected in the ABTR. The applicability evaluation identified code improvements and additional models needed to simulate the ABTR. The accuracy of the calculated thermodynamic and transport properties for sodium was also evaluated.

  5. Advances in modelling of condensation phenomena

    SciTech Connect

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

    1997-07-01

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

  6. Thermal-hydraulic assessment of concrete storage cubicle with horizontal 3013 canisters

    SciTech Connect

    HEARD, F.J.

    1999-04-08

    The FIDAP computer code was used to perform a series of analyses to assess the thermal-hydraulic performance characteristics of the concrete plutonium storage cubicles, as modified for the horizontal placement of 3013 canisters. Four separate models were developed ranging from a full height model of the storage cubicle to a very detailed standalone model of a horizontal 3013 canister.

  7. FFTF thermal-hydraulic testing results affecting piping and vessel component design in LMFBR's

    SciTech Connect

    Stover, R.L.; Beaver, T.R.; Chang, S.C.

    1983-01-01

    The Fast Flux Test Facility completed four years of pre-operational testing in April 1982. This paper describes thermal-hydraulic testing results from this period which impact piping and vessel component design in LMFBRs. Data discussed are piping flow oscillations, piping thermal stratification and vessel upper plenum stratification. Results from testing verified that plant design limits were met.

  8. Application of the ATHOS3 code for steam generator thermal hydraulics and fouling analysis

    SciTech Connect

    Srikantiah, G.S.; Chappidi, P.R.

    1996-09-01

    The steam generator is a most important component in the coolant loop of Pressurized Water Reactors. Although designed for a 30--40 year operating life, severe material degradation problems have occurred within the first ten years of operation. Performance and reliability evaluations are required on a continuing basis to develop solutions and design modifications to ensure reliable operation of these systems. Thermal hydraulic analysis provides basic information such as velocity and void fraction distributions within the secondary side of the steam generator needed for the evaluation of sludge deposition, bundle fouling, tube vibration, fretting, wear and fatigue. This paper presents detailed thermal hydraulic analysis of several steam generator designs, and analyzes the correlation between thermal hydraulic distributions, sludge deposition and bundle fouling using a recent model for sludge transport and deposition. The correlation between thermal hydraulic distributions and other degradation mechanisms such as circumferential cracking of tubes is also presented. The results show that there is a strong correlation between flow velocity, void fraction and sludge deposition. The calculated sludge deposit potential maps are in very good agreement with the observed results within operating steam generators.

  9. Thermal hydraulic study of the ESPRESSO blanket for a Tandem Mirror Reactor

    SciTech Connect

    Raffray, A.R.; Hoffman, M.A.

    1986-02-01

    This paper deals primarily with the thermal-hydraulic design and some critical thermomechanical aspects of the proposed ESPRESSO blanket for the Tandem Mirror Fusion Reactor. This conceptual design was based on the same physics as used in the MARS study.

  10. IAEA Coordinated Research Project on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis

    SciTech Connect

    Strydom, Gerhard; Bostelmann, F.

    2015-09-01

    The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal-hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, modelling and computational algorithms. (The interested reader is referred to the large body of published SA and UA literature for a more complete overview of the various types of uncertainties, methodologies and results obtained). SA is helpful for ranking the various sources of uncertainty and error in the results of core analyses. SA and UA are required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. SA and UA can guide experimental, modelling, and algorithm research and development. Current SA and UA rely either on derivative-based methods such as stochastic sampling methods or on generalized perturbation theory to obtain sensitivity coefficients. Neither approach addresses all needs. In order to benefit from recent advances in modelling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more robust and well validated calculation tools to meet design target accuracies. In February 2009, the Technical Working Group on Gas-Cooled Reactors (TWG-GCR) of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Program (CRP) on

  11. THERMAL HYDRAULIC ANALYSIS OF A LIQUID-METAL-COOLED NEUTRON SPALLATION TARGET

    SciTech Connect

    W. GREGORY; R. MARTIN; T. VALACHOVIC

    2000-07-01

    We have carried out numerical simulations of the thermal hydraulic behavior of a neutron spallation target where liquid metal lead-bismuth serves as both coolant and as a neutron spallation source. The target is one of three designs provided by the Institute of Physics and Power Engineering (IPPE) in Russia. This type of target is proposed for Accelerator-driven Transmutation of Waste (ATW) to eliminate plutonium from hazardous fission products. The thermal hydraulic behavior was simulated by use of a commercial CFD computer code called CFX. Maximum temperatures in the diaphragm window and in the liquid lead were determined. In addition the total pressure drop through the target was predicted. The results of the CFX analysis were close to those results predicted by IPPE in their preliminary analysis.

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

    SciTech Connect

    Smith, Thomas Michael; Shadid, John N.; Pawlowski, Roger P.; Cyr, Eric C.; Wildey, Timothy Michael

    2014-01-01

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

  13. Computational Fluid Dynamics in Support of the SNS Liquid Mercury Thermal-Hydraulic Analysis

    SciTech Connect

    Siman-Tov, M.; Wendel, M.W.; Yoder, G.L.

    1999-11-14

    Experimental and computational thermal-hydraulic research is underway to support the liquid mercury target design for the Spallation Neutron Source (SNS) facility. The SNS target will be subjected to internal nuclear heat generation that results from pulsed proton beam collisions with the mercury nuclei. Recirculation and stagnation zones within the target are of particular concern because of the likelihood that they will result in local hot spots and diminished heat removal from the target structure. Computational fluid dynamics (CFD) models are being used as a part of this research. Recent improvements to the 3D target model include the addition of the flow adapter which joins the inlet/outlet coolant pipes to the target body and an updated heat load distribution at the new baseline proton beam power level of 2 MW. Two thermal-hydraulic experiments are planned to validate the CFD model.

  14. Thermal Hydraulic Characteristics of Fuel Defects in Plate Type Nuclear Research Reactors

    SciTech Connect

    Bodey, Isaac T

    2014-05-01

    Turbulent flow coupled with heat transfer is investigated for a High Flux Isotope Reactor (HFIR) fuel plate. The Reynolds Averaged Navier-Stokes Models are used for fluid dynamics and the transfer of heat from a thermal nuclear fuel plate using the Multi-physics code COMSOL. Simulation outcomes are compared with experimental data from the Advanced Neutron Source Reactor Thermal Hydraulic Test Loop. The computational results for the High Flux Isotope Reactor core system provide a more physically accurate simulation of this system by modeling the turbulent flow field in conjunction with the diffusion of thermal energy within the solid and fluid phases of the model domain. Recommendations are made regarding Nusselt number correlations and material properties for future thermal hydraulic modeling efforts

  15. RAMONA-3B/MINET composite representation of BWR thermal-hydraulic systems

    SciTech Connect

    Van Tuyle, G.J.; Slovik, G.; Cazzoli, E.G.; Nepsee, T.C.; Guppy, J.G.

    1985-01-01

    The modification and interfacing of two computer codes, RAMONA-3B and MINET, for the thermal hydraulic transient analysis of a Boiling Water Reactor nuclear steam supply system, is described. The RAMONA-3B code provides for multi-channel thermal hydraulics and three-dimensional (or one-dimensional) neutron kinetics analysis of a boiling water reactor core. The RAMONA-3B system representation terminates at the end of the steam line and at the junction of the feedwater line at the vessel inlet. By interfacing RAMONA-3B with MINET, a generic balance-of-plant systems analysis code, a complete BWR systems code with detailed core modeling was obtained. The result is a code of particular importance to the analysis of transients such as ATWS. A comparison between the 3-D and 1-D neutronics representation is provided, along with a test case utilizing the composite RAMONA-3B/MINET code.

  16. Cold source moderator vessel development for the High Flux Isotope Reactor: Thermal-hydraulic studies

    SciTech Connect

    Williams, P.T.; Lucas, A.T.; Wendel, M.W.

    1998-07-01

    A project is underway at Oak Ridge National Laboratory (ORNL) to design, test, and install a cold neutron source facility in the High Flux Isotope Reactor (HFIR). This new cold source employs supercritical hydrogen at cryogenic temperatures both as the medium for neutron moderation and as the working fluid for removal of internally-generated nuclear heating. The competing design goals of minimizing moderator vessel mass and providing adequate structural integrity for the vessel motivated the requirement of detailed multidimensional thermal-hydraulic analyses of the moderator vessel as a critical design subtask. This paper provides a summary review of the HFIR cold source moderator vessel design and a description of the thermal-hydraulic studies that were carried out to support the vessel development.

  17. Condensation heat transfer coefficient with noncondensible gases for heat transfer in thermal hydraulic codes

    SciTech Connect

    Banerjee, S.; Hassan, Y.A.

    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.

  18. CFX Analysis of the CANDU Moderator Thermal-Hydraulics in the Stern Lab. Test Facility

    NASA Astrophysics Data System (ADS)

    Kim, Hyoung Tae

    2014-06-01

    A numerical calculation with the commercial CFD code CFX is conducted for a test facility simulating the CANDU moderator thermal-hydraulics. Two kinds of moderator thermal-hydraulic tests at Stern Laboratories Inc. were performed in the full geometric configuration of the CANDU moderator circulating vessel, which is called a Calandria, housing a matrix of horizontal rod bundles simulating the Calandria tubes. The first of these tests is the pressure drop measurement of a cross flow in the horizontal rod bundles. The other is the local temperature measurement on the cross section of the horizontal cylinder vessel simulating the Calandria. In the present study the full geometric details of the Calandria are incorporated in the grid generation of the computational domain to which the boundary conditions for each experiment are applied. The numerical solutions are reviewed and compared with the available test data.

  19. Current and anticipated uses of thermal-hydraulic codes in Germany

    SciTech Connect

    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.

  20. Two-dimensional thermal-hydraulics analyses of the Pellet Bed Reactor for Nuclear Thermal Propulsion

    NASA Technical Reports Server (NTRS)

    Morley, Nicholas J.; El-Genk, Mohamed S.

    1993-01-01

    Thermal-hydraulics design and analyses of the Pellet Bed Reactor for nuclear thermal propulsion are performed using the nuclear propulsion thermal-hydraulic analysis model to determine the 2D steady-state temperature, pressure, and flow fields in the core and optimize the orificing in the hot-frit to avoid hot spots in the core at full power operation. Results show that by properly adjusting the axial porosity profile in the hot frit, hot spots in the core can be essentially eliminated during full power operation. This important accomplishment is achieved at the expense of slightly larger pressure losses in the core because of flow restriction at the hot frit. However, the overall pressure losses is only about 11 percent of the propellant inlet pressure.

  1. TWIST: a transient two-dimensional intra-subassembly thermal hydraulics model for LMFBRs

    SciTech Connect

    Khatib-Rahbar, M.; Cazzoli, E.G.

    1984-06-03

    Mathematical models and numerical methods for a two-dimensional porous body simulation of steady state and transient thermal-hydraulics conditions in LMFBR subassemblies resulting in the TWIST computer code are presented. Comparison of calculated results to steady state and transient out-of-pile sodium experiments show good agreement for cross-assembly temperature distributions for a wide range of heat transfer and flow conditions.

  2. Test program element II blanket and shield thermal-hydraulic and thermomechanical testing, experimental facility survey

    SciTech Connect

    Ware, A.G.; Longhurst, G.R.

    1981-12-01

    This report presents results of a survey conducted by EG and G Idaho to determine facilities available to conduct thermal-hydraulic and thermomechanical testing for the Department of Energy Office of Fusion Energy First Wall/Blanket/Shield Engineering Test Program. In response to EG and G queries, twelve organizations (in addition to EG and G and General Atomic) expressed interest in providing experimental facilities. A variety of methods of supplying heat is available.

  3. Reactor thermal-hydraulic FY 1986 status report for the multimegawatt Space Nuclear Power Program

    SciTech Connect

    Krotiuk, W.J.; Antoniak, Z.I.

    1986-10-01

    PNL's 1986 activities can be divided into three basic areas: code assessment, correlation assessment and experimental activities. The ultimate goal of all these activities is developing computer codes and verifying their use to perform the thermal-hydraulic analysis and design of the reactor core and plenum of the various proposed concepts. To perform this task as assessment is made of existing computer codes, models, correlations, and microgravity experimental data.

  4. Visualization tools for uncertainty and sensitivity analyses on thermal-hydraulic transients

    NASA Astrophysics Data System (ADS)

    Popelin, Anne-Laure; Iooss, Bertrand

    2014-06-01

    In nuclear engineering studies, uncertainty and sensitivity analyses of simulation computer codes can be faced to the complexity of the input and/or the output variables. If these variables represent a transient or a spatial phenomenon, the difficulty is to provide tool adapted to their functional nature. In this paper, we describe useful visualization tools in the context of uncertainty analysis of model transient outputs. Our application involves thermal-hydraulic computations for safety studies of nuclear pressurized water reactors.

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

    SciTech Connect

    Robert C. O'Brien; Andrew C. Klein; William T. Taitano; Justice Gibson; Brian Myers; Steven D. Howe

    2011-02-01

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

  6. Thermal hydraulic characteristics study of prototype NET and CEA cable-in-conduit conductors (CICCs)

    SciTech Connect

    Maekawa, Ryuji

    1995-10-31

    The thermal hydraulic characteristics of low temperature helium in a Cable-in-Conduit Conductor (CICC) significantly affects the overall design and performance of the associated large scale superconducting magnet system. It is essential to understand the transient and steady state behavior of the helium in the conductor. Throughout the development of CICCs, the reduction of flow impedance has been one of the key factors to improving the overall pressure drop. The newly developed CICC for the ITER project has a hybrid cooling scheme: a central channel that is surrounded by bundles, for which the thermal hydraulic characteristics are not well understood. This thesis describes an experimental and analytical investigation of thermal hydraulic characteristics of low temperature helium in conventional and hybrid CICCS. Pressure drop measurements for both NET and CEA conductors have been conducted, using low temperature helium and liquid nitrogen to obtain a range of Reynolds numbers. The results are correlated with classical friction factor and Reynolds number analysis. The flow impedance reduction of the CEA conductor is described by measures of a developed flow model. Thermally induced flow in the CEA conductor has been studied with an inductive heating method. The induced velocity in the central channel is measured by a Pitot tube with steady state Reynolds number up to {approximately}7000. The transient pressure wave propagation has been recorded with pressure transducers placed equally along the conductor. The supercritical helium temperature in the central channel has been measured with the thermometer probe. However, the reduction of the central channel area significantly affects the overall thermal hydraulic characteristics of the conductor. The results suggest the importance of the central channel. A transient heat transfer experiment studied the.transverse heat transfer mechanism in the CEA conductor. The temperatures in the central channel and bundle region

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

    NASA Astrophysics Data System (ADS)

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

    2015-09-01

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

  8. Thermal hydraulic behavior and efficiency analysis of an all-vanadium redox flow battery

    NASA Astrophysics Data System (ADS)

    Xiong, Binyu; Zhao, Jiyun; Tseng, K. J.; Skyllas-Kazacos, Maria; Lim, Tuti Mariana; Zhang, Yu

    2013-11-01

    Vanadium redox flow batteries (VRBs) are very competitive for large-capacity energy storage in power grids and in smart buildings due to low maintenance costs, high design flexibility, and long cycle life. Thermal hydraulic modeling of VRB energy storage systems is an important issue and temperature has remarkable impacts on the battery efficiency, the lifetime of material and the stability of the electrolytes. In this paper, a lumped model including auxiliary pump effect is developed to investigate the VRB temperature responses under different operating and surrounding environmental conditions. The impact of electrolyte flow rate and temperature on the battery electrical characteristics and efficiencies are also investigated. A one kilowatt VRB system is selected to conduct numerical simulations. The thermal hydraulic model is benchmarked with experimental data and good agreement is found. Simulation results show that pump power is sensitive to hydraulic design and flow rates. The temperature in the stack and tanks rises up about 10 °C under normal operating conditions for the stack design and electrolyte volume selected. An optimal flow rate of around 90 cm3 s-1 is obtained for the proposed battery configuration to maximize battery efficiency. The models developed in this paper can also be used for the development of a battery control strategy to achieve satisfactory thermal hydraulic performance and maximize energy efficiency.

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

    SciTech Connect

    Strizhov, V.; Kanukova, V.; Vinogradova, T.; Askenov, E.; Nikulshin, V.

    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.

  10. Feasibility Study on Thermal-Hydraulic Performance of Innovative Water Reactor for Flexible Fuel Cycle (FLWR)

    SciTech Connect

    Akira, Ohnuki; Kazuyuki, Takase; Masatoshi, Kureta; Hiroyuki, Yoshida; Hidesada, Tamai; Wei, Liu; Toru, Nakatsuka; Takeharu, Misawa; Hajime, Akimoto

    2006-07-01

    R and D project to investigate thermal-hydraulic performance in tight-lattice rod bundles of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) is started at Japan Atomic Energy Agency (JAEA) in collaboration with power company, reactor vendors, universities since 2002. The FLWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured LWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R and D items for the FLWR because of the tight lattice configuration. In this paper, we will show the R and D plan and summarize experimental studies. The experimental study is performed mainly using large-scale (37-rod bundle) test facility. Most important objective of the large-scale test is to resolve a fundamental subject whether the core cooling under a tight-lattice configuration is feasible. The characteristics of critical power and flow behavior are investigated under different geometrical configuration and boundary conditions. The configuration parameter is the gap between rods (FY2004) and the rod bowing (FY2005). We have confirmed the thermal-hydraulic feasibility from the experimental results. (authors)

  11. Evaluation of operational safety at Babcock and Wilcox Plants: Volume 2, Thermal-hydraulic results

    SciTech Connect

    Wheatley, P.D.; Davis, C.B.; Callow, R.A.; Fletcher, C.D.; Dobbe, C.A.; Beelman, R.J.

    1987-11-01

    The Nuclear Regulatory Commission has initiated a research program to develop a methodology to assess the operational performance of Babcock and Wilcox plants and to apply this methodology on a trial basis. The methodology developed for analyzing Babcock and Wilcox plants integrated methods used in both thermal-hydraulics and human factors and compared results with information used in the assessment of risk. The integrated methodology involved an evaluation of a selected plant for each pressurized water reactor vendor during a limited number of transients. A plant was selected to represent each vendor, and three transients were identified for analysis. The plants were Oconee Unit 1 for Babcock and Wilcox, H.B. Robinson Unit 2 for Westinghouse, and Calvert Cliffs Unit 1 for Combustion Engineering. The three transients were a complete loss of all feedwater, a small-break loss-of-coolant accident, and a steam-generator overfill with auxiliary feedwater. Included in the integrated methodology was an assessment of the thermal-hydraulic behavior, including event timing, of the plants during the three transients. Thermal-hydraulic results are presented in this volume (Volume 2) of the report. 26 refs., 30 figs., 7 tabs.

  12. Lead Coolant Test Facility Systems Design, Thermal Hydraulic Analysis and Cost Estimate

    SciTech Connect

    Soli Khericha; Edwin Harvego; John Svoboda; Ryan Dalling

    2012-01-01

    The Idaho National Laboratory prepared a preliminary technical and functional requirements (T&FR), thermal hydraulic design and cost estimate for a lead coolant test facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic coolant. Based on review of current world lead or lead-bismuth test facilities and research needs listed in the Generation IV Roadmap, five broad areas of requirements were identified as listed: (1) Develop and Demonstrate Feasibility of Submerged Heat Exchanger; (2) Develop and Demonstrate Open-lattice Flow in Electrically Heated Core; (3) Develop and Demonstrate Chemistry Control; (4) Demonstrate Safe Operation; and (5) Provision for Future Testing. This paper discusses the preliminary design of systems, thermal hydraulic analysis, and simplified cost estimate. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 4200 C. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M (in 2006 $). It is also estimated that the facility will require two years to be constructed and ready for operation.

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

    NASA Technical Reports Server (NTRS)

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

    2007-01-01

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

  14. VIPRE-01. a thermal-hydraulic analysis code for reactor cores. Volume 1. Mathematical modeling. [PWR; BWR

    SciTech Connect

    Stewart, C.W.; Cuta, J.M.; Koontz, A.S.; Kelly, J.M.; Basehore, K.L.; George, T.L.; Rowe, D.S.

    1983-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 1: Mathematical Modeling) explains the major thermal hydraulic models and supporting correlations in detail.

  15. Strategic Need for Multi-Purpose Thermal Hydraulic Loop for Support of Advanced Reactor Technologies

    SciTech Connect

    James E. O'Brien; Piyush Sabharwall; Su-Jong Yoon; Gregory K. Housley

    2014-09-01

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

  16. Monte Carlo Neutronics and Thermal Hydraulics Analysis of Reactor Cores with Multilevel Grids

    NASA Astrophysics Data System (ADS)

    Bernnat, W.; Mattes, M.; Guilliard, N.; Lapins, J.; Zwermann, W.; Pasichnyk, I.; Velkov, K.

    2014-06-01

    Power reactors are composed of assemblies with fuel pin lattices or other repeated structures with several grid levels, which can be modeled in detail by Monte Carlo neutronics codes such as MCNP6 using corresponding lattice options, even for large cores. Except for fresh cores at beginning of life, there is a varying material distribution due to burnup in the different fuel pins. Additionally, for power states the fuel and moderator temperatures and moderator densities vary according to the power distribution and cooling conditions. Therefore, a coupling of the neutronics code with a thermal hydraulics code is necessary. Depending on the level of detail of the analysis, a very large number of cells with different materials and temperatures must be regarded. The assignment of different material properties to all elements of a multilevel grid is very elaborate and may exceed program limits if the standard input procedure is used. Therefore, an internal assignment is used which overrides uniform input parameters. The temperature dependency of continuous energy cross sections, probability tables for the unresolved resonance region and thermal neutron scattering laws is taken into account by interpolation, requiring only a limited number of data sets generated for different temperatures. The method is applied with MCNP6 and proven for several full core reactor models. For the coupling of MCNP6 with thermal hydraulics appropriate interfaces were developed for the GRS system code ATHLET for liquid coolant and the IKE thermal hydraulics code ATTICA-3D for gaseous coolant. Examples will be shown for different applications for PWRs with square and hexagonal lattices, fast reactors (SFR) with hexagonal lattices and HTRs with pebble bed and prismatic lattices.

  17. An effective thermal-hydraulics methodology for prismatic core HTGR and VHTR

    SciTech Connect

    Travis, B. W.; El-Genk, M. S.

    2012-07-01

    Optimizing the performance and design of a prismatic core HTGR or VHTR requires a full core thermal-hydraulics analysis. Owing to the complexity and massive core structure, such analysis requires extensive and massively parallelized computation capabilities and a relatively long time (weeks to months) to complete. These demanding requirements are not due to the 3-D simulation of heat conduction in the annular core of the reactor, but rather the 3-D computational fluid dynamics (CFD) simulation of the helium gas flow in the 10-m long cooling channels in the 102 hexagonal fuel elements and the axial graphite reflector blocks in the core. This paper applies and examines the effectiveness of using a 1-D simulation of the helium flow in the core coolant channels, coupled to a 3-D simulation of the heat conduction in the graphite and fuel compacts, to perform thermal-hydraulics analysis of a hexagonal fuel element and of a 1/6 full core. This methodology employs typical cosine and constant axial power profiles and an applicable convective heat transfer correlation for the helium flow in the coolant channels. The correlation has recently been validated for a 10 m tall, single channel fuel module and shown to significantly reduce the computation time and memory requirements without compromising the accuracy of the calculations. The fidelity and accuracy of the present results for a hexagonal fuel element are verified by comparing them to those of a full 3-D numerical analysis. In addition to the temperature fields, results compare the computation time and number of numerical grid elements for implementing the two numerical simulation methods. The results of the thermal-hydraulics analysis of a 1/6 full core with the simplified methodology are also presented. All performed analysis accounts for the temperature dependent properties of helium, graphite in the reactor core and reflector blocks and the TRISO particle fuel compacts. (authors)

  18. Resolution of thermal-hydraulic safety and licensing issues for the system 80+{sup {trademark}} design

    SciTech Connect

    Carpentino, S.E.; Ritterbusch, S.E.; Schneider, R.E.

    1995-09-01

    The System 80+{sup {trademark}} Standard Design is an evolutionary Advanced Light Water Reactor (ALWR) with a generating capacity of 3931 MWt (1350 MWe). The Final Design Approval (FDA) for this design was issued by the Nuclear Regulatory Commission (NRC) in July 1994. The design certification by the NRC is anticipated by the end of 1995 or early 1996. NRC review of the System 80+ design has involved several new safety issues never before addressed in a regulatory atmosphere. In addition, conformance with the Electric Power Research Institute (EPRI) ALWR Utility Requirements Document (URD) required that the System 80+ plant address nuclear industry concerns with regard to design, construction, operation and maintenance of nuclear power plants. A large number of these issues/concerns deals with previously unresolved generic thermal-hydraulic safety issues and severe accident prevention and mitigation. This paper discusses the thermal-hydraulic analyses and evaluations performed for the System 80+ design to resolve safety and licensing issues relevant to both the Nuclear Stream Supply System (NSSS) and containment designs. For the NSSS design, the Safety Depressurization System mitigation capability and resolution of the boron dilution concern are described. Examples of containment design issues dealing with containment shell strength, robustness of the reactor cavity walls and hydrogen mixing under severe accident conditions are also provided. Finally, the overall approach used in the application of NRC`s new (NUREG-1465) radiological source term for System 80+ evaluation is described. The robustness of the System 80+ containment design to withstand severe accident consequences was demonstrated through detailed thermal-hydraulic analyses and evaluations. This advanced design to shown to meet NRC severe accident policy goals and ALWR URD requirements without any special design features and unnecessary costs.

  19. A parallelization approach to the COBRA-TF thermal-hydraulic subchannel code

    NASA Astrophysics Data System (ADS)

    Ramos, Enrique; Abarca, Agustín; Roman, Jose E.; Miró, Rafael

    2014-06-01

    In order to reduce the response time when simulating large reactors in detail, we have developed a parallel version of the thermal-hydraulic subchannel code COBRA-TF, with standard message passing technology (MPI). The parallelization is oriented to reactor cells, so it is best suited for models consisting of many cells. The generation of the Jacobian is parallelized, in such a way that each processor is in charge of generating the data associated to a subset of cells. Also, the solution of the linear system of equations is done in parallel, using the PETSc toolkit.

  20. Thermal-hydraulic aspects of flow inversion in a research reactor

    SciTech Connect

    Smith, R.S.; Woodruff, W.L.

    1986-11-01

    PARET, a neutronics and thermal-hydraulics computer code, has been modified to account for natural convection in a reactor core. The code was then used to analyze the flow inversion that occurs in a reactor with heat removal by forced convection in the downward direction after a pump failure. Typical results are shown for a number of parameters. Research reactors normally operating much above ten MW are predicted to experience nucleate boiling in the event of a flow inversion. Comparison with experimental results from the Belgian BR2 reactor indicated general agreement although nucleate boiling that was analytically predicted was not noted in the BR2 data.

  1. Coupled neutronic and thermal-hydraulic code benchmark activities at the International Nuclear Safety Center.

    SciTech Connect

    Podlazov, L. N.

    1998-07-29

    Two realistic benchmark problems are defined and used to assess the performance of coupled thermal-hydraulic and neutronic codes used in simulating dynamic processes in VVER-1000 and RBMK reactor systems. One of the problems simulates a design basis accident involving the ejection of three control and protection system rods from a VVER-1000 reactor. The other is based on a postulated rod withdrawal from an operating RBMK reactor. Preliminary results calculated by various codes are compared. While these results show significant differences, the intercomparisons performed so far provide a basis for further evaluation of code limitations and modeling assumptions.

  2. Subchannel thermal-hydraulic modeling of an APT tungsten target rod bundle

    SciTech Connect

    Hamm, L.L.; Shadday, M.A. Jr.

    1997-09-01

    The planned target for the Accelerator Production of Tritium (APT) neutron source consists of an array of tungsten rod bundles through which D{sub 2}O coolant flows axially. Here, a scoping analysis of flow through an APT target rod bundle was conducted to demonstrate that lateral cross-flows are important, and therefore subchannel modeling is necessary to accurately predict thermal-hydraulic behavior under boiling conditions. A local reactor assembly code, FLOWTRAN, was modified to model axial flow along the rod bundle as flow through three concentric heated annular passages.

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

    SciTech Connect

    Richard, Joshua; Galloway, Jack; Fensin, Michael; Trellue, Holly

    2015-04-04

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

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

    DOE PAGESBeta

    Richard, Joshua; Galloway, Jack; Fensin, Michael; Trellue, Holly

    2015-04-04

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

  5. Thermal-Hydraulic Mockup Tests with Two-Phase Thermosyphon for Cold Neutron Source

    SciTech Connect

    Lee, C.H.; Chan, Y.K.; Lee, D.J.; Chang, C.J.; Hong, W.T.

    2002-07-01

    The improvement and utilization promotion project of the Taiwan Research Reactor (TRR-II) is carrying out at the Institute of Nuclear Energy Research (INER). The Cold Neutron Source (CNS) with a two-phase thermosyphon will be installed in the heavy water reactor of TRR-II. The hydrogen cold loop of TRR-II CNS consists of a cylindrical moderator cell, a single transfer tube, and a condenser. The thermal-hydraulic characteristics of a two-phase thermosyphon are investigated against the variations of mass inventory, tube geometry and heat loads. The thermal-hydraulic experiments have been performed using a full-scale mockup loop and a Freon-11 as a working fluid. The scaling approach is that the mass-fluxes of the liquid and the vapor in the Wallis correlation are identical between hydrogen and Freon-11. So, the same density ratio and a scaling heat load are applied to the loop. The flooding limitations as a function of initial Freon-11 inventory, transfer tube diameter, transfer tube geometry, and heat loads are presented. (authors)

  6. Thermal-hydraulic criteria for the APT tungsten neutron source design

    SciTech Connect

    Pasamehmetoglu, K.

    1998-03-01

    This report presents the thermal-hydraulic design criteria (THDC) developed for the tungsten neutron source (TNS). The THDC are developed for the normal operations, operational transients, and design-basis accidents. The requirements of the safety analyses are incorporated into the design criteria, consistent with the integrated safety management and the safety-by-design philosophy implemented throughout the APT design process. The phenomenology limiting the thermal-hydraulic design and the confidence level requirements for each limit are discussed. The overall philosophy of the uncertainty analyses and the confidence level requirements also are presented. Different sets of criteria are developed for normal operations, operational transients, anticipated accidents, unlikely accidents, extremely unlikely accidents, and accidents during TNS replacement. In general, the philosophy is to use the strictest criteria for the high-frequency events. The criteria is relaxed as the event frequencies become smaller. The THDC must be considered as a guide for the design philosophy and not as a hard limit. When achievable, design margins greater than those required by the THDC must be used. However, if a specific event sequence cannot meet the THDC, expensive design changes are not necessary if the single event sequence results in sufficient margin to safety criteria and does not challenge the plant availability or investment protection considerations.

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

    NASA Astrophysics Data System (ADS)

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

    1988-09-01

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

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

    SciTech Connect

    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.

  9. Thermal hydraulic method for whole core design analysis of an HTGR

    SciTech Connect

    Huning, A. J.; Garimella, S.

    2013-07-01

    A new thermal hydraulic method and initial results are presented for core-wide steady state analysis of prismatic High Temperature Gas-Cooled Reactors (HTGR). The method allows for the complete solution of temperature and coolant mass flow distribution by solving quasi-steady energy balances for the discretized core. Assembly blocks are discretized into unit cells for which the average temperature of each unit cell is determined. Convective heat removal is coupled to the unit cell energy balances by a 1-D axial flow model. The flow model uses established correlations for friction factor and Nusselt number. Bypass flow is explicitly calculated by using an initial guess for mass flow distribution and determining the exit pressure of each flow channel. The mass flow distribution is updated until a uniform core exit pressure condition is reached. Results are obtained for the MHTGR-350 with emphasis on the change in thermal hydraulic parameters due to various steady state power profiles and bypass gap widths. Steady state temperature distribution and its variations are discussed. (authors)

  10. Shape optimization of a printed-circuit heat exchanger to enhance thermal-hydraulic performance

    SciTech Connect

    Lee, S. M.; Kim, K. Y.

    2012-07-01

    Printed circuit heat exchanger (PCHE) is recently considered as a recuperator for the high temperature gas cooled reactor. In this work, the zigzag-channels of a PCHE have been optimized by using three-dimensional Reynolds-Averaged Navier-Stokes (RANS) analysis and response surface approximation (RSA) modeling technique to enhance thermal-hydraulic performance. Shear stress transport turbulence model is used as a turbulence closure. The objective function is defined as a linear combination of the functions related to heat transfer and friction loss of the PCHE, respectively. Three geometric design variables viz., the ratio of the radius of the fillet to hydraulic diameter of the channels, the ratio of wavelength to hydraulic diameter of the channels, and the ratio of wave height to hydraulic diameter of the channels, are used for the optimization. Design points are selected through Latin-hypercube sampling. The optimal design is determined through the RSA model which uses RANS derived calculations at the design points. The results show that the optimum shape enhances considerably the thermal-hydraulic performance than a reference shape. (authors)

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

    SciTech Connect

    Feltus, M.A.; Rubinaccio, G.

    1996-12-31

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

  12. The Thermal Hydraulics of Tube Support Fouling in Nuclear Steam Generators

    SciTech Connect

    Rummens, Helena E.C.; Rogers, J.T.; Turner, C.W.

    2004-12-15

    It is hypothesized that the thermal-hydraulic environment plays a role in the fouling of tube supports in nuclear steam generators. Experiments were performed to simulate the thermal-hydraulic environment near various designs of supports. Pressure loss, local velocity, turbulence intensity, and local void fraction were measured to characterize the effect of the support. Fouling mechanisms specific to supports were inferred from these experimental data and from actual steam generator inspection results. An analytical model was developed to predict the rate of particulate deposition on the supports, to better understand the complex processes involved.This paper presents the following set of tools for assessing the fouling propensity of a given support design: (1) proposed fouling mechanisms, (2) criteria for support fouling propensity, (3) correlation of fouling with parameters such as mass flux and quality, (4) descriptions of experimental tools such as flow visualization and measurement of pressure-loss profiles, and (5) analytical tools.An important conclusion from this and our previous work is that the fouling propensity is greater with broached support plates, both trefoil and quatrefoil, than with lattice bar supports and formed bar supports, in which significant cross flows occur.

  13. Fundamental approaches for analysis thermal hydraulic parameter for Puspati Research Reactor

    NASA Astrophysics Data System (ADS)

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

    2016-01-01

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

  14. Thermal-hydraulic studies of the Advanced Neutron Source cold source

    SciTech Connect

    Williams, P.T.; Lucas, A.T.

    1995-08-01

    The Advanced Neutron Source (ANS), in its conceptual design phase at Oak Ridge National Laboratory, was to be a user-oriented neutron research facility producing the most intense steady-state flux of thermal and cold neutrons in the world. Among its many scientific applications, the production of cold neutrons was a significant research mission for the ANS. The cold neutrons come from two independent cold sources positioned near the reactor core. Contained by an aluminum alloy vessel, each cold source is a 410-mm-diam sphere of liquid deuterium that functions both as a neutron moderator and a cryogenic coolant. With nuclear heating of the containment vessel and internal baffling, steady-state operation requires close control of the liquid deuterium flow near the vessel`s inner surface. Preliminary thermal-hydraulic analyses supporting the cold source design were performed with heat conduction simulations of the vessel walls and multidimensional computational fluid dynamics simulations of the liquid deuterium flow and heat transfer. This report presents the starting phase of a challenging program and describes the cold source conceptual design, the thermal-hydraulic feasibility studies of the containment vessel, and the future computational and experimental studies that were planned to verify the final design.

  15. Thermal-hydraulic development a small, simplified, proliferation-resistant reactor.

    SciTech Connect

    Farmer, M. T.; Hill, D. J.; Sienicki, J. J.; Spencer, B. W.; Wade, D. C.

    1999-07-02

    This paper addresses thermal-hydraulics related criteria and preliminary concepts for a small (300 MWt), proliferation-resistant, liquid-metal-cooled reactor system. A main objective is to assess what extent of simplification is achievable in the concepts with the primary purpose of regaining economic competitiveness. The approach investigated features lead-bismuth eutectic (LBE) and a low power density core for ultra-long core lifetime (goal 15 years) with cartridge core replacement at end of life. This potentially introduces extensive simplifications resulting in capital cost and operating cost savings including: (1) compact, modular, pool-type configuration for factory fabrication, (2) 100+% natural circulation heat transport with the possibility of eliminating the main coolant pumps, (3) steam generator modules immersed directly in the primary coolant pool for elimination of the intermediate heat transport system, and (4) elimination of on-site fuel handling and storage provisions including rotating plug. Stage 1 natural circulation model and results are presented. Results suggest that 100+% natural circulation heat transport is readily achievable using LBE coolant and the long-life cartridge core approach; moreover, it is achievable in a compact pool configuration considerably smaller than PRISM A (for overland transportability) and with peak cladding temperature within the existing database range for ferritic steel with oxide layer surface passivation. Stage 2 analysis follows iteration with core designers. Other thermal hydraulic investigations are underway addressing passive, auxiliary heat removal by air cooling of the reactor vessel and the effects of steam generator tube rupture.

  16. Development of an integrated thermal-hydraulics capability incorporating RELAP5 and PANTHER neutronics code

    SciTech Connect

    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.

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

    SciTech Connect

    Yamada, K.; Aksan, S. N.

    2012-07-01

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

  18. KUGEL: a thermal, hydraulic, fuel performance, and gaseous fission product release code for pebble bed reactor core analysis

    SciTech Connect

    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.

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

    NASA Astrophysics Data System (ADS)

    Urquiza, Eugenio

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

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

    SciTech Connect

    Hirano, Masashi

    1997-07-01

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

  1. Thermal-Hydraulic Performance of Cross-Shaped Spiral Fuel in High-Power-Density BWRs

    SciTech Connect

    Conboy, Thomas; Hejzlar, Pavel

    2006-07-01

    Power up-rating of existing nuclear reactors promises to be an area of great study for years to come. One of the major approaches to efficiently increasing power density is by way of advanced fuel design, and cross-shaped spiral-fuel has shown such potential in previous studies. Our work aims to model the thermal-hydraulic consequences of filling a BWR core with these spiral-shaped pins. The helically-wound pins have a cross-section resembling a 4-petaled flower. They fill an assembly in a tight bundle, their dimensions chosen carefully such that the petals of neighboring pins contact each other at their outer-most extent in a self-supporting lattice, absent of grid spacers. Potential advantages of this design raise much optimism from a thermal-hydraulic perspective. These spiral rods possess about 40% larger surface area than traditional rods, resulting in increased cooling and a proportional reduction in average surface heat flux. The thin petal-like extensions help by lowering thermal resistance between the hot central region of the pin and the bulk coolant flow, decreasing the maximum fuel temperature by 200 deg. C according to Finite Element (COSMOS) models. However, COSMOS models also predict a potential problem area at the 'elbow' region of two adjoining petals, where heat flux peaking is twice that along the extensions. Preliminary VIPRE models, which account only for the surface area increase, predict a 22% increase in critical power. It is also anticipated that the spiral twist would provide the flowing coolant with an additional radial velocity component, and likely promote turbulence and mixing within an assembly. These factors are expected to provide further margin for increased power density, and are currently being incorporated into the VIPRE model. The reduction in pressure drop inherent in any core without grid-spacers is also expected to be significant in aiding core stability, though this has not yet been quantified. Spiral-fuel seems to be a

  2. Relationship of core exit-temperature noise to thermal-hydraulic conditions in PWRs

    SciTech Connect

    Sweeney, F.J.; Upadhyaya, B.R.

    1983-01-01

    Core exit thermocouple temperature noise and neutron detector noise measurements were performed at the Loss of Fluid Test Facility (LOFT) reactor and a Westinghouse, 1148 MW(e) PWR to relate temperature noise to core thermal-hydraulic conditions. The noise analysis results show that the RMS of the temperature noise increases linearly with increasing core ..delta..T at LOFT and the commercial PWR. Out-of-core test loop temperature noise has shown similar behavior. The phase angle between core exit temperature noise and in-core or ex-core neutron noise is directly related to the core coolant flow velocity. However, if the thermocouple response time is slow, compared to the coolant transit time between the sensors, velocities inferred from the phase angle are lower than measured coolant flow velocities.

  3. A comparison of the CHF between tubes and annuli under PWR thermal-hydraulic conditions

    SciTech Connect

    Herer, C.

    1995-09-01

    Critical Heat Flux (CHF) tests were carried out in three tubes with inside diameters of 8, 13, and 19.2 mm and in two annuli with an inner tube of 9.5 mm and an outer tube of 13 or 19.2 mm. All axial heat flux distributions in the test sections were uniform. The coolant fluid was Refrigerant 12 (Freon-12) under PWR thermal-hydraulic conditions (equivalent water conditions - Pressure: 7 to 20 MPa, Mass Velocity: 1000 to 6000 kg/m2/s, Local Quality: -75% to +45%). The effect of tube diameter is correlated for qualities under 15%. The change from the tube to the annulus configuration is correctly taken into account by the equivalent hydraulic diameter. Useful information is also provided concerning the effect of a cold wall in an annulus.

  4. Thermal-hydraulic tests of a recirculation cooling installation for the Rostov nuclear power station

    NASA Astrophysics Data System (ADS)

    Balunov, B. F.; Balashov, V. A.; Il'in, V. A.; Krayushnikov, V. V.; Lychakov, V. D.; Meshalkin, V. V.; Ustinov, A. N.; Shcheglov, A. A.

    2013-09-01

    Results obtained from thermal-hydraulic tests of the recirculation cooling installation used as part of the air cooling system under the containments of the Rostov nuclear power station Units 3 and 4 are presented. The operating modes of the installation during normal operation (air cooling on the surface of finned tubes), under the conditions of anticipated operational occurrences (air cooling and steam condensation from a steam-air mixture), and during an accident (condensation of pure steam) are considered. Agreement is obtained between the results of tests and calculations carried out according to the recommendations given in the relevant regulatory documents. A procedure of carrying out thermal calculation for the case of steam condensation from a steam-air mixture on the surface of fins is proposed. The possibility of efficient use of the recirculation cooling installation in the system for reducing emergency pressure under the containment of a nuclear power station is demonstrated.

  5. Thermal hydraulics of the impurity control system for FED/INTOR

    SciTech Connect

    Cha, Y.S.; Mattas, R.F.; Abdou, M.A.; Haines, J.R.

    1983-01-01

    This paper addresses two important aspects of thermal hydraulics related to the design of the impurity control system (limiter and divertor) of the Fusion Engineering Device (FED) and the International Tokamak Reactor (INTOR). The first part of the paper is devoted to the determination of temperature distributions in various combinations of the coating/structural materials proposed for the limiter/divertor of FED and INTOR. The second part of the paper describes the analysis of the tangential motion of the melt layer under the influence of magnetic force during plasma disruption. The results of both analysis provide inputs to the determination of the life time of the limiter (or divertor) which is the most critical problem for the impurity control system as far as engineering and materials consideration is concerned.

  6. Thermal-hydraulic-structural behavior of the EBR-II IHX for overpower transients

    SciTech Connect

    Mohr, D.; Chang, L.K.; Lee, M.J.; Feldman, E.E.

    1982-01-01

    A detailed study has been made of the effects of the Operational Reliability Testing (ORT) program on major plant components of the Experimental Breeder Reactor No. II (EBR-II). This paper describes the integrated thermal-hydraulic-structural analyses conducted for the intermediate heat exchanger (IHX) with the aid of the NATDEMO, THTB, and ANSYS codes. An extensive analysis revealed the stress limiting area to be the junction between the upper head and upper tube sheet. The analyses indicate, however, that the EBR-II IHX, the major plant component most affected by the ORT program, will be able to withstand the thermal stress and accumulated fatigue damage during the lifetime of the plant including the ORT program.

  7. 2D Thermal Hydraulic Analysis and Benchmark in Support of HFIR LEU Conversion using COMSOL

    SciTech Connect

    Freels, James D; Bodey, Isaac T; Lowe, Kirk T; Arimilli, Rao V

    2010-09-01

    The research documented herein was funded by a research contract between the Research Reactors Division (RRD) of Oak Ridge National Laboratory (ORNL) and the University of Tennessee, Knoxville (UTK) Mechanical, Aerospace and Biomedical Engineering Department (MABE). The research was governed by a statement of work (SOW) which clearly defines nine specific tasks. This report is outlined to follow and document the results of each of these nine specific tasks. The primary goal of this phase of the research is to demonstrate, through verification and validation methods, that COMSOL is a viable simulation tool for thermal-hydraulic modeling of the High Flux Isotope Reactor (HFIR) core. A secondary goal of this two-dimensional phase of the research is to establish methodology and data base libraries that are also needed in the full three-dimensional COMSOL simulation to follow. COMSOL version 3.5a was used for all of the models presented throughout this report.

  8. MNSR transient analyses and thermal-hydraulic safety margins for HEU and LEU cores using PARET

    SciTech Connect

    Olson, Arne P.; Jonah, S.A.

    2008-07-15

    Thermal-hydraulic performance characteristics of Miniature Neutron Source Reactors under long-term steady-state and transient conditions are investigated. Safety margins and limiting conditions attained during these events are determined. Modeling extensions are presented that enable the PARET/ANL code to realistically track primary loop heatup, heat exchange to the pool, and heat loss from the pool to air over the pool. Comparisons are made of temperature predictions for HEU and LEU fueled cores under transient conditions. Results are obtained using three different natural convection heat transfer correlations: the original (PARET/ANL version 5), Churchill-Chu, and an experiment- based correlation from the China Institute of Atomic Energy (CIAE). The MNSR, either fueled by HEU or by LEU, satisfies the design limits for long-term transient operation. (author)

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

    SciTech Connect

    Vondy, D.R.

    1981-09-01

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

  10. Use of separate-effects experiments in verification of system thermal-hydraulics

    SciTech Connect

    Saha, P.

    1982-01-01

    In recent years, a number of advanced, best-estimate systems codes such as TRAC and RELAP5 have been developed in order to accurately predict the consequences of various postulated accidents and transients in Light Water Reactor (LWR) systems. Although these codes had to go through some verification or assessment during the developmental stage, it has been recognized that an independent assessment of these codes is necessary before they should be applied to any decision making process. The USNRC is, therefore, sponsoring such efforts at several national laboratories including BNL. The overall assessment matrix includes separate-effects, integral and plant tests. However, this paper will focus on how the separate-effects tests can be utilized in verifying the thermal-hydraulic models that control the various stages of postulated accidents and/or transients in a LWR system.

  11. Conceptual design and thermal-hydraulic characteristics of natural circulation Boiling Water Reactors

    SciTech Connect

    Kataoka, Y.; Suzuki, H.; Murase, M. ); Horiuchi, T.; Miki, M. )

    1988-08-01

    A natural circulation boiling water reactor (BWR) with a rated capacity of 600 MW (electric) has been conceptually designed for small- and medium-sized light water reactors. The components and systems in the reactor are simplified by eliminating pumped recirculation systems and pumped emergency core cooling systems. Consequently, the volume of the reactor building is -- 50% of that for current BWRs with the same rated capacity; the construction period is also shorter. Its thermal-hydraulic characteristics, critical power ratio (CPR) and flow stability at steady state, decrease in the minimum CPR (..delta..MCPR) at transients, and the two-phase mixture level in the reactor pressure vessel (RPV) during accidents are investigated. The two-phase mixture level in the RPV during an accident does not decrease to lower than the top of the core; the core uncovery and heatup of fuel cladding would not occur during any loss-of-coolant accident.

  12. Thermal-hydraulic post-test analysis of OECD LOFT LP-FP-2 experiment

    SciTech Connect

    Pena, J.J. ); Enciso, S. ); Reventos, F. )

    1992-04-01

    An assessment of RELAP5/MOD2 and SCDAP/MOD1 against the OECD LOFT experiment LP-FP-2 is presented. LP-FP-2 studies the hypothetical release of fission products and their transport following a large-break LOCA scenario. The report comprises a general description of the LP-FP-2 experiment, a summary of thermal-hydraulic data, a simulation of the LP-FP-2 experiment, results of the RELAP5/MOD2 base calculation, the RELAP5/MOD2 sensitivity analysis, the SCDAP/MOD1 nodalization for an LP-FP-2 experiment, the results of the SCDAP/MOD1 calculation, and the summary and conclusions.

  13. Thermal-hydraulics for space power, propulsion, and thermal management system design

    SciTech Connect

    Krotiuk, W.J.

    1990-01-01

    The present volume discusses thermal-hydraulic aspects of current space projects, Space Station thermal management systems, the thermal design of the Space Station Free-Flying Platforms, the SP-100 Space Reactor Power System, advanced multi-MW space nuclear power concepts, chemical and electric propulsion systems, and such aspects of the Space Station two-phase thermal management system as its mechanical pumped loop and its capillary pumped loop's supporting technology. Also discussed are the startup thaw concept for the SP-100 Space Reactor Power System, calculational methods and experimental data for microgravity conditions, an isothermal gas-liquid flow at reduced gravity, low-gravity flow boiling, computations of Space Shuttle high pressure cryogenic turbopump ball bearing two-phase coolant flow, and reduced-gravity condensation.

  14. Prototypic Thermal-Hydraulic Experiment in NRU to Simulate Loss-of-Coolant Accidents

    SciTech Connect

    Mohr, C. L.; Hesson, G. M.; Russcher, G. E.; Marsh, R. K.; King, L. L.; Wildung, N. J.; Rausch, W. N.; Bennett, W. D.

    1981-04-01

    Quick-look test results are reported for the initial test series of the Loss-of-Coolant Accident (LOCA) Simulation in the National Research Universal {NRU) test program, conducted by Pacific Northwest Laboratory (PNL) for the U.S. Nuclear Regulatory Commission (NRC). This test was devoted to evaluating the thermal-hydraulic characteristics of a full-length light water reactor (LWR) fuel bundle during the heatup, reflood, and quench phases of a LOCA. Experimental results from 28 tests cover reflood rates of 0.74 in./sec to 11 in./sec and delay times to initiate reflood of 3 sec to 66 sec. The results indicate that current analysis methods can predict peak temperatures within 10% and measured quench times for the bundle were significantly less than predicted. For reflood rates of 1 in./sec where long quench times were predicted (>2000 sec}, measured quench times of 200 sec were found.

  15. Code System for 2-Group, 3D Neutronic Kinetics Calculations Coupled to Core Thermal Hydraulics.

    2000-05-12

    Version 00 QUARK is a combined computer program comprising a revised version of the QUANDRY three-dimensional, two-group neutron kinetics code and an upgraded version of the COBRA transient core analysis code (COBRA-EN). Starting from either a critical steady-state (k-effective or critical dilute Boron problem) or a subcritical steady-state (fixed source problem) in a PWR plant, the code allows one to simulate the neutronic and thermal-hydraulic core transient response to reactivity accidents initiated both inside themore » vessel (such as a control rod ejection) and outside the vessel (such as the sudden change of the Boron concentration in the coolant). QUARK output can be used as input to PSR-470/NORMA-FP to perform a subchannel analysis from converged coarse-mesh nodal solutions.« less

  16. Numerical simulation of combined natural and forced convection during thermal-hydraulic transients. [LMFBR

    SciTech Connect

    Domanus, H.M.; Sha, W.T.

    1981-01-01

    The single-phase COMMIX (COMponent MIXing) computer code performs fully three-dimensional, transient, thermal-hydraulic analyses of liquid-sodium LMFBR components. It solves the conservation equations of mass, momentum, and energy as a boundary-value problem in space and as an initial-value problem in time. The concepts of volume porosity, surface permeability and distributed resistance, and heat source have been employed in quasi-continuum (rod-bundle) applications. Results from three transient simulations involving forced and natural convection are presented: (1) a sodium-filled horizontal pipe initially of uniform temperature undergoing an inlet velocity rundown transient, as well as an inlet temperature transient; (2) a 19-pin LMFBR rod bundle undergoing a velocity transient; and, (3) a simulation of a water test of a 1/10-scale outlet plenum undergoing both velocity and temperature transients.

  17. Thermal hydraulic codes for LWR safety analysis - present status and future perspective

    SciTech Connect

    Staedtke, H.

    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.

  18. COMSOL Simulations for Steady State Thermal Hydraulics Analyses of ORNL s High Flux Isotope Reactor

    SciTech Connect

    Khane, Vaibhav B; Jain, Prashant K; Freels, James D

    2012-01-01

    Simulation models for steady state thermal hydraulics analyses of Oak Ridge National Laboratory s High Flux Isotope Reactor (HFIR) have been developed using the COMSOL Multiphysics simulation software. A single fuel plate and coolant channel of each type of HFIR fuel element was modeled in three dimensions; coupling to adjacent plates and channels was accounted for by using periodic boundary conditions. The standard k- turbulence model was used in simulating turbulent flow with conjugate heat transfer. The COMSOL models were developed to be fully parameterized to allow assessing impacts of fuel fabrication tolerances and uncertainties related to low enriched uranium (LEU) fuel design and reactor operating parameters. Heat source input for the simulations was obtained from separate Monte Carlo N Particle calculations for the axially non-contoured LEU fuel designs at the beginning of the reactor cycle. Mesh refinement studies have been performed to calibrate the models against the pressure drop measured across the HFIR core.

  19. Simulating HFIR Core Thermal Hydraulics Using 3D-2D Model Coupling

    SciTech Connect

    Travis, Adam R; Freels, James D; Ekici, Kivanc

    2013-01-01

    A model utilizing interdimensional variable coupling is presented for simulating the thermal hydraulic interactions of the High Flux Isotope Reactor (HFIR) core at Oak Ridge National Laboratory (ORNL). The model s domain consists of a single, explicitly represented three-dimensional fuel plate and a simplified two-dimensional coolant channel slice. In simplifying the coolant channel, and thus the number of mesh points in which the Navier-Stokes equations must be solved, the computational cost and solution time are both greatly reduced. In order for the reduced-dimension coolant channel to interact with the explicitly represented fuel plate, however, interdimensional variable coupling must be enacted along all shared boundaries. The primary focus of this paper is in detailing the collection, storage, passage, and application of variables across this interdimensional interface. Comparisons are made showing the general speed-up associated with this simplified coupled model.

  20. Neutron Tomography Using Mobile Neutron Generators for Assessment of Void Distributions in Thermal Hydraulic Test Loops

    NASA Astrophysics Data System (ADS)

    Andersson, P.; Bjelkenstedt, T.; Sundén, E. Andersson; Sjöstrand, H.; Jacobsson-Svärd, S.

    Detailed knowledge of the lateral distribution of steam (void) and water in a nuclear fuel assembly is of great value for nuclear reactor operators and fuel manufacturers, with consequences for both reactor safety and economy of operation. Therefore, nuclear relevant two-phase flows are being studied at dedicated thermal-hydraulic test loop, using two-phase flow systems ranging from simplified geometries such as heated circular pipes to full scale mock-ups of nuclear fuel assemblies. Neutron tomography (NT) has been suggested for assessment of the lateral distribution of steam and water in such test loops, motivated by a good ability of neutrons to penetrate the metallic structures of metal pipes and nuclear fuel rod mock-ups, as compared to e.g. conventional X-rays, while the liquid water simultaneously gives comparatively good contrast. However, these stationary test loops require the measurement setup to be mobile, which is often not the case for NT setups. Here, it is acknowledged that fast neutrons of 14 MeV from mobile neutron generators constitute a viable option for a mobile NT system. We present details of the development of neutron tomography for this purpose at the division of Applied Nuclear Physics at Uppsala University. Our concept contains a portable neutron generator, exploiting the fusion reaction of deuterium and tritium, and a detector with plastic scintillator elements designed to achieveadequate spatial and energy resolution, all mounted in a light-weight frame without collimators or bulky moderation to allow for a mobile instrument that can be moved about the stationary thermal hydraulic test sections. The detector system stores event-to-event pulse-height information to allow for discrimination based on the energy deposition in the scintillator elements.

  1. 4C code analysis of thermal-hydraulic transients in the KSTAR PF1 superconducting coil

    NASA Astrophysics Data System (ADS)

    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.

  2. Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2

    SciTech Connect

    Coryell, E.W.; Siefken, L.J.; Harvego, E.A.

    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.

  3. Current and anticipated uses of the thermal hydraulics codes at the NRC

    SciTech Connect

    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.

  4. Investigation of the possibility to use a fine-mesh solver for resolving coupled neutronics and thermal-hydraulics

    SciTech Connect

    Jareteg, K.; Vinai, P.; Demaziere, C.

    2013-07-01

    The development of a fine-mesh coupled neutronic/thermal-hydraulic solver is touched upon in this paper. The reported work investigates the feasibility of using finite volume techniques to discretize a set of conservation equations modeling neutron transport, fluid dynamics, and heat transfer within a single numerical tool. With the long-term objective of developing fine-mesh computing capabilities for a few selected fuel assemblies in a nuclear core, this preliminary study considers an infinite array of a single fuel assembly having a finite height. Thermal-hydraulic conditions close to the ones existing in PWRs are taken as a first test case. The neutronic modeling relies on the diffusion approximation in a multi-energy group formalism, with cross-sections pre-calculated and tabulated at the sub-pin level using a Monte Carlo technique. The thermal-hydraulics is based on the Navier-Stokes equations, complemented by an energy conservation equation. The non-linear coupling terms between the different conservation equations are fully resolved using classical iteration techniques. Early tests demonstrate that the numerical tool provides an unprecedented level of details of the coupled solution estimated within the same numerical tool and thus avoiding any external data transfer, using fully consistent models between the neutronics and the thermal-hydraulics. (authors)

  5. 75 FR 69140 - NUREG-1953, Confirmatory Thermal-Hydraulic Analysis To Support Specific Success Criteria in the...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-11-10

    ... COMMISSION NUREG-1953, Confirmatory Thermal-Hydraulic Analysis To Support Specific Success Criteria in the...- Hydraulic Analysis to Support Specific Success Criteria in the Standardized Plant Analysis Risk Models...-Hydraulic Analysis to Support Specific Success Criteria in the Standardized Plant Analysis Risk...

  6. A study of thermal stratification in the cold legs during the subcooled blowdown phase of a loss of coolant accident in the OSU APEX thermal hydraulic testing facility.

    SciTech Connect

    Wachs, D. M.

    1998-11-04

    Thermal stratification, which has been linked to the occurrence of pressurized thermal shock (PTS), is observed to occur during the early stages of simulated loss of coolant accidents (LOCAS) in the Oregon State University Advanced Plant Experiment (OSU APEX) Thermal Hydraulic Test Facility. The OSU APEX Test Facility is a scaled model of the Westinghouse AP600 nuclear power plant. Analysis of the OSU APEX facility data has allowed the determination of an onset criteria for thermal stratification and has provided support for the postulated mechanisms leading to thermal stratification. CFX 4.1, a computational fluid dynamics code, was used to generate a model of the cold legs and the downcomer that described the phenomena occurring within them. Some mixing phenomena were predicted that lead to non-uniformity between the two cold legs attached to the steam generator on the side of the facility containing the Passive Residual Heat Removal (PRHR) injection system. The stratification was found to be two phase and unlikely to be a factor in PTS.

  7. Integral Circulation Experiment: Thermal-hydraulic simulator of a heavy liquid metal reactor

    NASA Astrophysics Data System (ADS)

    Tarantino, M.; Agostini, P.; Benamati, G.; Coccoluto, G.; Gaggini, P.; Labanti, V.; Venturi, G.; Class, A.; Liftin, K.; Forgione, N.; Moreau, V.

    2011-08-01

    In the frame of the IP-EUROTRANS (6th Framework Program EU), domain DEMETRA, ENEA was involved in the Work Package 4.5 " Large Scale Integral Test", devoted to characterize a relevant portion of a sub-critical ADS reactor block (core, internals, heat exchanger, cladding for fuel elements) in steady state, transient and accidental conditions. More in details ENEA assumed the commitment to perform an integral experiment aiming to reproduce the primary flow path of the " European Transmutation Demonstrator (ETD)" pool-type nuclear reactor, cooled by Lead Bismuth Eutectics (LBE). This experimental activity, called " Integral Circulation Experiment (ICE)", has been implemented merging the efforts of several research institutes, among which, besides ENEA, FZK, CRS4 and University of Pisa, allowing to design an appropriate test section to be installed in the CIRCE facility. The goal of the experiments is therefore to demonstrate the technological feasibility of a heavy liquid metal (HLM) nuclear system pool-type in a relevant scale (1 MW), investigating the related thermal-hydraulic behaviour (heat source and heat exchanger coupling, primary system and downcomer coupling, gas trapping into the main stream, thermal stratification in the pool, forced and mixed convection in rod bundle) under both steady state and transient conditions. Moreover the preliminary as well as the planned experiments aims to address performance and reliability tests of some prototypical components, such as heat source, heat exchanger, chemistry control system. The paper reports a detailed description of the experiment, the design performed for the test section and its main components as well as the preliminary experimental results carried out in the first experimental campaign run on the CIRCE pool, which consists of a full power steady state test. The preliminary experimental results carried out have demonstrate the proper design of the test section trough the experiment goals as well as the HLM

  8. Summary of papers on current and anticipated uses of thermal-hydraulic codes

    SciTech Connect

    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).

  9. Theoretical investigation of the thermal hydraulic behaviour of a slab-type liquid metal target

    SciTech Connect

    Dury, T.V.; Smith, B.L.

    1996-06-01

    The thermal hydraulics codes CFDS-FLOW3D and ASTEC have been used to simulate a slabtype design of ESS spallation target. This design is single-skinned, and of tapering form (in the beam direction), with rounded sides in a cross-section through a plane normal to the beam. The coolant fluid used is mercury, under forced circulation, with an inlet temperature of 180{degrees}C. The goal of these computer studies was to understand the behaviour of the coolant flow, and hence to arrive at a design which optimises the heat extraction for a given beam power - in the sense of: (1) minimising the peak local fluid temperature within the target, (2) maintaining an acceptable temperature level and distribution over and through the target outer wall, (3) keeping the overall fluid pressure loss through the complete target to a minimum, (4) staying within the physical limits of overall size required, particularly in the region of primary spallation. Two- and three-dimensional models have been used, with different arrangements and design of internal baffles, and different coolant flow distributions at the target inlet. Nominal total inlet mass flow was 245 kg/s, and a heat deposition profile used which was based on the proton beam energy distribution. This gave a nominal total heat load of 3.23 MW - of which 8.2kW were deposited in the window steel.

  10. Thermal-Hydraulic Analyses of the Submersion-Subcritical Safe Space (S and 4) Reactor

    SciTech Connect

    King, Jeffrey C.; El-Genk, Mohamed S.

    2007-01-30

    Detailed thermal-hydraulic analyses of the S and 4 reactor are performed to reduce the maximum fuel temperature of the Submersion-Subcritical Safe Space (S and 4) reactor to below 1300 K. The fuel pellet diameter is reduced from 1.315 cm to 1.25 cm, decreasing the thermal resistance of the pellets and each of the 1.54 cm diameter coolant channels in the reactor core are replaced with several 0.3 cm ID channels to increase the effective heat transfer area and to encourage mixing of the flowing helium-28% xenon coolant. The calculated maximum fuel temperature decreased from more than 1900 K to 1302 K and the relative pressure drop across the reactor core increased from 1.98% to 2.57% of the inlet pressure. Moving the concentric inlet and outlet pipes 1 cm towards the center of the reactor core encouraged more flow through the center region, further reducing the maximum fuel temperature by 14 degrees to 1288 K, with a negligible effect on the core pressure losses.

  11. Oyster Creek fuel thermal margin during core thermal-hydraulic oscillations

    SciTech Connect

    Dougher, J.D.

    1990-01-01

    The Oyster Creek nuclear facility, a boiling water reactor (BWR)-2 plant type, has never experienced core thermal-hydraulic instability. Power oscillations, however, have been observed in other BWR cores both domestically and internationally. Two modes of oscillations have been observed, core wide and regional half-core. During core wide oscillations, the neutron flux in the core oscillates in the radial fundamental mode. During regional half-core oscillations, higher order harmonics in the radial plane result in out-of-phase oscillations with the neutron flux in one half of the core oscillating 180 deg out-of-phase with the neutron flux in the other half of the core. General Design Criteria 12 requires either prevention or detection and suppression of power oscillations which could result in violations of fuel design limits. Analyses performed by General Electric have demonstrated that for large-magnitude oscillations the potential exists for violation of the safety limit minimum critical power ratio (MCPR). However, for plants with a flow-biased neutron flux scram automatic mitigation of oscillations may be provided at an oscillation magnitude below that at which the safety limit is challenged. Plant-specific analysis for Oyster Creek demonstrates that the existing average power range monitor (APRM) system will sense and suppress power oscillations prior to violation of any safety limits.

  12. Thermal hydraulic evaluation of consolidating tank C-106 waste into tank AY-102

    SciTech Connect

    Sathyanarayana, K.

    1996-02-01

    This report describes the thermal hydraulic analysis performed to provide a technical basis in support of consolidation of tank C-106 waste into tank AY-102. Several parametric calculations were performed using the HUB and GOTH computer codes. First, the current heat load of tank AY-102 was determined. Potential quantities of waste transfer from tank C-106 were established to maintain the peak temperatures of consolidated sludge in tank AY-102 to remain within Operating Specification limits. For this purpose, it was shown that active cooling of the tank floor was essential and a secondary ventilation flow of 2,000 cfm should be maintained. Transient calculations were also conducted to evaluate the effects of ambient meteorological cyclic conditions on sludge peak temperature, and loss of ventilation systems. Detailed calculations were also performed to estimate the insulating concrete air channels cooling effectiveness and the resulting peak temperatures for the consolidated sludge in tank AY-102. Calculations are were also performed for a primary and secondary ventilation systems outage, both individually and combined to establish limits on outage duration. Because of its active cooling mode of operation, the secondary ventilation system limits the outage duration.

  13. An anisotropic numerical model for thermal hydraulic analyses: application to liquid metal flow in fuel assemblies

    NASA Astrophysics Data System (ADS)

    Vitillo, F.; Vitale Di Maio, D.; Galati, C.; Caruso, G.

    2015-11-01

    A CFD analysis has been carried out to study the thermal-hydraulic behavior of liquid metal coolant in a fuel assembly of triangular lattice. In order to obtain fast and accurate results, the isotropic two-equation RANS approach is often used in nuclear engineering applications. A different approach is provided by Non-Linear Eddy Viscosity Models (NLEVM), which try to take into account anisotropic effects by a nonlinear formulation of the Reynolds stress tensor. This approach is very promising, as it results in a very good numerical behavior and in a potentially better fluid flow description than classical isotropic models. An Anisotropic Shear Stress Transport (ASST) model, implemented into a commercial software, has been applied in previous studies, showing very trustful results for a large variety of flows and applications. In the paper, the ASST model has been used to perform an analysis of the fluid flow inside the fuel assembly of the ALFRED lead cooled fast reactor. Then, a comparison between the results of wall-resolved conjugated heat transfer computations and the results of a decoupled analysis using a suitable thermal wall-function previously implemented into the solver has been performed and presented.

  14. Thermal-Hydraulic Design of a Fluoride High-Temperature Demonstration Reactor

    SciTech Connect

    Carbajo, Juan J; Qualls, A L

    2016-01-01

    INTRODUCTION The Fluoride High-Temperature Reactor (FHR) named the Demonstration Reactor (DR) is a novel reactor concept using molten salt coolant and TRIstructural ISOtropic (TRISO) fuel that is being developed at Oak Ridge National Laboratory (ORNL). The objective of the FHR DR is to advance the technology readiness level of FHRs. The FHR DR will demonstrate technologies needed to close remaining gaps to commercial viability. The FHR DR has a thermal power of 100 MWt, very similar to the SmAHTR, another FHR ORNL concept (Refs. 1 and 2) with a power of 125 MWt. The FHR DR is also a small version of the Advanced High Temperature Reactor (AHTR), with a power of 3400 MWt, cooled by a molten salt and also being developed at ORNL (Ref. 3). The FHR DR combines three existing technologies: (1) high-temperature, low-pressure molten salt coolant, (2) high-temperature coated-particle TRISO fuel, (3) and passive decay heat cooling systems by using Direct Reactor Auxiliary Cooling Systems (DRACS). This paper presents FHR DR thermal-hydraulic design calculations.

  15. Methodology of Internal Assessment of Uncertainty and Extension to Neutron Kinetics/Thermal-Hydraulics Coupled Codes

    SciTech Connect

    Petruzzi, A.; D'Auria, F.; Giannotti, W.; Ivanov, K.

    2005-02-15

    The best-estimate calculation results from complex system codes are affected by approximations that are unpredictable without the use of computational tools that account for the various sources of uncertainty.The code with (the capability of) internal assessment of uncertainty (CIAU) has been previously proposed by the University of Pisa to realize the integration between a qualified system code and an uncertainty methodology and to supply proper uncertainty bands each time a nuclear power plant (NPP) transient scenario is calculated. The derivation of the methodology and the results achieved by the use of CIAU are discussed to demonstrate the main features and capabilities of the method.In a joint effort between the University of Pisa and The Pennsylvania State University, the CIAU method has been recently extended to evaluate the uncertainty of coupled three-dimensional neutronics/thermal-hydraulics calculations. The result is CIAU-TN. The feasibility of the approach has been demonstrated, and sample results related to the turbine trip transient in the Peach Bottom NPP are shown. Notwithstanding that the full implementation and use of the procedure requires a database of errors not available at the moment, the results give an idea of the errors expected from the present computational tools.

  16. Incorporating Artificial Neural Networks in the dynamic thermal-hydraulic model of a controlled cryogenic circuit

    NASA Astrophysics Data System (ADS)

    Carli, S.; Bonifetto, R.; Savoldi, L.; Zanino, R.

    2015-09-01

    A model based on Artificial Neural Networks (ANNs) is developed for the heated line portion of a cryogenic circuit, where supercritical helium (SHe) flows and that also includes a cold circulator, valves, pipes/cryolines and heat exchangers between the main loop and a saturated liquid helium (LHe) bath. The heated line mimics the heat load coming from the superconducting magnets to their cryogenic cooling circuits during the operation of a tokamak fusion reactor. An ANN is trained, using the output from simulations of the circuit performed with the 4C thermal-hydraulic (TH) code, to reproduce the dynamic behavior of the heated line, including for the first time also scenarios where different types of controls act on the circuit. The ANN is then implemented in the 4C circuit model as a new component, which substitutes the original 4C heated line model. For different operational scenarios and control strategies, a good agreement is shown between the simplified ANN model results and the original 4C results, as well as with experimental data from the HELIOS facility confirming the suitability of this new approach which, extended to an entire magnet systems, can lead to real-time control of the cooling loops and fast assessment of control strategies for heat load smoothing to the cryoplant.

  17. Scalable three-dimensional thermal-hydraulic best-estimate code BAGIRA

    SciTech Connect

    Vasenin, V. A.; Krivchikov, M. A.; Kroshilin, V. E.; Kroshilin, A. E.; Roganov, V. A.

    2012-07-01

    The three-dimensional thermal-hydraulic best-estimate code BAGIRA for modeling of multi-phase flows was developed without any artificial physical assumptions or simplifications. The mathematical model is based on numerical approximations of exact three-dimensional equations, including effective multi-dimensional models for turbulent heat and mass transfer. With use of BAGIRA All-Russian Scientific Research Inst. of Nuclear Power Plants (VNIIAES) has developed a full-scope and analytical simulators using BAGIRA for a number of power plants with VVER-1000 and RBMK type design, which are being used in Kalinin, Kursk, Smolensk, Chernobyl, and Bilibino NPPs. The comparison of calculated and experimental results shows that BAGIRA can successfully reproduce the most important processes observed in experiments. BAGIRA is implemented in FORTRAN. It is a relatively complicated code that tends to decompose task by aspects. Such a style is welcoming for extensions, which can be added without code redesign. We would like to present an aspect-oriented mix-in approach for BAGIRA code extension. It allows to make it scalable in number of directions leaving original code base untouched. It is possible to add new effects/units, and even to produce a supercomputer version of the code. The last is a key point today due to availability of low-cost compact supercomputers, which makes building compact NPP simulators possible. (authors)

  18. COBRA-SFS: A thermal-hydraulic analysis code for spent fuel storage and transportation casks

    SciTech Connect

    Michener, T.E.; Rector, D.R.; Cuta, J.M.; Dodge, R.E.; Enderlin, C.W.

    1995-09-01

    COBRA-SFS is a general thermal-hydraulic analysis computer code for prediction of material temperatures and fluid conditions in a wide variety of systems. The code has been validated for analysis of spent fuel storage systems, as part of the Commercial Spent Fuel Management Program of the US Department of Energy. The code solves finite volume equations representing the conservation equations for mass, moment, and energy for an incompressible single-phase heat transfer fluid. The fluid solution is coupled to a finite volume solution of the conduction equation in the solid structure of the system. This document presents a complete description of Cycle 2 of COBRA-SFS, and consists of three main parts. Part 1 describes the conservation equations, constitutive models, and solution methods used in the code. Part 2 presents the User Manual, with guidance on code applications, and complete input instructions. This part also includes a detailed description of the auxiliary code RADGEN, used to generate grey body view factors required as input for radiative heat transfer modeling in the code. Part 3 describes the code structure, platform dependent coding, and program hierarchy. Installation instructions are also given for the various platform versions of the code that are available.

  19. Thermal-hydraulic calculations for the conversion to LEU of a research reactor core

    SciTech Connect

    Grigoriadis, D.; Varvayanni, M.; Catsaros, N.; Stakakis, E.

    2008-07-15

    The thermal-hydraulic analysis performed for the needs of the conversion of the open pool 5MW Greek Research Reactor (GRR-1) to a pure Low Enrichment (LEU) configuration is presented. The methodology was based on a complete set of neutronic calculations performed for the new core configuration, in compliance with pre-defined Operation Limiting Conditions. The hottest channel analysis approach was adopted, and peaking factors were used to account for fabrication or measuring uncertainties. Calculations were carried out using the numerical codes NATCON, PLTEMP and PARET provided by Argonne National Laboratory (ANL). Two main different classes of conditions were considered, namely i) steady state normal operating conditions and ii) transient cases related to accidental events including reactivity feedback effects. For steady state operating conditions the behaviour of the new configuration was examined both for forced and natural convection cooling modes. Transient calculations considered several initiating events including reactivity insertion accidents (slow or fast reactivity insertion) and total or partial loss-of-flow accidents, i.e. in accordance to guidelines provided by the IAEA for research Reactors. (author)

  20. Investigation of approximations in thermal-hydraulic modeling of core conversions

    SciTech Connect

    Garner, Patrick L.; Hanan, Nelson A.

    2008-07-15

    Neutronics analyses for core conversions are usually fairly detailed, for example representing all 4 flats and all 4 corners of all 6 tubes of all 20 IRT-3M or -4M fuel assemblies in the core of the VVR-SM reactor in Uzbekistan. The coupled neutronics and thermal-hydraulic analysis for safety analysis transients is usually less detailed, for example modeling only a hot and an average fuel plate and the associated coolant. Several of the approximations have been studied using the RELAP5 and PARET computer codes in order to provide assurance that the lack of full detail is not important to the safety analysis. Two specific cases studied are (1) representation of a core of same- type fuel assemblies by a hot and an average assembly each having multiple channels as well as by merely a hot and average channel and (2) modeling a core containing multiple fuel types as the sum of fractional core models for each fuel type. (author)

  1. Multi-Function Waste Tank Facility thermal hydraulic analysis for Title II design

    SciTech Connect

    Cramer, E.R.

    1994-11-10

    The purpose of this work was to provide the thermal hydraulic analysis for the Multi-Function Waste Tank Facility (MWTF) Title II design. Temperature distributions throughout the tank structure were calculated for subsequent use in the structural analysis and in the safety evaluation. Calculated temperatures of critical areas were compared to design allowables. Expected operating parameters were calculated for use in the ventilation system design and in the environmental impact documentation. The design requirements were obtained from the MWTF Functional Design Criteria (FDC). The most restrictive temperature limit given in the FDC is the 200 limit for the haunch and dome steel and concrete. The temperature limit for the rest of the primary and secondary tanks and concrete base mat and supporting pad is 250 F. Also, the waste should not be allowed to boil. The tank geometry was taken from ICF Kaiser Engineers Hanford drawing ES-W236A-Z1, Revision 1, included here in Appendix B. Heat removal rates by evaporation from the waste surface were obtained from experimental data. It is concluded that the MWTF tank cooling system will meet the design temperature limits for the design heat load of 700,000 Btu/h, even if cooling flow is lost to the annulus region, and temperatures change very slowly during transients due to the high heat capacity of the tank structure and the waste. Accordingly, transients will not be a significant operational problem from the viewpoint of meeting the specified temperature limits.

  2. Thermal-hydraulics and safety analysis of sectored compact reactor for lunar surface power

    SciTech Connect

    Schriener, T. M.; El-Genk, M. S.

    2012-07-01

    The liquid NaK-cooled, fast-neutron spectrum, Sectored Compact Reactor (SCoRe-N 5) concept has been developed at the Univ. of New Mexico for lunar surface power applications. It is loaded with highly enriched UN fuel pins in a triangular lattice, and nominally operates at exit and inlet coolant temperatures of 850 K and 900 K. This long-life reactor generates up to 1 MWth continuously for {>=} 20 years. To avoid a single point failure in reactor cooling, the core is divided into 6 sectors that are neutronically and thermally coupled, but hydraulically independent. This paper performs a 3-D the thermal-hydraulic analysis of SCoRe--N 5 at nominal operation temperatures and a power level of 1 MWth. In addition, the paper investigates the potential of continuing reactor operation at a lower power in the unlikely event that one sector in the core experiences a loss of coolant (LOC). Redesigning the core with a contiguous steel matrix enhances the cooling of the sector experiencing a LOC. Results show that with a core sector experiencing a LOC, SCORE-N 5 could continue operating safely at a reduced power of 166.6 kWth. (authors)

  3. Independent assessment of MELCOR as a severe accident thermal-hydraulic/source term analysis tool

    SciTech Connect

    Madni, I.K.; Eltawila, F.

    1994-01-01

    MELCOR is a fully integrated computer code that models all phases of the progression of severe accidents in light water reactor nuclear power plants, and is being developed for the US Nuclear Regulatory Commission (NRC) by Sandia National Laboratories (SNL). Brookhaven National Laboratory (BNL) has a program with the NRC called ``MELCOR Verification, Benchmarking, and Applications,`` whose aim is to provide independent assessment of MELCOR as a severe accident thermal-hydraulic/source term analysis tool. The scope of this program is to perform quality control verification on all released versions of MELCOR, to benchmark MELCOR against more mechanistic codes and experimental data from severe fuel damage tests, and to evaluate the ability of MELCOR to simulate long-term severe accident transients in commercial LWRs, by applying the code to model both BWRs and PWRs. Under this program, BNL provided input to the NRC-sponsored MELCOR Peer Review, and is currently contributing to the MELCOR Cooperative Assessment Program (MCAP). This paper presents a summary of MELCOR assessment efforts at BNL and their contribution to NRC goals with respect to MELCOR.

  4. Current and anticipated uses of thermal-hydraulic codes in Spain

    SciTech Connect

    Pelayo, F.; Reventos, F.

    1997-07-01

    Spanish activities in the field of Applied Thermal-Hydraulics are steadily increasing as the codes are becoming practicable enough to efficiently sustain engineering decision in the Nuclear Power industry. Before reaching this point, a lot of effort has been devoted to achieve this goal. This paper briefly describes this process, points at the current applications and draws conclusions on the limitations. Finally it establishes the applications where the use of T-H codes would be worth in the future, this in turn implies further development of the codes to widen the scope of application and improve the general performance. Due to 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.

  5. Stabilized FE simulation of prototype thermal-hydraulics problems with integrated adjoint-based capabilities

    NASA Astrophysics Data System (ADS)

    Shadid, J. N.; Smith, T. M.; Cyr, E. C.; Wildey, T. M.; Pawlowski, R. P.

    2016-09-01

    A critical aspect of applying modern computational solution methods to complex multiphysics systems of relevance to nuclear reactor modeling, is the assessment of the predictive capability of specific proposed mathematical models. In this respect the understanding of numerical error, the sensitivity of the solution to parameters associated with input data, boundary condition uncertainty, and mathematical models is critical. Additionally, the ability to evaluate and or approximate the model efficiently, to allow development of a reasonable level of statistical diagnostics of the mathematical model and the physical system, is of central importance. In this study we report on initial efforts to apply integrated adjoint-based computational analysis and automatic differentiation tools to begin to address these issues. The study is carried out in the context of a Reynolds averaged Navier-Stokes approximation to turbulent fluid flow and heat transfer using a particular spatial discretization based on implicit fully-coupled stabilized FE methods. Initial results are presented that show the promise of these computational techniques in the context of nuclear reactor relevant prototype thermal-hydraulics problems.

  6. Development of a Spatially-Selective, Nonlinear Refinement Algorithm for Thermal-Hydraulic Safety Analysis

    NASA Astrophysics Data System (ADS)

    Lloyd, Lewis John

    This work focused on developing a novel method for solving the nonlinear partial differential equations associated with thermal-hydraulic safety analysis software. Traditional methods involve solving large systems of nonlinear equations. One class of methods linearizes the nonlinear equations and attempts to minimize the nonlinear truncation error with timestep size selection. These linearized methods are characterized by low computational cost but reduced accuracy. Another class resolves those nonlinearities by using an iterative nonlinear refinement technique. However, these iterative methods are computationally expensive when multiple iterates are required to resolve the nonlinearities. These two paradigms stand at the opposite ends of a spectrum, and the middle ground had yet to be investigated. This research sought to find that middle ground, a balance between the competing incentives of computational cost and accuracy, by creating a hybrid method: a spatially-selective, nonlinear refinement (SNR) algorithm. As part of this work, the two-phase, three-field software COBRA was converted from a linearized semi-implicit solver to a nonlinearly convergent solver; an operator-based scaling that provides a physically meaningful convergence measure was developed and implemented; and the SNR algorithm was developed to enable a subdomain of the simulation to be subjected to multiple nonlinear iterates while maintaining global consistency. By selecting those areas of the computational domain where nonlinearities are expected to be high and subjecting only them to multiple nonlinear iterations, the accuracy of the nonlinear solver may be obtained without its associated computational cost.

  7. COBRA-SFS CYCLE 3. Thermal Hydraulic Analysis of Spent Fuel Casks

    SciTech Connect

    Rector, D.R.; Cuta, J.M.; Enderlin, C.W.

    1995-09-01

    COBRA-SFS (Spent Fuel Storage) is a code for thermal-hydraulic analysis of multi-assembly spent fuel storage and transportation systems. It uses a lumped parameter finite difference approach to predict flow and temperature distributions in spent fuel storage systems and fuel assemblies, under forced and natural convection heat transfer conditions. Derived from the COBRA family of codes, which have been extensively evaluated against in-pile and out-of-pile data, COBRA-SFS retains all the important features of the COBRA codes for single phase fluid analysis, and extends the range application to include problems with two-dimensional radiative and three-dimensional conductive heat transfer. COBRA-SFS has been used to analyze various single- and multi-assembly spent fuel storage systems containing unconsolidated and consolidated fuel rods, with a variety of fill media, including air, helium and vacuum. Cycle 0 of COBRA-SFS was released in 1986. Subsequent applications of the code led to development of additional capabilities, which resulted in the release of Cycle 1 in February 1989. Since then, the code has undergone an independent technical review as part of a submittal to the Nuclear Regulatory Commission for a generic license to apply the code to spent fuel storage system analysis. Modifications and improvements to the code have been combined to form Cycle 2. Cycle 3., the newest version of COBRA-SFS, has been validated and verified for transient applications, such as a storage cask thermal response to a pool fire.

  8. Thermal-hydraulic characteristics of a Westinghouse Model 51 steam generator. Volume 2. Appendix A, numerical results. Interim report. [CALIPSOS code numerical data

    SciTech Connect

    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.

  9. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-5

    SciTech Connect

    Block, R.C.; Feiner, F.

    1995-09-01

    This document, Volume 1, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  10. Thermal-Hydraulics and Electrochemistry of a Boiling Solution in a Porous Sludge Pile A Test Methodology

    SciTech Connect

    R.F. Voelker

    2001-05-03

    When boiling occurs in a pile of porous corrosion products (sludge), chemical species can concentrate. These species can react with the corrosion products and transform the sludge into a rock hard mass and/or create a corrosive environment. In-situ measurements are required to improve the understanding of this process, and the thermal-hydraulic and electrochemical environment in the pile. A test method is described that utilizes a water heated instrumented tube array in an autoclave to perform the in-situ measurements. As a proof of method feasibility, tests were performed in an alkaline phosphate solution. The test data is discussed. Temperature changes and electrochemical potential shifts were used to indicate when chemicals concentrate and if/when the pile hardens. Post-test examinations confirmed hardening occurred. Experiments were performed to reverse the hardening process. A one-dimensional model, utilizing capillary forces, was developed to understand the thermal-hydraulic measurements.

  11. Thermal Hydraulic Design and Analysis of a Water-Cooled Ceramic Breeder Blanket with Superheated Steam for CFETR

    NASA Astrophysics Data System (ADS)

    Cheng, Xiaoman; Ma, Xuebin; Jiang, Kecheng; Chen, Lei; Huang, Kai; Liu, Songlin

    2015-09-01

    The water-cooled ceramic breeder blanket (WCCB) is one of the blanket candidates for China fusion engineering test reactor (CFETR). In order to improve power generation efficiency and tritium breeding ratio, WCCB with superheated steam is under development. The thermal-hydraulic design is the key to achieve the purpose of safe heat removal and efficient power generation under normal and partial loading operation conditions. In this paper, the coolant flow scheme was designed and one self-developed analytical program was developed, based on a theoretical heat transfer model and empirical correlations. Employing this program, the design and analysis of related thermal-hydraulic parameters were performed under different fusion power conditions. The results indicated that the superheated steam water-cooled blanket is feasible. supported by the National Special Project for Magnetic Confined Nuclear Fusion Energy of China (Nos. 2013GB108004, 2014GB122000 and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  12. Overview of the Use of ATHENA for Thermal-Hydraulic Analysis of Systems with Lead-Bismuth Coolant

    SciTech Connect

    Davis, Cliff Bybee; Shieh, Arthur Shan Luk

    2000-04-01

    The INEEL and MIT are investigating the suitability of lead-bismuth cooled fast reactor for producing low-cost electricity as well as for actinide burning. This paper is concerned with the general area of thermal-hydraulics of lead-bismuth cooled reactors. The ATHENA code is being used in the thermal-hydraulic design and analysis of lead-bismuth cooled reactors. The ATHENA code was reviewed to determine its applicability for simulating lead-bismuth cooled reactors. Two modifications were made to the code as a result of this review. Specifically, a correlation to represent heat transfer from rod bundles to a liquid metal and a void correlation based on data taken in a mixture of lead-bismuth and steam were added the code. The paper also summarizes the analytical work that is being performed with the code and plans for future analytical work.

  13. Overview of the use of ATHENA for thermal-hydraulic analysis of systems with lead-bismuth coolant

    SciTech Connect

    C. B. Davis; A. S. Shieh

    2000-04-02

    The INEEL and MIT are investigating the suitability of lead-bismuth cooled fast reactor for producing low-cost electricity as well as for actinide burning. This paper is concerned with the general area of thermal-hydraulics of lead-bismuth cooled reactors. The ATHENA code is being used in the thermal-hydraulic design and analysis of lead-bismuth cooled reactors. The ATHENA code was reviewed to determine its applicability for simulating lead-bismuth cooled reactors. Two modifications were made to the code as a result of this review. Specifically, a correlation to represent heat transfer from rod bundles to a liquid metal and a void correlation based on data taken in a mixture of lead-bismuth and steam were added the code. The paper also summarizes the analytical work that is being performed with the code and plans for future analytical work.

  14. ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) simulation of a loss of coolant accident in a space reactor

    SciTech Connect

    Roth, P.A.; Shumway, R.W.

    1988-01-01

    The Advanced Thermal Hydraulic Energy Network Analyzer (ATHENA) code was used to simulate a loss-of-coolant accident (LOCA) in a conceptual space reactor design. ATHENA provides the capability of simulating the thermal-hydraulic behavior of the wide variety of systems which are being considered for use in space reactors. Flow loops containing any one of several available working fluids may interact through thermal connections with other loops containing the same or a different working fluid. The code can be used to model special systems such as: heat pipes, point reactor kinetics, plant control systems, turbines, valves, and pumps. This work demonstrates the application of the thermal radiation model which has been recently incorporated into ATHENA and verifies the need for supplemental reactor cooling to prevent reactor fuel damage in the event of a LOCA.

  15. Program ELM: A tool for rapid thermal-hydraulic analysis of solid-core nuclear rocket fuel elements

    NASA Technical Reports Server (NTRS)

    Walton, James T.

    1992-01-01

    This report reviews the state of the art of thermal-hydraulic analysis codes and presents a new code, Program ELM, for analysis of fuel elements. ELM is a concise computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in a nuclear thermal rocket reactor with axial coolant passages. The program was developed as a tool to swiftly evaluate various heat transfer coefficient and friction factor correlations generated for turbulent pipe flow with heat addition which have been used in previous programs. Thus, a consistent comparison of these correlations was performed, as well as a comparison with data from the NRX reactor experiments from the Nuclear Engine for Rocket Vehicle Applications (NERVA) project. This report describes the ELM Program algorithm, input/output, and validation efforts and provides a listing of the code.

  16. History of the 185-/189-D thermal hydraulics laboratory and its effects on reactor operations at the Hanford Site

    SciTech Connect

    Gerber, M.S.

    1994-09-01

    The 185-D deaeration building and the 189-D refrigeration building were constructed at Hanford during 1943 and 1944. Both buildings were constructed as part of the influent water cooling system for D reactor. The CMS studies eliminated the need for 185-D function. Early gains in knowledge ended the original function of the 189-D building mission. In 1951, 185-D and 189-D were converted to a thermal-hydraulic laboratory. The experiments held in the thermal-hydraulic lab lead to historic changes in Hanford reactor operations. In late 1951, the exponential physics experiments were moved to the 189-D building. In 1958, new production reactor experiments were begun in 185/189-D. In 1959, Plutonium Recycle Test Reactor experiments were added to the 185/189-D facility. By 1960, the 185/189-D thermal hydraulics laboratory was one of the few full service facilities of its type in the nation. During the years 1961--1963 tests continued in the facility in support of existing reactors, new production reactors, and the Plutonium Recycle Test Reactor. In 1969, Fast Flux Test Facility developmental testings began in the facility. Simulations in 185/189-D building aided in the N Reactor repairs in the 1980`s. In 1994 the facility was nominated to the National Register of Historic Places, because of its pioneering role over many years in thermal hydraulics, flow studies, heat transfer, and other reactor coolant support work. During 1994 and 1995 it was demolished in the largest decontamination and decommissioning project thus far in Hanford Site history.

  17. VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 3. Programmer's manual. Final report. [PWR; BWR

    SciTech Connect

    Stewart, C.W.; Koontz, A.S.; Cuta, J.M.; Montgomery, S.D.

    1983-05-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear-reactor-core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This is Volume 3, the Programmer's Manual. It explains the codes' structures and the computer interfaces.

  18. The Euratom-Rosatom ERCOSAM-SAMARA projects on containment thermal-hydraulics of current and future LWRs for severe accident management

    SciTech Connect

    Paladino, D.; Guentay, S.; Andreani, M.; Tkatschenko, I.; Brinster, J.; Dabbene, F.; Kelm, S.; Allelein, H. J.; Visser, D. C.; Benz, S.; Jordan, T.; Liang, Z.; Porcheron, E.; Malet, J.; Bentaib, A.; Kiselev, A.; Yudina, T.; Filippov, A.; Khizbullin, A.; Kamnev, M.; Zaytsev, A.; Loukianov, A.

    2012-07-01

    During a postulated severe accident with core degradation, hydrogen would form in the reactor pressure vessel mainly due to high temperatures zirconium-steam reaction and flow together with steam into the containment where it will mix with the containment atmosphere (steam-air). The hydrogen transport into the containment is a safety concern because it can lead to explosive mixtures through the associated phenomena of condensation, mixing and stratification. The ERCOSAM and SAMARA projects, co-financed by the European Union and the Russia, include various experiments addressing accident scenarios scaled down from existing plant calculations to different thermal-hydraulics facilities (TOSQAN, MISTRA, PANDA, SPOT). The tests sequences aim to investigate hydrogen concentration build-up and stratification during a postulated accident and the effect of the activation of Severe Accident Management systems (SAMs), e.g. sprays, coolers and Passive Auto-catalytic Recombiners (PARs). Analytical activities, performed by the project participants, are an essential component of the projects, as they aim to improve and validate various computational methods. They accompany the projects in the various phases; plant calculations, scaling to generic containment and to the different facilities, planning pre-test and post-test simulations are performed. Code benchmark activities on the basis of conceptual near full scale HYMIX facility will finally provide a further opportunity to evaluate the applicability of the various methods to the study of scaling issues. (authors)

  19. RELAP5 model to simulate the thermal-hydraulic effects of grid spacers and cladding rupture during reflood

    SciTech Connect

    Nithianandan, C.K.; Klingenfus, J.A.; Reilly, S.S.

    1995-09-01

    Droplet breakup at spacer grids and a cladding swelled and ruptured locations plays an important role in the cooling of nuclear fuel rods during the reflooding period of a loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). During the reflood phase, a spacer grid affects the thermal-hydraulic system behavior through increased turbulence, droplet breakup due to impact on grid straps, grid rewetting, and liquid holdup due to grid form losses. Recently, models to simulate spacer grid effects and blockage and rupture effects on system thermal hydraulics were added to the B&W Nuclear Technologies (BWNT) version of the RELAP5/MOD2 computer code. Several FLECHT-SEASET forced reflood tests, CCTF Tests C1-19 and C2-6, SCTF Test S3-15, and G2 Test 561 were simulated using RELAP5/MOD2-B&W to verify the applicability of the model at the cladding swelled and rupture locations. The results demonstrate the importance of modeling the thermal-hydraulic effects due to grids, and clad swelling and rupture to correctly predict the clad temperature response during the reflood phase of large break LOCA. The RELAP5 models and the test results are described in this paper.

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

    NASA Astrophysics Data System (ADS)

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

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

  1. A coupled neutronic/thermal-hydraulic scheme between COBAYA3 and SUBCHANFLOW within the NURESIM simulation platform

    SciTech Connect

    Calleja, M.; Stieglitz, R.; Sanchez, V.; Jimenez, J.; Imke, U.

    2012-07-01

    Multi-scale, multi-physics problems reveal significant challenges while dealing with coupled neutronic/thermal-hydraulic solutions. Current generation of codes applied to Light Water Reactors (LWR) are based on 3D neutronic nodal methods coupled with one or two phase flow thermal-hydraulic system or sub-channel codes. In addition, spatial meshing and temporal schemes are crucial for the proper description of the non-symmetrical core behavior in case of transient and accidents e.g. reactivity insertion accidents. This paper describes the coupling approach between the 3D neutron diffusion code COBAYA3 and the sub-channel code SUBCHANFLOW within SALOME. The coupling is done inside the SALOME open source platform that is characterized by a powerful pre- and post-processing capabilities and a novel functionality for mapping of the neutronic and thermal hydraulic domains. The peculiar functionalities of SALOME and the steps required for the code integration and coupling are presented. The validation of the coupled codes is done based on two benchmarks the PWR MOX/UO{sub 2} RIA and the TMI-1 MSLB benchmark. A discussion of the prediction capability of COBAYA3/SUBCHANFLOW compared to other coupled solutions will be provided too. (authors)

  2. Simulating High Flux Isotope Reactor Core Thermal-Hydraulics via Interdimensional Model Coupling

    SciTech Connect

    Travis, Adam R

    2014-05-01

    A coupled interdimensional model is presented for the simulation of the thermal-hydraulic characteristics of the High Flux Isotope Reactor core at Oak Ridge National Laboratory. The model consists of two domains a solid involute fuel plate and the surrounding liquid coolant channel. The fuel plate is modeled explicitly in three-dimensions. The coolant channel is approximated as a twodimensional slice oriented perpendicular to the fuel plate s surface. The two dimensionally-inconsistent domains are linked to one another via interdimensional model coupling mechanisms. The coupled model is presented as a simplified alternative to a fully explicit, fully three-dimensional model. Involute geometries were constructed in SolidWorks. Derivations of the involute construction equations are presented. Geometries were then imported into COMSOL Multiphysics for simulation and modeling. Both models are described in detail so as to highlight their respective attributes in the 3D model, the pursuit of an accurate, reliable, and complete solution; in the coupled model, the intent to simplify the modeling domain as much as possible without affecting significant alterations to the solution. The coupled model was created with the goal of permitting larger portions of the reactor core to be modeled at once without a significant sacrifice to solution integrity. As such, particular care is given to validating incorporated model simplifications. To the greatest extent possible, the decrease in solution time as well as computational cost are quantified versus the effects such gains have on the solution quality. A variant of the coupled model which sufficiently balances these three solution characteristics is presented alongside the more comprehensive 3D model for comparison and validation.

  3. Thermal-hydraulic modeling of the Pennsylvania State University Breazeale Nuclear Reactor (PSBR)

    NASA Astrophysics Data System (ADS)

    Chang, Jong E.

    2005-11-01

    Earlier experiments determined that the Pennsylvania State University Breazeale Nuclear Reactor (PSBR) core is cooled, not by an axial flow, but rather by a strong cross flow due to the thermal expansion of the coolant. To further complicate the flow field, a nitrogen-16 (N-16) pump was installed above the PSBR core to mix the exiting core buoyant thermal plume in order to delay the rapid release of radioactive N-16 to the PSBR pool surface. Thus, the interaction between the N-16 jet flow and the buoyancy driven flow complicates the analysis of the flow distribution in the PSBR pool. The main objectives of this study is to model the thermal-hydraulic behavior of the PSBR core and pool. During this study four major things were performed including the Computational Fluid Dynamics (CFD) model for the PSBR pool, the stand-alone fuel rod model for a PSBR fuel rod, the velocity measurements in and around the PSBR core, and the temperature measurements in the PSBR pool. Once the flow field was predicted by the CFD model, the measurement devices were manufactured and calibrated based on the CFD results. The major contribution of this study is to understand and to explain the flow behavior in the PSBR subchannels and pool using the FLOW3D model. The stand-alone dynamic fuel rod model was developed to determine the temperature distribution inside a PSBR fuel rod. The stand-alone fuel rod model was coupled to the FLOW3D model and used to predict the temperature behavior during steady-state and pulsing. The heat transfer models in the stand-alone fuel rod code are used in order to overcome the disadvantage of the CFD code, which does not calculate the mechanical stress, the gap conductance, and the two phase heat transfer. (Abstract shortened by UMI.)

  4. Rod bundle thermal-hydraulic and melt progression analysis of CORA severe fuel damage experiments

    SciTech Connect

    Suh, K.Y. )

    1994-04-01

    An integral, fast-running computational model is developed to simulate the thermal-hydraulic and melt progression behavior in a nuclear reactor rod bundle under severe fuel damage conditions. This consists of the submodels for calculating steaming from the core, hydrogen formation, heat transfer in and out of the core, cooling from core spray or injection, and, most importantly, fuel melting, relocation, and freezing with chemical interactions taking place among the material constituents in a degrading core. The integral model is applied to three German severe fuel damage tests to analyze the core thermal and melt behavior: CORA-16 (18-rod bundle and slow cooling), CORA-17 (18-rod bundle and quenching), and CORA-18 (48-rod bundle and slow cooling). Results of the temperature response of the fuel rods, the channel box, and the absorber blade; hydrogen generation from the fuel rod and the channel box; and core material eutectic formation, melt relocation, and blockage formation are discussed. Reasonable agreement is observed for component temperatures at midelevation where prediction and measurement uncertainties are minimal. However, discrepancies or uncertainties are noticed for hydrogen generation and core-melt progression. The experimentally observed peak generation of hydrogen upon reflooding is not able to be reproduced, and the total amount generated is generally underpredicted primarily because of the early relocation of the Zircaloy fuel channel box and cladding. Also, difficulties are encountered in the process of assessing the core-melt formation and the relocation model because of either modeling uncertainties or a lack of definitive metallurgical data as a function of time throughout the transient.

  5. Thermal hydraulic design analysis of ternary carbide fueled square-lattice honeycomb nuclear rocket engine

    SciTech Connect

    Furman, Eric M.; Anghaie, Samim

    1999-01-22

    A computational analysis is conducted to determine the optimum thermal-hydraulic design parameters for a square-lattice honeycomb nuclear rocket engine core that will incorporate ternary carbide based uranium fuels. Recent studies at the Innovative Nuclear Space Power and Propulsion Institute (INSPI) have demonstrated the feasibility of processing solid solution, ternary carbide fuels such as (U, Zr, Nb)C, (U, Zr, Ta)C, (U, Zr, Hf)C and (U, Zr, W)C. The square-lattice honeycomb design provides high strength and is amenable to the processing complexities of these ultrahigh temperature fuels. A parametric analysis is conducted to examine how core geometry, fuel thickness and the propellant flow area effect the thermal performance of the nuclear rocket engine. The principal variables include core size (length and diameter) and fuel element dimensions. The optimum core configuration requires a balance between high specific impulse and thrust level performance, and maintaining the temperature and strength limits of the fuel. A nuclear rocket engine simulation code is developed and used to examine the system performance as well as the performance of the main reactor core components. The system simulation code was originally developed for analysis of NERVA-Derivative and Pratt and Whitney XNR-2000 nuclear thermal rockets. The code is modified and adopted to the square-lattice geometry of the new fuel design. Thrust levels ranging from 44,500 to 222,400 N (10,000 to 50,000 lbf) are considered. The average hydrogen exit temperature is kept at 2800 K, which is well below the melting point of these fuels. For a nozzle area ratio of 300 and a thrust chamber pressure of 4.8 Mpa (700 psi), the specific impulse is 930 s. Hydrogen temperature and pressure distributions in the core and the fuel maximum temperatures are calculated.

  6. Thermal Hydraulic Challenges of Gas Cooled Fast Reactors with Passive Safety Features

    SciTech Connect

    Michael Pope; Jeong-Ik Lee; Pavel Hejzlar; Michael J. Driscoll

    2009-05-01

    Transient response of a Gas cooled Fast Reactor (GFR) coupled to a recompression supercritical CO2 (S-CO2) power conversion system (PCS) in a direct cycle to Loss of Coolant and Loss of Generator Load Accidents is analyzed using RELAP5-3D. A number of thermal hydraulic challenges for GFR design are pointed out as the designers strive to accommodate cooling of the high power density core of a fast reactor by a gas with its inherently low heat transfer capability, in particular under post LOCA events when system pressure is lost and when reliance on passive decay heat removal is emphasized. Although it is possible to design a S-CO2 cooled GFR that can survive LOCA by cooling the core through natural circulating loops between the core and elevated emergency cooling heat exchangers, it is not an attractive approach because of various bypass paths that can, depending on break location, degrade core cooling. Moreover, natural circulation gas loops can operate in deteriorated heat transfer regimes with substantial reduction of heat transfer coefficient: as low as 30% of forced convection values, and data and correlations in these regimes carry large uncertainties. Therefore, reliable battery powered blowers for post-LOCA decay heat removal (DHR) that provide flow in well defined regimes with low uncertainty, and can be easily over-designed to accommodate bypass flows were selected. The results confirm that a GFR with such a DHR system and negative coolant void worth can withstand LOCA with and without scram as well as loss of electrical load without exceeding core temperature and turbomachinery overspeed limits.

  7. COBRA-SFS CYCLE 3: Code System for Thermal Hydraulic Analysis of Spent Fuel Casks

    2003-11-01

    COBRA-SFS (Spent Fuel Storage) is a code for thermal-hydraulic analysis of multi-assembly spent fuel storage and transportation systems. It uses a lumped parameter finite difference approach to predict flow and temperature distributions in spent fuel storage systems and fuel assemblies, under forced and natural convection heat transfer conditions. Derived from the COBRA family of codes, which have been extensively evaluated against in-pile and out-of-pile data, COBRA-SFS retains all the important features of the COBRA codesmore » for single phase fluid analysis and extends the range application to include problems with two-dimensional radiative and three-dimensional conductive heat transfer. COBRA-SFS has been used to analyze various single- and multi-assembly spent fuel storage systems containing unconsolidated and consolidated fuel rods, with a variety of fill media, including air, helium and vacuum. Cycle 0 of COBRA-SFS was released in 1986. Subsequent applications of the code led to development of additional capabilities, which resulted in the release of Cycle 1 in February 1989. Since then, the code has undergone an independent technical review as part of a submittal to the Nuclear Regulatory Commission for a generic license to apply the code to spent fuel storage system analysis. Modifications and improvements to the code have been combined to form Cycle 2. Cycle 3., the newest version of COBRA-SFS, has been validated and verified for transient applications, such as a storage cask thermal response to a pool fire.« less

  8. ORNL rod-bundle heat-transfer test data. Volume 7. Thermal-Hydraulic Test Facility experimental data report for test series 3. 07. 9 - steady-state film boiling in upflow

    SciTech Connect

    Mullins, C.B.; Felde, D.K.; Sutton, A.G.; Gould, S.S.; Morris, D.G.; Robinson, J.J.

    1982-05-01

    Thermal-Hydraulic Test Facility (THTF) test series 3.07.9 was conducted by members of the Oak Ridge National Laboratory Pressurized-Water Reactor (ORNL-PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on September 11, September 18, and October 1, 1980. The objective of the program is to investigate heat transfer phenomena believed to occur in PWRs during accidents, including small- and large-break loss-of-coolant accidents. Test series 3.07.9 was designed to provide steady-state film boiling data in rod bundle geometry under reactor accident-type conditions. This report presents the reduced instrument responses for THTF test series 3.07.9. Also included are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers.

  9. Thermal-hydraulic analysis of the liquid mercury target for the national spallation neutron source

    SciTech Connect

    Siman-Tov, M.; Wendel, M.W.; Haines, J.R.; Rogers, M.

    1997-04-01

    The National Spallation Neutron Source (NSNS) is a high-energy, accelerator-based spallation neutron source being designed by a multi-laboratory team led by Oak Ridge National Laboratory (ORNL) to achieve very high fluxes of neutrons for scientific experiments. The NSNS is proposed to have a 1 MW beam of high-energy ({approximately}1 GeV) protons upgradable to 5 MW and operating at 60 Hz with a pulse duration of 0.5 {mu}s. Peak steady-state power density in the target is about 640 MW/m{sup 3} for 1 MW, whereas the pulse instantaneous peak power density is as high as 22,000 GW/m{sup 3}. The local peak temperature rise for a single pulse over it`s time-averaged value is only 6{degrees}C, but the rate of this temperature rise during the pulse is extremely fast ({approximately}12 million {degrees}C/s). In addition to the resulting thermal shock and materials compatibility concerns, key feasibility issues for the target are related to its thermal-hydraulic performance. These include proper flow distribution, flow reversals and stagnation zones, possible {open_quotes}hot spots{close_quotes}, cooling of the beam {open_quotes}window{close_quotes}, and the challenge of mitigating the effects of thermal shock through possible injection of helium bubbles. An analytic approach was used on the PC spreadsheet EXCEL to evaluate target design options and to determine the global T/H parameters in the current concept. The general computational fluid dynamics (CFD) code CFX was used to simulate the detailed time-averaged two-dimensional thermal and flow distributions in the liquid mercury. In this paper, an overview of the project and the results of this preliminary work are presented. Heat transfer characteristics of liquid mercury under wetting and non-wetting conditions are discussed, and future directions of the program in T/H analysis and R&D are outlined.

  10. Design and Construction of Experiment for Direct Electron Irradiation of Uranyl Sulfate Solution: Bubble Formation and Thermal Hydraulics Studies

    SciTech Connect

    Chemerisov, Sergey; Gromov, Roman; Makarashvili, Vakho; Heltemes, Thad; Sun, Zaijing; Wardle, Kent E.; Bailey, James; Quigley, Kevin; Stepinski, Dominique; Vandegrift, George

    2014-10-01

    Argonne is assisting SHINE Medical Technologies in developing SHINE, a system for producing fission-product 99Mo using a D/T-accelerator to produce fission in a non-critical target solution of aqueous uranyl sulfate. We have developed an experimental setup for studying thermal-hydraulics and bubble formation in the uranyl sulfate solution to simulate conditions expected in the SHINE target solution during irradiation. A direct electron beam from the linac accelerator will be used to irradiate a 20 L solution (sector of the solution vessel). Because the solution will undergo radiolytic decomposition, we will be able to study bubble formation and dynamics and effects of convection and temperature on bubble behavior. These experiments will serve as a verification/ validation tool for the thermal-hydraulic model. Utilization of the direct electron beam for irradiation allows homogeneous heating of a large solution volume and simplifies observation of the bubble dynamics simultaneously with thermal-hydraulic data collection, which will complement data collected during operation of the miniSHINE experiment. Irradiation will be conducted using a 30-40 MeV electron beam from the high-power linac accelerator. The total electron-beam power will be 20 kW, which will yield a power density on the order of 1 kW/L. The solution volume will be cooled on the front and back surfaces and central tube to mimic the geometry of the proposed SHINE solution vessel. Also, multiple thermocouples will be inserted into the solution vessel to map thermal profiles. The experimental design is now complete, and installation and testing are in progress.

  11. Numerical Modeling of a Thermal-Hydraulic Loop and Test Section Design for Heat Transfer Studies in Supercritical Fluids

    NASA Astrophysics Data System (ADS)

    McGuire, Daniel

    A numerical tool for the simulation of the thermal dynamics of pipe networks with heat transfer has been developed with the novel capability of modeling supercritical fluids. The tool was developed to support the design and deployment of two thermal-hydraulic loops at Carleton University for the purpose of heat transfer studies in supercritical and near-critical fluids. First, the system was characterized based on its defining features; the characteristic length of the flow path is orders of magnitude larger than the other characteristic lengths that define the system's geometry; the behaviour of the working fluid in the supercritical thermodynamic state. An analysis of the transient thermal behaviour of the model's domains is then performed to determine the accuracy and range of validity of the modeling approach for simulating the transient thermal behaviour of a thermal-hydraulic loop. Preliminary designs of three test section geometries, for the purpose of heat transfer studies, are presented in support of the overall design of the Carleton supercritical thermal-hydraulic loops. A 7-rod-bundle, annular and tubular geometries are developed with support from the new numerical tool. Materials capable of meeting the experimental requirements while operating in supercritical water are determined. The necessary geometries to satisfy the experimental goals are then developed based on the material characteristics and predicted heat transfer behaviour from previous simulation results. An initial safety analysis is performed on the test section designs, where they are evaluated against the ASME Boiler, Pressure Vessel, and Pressure Piping Code standard, required for safe operation and certification.

  12. Thermal-hydraulic design of the target/blanket for the accelerator production of tritium conceptual design

    SciTech Connect

    Willcutt, G.J.E. Jr.; Kapernick, R.J.

    1997-11-01

    A conceptual design was developed for the target/blanket system of an accelerator-based system to produce tritium. The target/blanket system uses clad tungsten rods for a spallation target and clad lead rods as a neutron multiplier in a blanket surrounding the target. The neutrons produce tritium in {sup 3}He, which is contained in aluminum tubes located in the decoupler and blanket regions. This paper presents the thermal-hydraulic design of the target, decoupler, and blanket developed for the conceptual design of the Accelerator Production of Tritium Project, and demonstrates there is adequate margin in the design at full power operation.

  13. VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 2. User's manual. [PWR; BWR

    SciTech Connect

    Cuta, J.M.; Koontz, A.S.; Stewart, C.W.; Montgomery, S.D.

    1983-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear energy reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 2: User's Manual) describes the input requirements of VIPRE and its auxiliary programs, SPECSET, ASP and DECCON, and lists the input instructions for each code.

  14. Code System to Perform Neutronic and Thermal-Hydraulic Subchannel Analysis from Converged Coarse-Mesh Nodal Solutions.

    2000-06-14

    Version 00 NORMA-FP is an auxiliary program which can perform a neutronic and thermal-hydraulic subchannel analysis, starting from global core calculations carried out by both PSR-471/NORMA or PSR-492/QUARK codes. Detailed flux and power distributions inside homogenized nodes are computed by a two-stage bivariate interpolation method, upon separation of the axial variable for which an analytical solution is adopted. The actual heterogeneous structure of a node is accounted for by fuel rod power factors computed asmore » functions of burnup, burnup-weighted coolant density, and instantaneous coolant density.« less

  15. Thermal hydraulic calculations to support increase in operating power in McClellen Nuclear Radiation Center(MNRC) TRIGA reactor.

    SciTech Connect

    Jensen, R. T.

    1998-05-05

    The RELAP5/Mod3.1 computer program has been used to successfully perform thermal-hydraulic analyses to support the Safety Analysis for increasing the MNRC reactor from 1.0 MW to 2.0 MW. The calculation results show the reactor to have operating margin for both the fuel temperature and critical heat flux limits. The calculated maximum fuel temperature of 705 C is well below the 750 C operating limit. The critical heat flux ratio was calculated to be 2.51.

  16. 3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

    SciTech Connect

    Langenbuch, S.; Velkov, K.; Lizorkin, M.

    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.

  17. Nuclear-coupled thermal-hydraulic stability analysis of boiling water reactors

    NASA Astrophysics Data System (ADS)

    Karve, Atul A.

    We have studied the nuclear-coupled thermal-hydraulic stability of boiling water reactors (BWRs) using a model we developed from: the space-time modal neutron kinetics equations based on spatial omega-modes, the equations for two-phase flow in parallel boiling channels, the fuel rod heat conduction equations, and a simple model for the recirculation loop. The model is represented as a dynamical system comprised of time-dependent nonlinear ordinary differential equations, and it is studied using stability analysis, modern bifurcation theory, and numerical simulations. We first determine the stability boundary (SB) in the most relevant parameter plane, the inlet-subcooling-number/external-pressure-drop plane, for a fixed control rod induced external reactivity equal to the 100% rod line value and then transform the SB to the practical power-flow map. Using this SB, we show that the normal operating point at 100% power is very stable, stability of points on the 100% rod line decreases as the flow rate is reduced, and that points are least stable in the low-flow/high-power region. We also determine the SB when the modal kinetics is replaced by simple point reactor kinetics and show that the first harmonic mode has no significant effect on the SB. Later we carry out the relevant numerical simulations where we first show that the Hopf bifurcation, that occurs as a parameter is varied across the SB is subcritical, and that, in the important low-flow/high-power region, growing oscillations can result following small finite perturbations of stable steady-states on the 100% rod line. Hence, a point on the 100% rod line in the low-flow/high-power region, although stable, may nevertheless be a point at which a BWR should not be operated. Numerical simulations are then done to calculate the decay ratios (DRs) and frequencies of oscillations for various points on the 100% rod line. It is determined that the NRC requirement of DR < 0.75-0.8 is not rigorously satisfied in the low

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

    SciTech Connect

    Lu, S.

    2002-07-01

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

  19. Dynamic thermal-hydraulic modeling and stack flow pattern analysis for all-vanadium redox flow battery

    NASA Astrophysics Data System (ADS)

    Wei, Zhongbao; Zhao, Jiyun; Skyllas-Kazacos, Maria; Xiong, Binyu

    2014-08-01

    The present study focuses on dynamic thermal-hydraulic modeling for the all-vanadium flow battery and investigations on the impact of stack flow patterns on battery performance. The inhomogeneity of flow rate distribution and reversible entropic heat are included in the thermal-hydraulic model. The electrolyte temperature in tanks is modeled with the finite element modeling (FEM) technique considering the possible non-uniform distribution of electrolyte temperature. Results show that the established model predicts electrolyte temperature accurately under various ambient temperatures and current densities. Significant temperature gradients exist in the battery system at extremely low flow rates, while the electrolyte temperature tends to be the same in different components under relatively high flow rates. Three stack flow patterns including flow without distribution channels and two cases of flow with distribution channels are compared to investigate their effects on battery performance. It is found that the flow rates are not uniformly distributed in cells especially when the stack is not well designed, while adding distribution channels alleviates the inhomogeneous phenomenon. By comparing the three flow patterns, it is found that the serpentine-parallel pattern is preferable and effectively controls the uniformity of flow rates, pressure drop and electrolyte temperature all at expected levels.

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

    SciTech Connect

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

    2007-07-01

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

  1. RELAP5-3D code validation for RBMK phenomena

    SciTech Connect

    Fisher, J.E.

    1999-09-01

    The RELAP5-3D thermal-hydraulic code was assessed against Japanese Safety Experiment Loop (SEL) and Heat Transfer Loop (HTL) tests. These tests were chosen because the phenomena present are applicable to analyses of Russian RBMK reactor designs. The assessment cases included parallel channel flow fluctuation tests at reduced and normal water levels, a channel inlet pipe rupture test, and a high power, density wave oscillation test. The results showed that RELAP5-3D has the capability to adequately represent these RBMK-related phenomena.

  2. RELAP5-3D Code Validation for RBMK Phenomena

    SciTech Connect

    Fisher, James Ebberly

    1999-09-01

    The RELAP5-3D thermal-hydraulic code was assessed against Japanese Safety Experiment Loop (SEL) and Heat Transfer Loop (HTL) tests. These tests were chosen because the phenomena present are applicable to analyses of Russian RBMK reactor designs. The assessment cases included parallel channel flow fluctuation tests at reduced and normal water levels, a channel inlet pipe rupture test, and a high power, density wave oscillation test. The results showed that RELAP5-3D has the capability to adequately represent these RBMK-related phenomena.

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

    NASA Astrophysics Data System (ADS)

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

    2016-02-01

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

  4. Scaling approach and thermal-hydraulic analysis in the reactor cavity cooling system of a high temperature gas -cooled reactor and thermal-jet mixing in a sodium fast reactor

    NASA Astrophysics Data System (ADS)

    Omotowa, Olumuyiwa A.

    This dissertation develops and demonstrates the application of the top-down and bottom-up scaling methodologies to thermal-hydraulic flows in the reactor cavity cooling system (RCCS) of the high temperature gas reactor (HTGR) and upper plenum of the sodium fast reactor (SFR), respectively. The need to integrate scaled separate effects and integral tests was identified. Experimental studies and computational tools (CFD) have been integrated to guide the engineering design, analysis and assessment of this scaling methods under single and two-phase flow conditions. To test this methods, two applicable case studies are considered, and original contributions are noted. Case 1: "Experimental Study of RCCS for the HTGR". Contributions include validation of scaling analysis using the top-down approach as guide to a ¼-scale integral test facility. System code, RELAP5, was developed based on the derived scaling parameters. Tests performed included system sensitivity to decay heat load and heat sink inventory variations. System behavior under steady-state and transient scenarios were predicted. Results show that the system has the capacity to protect the cavity walls from over-heating during normal operations and provide a means for decay heat removal under accident scenarios. A full width half maximum statistical method was devised to characterize the thermal-hydraulics of the non-linear two-phase oscillatory behavior. This facilitated understanding of the thermal hydraulic coupling of the loop segments of the RCCS, the heat transfer, and the two-phase flashing flow phenomena; thus the impact of scaling overall. Case 2: "Computational Studies of Thermal Jet Mixing in SFR". In the pool-type SFR, susceptible regions to thermal striping are the upper instrumentation structure and the intermediate heat exchanger (IHX). We investigated the thermal mixing above the core to UIS and the potential impact due to poor mixing. The thermal mixing of dual-jet flows at different

  5. VIPRE (Versatile Internals and Component Program for Reactors; EPRI)-01: A thermal-hydraulic code for reactor cores: Volume 4, Applications: Final report

    SciTech Connect

    Cuta, J.M.; Stewart, C.W.; Koontz, A.S.; Montgomery, S.D.

    1987-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 4: Applications) contains extensive comparisons of VIPRE calculations to experimental data. There are also sensitivity studies and evaluations of code numerical and computational performance. In addition, calculations performed by member utilities using VIPRE for comparisons with transient CHF data, and FSAR plant analyses are presented. Comparisons are also presented of plant thermal-hydraulic calculations with VIPRE and other COBRA codes. These calculations demonstrate the suitability of VIPRE for PWR core thermal-hydraulic analysis.

  6. MASFLO: a computer code to calculate mass flow rates in the Thermal-Hydraulic Test Facility (THTF). Technical report

    SciTech Connect

    White, M.D.

    1980-05-01

    This report documents a modular data interpretation computer code. The MASFLO code is a Fortran code used in the Oak Ridge National Laboratory Blowdown Heat Transfer Program to convert measured quantities of density, volumetric flow, and momentum flux into a calculated quantity: mass flow rate. The code performs both homogeneous and two-velocity calculations. The homogeneous models incorporate various combinations of the Thermal-Hydraulic Test Facility instrumented spool piece turbine flow meter, gamma densitometer, and drag disk readings. The two-velocity calculations also incorporate these instruments, but in models developed by Aya, Rouhani, and Popper. Each subroutine is described briefly, and input instructions are provided in the appendix along with a sample of the code output.

  7. COBRA-SFS (Spent Fuel Storage): A thermal-hydraulic analysis computer code: Volume 2, User's manual

    SciTech Connect

    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.

  8. RELAP5-3D thermal hydraulic analysis of the target cooling system in the SPES experimental facility

    NASA Astrophysics Data System (ADS)

    Giardina, M.; Castiglia, F.; Buffa, P.; Palermo, G.; Prete, G.

    2014-11-01

    The SPES (Selective Production of Exotic Species) experimental facility, under construction at the Italian National Institute of Nuclear Physics (INFN) Laboratories of Legnaro, Italy, is a second generation Isotope Separation On Line (ISOL) plant for advanced nuclear physic studies. The UCx target-ion source system works at temperature of about 2273 K, producing a high level of radiation (105 Sv/h), for this reason a careful risk analysis for the target chamber is among the major safety issues. In this paper, the obtained results of thermofluid-dynamics simulations of accidental transients in the SPES target cooling system are reported. The analysis, performed by using the RELAP5-3D 2.4.2 qualified thermal-hydraulic system code, proves good safety performance of this system during different accidental conditions.

  9. Some computational challenges of developing efficient parallel algorithms for data-dependent computations in thermal-hydraulics supercomputer applications

    SciTech Connect

    Woodruff, S.B.

    1992-05-01

    The Transient Reactor Analysis Code (TRAC), which features a two- fluid treatment of thermal-hydraulics, is designed to model transients in water reactors and related facilities. One of the major computational costs associated with TRAC and similar codes is calculating constitutive coefficients. Although the formulations for these coefficients are local the costs are flow-regime- or data-dependent; i.e., the computations needed for a given spatial node often vary widely as a function of time. Consequently, poor load balancing will degrade efficiency on either vector or data parallel architectures when the data are organized according to spatial location. Unfortunately, a general automatic solution to the load-balancing problem associated with data-dependent computations is not yet available for massively parallel architectures. This document discusses why developers algorithms, such as a neural net representation, that do not exhibit algorithms, such as a neural net representation, that do not exhibit load-balancing problems.

  10. The development of a preliminary correlation of data on oxide growth on 6061 aluminum under ANS thermal-hydraulic conditions

    SciTech Connect

    Pawel, R.E.; Yoder, G.L.; West, C.D.; Montgomery, B.H.

    1990-06-01

    The corrosion of aluminum alloy 6061 is being studied in a special test loop facility under the range of thermal-hydraulic conditions appropriate for fuel plate operation in the Advanced Neutron Source (ANS) reactor core. Experimental measurements describing the growth of the boehmite (Al{sub 2}O{sub 3}H{sub 2}O) films on the exposed aluminum surfaces are now available for a range of coolant conditions and heat fluxes, and these results have been analyzed to demonstrate the influence of several important experimental variables. A subset of our data base particularly appropriate to the ANS conditions presently anticipated was used to develop a preliminary correlation based on an empirical oxidation model.

  11. The OECD/NEA/NSC PBMR coupled neutronics/thermal hydraulics transient benchmark: The PBMR-400 core design

    SciTech Connect

    Reitsma, F.; Ivanov, K.; Downar, T.; De Haas, H.; Gougar, H. D.

    2006-07-01

    The Pebble Bed Modular Reactor (PBMR) is a High-Temperature Gas-cooled Reactor (HTGR) concept to be built in South Africa. As part of the verification and validation program the definition and execution of code-to-code benchmark exercises are important. The Nuclear Energy Agency (NEA) of the Organisation for Economic Cooperation and Development (OECD) has accepted, through the Nuclear Science Committee (NSC), the inclusion of the Pebble-Bed Modular Reactor (PBMR) coupled neutronics/thermal hydraulics transient benchmark problem in its program. The OECD benchmark defines steady-state and transients cases, including reactivity insertion transients. It makes use of a common set of cross sections (to eliminate uncertainties between different codes) and includes specific simplifications to the design to limit the need for participants to introduce approximations in their models. In this paper the detailed specification is explained, including the test cases to be calculated and the results required from participants. (authors)

  12. Feasibility study Part I - Thermal hydraulic analysis of LEU target for {sup 99}Mo production in Tajoura reactor

    SciTech Connect

    Bsebsu, F.M.; Abotweirat, F. E-mail: abutweirat@yahoo.com; Elwaer, S.

    2008-07-15

    The Renewable Energies and Water Desalination Research Center (REWDRC), Libya, will implement the technology for {sup 99}Mo isotope production using LEU foil target, to obtain new revenue streams for the Tajoura nuclear research reactor and desiring to serve the Libyan hospitals by providing the medical radioisotopes. Design information is presented for LEU target with irradiation device and irradiation Beryllium (Be) unit in the Tajoura reactor core. Calculated results for the reactor core with LEU target at different level of power are presented for steady state and several reactivity induced accident situations. This paper will present the steady state thermal hydraulic design and transient analysis of Tajoura reactor was loaded with LEU foil target for {sup 99}Mo production. The results of these calculations show that the reactor with LEU target during the several cases of transient are in safe and no problems will occur. (author)

  13. COBRA-SFS (Spent Fuel Storage): A thermal-hydraulic analysis computer code: Volume 1, Mathematical models and solution method

    SciTech Connect

    Rector, D.R.; Wheeler, C.L.; Lombardo, N.J.

    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 describes the finite-volume equations and the method used to solve these equations. It is directed toward the user who is interested in gaining a more complete understanding of these methods.

  14. LMFBR in-core thermal-hydraulics: the state of the art and US research and development needs

    SciTech Connect

    Khan, E.U.

    1980-04-01

    A detailed critical review is presented of the literature relevant to predicting coolant flow and temperature fields in LMFBR core assemblies for nominal and non-nominal rod bundle geometries and reactor operating conditions. The review covers existing thermal-hydraulic models, computational methods, and experimental data useful for the design of an LMFBR core. The literature search made for this review included publications listed by Nuclear Science Abstracts and Energy Data Base as well as papers presented at key nuclear conferences. Based on this extensive review, the report discusses the accuracy with which the models predict flow and temperature fields in rod assemblies, identifying areas where analytical, experimental, and model development needs exist.

  15. Coupled mechanical electromagnetic thermal hydraulic effects in Nb3Sn cable-in-conduit conductors for ITER

    NASA Astrophysics Data System (ADS)

    Zanino, R.; Ciazynski, D.; Mitchell, N.; Savoldi Richard, L.

    2005-12-01

    The crucial multi-physics problem of how to extrapolate from the performance of an isolated Nb3Sn strand measured in the laboratory to the performance of a superconducting coil using multi-strand twisted cables is addressed here. We consider the particular case of the path going from the LMI strand to the international thermonuclear experimental reactor (ITER) toroidal field model coil (TFMC), through its associated Full Size Joint Sample, the TFMC-FSJS. Mechanical, electromagnetic and thermal-hydraulic conditions are simulated using the ANSYS, ENSIC and Mithrandir/M&M codes, respectively. At least in this case, the DC performance of the short sample turns out to be relatively close to (considering error bars) but not fully representative of that of the coil, showing higher (less compressive) effective thermal strain but also higher sensitivity to the electromechanical load.

  16. Thermal-hydraulic/heat transfer code development for sphere-pac-fueled LMFBRs. [COBRA-3SP code

    SciTech Connect

    Morris, D.G.

    1980-06-01

    Sphere-pac fuel has received much attention recently in light of the development of proliferation-resistant fuel cycles for the Fast Breeder Reactor Program in the United States. However, for sphere-pac fuel to be a viable alternative to conventional pellet fuel, a means to analyze the thermal behavior of sphere-pac-fueled pin bundles is needed. To meet this need, a thermal-hydraulic/heat transfer computer code has been developed for sphere-pac-fueled fast breeder reactors. The code, COBRA-3SP, is a modified version of COBRA-3M incorporating a three-region sphere-pac fuel pin model which permits fuel restructuring. With COBRA-3SP, steady-state and transient analysis of sphere-pac-fueled pin bundles is possible. The validity of the sphere-pac fuel pin model has been verified using experimental results of irradiated sphere-pac fuel.

  17. Paranormal phenomena

    NASA Astrophysics Data System (ADS)

    Gaina, Alex

    1996-08-01

    Critical analysis is given of some paranormal phenomena events (UFO, healers, psychokinesis (telekinesis))reported in Moldova. It is argued that correct analysis of paranormal phenomena should be made in the framework of electromagnetism.

  18. Integral and Separate Effects Tests for Thermal Hydraulics Code Validation for Liquid-Salt Cooled Nuclear Reactors

    SciTech Connect

    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

  19. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    SciTech Connect

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B.

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked & influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs.

  20. A three-dimensional transient neutronics routine for the TRAC-PF1 reactor thermal hydraulic computer code

    SciTech Connect

    Bandini, B.R. Los Alamos National Lab., NM

    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.

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

    SciTech Connect

    Licht, J.; Bergeron, A.; Dionne, B.; Van den Branden, G; Kalcheva, S; Sikik, E; Koonen, E

    2015-01-01

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

  2. An analysis of the proposed MITR-III core to establish thermal-hydraulic limits at 10 MW. Final report

    SciTech Connect

    Harling, O.K.; Lanning, D.D.; Bernard, J.A.; Meyer, J.E.; Henry, A.F.

    1997-06-01

    The 5 MW Massachusetts Institute of Technology Research Reactor (MITR-II) is expected to operate under a new license beginning in 1999. Among the options being considered is an upgrade in the heat removal system to allow operation at 10 MW. The purpose of this study is to predict the Limiting Safety System Settings and Safety Limits for the upgraded reactor (MITR-III). The MITR Multi-Channel Analysis Code was written to analyze the response of the MITR system to a series of anticipated transients in order to determine the Limiting Safety System Settings and Safety Limits under various operating conditions. The MIT Multi-Channel Analysis Code models the primary and secondary systems, with special emphasis placed on analyzing the thermal-hydraulic conditions in the core. The code models each MITR fuel element explicitly in order to predict the behavior of the system during flow instabilities. The results of the code are compared to experimental data from MITR-II and other sources. New definitions are suggested for the Limiting Safety System Settings and Safety Limits. MITR Limit Diagrams are included for three different heat removal system configurations. It is concluded that safe, year-round operating at 10 MW is possible, given that the primary and secondary flow rates are both increased by approximately 40%.

  3. Steady-state thermal-hydraulic analysis of the pellet-bed reactor for nuclear thermal propulsion

    SciTech Connect

    El-Genk, M.S.; Morley, N.J.; Yang, J.Y. )

    1992-01-01

    The pellet-bed reactor (PBR) for nuclear thermal propulsion is a hydrogen-cooled, BeO-reflected, fast reactor, consisting of an annular core region filled with randomly packed, spherical fuel pellets. The fuel pellets in the PBR are self-supported, eliminating the need for internal core structure, which simplifies the core design and reduces the size and mass of the reactor. Each spherical fuel pellet is composed of hundreds of fuel microspheres embedded in a zirconium carbide (ZrC) matrix. Each fuel microsphere is composed of a UC-NbC fuel kernel surrounded by two consecutive layers of the NbC and ZrC. Gaseous hydrogen serves both as core coolant and as the propellant for the PBR rocket engine. The cold hydrogen flows axially down the inlet channel situated between the core and the external BeO reflector and radially through the orifices in the cold frit, the core, and the orifices in the hot frit. Finally, the hot hydrogen flows axially out the central channel and exits through converging-diverging nozzle. A thermal-hydraulic analysis of the PBR core was performed with an emphasis on optimizing the size and axial distribution of the orifices in the hot and cold frits to ensure that hot spots would not develop in the core during full-power operation. Also investigated was the validity of the assumptions of neglecting the axial conduction and axial cross flow in the core.

  4. Scaling of Thermal-Hydraulic Experiments for a Space Rankine Cycle and Selection of a Preconceptual Scaled Experiment Design

    SciTech Connect

    Sulfredge, CD

    2006-01-27

    To assist with the development of a space-based Rankine cycle power system using liquid potassium as the working fluid, a study has been conducted on possible scaled experiments with simulant fluids. This report will consider several possible working fluids and describe a scaling methodology to achieve thermal-hydraulic similarity between an actual potassium system and scaled representations of the Rankine cycle boiler or condenser. The most practical scaling approach examined is based on the selection of perfluorohexane (FC-72) as the simulant. Using the scaling methodology, a series of possible solutions have been calculated for the FC-72 boiler and condenser. The possible scaled systems will then be compared and preconceptual specifications and drawings given for the most promising design. The preconceptual design concept will also include integrating the scaled boiler and scaled condenser into a single experimental loop. All the preconceptual system specifications appear practical from a fabrication and experimental standpoint, but further work will be needed to arrive at a final experiment design.

  5. Review of thermal-hydraulic calculations for Calvert Cliffs and H. B. Robinson PTS study. [Pressurized thermal shock

    SciTech Connect

    Jo, J.H.; Yuelys-Miksis, C.; Rohatgi, U.S.

    1984-01-01

    Thermal-hydraulic transient calculations performed by LANL using the TRAC-PF1 code and by INEL using the RELAP5 code for the USNRC pressurized thermal shock (PTS) study of the Calvert Cliffs and H.B. Robinson Nuclear Power Plants have been reviewed at BNL including the input decks and steady state calculations. Furthermore, six transients for each plant have been selected for the in-depth review. Simple hand calculations based on the mass and energy balances of the entire reactor system, have been performed to predict the temperature and pressure of the reactor system, and the results have been compared with those obtained by the code calculation. In general, the temperatures and pressures of the primary system calculated by the codes have been very reasonable. The secondary pressures calculated by TRAC appear to indicate that the codes have some difficulty with the condensation model and further work is needed to assess the code calculation of the U-tube steam generator pressure when the cold auxiliary feedwater is introduced to the steam generator. However, it is not expected that this uncertainty would affect the transient calculations significantly.

  6. Thermal Hydraulics Design and Analysis Methodology for a Solid-Core Nuclear Thermal Rocket Engine Thrust Chamber

    NASA Technical Reports Server (NTRS)

    Wang, Ten-See; Canabal, Francisco; Chen, Yen-Sen; Cheng, Gary; Ito, Yasushi

    2013-01-01

    Nuclear thermal propulsion is a leading candidate for in-space propulsion for human Mars missions. This chapter describes a thermal hydraulics design and analysis methodology developed at the NASA Marshall Space Flight Center, in support of the nuclear thermal propulsion development effort. The objective of this campaign is to bridge the design methods in the Rover/NERVA era, with a modern computational fluid dynamics and heat transfer methodology, to predict thermal, fluid, and hydrogen environments of a hypothetical solid-core, nuclear thermal engine the Small Engine, designed in the 1960s. The computational methodology is based on an unstructured-grid, pressure-based, all speeds, chemically reacting, computational fluid dynamics and heat transfer platform, while formulations of flow and heat transfer through porous and solid media were implemented to describe those of hydrogen flow channels inside the solid24 core. Design analyses of a single flow element and the entire solid-core thrust chamber of the Small Engine were performed and the results are presented herein

  7. Analysis of the OECD/NRC BWR Turbine Trip Transient Benchmark with the Coupled Thermal-Hydraulics and Neutronics Code TRAC-M/PARCS

    SciTech Connect

    Lee, Deokjung; Downar, Thomas J.; Ulses, Anthony; Akdeniz, Bedirhan; Ivanov, Kostadin N.

    2004-10-15

    An analysis of the Peach Bottom Unit 2 Turbine Trip 2 (TT2) experiment has been performed using the U.S. Nuclear Regulatory Commission coupled thermal-hydraulics and neutronics code TRAC-M/PARCS. The objective of the analysis was to assess the performance of TRAC-M/PARCS on a BWR transient with significance in two-phase flow and spatial variations of the neutron flux. TRAC-M/PARCS results are found to be in good agreement with measured plant data for both steady-state and transient phases of the benchmark. Additional analyses of four fictitious extreme scenarios are performed to provide a basis for code-to-code comparisons and comprehensive testing of the thermal-hydraulics/neutronics coupling. The obtained results of sensitivity studies on the effect of direct moderator heating on transient simulation indicate the importance of this modeling aspect.

  8. Physics and thermal hydraulics design of a small water cooled reactor fuelled with plutonium in rock-like oxide (ROX) form

    SciTech Connect

    Gaultier, M.; Danguy, G.; Perry, A.; Williams, A.; Brushwood, J.; Thompson, A.; Beeley, P. A.

    2006-07-01

    This paper describes the Physics and Thermal Hydraulics areas of a design study for a small water-cooled reactor. The aim was to design a Pressurised Water Reactor (PWR) of maximum power 80 MWt, using a dispersed layout, capable of maximising primary natural circulation flow. The reactor fuel consists of plutonium contained in granular form within a Rock-like Oxide (ROX) pellet structure. (authors)

  9. Statistical Safety Evaluation of BWR Turbine Trip Scenario Using Coupled Neutron Kinetics and Thermal Hydraulics Analysis Code SKETCH-INS/TRACE5.0

    NASA Astrophysics Data System (ADS)

    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.

  10. Investigation of the MTC noise estimation with a coupled neutronic/thermal-hydraulic dedicated model - 'Closing the loop'

    SciTech Connect

    Demaziere, C.; Larsson, V.

    2012-07-01

    This paper investigates the reliability of different noise estimators aimed at determining the Moderator Temperature Coefficient (MTC) of reactivity in Pressurized Water Reactors. By monitoring the inherent fluctuations in the neutron flux and moderator temperature, an on-line monitoring of the MTC without perturbing reactor operation is possible. In order to get an accurate estimation of the MTC by noise analysis, the point-kinetic component of the neutron noise and the core-averaged moderator temperature noise have to be used. Because of the scarcity of the in-core instrumentation, the determination of these quantities is difficult, and several possibilities thus exist for estimating the MTC by noise analysis. Furthermore, the effect of feedback has to be negligible at the frequency chosen for estimating the MTC in order to get a proper determination of the MTC. By using an integrated neutronic/thermal- hydraulic model specifically developed for estimating the three-dimensional distributions of the fluctuations in neutron flux, moderator properties, and fuel temperature, different approaches for estimating the MTC by noise analysis can be tested individually. It is demonstrated that a reliable MTC estimation can only be provided if the core is equipped with a sufficient number of both neutron detectors and temperature sensors, i.e. if the core contain in-core detectors monitoring both the axial and radial distributions of the fluctuations in neutron flux and moderator temperature. It is further proven that the effect of feedback is negligible for frequencies higher than 0.1 Hz, and thus the MTC noise estimations have to be performed at higher frequencies. (authors)

  11. COBRA-SFS (Spent Fuel Storage): A thermal-hydraulic analysis computer code: Volume 3, Validation assessments

    SciTech Connect

    Lombardo, N.J.; Cuta, J.M.; Michener, T.E.; Rector, D.R.; Wheeler, C.L.

    1986-12-01

    This report presents the results of the COBRA-SFS (Spent Fuel Storage) computer code validation effort. COBRA-SFS, while refined and specialized for spent fuel storage system analyses, is a lumped-volume thermal-hydraulic analysis computer code that predicts temperature and velocity distributions in a wide variety of systems. Through comparisons of code predictions with spent fuel storage system test data, the code's mathematical, physical, and mechanistic models are assessed, and empirical relations defined. The six test cases used to validate the code and code models include single-assembly and multiassembly storage systems under a variety of fill media and system orientations and include unconsolidated and consolidated spent fuel. In its entirety, the test matrix investigates the contributions of convection, conduction, and radiation heat transfer in spent fuel storage systems. To demonstrate the code's performance for a wide variety of storage systems and conditions, comparisons of code predictions with data are made for 14 runs from the experimental data base. The cases selected exercise the important code models and code logic pathways and are representative of the types of simulations required for spent fuel storage system design and licensing safety analyses. For each test, a test description, a summary of the COBRA-SFS computational model, assumptions, and correlations employed are presented. For the cases selected, axial and radial temperature profile comparisons of code predictions with test data are provided, and conclusions drawn concerning the code models and the ability to predict the data and data trends. Comparisons of code predictions with test data demonstrate the ability of COBRA-SFS to successfully predict temperature distributions in unconsolidated or consolidated single and multiassembly spent fuel storage systems.

  12. Neutronic and thermal-hydraulic analysis of new irradiation channels inside the Moroccan TRIGA Mark II research reactor core.

    PubMed

    Chham, E; El Bardouni, T; Benaalilou, K; Boukhal, H; El Bakkari, B; Boulaich, Y; El Younoussi, C; Nacir, B

    2016-10-01

    This study was conducted to improve the capacity of radioisotope production in the Moroccan TRIGA Mark II research reactor, which is considered as one of the most important applications of research reactors. The aim of this study is to enhance the utilization of TRIGA core in the field of neutron activation and ensure an economic use of the fuel. The main idea was to create an additional irradiation channel (IC) inside the core. For this purpose, three new core configurations are proposed, which differ according to the IC position in the core. Thermal neutron flux distribution and other neutronic safety parameters such as power peaking factors, excess reactivity, and control rods worth reactivity were calculated using the Monte Carlo N-Particle Transport (MCNP) code and neutron cross-section library based on ENDF/B-VII evaluation. The calculated thermal flux in the central thimble (CT) and in the added IC for the reconfigured core is compared with the thermal flux in the CT of the existing core, which is taken as a reference. The results show that all the obtained fluxes in CTs are very close to the reference value, while a remarkable difference is observed between the fluxes in the new ICs and reference. This difference depends on the position of IC in the reactor core. To demonstrate that the Moroccan TRIGA reactor could safely operate at 2MW, with new configurations based on new ICs, different safety-related thermal-hydraulic parameters were investigated. The PARET model was used in this study to verify whether the safety margins are met despite the new modifications of the core. The results show that it is possible to introduce new ICs safely in the reactor core, because the obtained values of the parameters are largely far from compromising the safety of the reactor. PMID:27552124

  13. ELM - A SIMPLE TOOL FOR THERMAL-HYDRAULIC ANALYSIS OF SOLID-CORE NUCLEAR ROCKET FUEL ELEMENTS

    NASA Technical Reports Server (NTRS)

    Walton, J. T.

    1994-01-01

    ELM is a simple computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in nuclear thermal rockets. Written for the nuclear propulsion project of the Space Exploration Initiative, ELM evaluates the various heat transfer coefficient and friction factor correlations available for turbulent pipe flow with heat addition. In the past, these correlations were found in different reactor analysis codes, but now comparisons are possible within one program. The logic of ELM is based on the one-dimensional conservation of energy in combination with Newton's Law of Cooling to determine the bulk flow temperature and the wall temperature across a control volume. Since the control volume is an incremental length of tube, the corresponding pressure drop is determined by application of the Law of Conservation of Momentum. The size, speed, and accuracy of ELM make it a simple tool for use in fuel element parametric studies. ELM is a machine independent program written in FORTRAN 77. It has been successfully compiled on an IBM PC compatible running MS-DOS using Lahey FORTRAN 77, a DEC VAX series computer running VMS, and a Sun4 series computer running SunOS UNIX. ELM requires 565K of RAM under SunOS 4.1, 360K of RAM under VMS 5.4, and 406K of RAM under MS-DOS. Because this program is machine independent, no executable is provided on the distribution media. The standard distribution medium for ELM is one 5.25 inch 360K MS-DOS format diskette. ELM was developed in 1991. DEC, VAX, and VMS are trademarks of Digital Equipment Corporation. Sun4 and SunOS are trademarks of Sun Microsystems, Inc. IBM PC is a registered trademark of International Business Machines. MS-DOS is a registered trademark of Microsoft Corporation.

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

    SciTech Connect

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

    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.

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

    NASA Astrophysics Data System (ADS)

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

    2014-02-01

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

  16. Colloidal Phenomena.

    ERIC Educational Resources Information Center

    Russel, William B.; And Others

    1979-01-01

    Described is a graduate level engineering course offered at Princeton University in colloidal phenomena stressing the physical and dynamical side of colloid science. The course outline, reading list, and requirements are presented. (BT)

  17. Transport Phenomena.

    ERIC Educational Resources Information Center

    McCready, Mark J.; Leighton, David T.

    1987-01-01

    Discusses the problems created in graduate chemical engineering programs when students enter with a wide diversity of understandings of transport phenomena. Describes a two-semester graduate transport course sequence at the University of Notre Dame which focuses on fluid mechanics and heat and mass transfer. (TW)

  18. Innovative and Advanced Coupled Neutron Transport and Thermal Hydraulic Method (Tool) for the Design, Analysis and Optimization of VHTR/NGNP Prismatic Reactors

    SciTech Connect

    Rahnema, Farzad; Garimeela, Srinivas; Ougouag, Abderrafi; Zhang, Dingkang

    2013-11-29

    This project will develop a 3D, advanced coarse mesh transport method (COMET-Hex) for steady- state and transient analyses in advanced very high-temperature reactors (VHTRs). The project will lead to a coupled neutronics and thermal hydraulic (T/H) core simulation tool with fuel depletion capability. The computational tool will be developed in hexagonal geometry, based solely on transport theory without (spatial) homogenization in complicated 3D geometries. In addition to the hexagonal geometry extension, collaborators will concurrently develop three additional capabilities to increase the code’s versatility as an advanced and robust core simulator for VHTRs. First, the project team will develop and implement a depletion method within the core simulator. Second, the team will develop an elementary (proof-of-concept) 1D time-dependent transport method for efficient transient analyses. The third capability will be a thermal hydraulic method coupled to the neutronics transport module for VHTRs. Current advancements in reactor core design are pushing VHTRs toward greater core and fuel heterogeneity to pursue higher burn-ups, efficiently transmute used fuel, maximize energy production, and improve plant economics and safety. As a result, an accurate and efficient neutron transport, with capabilities to treat heterogeneous burnable poison effects, is highly desirable for predicting VHTR neutronics performance. This research project’s primary objective is to advance the state of the art for reactor analysis.

  19. Development of Design Technology on Thermal-Hydraulic Performance in Tight-Lattice Rod Bundles: II - Rod Bowing Effect on Boiling Transition under Transient Conditions

    NASA Astrophysics Data System (ADS)

    Liu, Wei; Tamai, Hidesada; Kureta, Masatoshi; Ohnuki, Akira; Akimoto, Hajime

    A thermal-hydraulic feasibility project for an Innovative Water Reactor for Flexible fuel cycle (FLWR) has been performed since 2002. In this R&D project, large-scale thermal-hydraulic tests, several model experiments and development of advanced numerical analysis codes have been carried out. In this paper, we describe the critical power characteristics in a 37-rod tight-lattice bundle with rod bowing under transient states. It is observed that transient Boiling Transition (BT) always occurs axially at exit elevation of upper high-heat-flux region and transversely in the central area of the bundle, which is same as that under steady state. For the postulated power increase and flow decrease cases that may be possibly met in a normal operation of the FLWR, it is confirmed that no BT occurs when Initial Critical Power Ratio (ICPR) is 1.3. Moreover, when the transients are run under severer ICPR that causes BT, the transient critical powers are generally same as the steady ones. The experiments are analyzed with a modified TRAC-BFI code, where Japan Atomic Energy Agency (JAEA) newest critical power correlation is implemented for the BT judgement. The code shows good prediction for the occurrence or the non occurrence of the BT and predicts the BT starting time conservatively. Traditional quasi-steady state prediction of the transient BT is confirmed being applicable for the postulated abnormal transient processes in the tight-lattice bundle with rod bowing.

  20. Analysis of Thermal-Hydraulic Gravity/ Buoyancy Effects in the Testing of the ITER Poloidal Field Full Size Joint Sample (PF-FSJS)

    SciTech Connect

    Zanino, R.; Savoldi Richard, L.; Bruzzone, P.; Ciazynski, D.; Nicollet, S.

    2004-06-23

    The PF-FSJS is a full-size joint sample, based on the NbTi dual-channel cable-in-conduit conductor (CICC) design currently foreseen for the International Thermonuclear Experimental Reactor (ITER) Poloidal Field coil system. It was tested during the summer of 2002 in the Sultan facility of CRPP at a background peak magnetic field of typically 6 T. It includes about 3 m of two jointed conductor sections, using different strands but with identical layout. The sample was cooled by supercritical helium at nominal 4.5-5.0 K and 0.9-1.0 MPa, in forced convection from the top to the bottom of the vertical configuration. A pulsed coil was used to test AC losses in the two legs resulting, above a certain input power threshold, in bundle helium backflow from the heated region. Here we study the thermal-hydraulics of the phenomenon with the M and M code, with particular emphasis on the effects of buoyancy on the helium dynamics, as well as on the thermal-hydraulic coupling between the wrapped bundles of strands in the annular cable region and the central cooling channel. Both issues are ITER relevant, as they affect the more general question of the heat removal capability of the helium in this type of conductors.

  1. Benchmarking of thermal hydraulic loop models for Lead-Alloy Cooled Advanced Nuclear Energy System (LACANES), phase-I: Isothermal steady state forced convection

    NASA Astrophysics Data System (ADS)

    Cho, Jae Hyun; Batta, A.; Casamassima, V.; Cheng, X.; Choi, Yong Joon; Hwang, Il Soon; Lim, Jun; Meloni, P.; Nitti, F. S.; Dedul, V.; Kuznetsov, V.; Komlev, O.; Jaeger, W.; Sedov, A.; Kim, Ji Hak; Puspitarini, D.

    2011-08-01

    As highly promising coolant for new generation nuclear reactors, liquid Lead-Bismuth Eutectic has been extensively worldwide investigated. With high expectation about this advanced coolant, a multi-national systematic study on LBE was proposed in 2007, which covers benchmarking of thermal hydraulic prediction models for Lead-Alloy Cooled Advanced Nuclear Energy System (LACANES). This international collaboration has been organized by OECD/NEA, and nine organizations - ENEA, ERSE, GIDROPRESS, IAEA, IPPE, KIT/IKET, KIT/INR, NUTRECK, and RRC KI - contribute their efforts to LACANES benchmarking. To produce experimental data for LACANES benchmarking, thermal-hydraulic tests were conducted by using a 12-m tall LBE integral test facility, named as Heavy Eutectic liquid metal loop for integral test of Operability and Safety of PEACER (HELIOS) which has been constructed in 2005 at the Seoul National University in the Republic of Korea. LACANES benchmark campaigns consist of a forced convection (phase-I) and a natural circulation (phase-II). In the forced convection case, the predictions of pressure losses based on handbook correlations and that obtained by Computational Fluid Dynamics code simulation were compared with the measured data for various components of the HELIOS test facility. Based on comparative analyses of the predictions and the measured data, recommendations for the prediction methods of a pressure loss in LACANES were obtained. In this paper, results for the forced convection case (phase-I) of LACANES benchmarking are described.

  2. Thermal-Hydraulic Analysis of the 3-MW TRIGA MARK-II Research Reactor Under Steady-State and Transient Conditions

    SciTech Connect

    Huda, M.Q.; Bhuiyan, S.I.; Chakrobortty, T.K.; Sarker, M.M.; Mondal, M.A.W

    2001-07-15

    Important thermal-hydraulic parameters of the 3-MW TRIGA MARK-II research reactor operating under both steady-state and transient conditions are reported. Neutronic analyses were performed by using the CITATION diffusion code and the MCNP4B2 Monte Carlo code. The output of CITATION and MCNP4B2 were input to the PARET thermal-hydraulic code to study the steady-state and transient thermal-hydraulic behavior of the reactor. To benchmark the PARET model, data were obtained from different measurements performed by thermocouples in the instrumented fuel (IF) rod during the steady-state operation both under forced- and natural-convection mode and compared with the calculation. The mass flow rates needed for input to PARET were taken from the Final Safety Analysis Report for a downward forced coolant flow equivalent to 3500 gal/min. For natural convection cooling of the reactor, the mass flow rate was generated using the NCTRIGA code. Peak fuel temperatures measured by the thermocouples in the IF rods at different power levels of the TRIGA core were compared with the values calculated by PARET. The axial distribution of the temperatures of the fuel centerline, fuel surface, and the cladding surface in the hot channel were calculated for the reactor operating at the full-power level. Fuel surface heat flux and heat transfer coefficients for the hot channel were also calculated for the reactor operating at the full-power level. The investigated results were found to be in good agreement with the experimental and operational values. The testing of the PARET model calculations through benchmarking the available TRIGA experimental and operational data for pulse-mode operations showed that PARET can successfully be used to analyze the transient behavior of the reactor. Major transient parameters, such as peak power and prompt energy released after pulse, full-width at half-maximum of pulse peak, and maximum fuel centerline temperatures for different fuel elements at different

  3. TWO-PHASE FLOW STUDIES IN NUCLEAR POWER PLANT PRIMARY CIRCUITS USING THE THREE-DIMENSIONAL THERMAL-HYDRAULIC CODE BAGIRA.

    SciTech Connect

    KOHURT, P. , KALINICHENKO, S.D.; KROSHILIN, A.E.; KROSHILIN, V.E.; SMIRNOV, A.V.

    2006-06-04

    In this paper we present recent results of the application of the thermal-hydraulic code BAGIRA to the analysis of complex two-phase flows in nuclear power plants primary loops. In particular, we performed benchmark numerical simulation of an integral LOCA experiment performed on a test facility modeling the primary circuit of VVER-1000. In addition, we have also analyzed the flow patterns in the VVER-1000 steam generator vessel for stationary and transient operation regimes. For both of these experiments we have compared the numerical results with measured data. Finally, we demonstrate the capabilities of BAGIRA by modeling a hypothetical severe accident for a VVER-1000 type nuclear reactor. The numerical analysis, which modeled all stages of the hypothetical severe accident up to the complete ablation of the reactor cavity bottom, shows the importance of multi-dimensional flow effects.

  4. Simplified modeling of liquid sodium medium with temperature and velocity gradient using real thermal-hydraulic data. Application to ultrasonic thermometry in sodium fast reactor

    NASA Astrophysics Data System (ADS)

    Massacret, N.; Moysan, J.; Ploix, M. A.; Jeannot, J. P.; Corneloup, G.

    2013-01-01

    In the framework of the French R&D program for the Generation IV reactors and specifically for the sodium cooled fast reactors (SFR), studies are carried out on innovative instrumentation methods in order to improve safety and to simplify the monitoring of fundamental physical parameters during reactor operation. The aim of the present work is to develop an acoustic thermometry method to follow up the sodium temperature at the outlet of subassemblies. The medium is a turbulent flow of liquid sodium at 550 °C with temperature inhomogeneities. To understand the effect of disturbance created by this medium, numerical simulations are proposed. A ray tracing code has been developed with Matlabin order to predict acoustic paths in this medium. This complex medium is accurately described by thermal-hydraulic data which are issued from a simulation of a real experiment in Japan. The analysis of these results allows understanding the effects of medium inhomogeneities on the further thermometric acoustic measurement.

  5. Experimental investigations of thermal-hydraulic processes arising during operation of the passive safety systems used in new projects of nuclear power plants equipped with VVER reactors

    NASA Astrophysics Data System (ADS)

    Morozov, A. V.; Remizov, O. V.; Kalyakin, D. S.

    2014-05-01

    The results obtained from experimental investigations into thermal-hydraulic processes that take place during operation of the passive safety systems used in new-generation reactor plants constructed on the basis of VVER technology are presented. The experiments were carried out on the model rigs available at the Leipunskii Institute for Physics and Power Engineering. The processes through which interaction occurs between the opposite flows of saturated steam and cold water moving in the vertical steam line of the additional system for passively flooding the core from the second-stage hydro accumulators are studied. The specific features pertinent to undeveloped boiling of liquid on a single horizontal tube heated by steam and steam-gas mixture that is typical for of the condensing operating mode of a VVER reactor steam generator are investigated.

  6. Analysis of the OECD Main Steam Line Break Benchmark Problem Using the Refined Core Thermal-Hydraulic Nodalization Feature of the MARS/MASTER Code

    SciTech Connect

    Joo, Han Gyu; Jeong, Jae-Jun; Cho, Byung-Oh; Lee, Won Jae; Zee, Sung Quun

    2003-05-15

    The refined core thermal-hydraulics (T-H) nodalization feature of the MARS/MASTER code is used to generate a high-fidelity solution to the OECD main steam line break benchmark problem and to investigate the effects of core T-H nodalization. The MARS/MASTER coupling scheme is introduced first that enables efficient refined node core T-H calculations via the COBRA-III module. The base solution is generated using a fine T-H nodalization consisting of fuel assembly-sized radial nodes. Sensitivity studies are performed on core T-H nodalization to examine the impacts on core reactivity, power distribution, and transient behavior. The results indicate that the error in the peak local power can be very large (up to 25%) with a coarse T-H nodalization because of the inability to incorporate detailed thermal feedback. A demonstrative departure from nucleate boiling (DNB) calculation shows no occurrence of DNB in this problem.

  7. Incorporation of a Helical Tube Heat Transfer Model in the MARS Thermal Hydraulic Systems Analysis Code for the T/H Analyses of the SMART Reactor

    SciTech Connect

    Young Jin Lee; Bub Dong Chung; Jong Chull Jo; Hho Jung Kim; Un Chul Lee

    2004-07-01

    SMART is a medium sized integral type advanced pressurized water reactor currently under development at KAERI. The steam generators of SMART are designed with helically coiled tubes and these are designed to produce superheated steam. The helical shape of the tubes can induce strong centrifugal effect on the secondary coolant as it flows inside the tubes. The presence of centrifugal effect is expected to enhance the formation of cross-sectional circulation flows within the tubes that will increase the overall heat transfer. Furthermore, the centrifugal effect is expected to enhance the moisture separation and thus make it easier to produce superheated steam. MARS is a best-estimate thermal-hydraulic systems analysis code with multi-phase, multi-dimensional analysis capability. The MARS code was produced by restructuring and merging the RELAP5 and the COBRA-TF codes. However, MARS as well as most other best-estimate systems analysis codes in current use lack the detailed models needed to describe the thermal hydraulics of helically coiled tubes. In this study, the heat transfer characteristics and relevant correlations for both the tube and shell sides of helical tubes have been investigated, and the appropriate models have been incorporated into the MARS code. The newly incorporated helical tube heat transfer package is available to the MARS users via selection of the appropriate option in the input. A performance analysis on the steam generator of SMART under full power operation was carried out using the modified MARS code. The results of the analysis indicate that there is a significant improvement in the code predictability. (authors)

  8. Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs

    SciTech Connect

    Monteleone, S.

    1994-04-01

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors.

  9. Preliminary investigation of the potential for transient vapor release events during in situ vitrification based on thermal- hydraulic modeling

    SciTech Connect

    Roberts, J.S.; Woosley, S.L.; Lessor, D.L.; Strachan, C.

    1992-07-01

    This study investigates a possible cause of molten glass displacements that occurred during two recent in situ vitrification (ISV) tests. The study was conducted for the US Department of Energy by Pacific Northwest Laboratory. It is hypothesized that these glass displacements are caused by large gas bubbles rising up through the ISV melt and bursting at its surface. These bubbles cause the molten surface to upwell and possibly overflow. When the bubbles burst, molten glass is thrown from the melt surface and the volume of gas contained in the bubble is released into the hood. Both of these phenomena are undesirable because the molten soil ejected from the melt is dangerous to operating personnel and can damage equipment. The sudden gas release can cause a temporary pressurization of the hood, allowing potentially contaminated gas to escape to the atmosphere. This study attempts to explain the conditions necessary for formation of large gas bubbles in the melt so that future glass displacements can be avoided.

  10. Thermal - Hydraulic Behavior of Unsaturated Bentonite and Sand-Bentonite Material as Seal for Nuclear Waste Repository: Numerical Simulation of Column Experiments

    NASA Astrophysics Data System (ADS)

    Ballarini, E.; Graupner, B.; Bauer, S.

    2015-12-01

    For deep geological repositories of high-level radioactive waste (HLRW), bentonite and sand bentonite mixtures are investigated as buffer materials to form a a sealing layer. This sealing layer surrounds the canisters and experiences an initial drying due to the heat produced by HLRW and a successive re-saturation with fluid from the host rock. These complex thermal, hydraulic and mechanical processes interact and were investigated in laboratory column experiments using MX-80 clay pellets as well as a mixture of 35% sand and 65% bentonite. The aim of this study is to both understand the individual processes taking place in the buffer materials and to identify the key physical parameters that determine the material behavior under heating and hydrating conditions. For this end, detailed and process-oriented numerical modelling was applied to the experiments, simulating heat transport, multiphase flow and mechanical effects from swelling. For both columns, the same set of parameters was assigned to the experimental set-up (i.e. insulation, heater and hydration system), while the parameters of the buffer material were adapted during model calibration. A good fit between model results and data was achieved for temperature, relative humidity, water intake and swelling pressure, thus explaining the material behavior. The key variables identified by the model are the permeability and relative permeability, the water retention curve and the thermal conductivity of the buffer material. The different hydraulic and thermal behavior of the two buffer materials observed in the laboratory observations was well reproduced by the numerical model.

  11. TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

    SciTech Connect

    Liles, D.R.; Mahaffy, J.H.

    1986-07-01

    The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients.

  12. Simplified modeling of liquid sodium medium with temperature and velocity gradient using real thermal-hydraulic data. Application to ultrasonic thermometry in sodium fast reactor

    SciTech Connect

    Massacret, N.; Jeannot, J. P.

    2013-01-25

    In the framework of the French R and D program for the Generation IV reactors and specifically for the sodium cooled fast reactors (SFR), studies are carried out on innovative instrumentation methods in order to improve safety and to simplify the monitoring of fundamental physical parameters during reactor operation. The aim of the present work is to develop an acoustic thermometry method to follow up the sodium temperature at the outlet of subassemblies. The medium is a turbulent flow of liquid sodium at 550 Degree-Sign C with temperature inhomogeneities. To understand the effect of disturbance created by this medium, numerical simulations are proposed. A ray tracing code has been developed with Matlab Copyright-Sign in order to predict acoustic paths in this medium. This complex medium is accurately described by thermal-hydraulic data which are issued from a simulation of a real experiment in Japan. The analysis of these results allows understanding the effects of medium inhomogeneities on the further thermometric acoustic measurement.

  13. Conversion of Molybdenum-99 production process to low enriched uranium: Neutronic and thermal hydraulic analyses of HEU and LEU target plates for irradiation in Pakistan Research Reactor-1

    NASA Astrophysics Data System (ADS)

    Mushtaq, Ahmad; Iqbal, Masood; Bokhari, Ishtiaq Hussain; Mahmood, Tayyab; Muhammad, Atta

    2012-09-01

    Technetium-99m, the daughter product of Molybdenum-99 is the most widely needed radionuclide for diagnostic studies in Pakistan. Molybdenum-99 Production Facility has been established at PINSTECH. Highly enriched uranium (93% 235U) U/Al alloy targets have been irradiated in Pakistan Research Reactor-1 (PARR-1) for the generation of fission Mo-99, while basic dissolution technique is used for separation of Mo-99 from target matrix activity. In line with the international objective of minimizing and eventually eliminating the use of HEU in civil commerce, national and international efforts have been underway to shift the production of medical isotopes from HEU to LEU (LEU; <20% 235U enrichment) targets. To achieve the equivalent amount of 99Mo with LEU targets, approximately 5 times uranium is needed. LEU aluminum uranium dispersion target has been developed, which may replace existing HEU aluminum/uranium alloy targets for production of 99Mo using basic dissolution technique. Neutronic and thermal hydraulic calculations were performed for safe irradiation of targets in the core of PARR-1.

  14. Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

    SciTech Connect

    Corradin, Michael; Anderson, M.; Muci, M.; Hassan, Yassin; Dominguez, A.; Tokuhiro, Akira; Hamman, K.

    2014-10-15

    This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

  15. Numerical modeling of the thermal-hydraulic behavior of wire-on-tube condensers operating with HFC-134a using homogeneous equilibrium model: evaluation of some void fraction correlations

    NASA Astrophysics Data System (ADS)

    Guzella, Matheus dos Santos; Cabezas-Gómez, Luben; da Silva, José Antônio; Maia, Cristiana Brasil; Hanriot, Sérgio de Morais

    2016-02-01

    This study presents a numerical evaluation of the influence of some void fraction correlations over the thermal-hydraulic behavior of wire-on-tube condensers operating with HFC-134a. The numerical model is based on finite volume method considering the homogeneous equilibrium model. Empirical correlations are applied to provide closure relations. Results show that the choice of void fraction correlation influences the refrigerant charge and pressure drop calculations, while no influences the heat transfer rate.

  16. COBRA-SFS (Spent-Fuel Storage) thermal-hydraulic analyses of the CASTOR-1C and REA 2023 BWR storage casks containing consolidated spent fuel

    SciTech Connect

    Rector, D.R.; Cuta, J.M.; Lombardo, N.J.

    1986-12-01

    Consolidation of spent nuclear fuel rods is being considered as one option for more efficient and compact storage of reactor spent fuel assemblies. In this concept, rods from two disassembled spent fuel assemblies will be consolidated in a space originally intended to store a single unconsolidated assembly. The thermal performance of consolidated fuel rods in dry storage, especially in multiassembly storage systems, is one of the major issues that must be addressed prior to implementation. In this study, Pacific Northwest Laboratory researchers performed thermal-hydraulic analyses for both the REA 2023 cask and the CASTOR-1C cask containing either unconsolidated or consolidated BWR spent fuel assemblies. The objective was to determine the effect of consolidating spent fuel assemblies on the temperature distributions within both types of casks. Two major conclusions resulted from this study. First, a lumping technique (combining rods and flow channels), which reduces the number of computational nodes required to model complex multiassembly geometries, could be used for both unconsolidated and consolidated rods with negligible effect on prediction accuracies. Second, with a relatively high thermal conductivity backfill gas (e.g., helium), the predicted peak fuel rod temperature in a canister of consolidated rods generating the same amount of heat as an unconsolidated assembly is essentially the same as the peak temperature in the unconsolidated assembly. In contrast, with a relatively low thermal conductivity backfill gas (e.g., nitrogen), the opposite is true and the predicted peak temperature in a consolidated canister is significantly higher than in an unconsolidated assembly. Therefore, when rods are consolidated, selection of the backfill gas is important in maintaining peak rod temperatures below allowable values for rods with relatively high decay heat generation rates.

  17. Peach Bottom 2 Turbine Trip Simulation Using TRAC-BF1/COS3D, a Best-Estimate Coupled 3-D Core and Thermal-Hydraulic Code System

    SciTech Connect

    Ui, Atsushi; Miyaji, Takamasa

    2004-10-15

    The best-estimate coupled three-dimensional (3-D) core and thermal-hydraulic code system TRAC-BF1/COS3D has been developed. COS3D, based on a modified one-group neutronic model, is a 3-D core simulator used for licensing analyses and core management of commercial boiling water reactor (BWR) plants in Japan. TRAC-BF1 is a plant simulator based on a two-fluid model. TRAC-BF1/COS3D is a coupled system of both codes, which are connected using a parallel computing tool. This code system was applied to the OECD/NRC BWR Turbine Trip Benchmark. Since the two-group cross-section tables are provided by the benchmark team, COS3D was modified to apply to this specification. Three best-estimate scenarios and four hypothetical scenarios were calculated using this code system. In the best-estimate scenario, the predicted core power with TRAC-BF1/COS3D is slightly underestimated compared with the measured data. The reason seems to be a slight difference in the core boundary conditions, that is, pressure changes and the core inlet flow distribution, because the peak in this analysis is sensitive to them. However, the results of this benchmark analysis show that TRAC-BF1/COS3D gives good precision for the prediction of the actual BWR transient behavior on the whole. Furthermore, the results with the modified one-group model and the two-group model were compared to verify the application of the modified one-group model to this benchmark. This comparison shows that the results of the modified one-group model are appropriate and sufficiently precise.

  18. Thermal-Hydraulic Analyses of Heat Transfer Fluid Requirements and Characteristics for Coupling A Hydrogen Production Plant to a High-Temperature Nuclear Reactor

    SciTech Connect

    C. B. Davis; C. H. Oh; R. B. Barner; D. F. Wilson

    2005-06-01

    evaluations determined which configurations and coolants are the most promising from thermal-hydraulic and efficiency points of view. These evaluations also determined which configurations and options do not appear to be feasible at the current time.

  19. An assessment of the critical heat flux approaches of thermal-hydraulic system analysis codes using bundle data from the Heat Transfer Research Facility

    SciTech Connect

    Min Lee . Dept. of Nuclear Engineering); Lihyih Liao )

    1994-02-01

    Critical heat flux (CHF) bundle data from the Heat Transfer Research Facility of Columbia University are used to check the validity of the CHF approaches used in thermal-hydraulic system analysis codes for light water reactors. The CHF approaches assessed include the Biasi et al. correlation of TRAC, the Groeneveld et al. CHF table lookup approach of RELAP5/MOD3, the CHF table lookup approach of CATHARE, and the CHF approach of RETRAN. Depending on system pressure, RETRAN uses the B and W2, Barnett, and modified Barnett correlations and a linear interpolation scheme to predict CHF. Results show that among these CHF approaches, the Groeneveld et al. approach has the best prediction accuracy and the smallest uncertainty in the estimation of the HTRF bundle data. On the average, the Groeneveld et al. approach overpredicts the uniform axial heat flux distribution by 3.6% and the nonuniform axial heat flux distribution by 0.9%. The performance of the RETRAN approach is comparable with that of the Groenevel et al. Approach for uniform axial heat flux. In general, the accuracy and the uncertainty of all the approaches, except that of CATHARE, are worse under a nonuniform axial heat distribution than under a uniform axial heat distribution. All the CHF approaches assessed have a tendency to overpredict the HTRF bundle data at low pressure, low measured CHF, and high CHF quality. The performance of the Groenevel et al. approach is improved through a CHF table update and modification of the bundle correction factor using the HTRF bundle data.

  20. The IAEA Coordinated Research Program on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis: Description of the Benchmark Test Cases and Phases

    SciTech Connect

    Frederik Reitsma; Gerhard Strydom; Bismark Tyobeka; Kostadin Ivanov

    2012-10-01

    The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The uncertainties in the HTR analysis tools are today typically assessed with sensitivity analysis and then a few important input uncertainties (typically based on a PIRT process) are varied in the analysis to find a spread in the parameter of importance. However, one wish to apply a more fundamental approach to determine the predictive capability and accuracies of coupled neutronics/thermal-hydraulics and depletion simulations used for reactor design and safety assessment. Today there is a broader acceptance of the use of uncertainty analysis even in safety studies and it has been accepted by regulators in some cases to replace the traditional conservative analysis. Finally, there is also a renewed focus in supplying reliable covariance data (nuclear data uncertainties) that can then be used in uncertainty methods. Uncertainty and sensitivity studies are therefore becoming an essential component of any significant effort in data and simulation improvement. In order to address uncertainty in analysis and methods in the HTGR community the IAEA launched a Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modelling early in 2012. The project is built on the experience of the OECD/NEA Light Water Reactor (LWR) Uncertainty Analysis in Best-Estimate Modelling (UAM) benchmark activity, but focuses specifically on the peculiarities of HTGR designs and its simulation requirements. Two benchmark problems were defined with the prismatic type design represented by the MHTGR-350 design from General Atomics (GA) while a 250 MW modular pebble bed design, similar to the INET (China) and indirect-cycle PBMR (South Africa) designs are also included. In the paper more detail on the benchmark cases, the different specific phases and tasks and the latest

  1. CFD Analysis of Core Bypass Phenomena

    SciTech Connect

    Richard W. Johnson; Hiroyuki Sato; Richard R. Schultz

    2010-03-01

    The U.S. Department of Energy is exploring the potential for the VHTR which will be either of a prismatic or a pebble-bed type. One important design consideration for the reactor core of a prismatic VHTR is coolant bypass flow which occurs in the interstitial regions between fuel blocks. Such gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The existence of the gaps induces a flow bias in the fuel blocks and results in unexpected increase of maximum fuel temperature. Traditionally, simplified methods such as flow network calculations employing experimental correlations are used to estimate flow and temperature distributions in the core design. However, the distribution of temperature in the fuel pins and graphite blocks as well as coolant outlet temperatures are strongly coupled with the local heat generation rate within fuel blocks which is not uniformly distributed in the core. Hence, it is crucial to establish mechanistic based methods which can be applied to the reactor core thermal hydraulic design and safety analysis. Computational Fluid Dynamics (CFD) codes, which have a capability of local physics based simulation, are widely used in various industrial fields. This study investigates core bypass flow phenomena with the assistance of commercial CFD codes and establishes a baseline for evaluation methods. A one-twelfth sector of the hexagonal block surface is modeled and extruded down to whole core length of 10.704m. The computational domain is divided vertically with an upper reflector, a fuel section and a lower reflector. Each side of the sector grid can be set as a symmetry boundary

  2. CFD Analysis of Core Bypass Phenomena

    SciTech Connect

    Richard W. Johnson; Hiroyuki Sato; Richard R. Schultz

    2009-11-01

    The U.S. Department of Energy is exploring the potential for the VHTR which will be either of a prismatic or a pebble-bed type. One important design consideration for the reactor core of a prismatic VHTR is coolant bypass flow which occurs in the interstitial regions between fuel blocks. Such gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The existence of the gaps induces a flow bias in the fuel blocks and results in unexpected increase of maximum fuel temperature. Traditionally, simplified methods such as flow network calculations employing experimental correlations are used to estimate flow and temperature distributions in the core design. However, the distribution of temperature in the fuel pins and graphite blocks as well as coolant outlet temperatures are strongly coupled with the local heat generation rate within fuel blocks which is not uniformly distributed in the core. Hence, it is crucial to establish mechanistic based methods which can be applied to the reactor core thermal hydraulic design and safety analysis. Computational Fluid Dynamics (CFD) codes, which have a capability of local physics based simulation, are widely used in various industrial fields. This study investigates core bypass flow phenomena with the assistance of commercial CFD codes and establishes a baseline for evaluation methods. A one-twelfth sector of the hexagonal block surface is modeled and extruded down to whole core length of 10.704m. The computational domain is divided vertically with an upper reflector, a fuel section and a lower reflector. Each side of the one-twelfth grid can be set as a symmetry boundary

  3. Scaling of Quench Front and Entrainment-Related Phenomena

    SciTech Connect

    Aumiller, D. L.; Hourser, R. J.; Holowach, M. J.; Hochreiter, L. E.; Cheung, F-B.

    2002-04-01

    The scaling of thermal hydraulic systems is of great importance in the development of experiments in laboratory-scale test facilities that are used to replicate the response of full-size prototypical designs. One particular phenomenon that is of interest in experimental modeling is the quench front that develops during the reflood phase in a PWR (Pressurized Water Reactor) following a large-break LOCA (Loss of Coolant Accident). The purpose of this study is to develop a scaling methodology such that the prototypical quench front related phenomena can be preserved in a laboratory-scale test facility which may have material, geometrical, fluid, and flow differences as compared to the prototypical case. A mass and energy balance on a Lagrangian quench front control volume along with temporal scaling methods are utilized in developing the quench front scaling groups for a phenomena-specific second-tier scaling analysis. A sample calculation is presented comparing the quench front scaling groups calculated for a prototypical Westinghouse 17 x 17 PWR fuel design and that of the geometry and material configuration used in the FLECHT SEASET series of experiments.

  4. The modeling of core melting and in-vessel corium relocation in the APRIL code

    SciTech Connect

    Kim. S.W.; Podowski, M.Z.; Lahey, R.T.

    1995-09-01

    This paper is concerned with the modeling of severe accident phenomena in boiling water reactors (BWR). New models of core melting and in-vessel corium debris relocation are presented, developed for implementation in the APRIL computer code. The results of model testing and validations are given, including comparisons against available experimental data and parametric/sensitivity studies. Also, the application of these models, as parts of the APRIL code, is presented to simulate accident progression in a typical BWR reactor.

  5. Thermal Hydraulic Computer Code System.

    1999-07-16

    Version 00 RELAP5 was developed to describe the behavior of a light water reactor (LWR) subjected to postulated transients such as loss of coolant from large or small pipe breaks, pump failures, etc. RELAP5 calculates fluid conditions such as velocities, pressures, densities, qualities, temperatures; thermal conditions such as surface temperatures, temperature distributions, heat fluxes; pump conditions; trip conditions; reactor power and reactivity from point reactor kinetics; and control system variables. In addition to reactor applications,more » the program can be applied to transient analysis of other thermal‑hydraulic systems with water as the fluid. This package contains RELAP5/MOD1/029 for CDC computers and RELAP5/MOD1/025 for VAX or IBM mainframe computers.« less

  6. Optimization and Parallelization of the Thermal-Hydraulic Sub-channel Code CTF for High-Fidelity Multi-physics Applications

    SciTech Connect

    Salko, Robert K; Schmidt, Rodney; Avramova, Maria N

    2014-01-01

    assemblies, ~56,000 pins, ~59,000 sub-channels, ~2.8 million thermal-hydraulic (TH) control volumes). Results demonstrate that CTF can now perform full core analysis (not previously possible due to excessively long runtimes and memory requirements) on the order of 20 minutes. This new capability is not only useful to standalone CTF users, but is also being leveraged in support of coupled code multi-physics calculations being done in the CASL program.

  7. Coupled 3D-neutronics / thermal-hydraulics analysis of an unprotected loss-of-flow accident for a 3600 MWth SFR core

    SciTech Connect

    Sun, K.; Chenu, A.; Mikityuk, K.; Krepel, J.; Chawla, R.

    2012-07-01

    The core behaviour of a large (3600 MWth) sodium-cooled fast reactor (SFR) is investigated in this paper with the use of a coupled TRACE/PARCS model. The SFR neutron spectrum is characterized by several performance advantages, but also leads to one dominating neutronics drawback - a positive sodium void reactivity. This implies a positive reactivity effect when sodium coolant is removed from the core. In order to evaluate such feedback in terms of the dynamics, a representative unprotected loss-of-flow (ULOF) transient, i.e. flow run-down without SCRAM in which sodium boiling occurs, is analyzed. Although analysis of a single transient cannot allow general conclusions to be drawn, it does allow better understanding of the underlying physics and can lead to proposals for improving the core response during such an accident. The starting point of this study is the reference core design considered in the framework of the Collaborative Project on the European Sodium Fast Reactor (CP-ESFR). To reduce the void effect, the core has been modified by introducing an upper sodium plenum (along with a boron layer) and by reducing the core height-to-diameter ratio. For the ULOF considered, a sharp increase in core power results in melting of the fuel in the case of the reference core. In the modified core, a large dryout leads to melting of the clad. It seems that, for the hypothetical event considered, fuel failure cannot be avoided with just improvement of the neutronics design; therefore, thermal-hydraulics optimization has been considered. An innovative assembly design is proposed to prevent sodium vapour blocking the fuel channel. This results in preventing a downward propagation of the sodium boiling to the core center, thus limiting it to the upper region. Such a void map introduces a negative coolant density reactivity feedback, which dominates the total reactivity change. As a result, the power level and the fuel temperature are effectively reduced, and a large dryout

  8. Experimental Validation of Stratified Flow Phenomena, Graphite Oxidation, and Mitigation Strategies of Air Ingress Accidents

    SciTech Connect

    Chang Ho Oh; Eung Soo Kim; Hee Cheon No; Nam Zin Cho

    2008-12-01

    The US Department of Energy is performing research and development (R&D) that focuses on key phenomena that are important during challenging scenarios that may occur in the Next Generation Nuclear Plant (NGNP) Program / GEN-IV Very High Temperature Reactor (VHTR). Phenomena identification and ranking studies (PIRT) to date have identified the air ingress event, following on the heels of a VHTR depressurization, as very important (Schultz et al., 2006). Consequently, the development of advanced air ingress-related models and verification and validation (V&V) are very high priority for the NGNP program. Following a loss of coolant and system depressurization, air will enter the core through the break. Air ingress leads to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heat-up of the bottom reflector and the reactor core and will cause the release of fission products eventually. The potential collapse of the bottom reflector because of burn-off and the release of CO lead to serious safety problems. For estimation of the proper safety margin we need experimental data and tools, including accurate multi-dimensional thermal-hydraulic and reactor physics models, a burn-off model, and a fracture model. We also need to develop effective strategies to mitigate the effects of oxidation. The results from this research will provide crucial inputs to the INL NGNP/VHTR Methods R&D project. This project is focused on (a) analytical and experimental study of air ingress caused by density-driven, stratified, countercurrent flow, (b) advanced graphite oxidation experiments, (c) experimental study of burn-off in the bottom reflector, (d) structural tests of the burnt-off bottom reflector, (e) implementation of advanced models developed during the previous tasks into the GAMMA code, (f) full air ingress and oxidation mitigation analyses, (g) development of core neutronic models, (h) coupling of the core neutronic and thermal hydraulic models, and (i

  9. Coupled Phenomena in Chemistry.

    ERIC Educational Resources Information Center

    Matsubara, Akira; Nomura, Kazuo

    1979-01-01

    Various phenomena in chemistry and biology can be understood through Gibbs energy utilization. Some common phenomena in chemistry are explained including neutralization, hydrolysis, oxidation and reaction, simultaneous dissociation equilibrium of two weak acids, and common ion effect on solubility. (Author/SA)

  10. The ITER in-vessel system

    SciTech Connect

    Lousteau, D.C.

    1994-09-01

    The overall programmatic objective, as defined in the ITER Engineering Design Activities (EDA) Agreement, is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. The ITER EDA Phase, due to last until July 1998, will encompass the design of the device and its auxiliary systems and facilities, including the preparation of engineering drawings. The EDA also incorporates validating research and development (R&D) work, including the development and testing of key components. The purpose of this paper is to review the status of the design, as it has been developed so far, emphasizing the design and integration of those components contained within the vacuum vessel of the ITER device. The components included in the in-vessel systems are divertor and first wall; blanket and shield; plasma heating, fueling, and vacuum pumping equipment; and remote handling equipment.

  11. FINAL REPORT on Experimental Validation of Stratified Flow Phenomena, Graphite Oxidation, and Mitigation Strategies of Air Ingress Accidents

    SciTech Connect

    Chang H. Oh; Eung S. Kim; Hee C. NO; Nam Z. Cho

    2011-01-01

    The U.S. Department of Energy is performing research and development that focuses on key phenomena that are important during challenging scenarios that may occur in the Next Generation Nuclear Plant (NGNP)/Generation IV very high temperature reactor (VHTR). Phenomena Identification and Ranking studies to date have identified the air ingress event, following on the heels of a VHTR depressurization, as very important. Consequently, the development of advanced air ingress-related models and verification & validation are of very high priority for the NGNP Project. Following a loss of coolant and system depressurization incident, air ingress will occur through the break, leading to oxidation of the in-core graphite structure and fuel. This study indicates that depending on the location and the size of the pipe break, the air ingress phenomena are different. In an effort to estimate the proper safety margin, experimental data and tools, including accurate multidimensional thermal-hydraulic and reactor physics models, a burn-off model, and a fracture model are required. It will also require effective strategies to mitigate the effects of oxidation, eventually. This 3-year project (FY 2008–FY 2010) is focused on various issues related to the VHTR air-ingress accident, including (a) analytical and experimental study of air ingress caused by density-driven, stratified, countercurrent flow, (b) advanced graphite oxidation experiments, (c) experimental study of burn-off in the core bottom structures, (d) structural tests of the oxidized core bottom structures, (e) implementation of advanced models developed during the previous tasks into the GAMMA code, (f) full air ingress and oxidation mitigation analyses, (g) development of core neutronic models, (h) coupling of the core neutronic and thermal hydraulic models, and (i) verification and validation of the coupled models.

  12. Ion exchange phenomena

    SciTech Connect

    Bourg, I.C.; Sposito, G.

    2011-05-01

    Ion exchange phenomena involve the population of readily exchangeable ions, the subset of adsorbed solutes that balance the intrinsic surface charge and can be readily replaced by major background electrolyte ions (Sposito, 2008). These phenomena have occupied a central place in soil chemistry research since Way (1850) first showed that potassium uptake by soils resulted in the release of an equal quantity of moles of charge of calcium and magnesium. Ion exchange phenomena are now routinely modeled in studies of soil formation (White et al., 2005), soil reclamation (Kopittke et al., 2006), soil fertilitization (Agbenin and Yakubu, 2006), colloidal dispersion/flocculation (Charlet and Tournassat, 2005), the mechanics of argillaceous media (Gajo and Loret, 2007), aquitard pore water chemistry (Tournassat et al., 2008), and groundwater (Timms and Hendry, 2007; McNab et al., 2009) and contaminant hydrology (Chatterjee et al., 2008; van Oploo et al., 2008; Serrano et al., 2009).

  13. Stress pulse phenomena

    SciTech Connect

    McGlaun, M.

    1993-08-01

    This paper is an introductory discussion of stress pulse phenomena in simple solids and fluids. Stress pulse phenomena is a very rich and complex field that has been studied by many scientists and engineers. This paper describes the behavior of stress pulses in idealized materials. Inviscid fluids and simple solids are realistic enough to illustrate the basic behavior of stress pulses. Sections 2 through 8 deal with the behavior of pressure pulses. Pressure is best thought of as the average stress at a point. Section 9 deals with shear stresses which are most important in studying solids.

  14. Imaging of snapping phenomena

    PubMed Central

    Guillin, R; Marchand, A J; Roux, A; Niederberger, E; Duvauferrier, R

    2012-01-01

    Snapping phenomena result from the sudden impingement between anatomical and/or heterotopical structures with subsequent abrupt movement and noise. Snaps are variously perceived by patients, from mild discomfort to significant pain requiring surgical management. Identifying the precise cause of snaps may be challenging when no abnormality is encountered on routinely performed static examinations. In this regard, dynamic imaging techniques have been developed over time, with various degrees of success. This review encompasses the main features of each imaging technique and proposes an overview of the main snapping phenomena in the musculoskeletal system. PMID:22744321

  15. Quantum phenomena in superconductors

    SciTech Connect

    Clarke, J.

    1987-08-01

    This paper contains remarks by the author on aspects of macroscopic quantum phenomena in superconductors. Some topics discussed are: Superconducting low-inductance undulatory galvanometer (SLUGS), charge imbalance, cylindrical dc superconducting quantum interference device (SQUIDS), Geophysics, noise theory, magnetic resonance with SQUIDS, and macroscopic quantum tunneling. 23 refs., 4 figs. (LSP)

  16. Processing and analysis techniques involving in-vessel material generation

    DOEpatents

    Schabron, John F.; Rovani, Jr., Joseph F.

    2011-01-25

    In at least one embodiment, the inventive technology relates to in-vessel generation of a material from a solution of interest as part of a processing and/or analysis operation. Preferred embodiments of the in-vessel material generation (e.g., in-vessel solid material generation) include precipitation; in certain embodiments, analysis and/or processing of the solution of interest may include dissolution of the material, perhaps as part of a successive dissolution protocol using solvents of increasing ability to dissolve. Applications include, but are by no means limited to estimation of a coking onset and solution (e.g., oil) fractionating.

  17. Processing and analysis techniques involving in-vessel material generation

    DOEpatents

    Schabron, John F.; Rovani, Jr., Joseph F.

    2012-09-25

    In at least one embodiment, the inventive technology relates to in-vessel generation of a material from a solution of interest as part of a processing and/or analysis operation. Preferred embodiments of the in-vessel material generation (e.g., in-vessel solid material generation) include precipitation; in certain embodiments, analysis and/or processing of the solution of interest may include dissolution of the material, perhaps as part of a successive dissolution protocol using solvents of increasing ability to dissolve. Applications include, but are by no means limited to estimation of a coking onset and solution (e.g., oil) fractionating.

  18. Membrane Transport Phenomena (MTP)

    NASA Technical Reports Server (NTRS)

    Mason, Larry W.

    1997-01-01

    The third semi-annual period of the MTP project has been involved with performing experiments using the Membrane Transport Apparatus (MTA), development of analysis techniques for the experiment results, analytical modeling of the osmotic transport phenomena, and completion of a DC-9 microgravity flight to test candidate fluid cell geometries. Preparations were also made for the MTP Science Concept Review (SCR), held on 13 June 1997 at Lockheed Martin Astronautics in Denver. These activities are detailed in the report.

  19. Thermal-hydraulic simulation of natural convection decay heat removal in the High Flux Isotope Reactor (HFIR) using RELAP5 and TEMPEST: Part 2, Interpretation and validation of results

    SciTech Connect

    Ruggles, A.E.; Morris, D.G.

    1989-01-01

    The RELAP5/MOD2 code was used to predict the thermal-hydraulic behavior of the HFIR core during decay heat removal through boiling natural circulation. The low system pressure and low mass flux values associated with boiling natural circulation are far from conditions for which RELAP5 is well exercised. Therefore, some simple hand calculations are used herein to establish the physics of the results. The interpretation and validation effort is divided between the time average flow conditions and the time varying flow conditions. The time average flow conditions are evaluated using a lumped parameter model and heat balance. The Martinelli-Nelson correlations are used to model the two-phase pressure drop and void fraction vs flow quality relationship within the core region. Systems of parallel channels are susceptible to both density wave oscillations and pressure drop oscillations. Periodic variations in the mass flux and exit flow quality of individual core channels are predicted by RELAP5. These oscillations are consistent with those observed experimentally and are of the density wave type. The impact of the time varying flow properties on local wall superheat is bounded herein. The conditions necessary for Ledinegg flow excursions are identified. These conditions do not fall within the envelope of decay heat levels relevant to HFIR in boiling natural circulation. 14 refs., 5 figs., 1 tab.

  20. Wolf-Rayet phenomena

    NASA Technical Reports Server (NTRS)

    Conti, P. S.

    1982-01-01

    The properties of stars showing Wolf-Rayet phenomena are outlined along with the direction of future work. Emphasis is placed on the characteristics of W-R spectra. Specifically the following topics are covered: the absolute visual magnitudes; the heterogeneity of WN spectra; the existence of transition type spectra and compositions the mass loss rates; and the existence of very luminous and possibly very massive W-R stars. Also, a brief overview of current understanding of the theoretical aspects of stellar evolution and stellar winds and the various scenarios that have been proposed to understand W-R spectra are included.

  1. MULTISCALE PHENOMENA IN MATERIALS

    SciTech Connect

    A. BISHOP

    2000-09-01

    This project developed and supported a technology base in nonequilibrium phenomena underpinning fundamental issues in condensed matter and materials science, and applied this technology to selected problems. In this way the increasingly sophisticated synthesis and characterization available for classes of complex electronic and structural materials provided a testbed for nonlinear science, while nonlinear and nonequilibrium techniques helped advance our understanding of the scientific principles underlying the control of material microstructure, their evolution, fundamental to macroscopic functionalities. The project focused on overlapping areas of emerging thrusts and programs in the Los Alamos materials community for which nonlinear and nonequilibrium approaches will have decisive roles and where productive teamwork among elements of modeling, simulations, synthesis, characterization and applications could be anticipated--particularly multiscale and nonequilibrium phenomena, and complex matter in and between fields of soft, hard and biomimetic materials. Principal topics were: (i) Complex organic and inorganic electronic materials, including hard, soft and biomimetic materials, self-assembly processes and photophysics; (ii) Microstructure and evolution in multiscale and hierarchical materials, including dynamic fracture and friction, dislocation and large-scale deformation, metastability, and inhomogeneity; and (iii) Equilibrium and nonequilibrium phases and phase transformations, emphasizing competing interactions, frustration, landscapes, glassy and stochastic dynamics, and energy focusing.

  2. Phenomena Associated with EIT Waves

    NASA Technical Reports Server (NTRS)

    Thompson, B. J.; Biesecker, D. A.; Gopalswamy, N.; Fisher, Richard R. (Technical Monitor)

    2002-01-01

    We discuss phenomena associated with 'EIT Wave' transients. These phenomena include coronal mass ejections, flares, EUV/SXR dimmings, chromospheric waves, Moreton waves, solar energetic particle events, energetic electron events, and radio signatures. Although the occurrence of many phenomena correlate with the appearance of EIT waves, it is difficult to infer which associations are causal. The presentation will include a discussion of correlation surveys of these phenomena.

  3. Phenomena Associated With EIT Waves

    NASA Technical Reports Server (NTRS)

    Thompson, B. J.; Biesecker, D. A.; Gopalswamy, N.

    2003-01-01

    We discuss phenomena associated with "EIT Wave" transients. These phenomena include coronal mass ejections, flares, EUV/SXR dimmings, chromospheric waves, Moreton waves, solar energetic particle events, energetic electron events, and radio signatures. Although the occurrence of many phenomena correlate with the appearance of EIT waves, it is difficult to mfer which associations are causal. The presentation will include a discussion of correlation surveys of these phenomena.

  4. Crystallization phenomena in slags

    NASA Astrophysics Data System (ADS)

    Orrling, Carl Folke

    2000-09-01

    The crystallization of the mold slag affects both the heat transfer and the lubrication between the mold and the strand in continuous casting of steel. In order for mold slag design to become an engineering science rather than an empirical exercise, a fundamental understanding of the melting and solidification behavior of a slag must be developed. Thus it is necessary to be able to quantify the phenomena that occur under the thermal conditions that are found in the mold of a continuous caster. The double hot thermocouple technique (DHTT) and the Confocal Laser Scanning Microscope used in this study are two novel techniques for investigating melting and solidification phenomena of transparent slags. Results from these techniques are useful in defining the phenomena that occur when the slag film infiltrates between the mold and the shell of the casting. TTT diagrams were obtained for various slags and indicated that the onset of crystallization is a function of cooling rate and slag chemistry. Crystal morphology was found to be dependent upon the experimental temperature and four different morphologies were classified based upon the degree of melt undercooling. Continuous cooling experiments were carried out to develop CCT diagrams and it was found that the amount and appearance of the crystalline fraction greatly depends on the cooling conditions. The DHTT can also be used to mimic the cooling profile encountered by the slag in the mold of a continuous caster. In this differential cooling mode (DCT), it was found that the details of the cooling rate determine the actual response of the slag to a thermal gradient and small changes can lead to significantly different results. Crystal growth rates were measured and found to be in the range between 0.11 mum/s to 11.73 mum/s depending on temperature and slag chemistry. Alumina particles were found to be effective innoculants in oxide melts reducing the incubation time for the onset of crystallization and also extending

  5. Weld pool phenomena

    SciTech Connect

    David, S.A.; Vitek, J.M.; Zacharia, T.; DebRoy, T.

    1994-09-01

    During welding, the composition, structure and properties of the welded structure are affected by the interaction of the heat source with the metal. The interaction affects the fluid flow, heat transfer and mass transfer in the weld pool, and the solidification behavior of the weld metal. In recent years, there has been a growing recognition of the importance of the weld pool transport processes and the solid state transformation reactions in determining the composition, structure and properties of the welded structure. The relation between the weld pool transport processes and the composition and structure is reviewed. Recent applications of various solidification theories to welding are examined to understand the special problems of weld metal solidification. The discussion is focussed on the important problems and issues related to weld pool transport phenomena and solidification. Resolution of these problems would be an important step towards a science based control of composition, structure and properties of the weld metal.

  6. Wave propagation phenomena

    NASA Astrophysics Data System (ADS)

    Groenenboom, P. H. L.

    The phenomenon of wave propagation is encountered frequently in a variety of engineering disciplines. It has been realized that for a growing number of problems the solution can only be obtained by discretization of the boundary. Advantages of the Boundary Element Method (BEM) over domain-type methods are related to the reduction of the number of space dimensions and of the modelling effort. It is demonstrated how the BEM can be applied to wave propagation phenomena by establishing the fundamental relationships. A numerical solution procedure is also suggested. In connection with a discussion of the retarded potential formulation, it is shown how the wave propagation problem can be cast into a Boundary Integral Formulation (BIF). The wave propagation problem in the BIF can be solved by time-successive evaluation of the boundary integrals. The example of pressure wave propagation following a sodium-water reaction in a Liquid Metal cooled Fast Breeder Reactor steam generator is discussed.

  7. Thermal Wave Phenomena

    NASA Technical Reports Server (NTRS)

    1999-01-01

    This map from the MGS Horizon Sensor Assembly (HORSE) shows middle atmospheric temperatures near the 1 mbar level of Mars between Ls 170 to 175 (approx. July 14 - 23, 1999). Local Mars times between 1:30 and 4:30 AM are included. Infrared radiation measured by the Mars Horizon Sensor Assembly was used to make the map. That device continuously views the 'limb' of Mars in four directions, to help orient the spacecraft instruments to the nadir: straight down.

    The map shows thermal wave phenomena that are caused by the large topographic variety of Mars' surface, as well the latitudinally symmetric behavior expected at this time of year near the equinox.

  8. Development of a multiphysics analysis system for sodium-water reaction phenomena in steam generators of sodium-cooled fast reactors

    NASA Astrophysics Data System (ADS)

    Uchibori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki

    2015-12-01

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

  9. Development of a multiphysics analysis system for sodium-water reaction phenomena in steam generators of sodium-cooled fast reactors

    SciTech Connect

    Uchibori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki

    2015-12-31

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

  10. ON DETECTING TRANSIENT PHENOMENA

    SciTech Connect

    Belanger, G.

    2013-08-10

    Transient phenomena are interesting and potentially highly revealing of details about the processes under observation and study that could otherwise go unnoticed. It is therefore important to maximize the sensitivity of the method used to identify such events. In this article, we present a general procedure based on the use of the likelihood function for identifying transients which is particularly suited for real-time applications because it requires no grouping or pre-processing of the data. The method makes use of all the information that is available in the data throughout the statistical decision-making process, and is suitable for a wide range of applications. Here we consider those most common in astrophysics, which involve searching for transient sources, events or features in images, time series, energy spectra, and power spectra, and demonstrate the use of the method in the case of a weak X-ray flare in a time series and a short-lived quasi-periodic oscillation in a power spectrum. We derive a fit statistic that is ideal for fitting arbitrarily shaped models to a power density distribution, which is of general interest in all applications involving periodogram analysis.

  11. Arcjet Cathode Phenomena

    NASA Technical Reports Server (NTRS)

    Curran, Francis M.; Haag, Thomas W.; Raquet, John F.

    1989-01-01

    Cathode tips made from a number of different materials were tested in a modular arcjet thruster in order to examine cathode phenomena. Periodic disassembly and examination, along with the data collected during testing, indicated that all of the tungsten-based materials behaved similarly despite the fact that in one of these samples the percentage of thorium oxide was doubled and another was 25 percent rhenium. The mass loss rate from a 2 percent thoriated rhenium cathode was found to be an order of magnitude greater than that observed using 2 percent thoriated tungsten. Detailed analysis of one of these cathode tips showed that the molten crater contained pure tungsten to a depth of about 150 microns. Problems with thermal stress cracking were encountered in the testing of a hafnium carbide tip. Post test analysis showed that the active area of the tip had chemically reacted with the propellant. A 100 hour continuous test was run at about 1 kW. Post test analysis revealed no dendrite formation, such as observed in a 30 kW arcjet lifetest, near the cathode crater. The cathodes from both this test and a previously run 1000 hour cycled test displayed nearly identical arc craters. Data and calculations indicate that the mass losses observed in testing can be explained by evaporation.

  12. Arcjet cathode phenomena

    NASA Technical Reports Server (NTRS)

    Curran, Francis M.; Haag, Thomas W.; Raquet, John F.

    1989-01-01

    Cathode tips made from a number of different materials were tested in a modular arcjet thruster in order to examine cathode phenomena. Periodic disassembly and examination, along with the data collected during testing, indicated that all of the tungsten-based materials behaved similarly despite the fact that in one of these samples the percentage of thorium oxide was doubled and another was 25 percent rhenium. The mass loss rate from a 2 percent thoriated rhenium cathode was found to be an order of magnitude greater than that observed using 2 percent thoriated tungsten. Detailed analysis of one of these cathode tips showed that the molten crater contained pure tungsten to a depth of about 150 microns. Problems with thermal stress cracking were encountered in the testing of a hafnium carbide tip. Post test analysis showed that the active area of the tip had chemically reacted with the propellant. A 100 hour continuous test was run at about 1 kW. Post test analysis revealed no dendrite formation, such as observed in a 30 kW arcjet lifetest, near the cathode crater. The cathodes from both this test and a previously run 1000 hour cycled test displayed nearly identical arc craters. Data and calculations indicate that the mass losses observed in testing can be explained by evaporation.

  13. SUMMARY REPORT: IN-VESSEL COMPOSTING OF MUNICIPAL WASTEWATER SLUDGE

    EPA Science Inventory

    This 177-page Technology Transfer Summary Report highlights design and operating considerations for possible incorporation into future in-vessel and other sludge composting systems. It is not meant to single out one design as superior to another. The document also aims to heighte...

  14. THERMAL HYDRAULIC ANALYSIS OF FIRE DIVERTOR

    SciTech Connect

    C.B. bAXI; M.A. ULRICKSON; D.E. DRIMEYER; P. HEITZENROEDER

    2000-10-01

    The Fusion Ignition Research Experiment (FIRE) is being designed as a next step in the US magnetic fusion program. The FIRE tokamak has a major radius of 2 m, a minor radius of 0.525 m, and liquid nitrogen cooled copper coils. The aim is to produce a pulse length of 20 s with a plasma current of 6.6 MA and with alpha dominated heating. The outer divertor and baffle of FIRE are water cooled. The worst thermal condition for the outer divertor and baffle is the baseline D-T operating mode (10 T, 6.6 MA, 20 s) with a plasma exhaust power of 67 MW and a peak heat flux of 20 MW/m{sup 2}. A swirl tape (ST) heat transfer enhancement method is used in the outer divertor cooling channels to increase the heat transfer coefficient and the critical heat flux (CHF). The plasma-facing surface consists of tungsten brush. The finite element (FE) analysis shows that for an inlet water temperature of 30 C, inlet pressure of 1.5 MPa and a flow velocity of 10 m/s, the incident critical heat flux is greater than 30 MW/m{sup 2}. The peak copper temperature is 490 C, peak tungsten temperature is 1560 C, and the pressure drop is less than 0.5 MPa. All these results fulfill the design requirements.

  15. Thermal hydraulic features of the TMI accident

    NASA Astrophysics Data System (ADS)

    Tolman, B.

    1985-10-01

    The Three Mile island (TMI)-2 accident resulted in extensive core damage and recent data confirms that the reactor vessel was challenged from molten core materials. A hypothesized TMI accident scenario is presented that consistently explains the TMI data and is also consistent with research findings from independent severe fuel damage experiments. The TMI data will prove useful in confirming our understanding of severe core damage accidents under realistic reactor systems conditions. This understanding will aid in addressing safety and regulatory issues related to severe core damage accidents in light water reactors.

  16. Relaxation phenomena in disordered systems

    NASA Astrophysics Data System (ADS)

    Sciortino, F.; Tartaglia, P.

    1997-02-01

    In this article we discuss how the assumptions of self-similarity imposed on the distribution of independently relaxing modes, as well as on their amplitude and characteristic times, manifest in the global relaxation phenomena. We also review recent applications of such approach to the description of relaxation phenomena in microemulsions and molecular glasses.

  17. Teaching Optical Phenomena with Tracker

    ERIC Educational Resources Information Center

    Rodrigues, M.; Carvalho, P. Simeão

    2014-01-01

    Since the invention and dissemination of domestic laser pointers, observing optical phenomena is a relatively easy task. Any student can buy a laser and experience at home, in a qualitative way, the reflection, refraction and even diffraction phenomena of light. However, quantitative experiments need instruments of high precision that have a…

  18. Wave phenomena in sunspots

    NASA Astrophysics Data System (ADS)

    Löhner-Böttcher, Johannes

    2016-03-01

    Context: The dynamic atmosphere of the Sun exhibits a wealth of magnetohydrodynamic (MHD) waves. In the presence of strong magnetic fields, most spectacular and powerful waves evolve in the sunspot atmosphere. Allover the sunspot area, continuously propagating waves generate strong oscillations in spectral intensity and velocity. The most prominent and fascinating phenomena are the 'umbral flashes' and 'running penumbral waves' as seen in the sunspot chromosphere. Their nature and relation have been under intense discussion in the last decades. Aims: Waves are suggested to propagate upward along the magnetic field lines of sunspots. An observational study is performed to prove or disprove the field-guided nature and coupling of the prevalent umbral and penumbral waves. Comprehensive spectroscopic observations at high resolution shall provide new insights into the wave characteristics and distribution across the sunspot atmosphere. Methods: Two prime sunspot observations were carried out with the Dunn Solar Telescope at the National Solar Observatory in New Mexico and with the Vacuum Tower Telescope at the Teide Observatory on Tenerife. The two-dimensional spectroscopic observations were performed with the interferometric spectrometers IBIS and TESOS. Multiple spectral lines are scanned co-temporally to sample the dynamics at the photospheric and chromospheric layers. The time series (1 – 2.5 h) taken at high spatial and temporal resolution are analyzed according to their evolution in spectral intensities and Doppler velocities. A wavelet analysis was used to obtain the wave power and dominating wave periods. A reconstruction of the magnetic field inclination based on sunspot oscillations was developed. Results and conclusions: Sunspot oscillations occur continuously in spectral intensity and velocity. The obtained wave characteristics of umbral flashes and running penumbral waves strongly support the scenario of slow-mode magnetoacoustic wave propagation along

  19. DESIGN OF THE ITER IN-VESSEL COILS

    SciTech Connect

    Neumeyer, C; Bryant, L; Chrzanowski, J; Feder, R; Gomez, M; Heitzenroeder, P; Kalish, M; Lipski, A; Mardenfeld, M; Simmons, R; Titus, P; Zatz, I; Daly, E; Martin, A; Nakahira, M; Pillsbury, R; Feng, J; Bohm, T; Sawan, M; Stone, H; Griffiths, I; Schaffer, M

    2010-11-27

    The ITER project is considering the inclusion of two sets of in-vessel coils, one to mitigate the effect of Edge Localized Modes (ELMs) and another to provide vertical stabilization (VS). The in-vessel location (behind the blanket shield modules, mounted to the vacuum vessel inner wall) presents special challenges in terms of nuclear radiation (~3000 MGy) and temperature (100oC vessel during operations, 200oC during bakeout). Mineral insulated conductors are well suited to this environment but are not commercially available in the large cross section required. An R&D program is underway to demonstrate the production of mineral insulated (MgO or Spinel) hollow copper conductor with stainless steel jacketing needed for these coils. A preliminary design based on this conductor technology has been developed and is presented herein.

  20. An Overview Of The ITER In-Vessel Coil Systems

    SciTech Connect

    Heitzenroeder, P J; Chrzanowski, J H; Dahlgren, F; Hawryluk, R J; Loesser, G D; Neumeyer, C; Mansfield, C; Smith, J P; Schaffer, M; Humphreys, D; Cordier, J J; Campbell, D; Johnson, G A; Martin, A; Rebut, P H; Tao, J O; Fogarty, P J; Nelson, B E; Reed, R P

    2009-09-24

    ELM mitigation is of particular importance in ITER in order to prevent rapid erosion or melting of the divertor surface, with the consequent risk of water leaks, increased plasma impurity content and disruptivity. Exploitable "natural" small or no ELM regimes might yet be found which extrapolate to ITER but this cannot be depended upon. Resonant Magnetic Perturbation has been added to pellet pacing as a tool for ITER to mitigate ELMs. Both are required, since neither method is fully developed and much work remains to be done. In addition, in-vessel coils enable vertical stabilization and RWM control. For these reasons, in-vessel coils (IVCs) are being designed for ITER to provide control of Edge Localized Modes (ELMs) in addition to providing control of moderately unstable resistive wall modes (RWMs) and the vertical stability (VS) of the plasma.

  1. Misconceptions of Emergent Semiconductor Phenomena

    NASA Astrophysics Data System (ADS)

    Nelson, Katherine G.

    The semiconductor field of Photovoltaics (PV) has experienced tremendous growth, requiring curricula to consider ways to promote student success. One major barrier to success students may face when learning PV is the development of misconceptions. The purpose of this work was to determine the presence and prevalence of misconceptions students may have for three PV semiconductor phenomena; Diffusion, Drift and Excitation. These phenomena are emergent, a class of phenomena that have certain characteristics. In emergent phenomena, the individual entities in the phenomena interact and aggregate to form a self-organizing pattern that can be observed at a higher level. Learners develop a different type of misconception for these phenomena, an emergent misconception. Participants (N=41) completed a written protocol. The pilot study utilized half of these protocols (n = 20) to determine the presence of both general and emergent misconceptions for the three phenomena. Once the presence of both general and emergent misconceptions was confirmed, all protocols (N=41) were analyzed to determine the presence and prevalence of general and emergent misconceptions, and to note any relationships among these misconceptions (full study). Through written protocol analysis of participants' responses, numerous codes emerged from the data for both general and emergent misconceptions. General and emergent misconceptions were found in 80% and 55% of participants' responses, respectively. General misconceptions indicated limited understandings of chemical bonding, electricity and magnetism, energy, and the nature of science. Participants also described the phenomena using teleological, predictable, and causal traits, indicating participants had misconceptions regarding the emergent aspects of the phenomena. For both general and emergent misconceptions, relationships were observed between similar misconceptions within and across the three phenomena, and differences in misconceptions were

  2. Uncertainty methodology for the strongly coupled physical phenomena associated with annular flow

    SciTech Connect

    Lane, J. W.; Aumiller Jr, D. L.

    2012-07-01

    Best-Estimate plus Uncertainty (BEPU) methods are slowly supplanting the use of deterministic analysis methods for thermal-hydraulic analyses. As the uncertainty methodologies evolve it is expected that, where both experimental techniques allow and data are available, there will be a shift to quantifying the uncertainty in increasingly more fundamental parameters. For example, for annular flow in a three-field analysis environment (vapor, liquid film, droplet), the driving parameters would be: a) film interfacial shear stress, b) droplet drag, c) droplet entrainment rate and d) droplet deposition rate. An improved annular flow modeling package was recently developed and implemented in an in-house version of the COBRA-TF best-estimate subchannel analysis tool (Lane, 2009). Significant improvement was observed in the code-to-data predictions of several steam-water annular flow tests following the implementation of this modeling package; however, to apply this model set in formal BEPU analysis requires uncertainty distributions to be determined. The unique aspect of annular flow, and the topic of the present work, is the strong coupling between the interfacial drag, entrainment and deposition phenomena. Ideally the uncertainty in each phenomenon would be isolated; however, the situation is further complicated by an inability to experimentally isolate and measure the individual rate processes (particularly entrainment rate), which results in available experimental data that are inherently integral in nature. This paper presents a methodology for isolating the individual physical phenomena of interest, to the extent that the currently available experimental data allow, and developing the corresponding uncertainty distributions for annular flow. (authors)

  3. Critical velocity phenomena and the LTP. [Lunar Transient Phenomena

    NASA Technical Reports Server (NTRS)

    Srnka, L. J.

    1977-01-01

    When the relative velocity between magnetized plasma and neutral gas exceeds a critical value, the gas-plasma interaction is dominated by collective phenomena which rapidly excite and ionize the neutrals. The interaction of the solar wind with a large cloud (between 10 to the 24th and 10 to the 28th power neutrals) vented from the moon should be of this type. Line radiation from such an interaction can yield an apparent lunar surface brightness rivaling reflected sunlight levels over small areas, if the kinetic-energy flow density of the gas is sufficiently high. The aberrated solar-wind flow past the moon would enhance the visibility of such interactions near the lunar sunrise terminator, supporting the statistical studies which indicate that the 'Lunar Transient Phenomena' (anomalous optical phenomena on the moon) are significantly correlated with the position of the terminator on the lunar surface.

  4. Teaching optical phenomena with Tracker

    NASA Astrophysics Data System (ADS)

    Rodrigues, M.; Simeão Carvalho, P.

    2014-11-01

    Since the invention and dissemination of domestic laser pointers, observing optical phenomena is a relatively easy task. Any student can buy a laser and experience at home, in a qualitative way, the reflection, refraction and even diffraction phenomena of light. However, quantitative experiments need instruments of high precision that have a relatively complex setup. Fortunately, nowadays it is possible to analyse optical phenomena in a simple and quantitative way using the freeware video analysis software ‘Tracker’. In this paper, we show the advantages of video-based experimental activities for teaching concepts in optics. We intend to show: (a) how easy the study of such phenomena can be, even at home, because only simple materials are needed, and Tracker provides the necessary measuring instruments; and (b) how we can use Tracker to improve students’ understanding of some optical concepts. We give examples using video modelling to study the laws of reflection, Snell’s laws, focal distances in lenses and mirrors, and diffraction phenomena, which we hope will motivate teachers to implement it in their own classes and schools.

  5. Abnormal pressures as hydrodynamic phenomena

    USGS Publications Warehouse

    Neuzil, C.E.

    1995-01-01

    So-called abnormal pressures, subsurface fluid pressures significantly higher or lower than hydrostatic, have excited speculation about their origin since subsurface exploration first encountered them. Two distinct conceptual models for abnormal pressures have gained currency among earth scientists. The static model sees abnormal pressures generally as relict features preserved by a virtual absence of fluid flow over geologic time. The hydrodynamic model instead envisions abnormal pressures as phenomena in which flow usually plays an important role. This paper develops the theoretical framework for abnormal pressures as hydrodynamic phenomena, shows that it explains the manifold occurrences of abnormal pressures, and examines the implications of this approach. -from Author

  6. Undergraduates' understanding of cardiovascular phenomena.

    PubMed

    Michael, Joel A; Wenderoth, Mary Pat; Modell, Harold I; Cliff, William; Horwitz, Barbara; McHale, Philip; Richardson, Daniel; Silverthorn, Dee; Williams, Stephen; Whitescarver, Shirley

    2002-12-01

    Undergraduates students in 12 courses at 8 different institutions were surveyed to determine the prevalence of 13 different misconceptions (conceptual difficulties) about cardiovascular function. The prevalence of these misconceptions ranged from 20 to 81% and, for each misconception, was consistent across the different student populations. We also obtained explanations for the students' answers either as free responses or with follow-up multiple-choice questions. These results suggest that students have a number of underlying conceptual difficulties about cardiovascular phenomena. One possible source of some misconceptions is the students' inability to apply simple general models to specific cardiovascular phenomena. Some implications of these results for teachers of physiology are discussed. PMID:12031940

  7. In-vessel remote maintenance of the Compact Ignition Tokamak

    SciTech Connect

    Tabor, M.A.; Hager, E.R.; Creedon, R.L.; Fisher, M.V.; Atkin, S.D.

    1987-01-01

    The Compact Ignition Tokamak (CIT) is the first deuterium-tritium (D-T) fusion device that will study the physics of an ignited plasma. The ability of the tokamak vacuum vessel to be maintained remotely while under vacuum has not been fully demonstrated on previous machines, and this ability will be critical to the efficient and safe operation of ignition devices. Although manned entry into the CIT vacuum vessel will be possible during the nonactivated stages of operation, remotely automated equipment will be used to assist in initial assembly of the vessel as well as to maintain all in-vessel components once the D-T burn is achieved. Remote maintenance and operation will be routinely required for replacement of thermal protection tiles, inspection of components, leak detection, and repair welding activities. Conceptual design to support these remote maintenance activities has been integrated with the conceptual design of the in-vessel components to provide a complete and practical remote maintenance system for CIT. The primary remote assembly and maintenance operations on CIT will be accomplished through two dedicated 37- x 100-cm ports on the main toroidal vessel. Each port contains a single articulated boom manipulator (ABM), which is capable of accessing half of the torus. The proposed ABM consists of a movable carriage assembly, telescoping two-part mast, and articulated link sections. 1 ref.

  8. Visualizing Chemical Phenomena in Microdroplets

    ERIC Educational Resources Information Center

    Lee, Sunghee; Wiener, Joseph

    2011-01-01

    Phenomena that occur in microdroplets are described to the undergraduate chemistry community. Droplets having a diameter in the micrometer range can have unique and interesting properties, which arise because of their small size and, especially, their high surface area-to-volume ratio. Students are generally unfamiliar with the characteristics of…

  9. Experimental Study of the Effect of Graphite Dispersion on the Heat Transfer Phenomena in a Reactor Cavity Cooling System

    SciTech Connect

    Vaghetto, Rodolfo; Capone, Luigi; Hassan, Yassin A

    2011-05-31

    An experimental activity was performed to observe and study the effects of graphite dispersion and deposition on thermal-hydraulic phenomena in a reactor cavity cooling system (RCCS). The small-scale RCCS experimental facility (16.5 x 16.5 x 30.4 cm) used for this activity represents half of the reactor cavity with an electrically heated vessel. Water flowing through five vertical pipes removes the heat produced in the vessel and releases it into the environment by mixing with cold water in a large tank. The particle image velocimetry technique was used to study the velocity field of the air inside the cavity. A set of 52 thermocouples was installed in the facility to monitor the temperature profiles of the vessel, pipe walls, and air. Ten grams of a fine graphite powder (average particle size 2 m) was injected into the cavity through a spraying nozzle placed at the bottom of the vessel. The temperatures and air velocity field were recorded and compared with the measurements obtained before the graphite dispersion, showing a decrease of the temperature surfaces that was related to an increase in their emissivity. The results contribute to the understanding of RCCS capability in an accident scenario.

  10. Studies on in-vessel debris coolability in ALPHA program

    SciTech Connect

    Maruyama, Yu; Yamano, Norihiro; Moriyama, Kiyofumi

    1997-02-01

    In-vessel debris coolability experiments have been performed in ALPHA Program at JAERI. Aluminum oxide (Al{sub 2}O{sub 3}) produced by a thermite reaction was applied as a debris simulant. Two scoping experiments using approximately 30 kg or 50 kg of Al{sub 2}O{sub 3} were conducted. In addition to post-test observations, temperature histories of the debris simulant and the lower head experimental vessel were evaluated. Rapid temperature reduction observed on the outer surface of the experimental vessel may imply that water penetration into a gap between the solidified debris and the experimental vessel occurred resulting in an effective cooling of once heated vessel wall. Preliminary measurement of a gap width was made with an ultrasonic device. Signals to show the existence of gaps, ranging from 0.7 mm to 1.4 mm, were detected at several locations.

  11. Investigation of Minimum Film boiling Phenomena on Fuel Rods Under Blowdown Cooling Conditions

    SciTech Connect

    Stephen M. Bajorek; Michael Gawron; Timothy Etzel; Lucas Peterson

    2003-06-30

    Blowdon cooling heat transfer is an important process that occurs early in a hypothetical large break loss-of-coolant accident (LOCA) in a pressurized water reactor. During blowdown, the flow through the hot assembly is a post-critical heat flux dispersed droplet flow. The heat transfer mechanisms that occur in blowdown cooling are complex and depend on droplet and heated surface interaction. In a safety analysis, it is of considerable importance to determine the thermal-hydraulic conditions leading to the minimum film boiling temperature, Tmin. A flow boiling rig for measurement of blowdown cooling heat transfer and quench phenomena on a nuclear fuel rod simulator was designed and constructed for operation at up to 12.4 MPa. The test section consisted of a concentric annulus, with a 9.5 mm OD nuclear fuel rod simulator at the center. The rod was contained within a 0.85 mm thick, 19 mm OD 316 stainless steel tube, forming the flow channel. Two types of rods were tested; one type was sheathed with Inconel 600 while the other was clad with Zircaloy-2. Water was injected into the test section at the top of the heated length through an injection header. This header was an annular sign that fit around the fuel rod simulator and within the stainless steel tube. Small spacers aligned the injection header and prevented contract with either the heater rod or the tube. A series of small diameter holes at the bottom of the header caused the formation of droplets that became entrained with the steam flow. The test section design was such that quench would take place on the rod, and not along the channel outer annulus.

  12. Statistical phenomena in particle beams

    SciTech Connect

    Bisognano, J.J.

    1984-09-01

    Particle beams are subject to a variety of apparently distinct statistical phenomena such as intrabeam scattering, stochastic cooling, electron cooling, coherent instabilities, and radiofrequency noise diffusion. In fact, both the physics and mathematical description of these mechanisms are quite similar, with the notion of correlation as a powerful unifying principle. In this presentation we will attempt to provide both a physical and a mathematical basis for understanding the wide range of statistical phenomena that have been discussed. In the course of this study the tools of the trade will be introduced, e.g., the Vlasov and Fokker-Planck equations, noise theory, correlation functions, and beam transfer functions. Although a major concern will be to provide equations for analyzing machine design, the primary goal is to introduce a basic set of physical concepts having a very broad range of applicability.

  13. New phenomena searches at CDF

    SciTech Connect

    Soha, Aron; /UC, Davis

    2006-04-01

    The authors report on recent results from the Collider Detector at Fermilab (CDF) experiment, which is accumulating data from proton-antiproton collisions with {radical}s = 1.96 TeV at Run II of the Fermilab Tevatron. The new phenomena being explored include Higgs, Supersymmetry, and large extra dimensions. They also present the latest results of searches for heavy objects, which would indicate physics beyond the Standard Model.

  14. Mathematical Modeling of Diverse Phenomena

    NASA Technical Reports Server (NTRS)

    Howard, J. C.

    1979-01-01

    Tensor calculus is applied to the formulation of mathematical models of diverse phenomena. Aeronautics, fluid dynamics, and cosmology are among the areas of application. The feasibility of combining tensor methods and computer capability to formulate problems is demonstrated. The techniques described are an attempt to simplify the formulation of mathematical models by reducing the modeling process to a series of routine operations, which can be performed either manually or by computer.

  15. Visualization of solidification front phenomena

    NASA Technical Reports Server (NTRS)

    Workman, Gary L.; Smith, Guy A.

    1993-01-01

    Directional solidification experiments have been utilized throughout the Materials Processing in Space Program to provide an experimental platform which minimizes variables in solidification experiments. Because of the wide-spread use of this experimental technique in space-based research, it has become apparent that a better understanding of all the phenomena occurring during solidification can be better understood if direct visualization of the solidification interface were possible.

  16. Cathodic phenomena in aluminum electrowinning

    NASA Astrophysics Data System (ADS)

    Bouteillon, J.; Poignet, J. C.; Rameau, J. J.

    1993-02-01

    Although aluminum is one of the world's highest production-volume primary metals, it is particularly costly to produce for a variety of factors, not the least of which are the expenses associated with electrolytic reduction. Based on the scale of global aluminum processing, even minor improvements in the electrowinning technology can result in significant savings of resources. Thus, from this perspective, the following reviews recent studies of cathodic phenomena in aluminum electrowinning.

  17. Correlated randomness and switching phenomena

    NASA Astrophysics Data System (ADS)

    Stanley, H. E.; Buldyrev, S. V.; Franzese, G.; Havlin, S.; Mallamace, F.; Kumar, P.; Plerou, V.; Preis, T.

    2010-08-01

    One challenge of biology, medicine, and economics is that the systems treated by these serious scientific disciplines have no perfect metronome in time and no perfect spatial architecture-crystalline or otherwise. Nonetheless, as if by magic, out of nothing but randomness one finds remarkably fine-tuned processes in time and remarkably fine-tuned structures in space. Further, many of these processes and structures have the remarkable feature of “switching” from one behavior to another as if by magic. The past century has, philosophically, been concerned with placing aside the human tendency to see the universe as a fine-tuned machine. Here we will address the challenge of uncovering how, through randomness (albeit, as we shall see, strongly correlated randomness), one can arrive at some of the many spatial and temporal patterns in biology, medicine, and economics and even begin to characterize the switching phenomena that enables a system to pass from one state to another. Inspired by principles developed by A. Nihat Berker and scores of other statistical physicists in recent years, we discuss some applications of correlated randomness to understand switching phenomena in various fields. Specifically, we present evidence from experiments and from computer simulations supporting the hypothesis that water’s anomalies are related to a switching point (which is not unlike the “tipping point” immortalized by Malcolm Gladwell), and that the bubbles in economic phenomena that occur on all scales are not “outliers” (another Gladwell immortalization). Though more speculative, we support the idea of disease as arising from some kind of yet-to-be-understood complex switching phenomenon, by discussing data on selected examples, including heart disease and Alzheimer disease.

  18. Phenomena and Diosignes of Aratous

    NASA Astrophysics Data System (ADS)

    Avgoloupis, S. I.

    2013-01-01

    Aratous (305-240B.C.) was a singular intellectual, writer and poet which engage himself to compose a very interesting astronomical poet, using the "Dactylous sixstage' style, the formal style of the ancient Greek Epic poetry. This astronomic poem of Aratous "Phenomena and Diosignes" became very favorite reading during the Alexandrine, the Romman and the Byzandin eras as well and had received many praises from significant poets and particularly from Hipparchous and from Theonas from Alexandria, an astronomer of 4rth century A.C.(in Greeks)

  19. In-vessel composting at the Hidden Valley Landfill

    SciTech Connect

    Cox, C.

    1998-01-01

    Yard waste composting is a simple and natural process. But left alone, natural decomposition takes years. With commercial composting, on the other hand, the process must be accelerated by workers and equipment. Moreover, it has to be accomplished in a relatively small space, it has to be accessible to trucks and other vehicles, it is subject to quality control standards, and it has to be free or relatively free of objectionable odor. Most importantly, for economic feasibility, it must find an end market. One facility that apparently has met those criteria is Land Recovery, Inc.`s (LRI, Tacoma, Wash.) Hidden Valley Landfill site in Puyallup, Wash., south of Tacoma. LRI is a fully integrated solid waste management company that operates a landfill, intermodal transfer site, and a recycling center. The Purdy facility has surpassed its designed average capacity of 80 tpd and designed peak capacity of 120 tpd, with peaks running as high as 200 tpd. LRI needed to expand, but there was very little room to do so at the Purdy site. LRI`s solution was to start an in-vessel composting operation adjacent to the Hidden Valley Landfill, using 50-cu.yd. modified roll-off containers as the composting enclosure and 20-cu.yd. containers to filter the odorous exhaust from the decomposing materials. The compost facility is a temporary measure until a new, fully enclosed facility is built in about another year.

  20. Uranium Pyrophoricity Phenomena and Prediction

    SciTech Connect

    DUNCAN, D.R.

    2000-04-20

    We have compiled a topical reference on the phenomena, experiences, experiments, and prediction of uranium pyrophoricity for the Hanford Spent Nuclear Fuel Project (SNFP) with specific applications to SNFP process and situations. The purpose of the compilation is to create a reference to integrate and preserve this knowledge. Decades ago, uranium and zirconium fires were commonplace at Atomic Energy Commission facilities, and good documentation of experiences is surprisingly sparse. Today, these phenomena are important to site remediation and analysis of packaging, transportation, and processing of unirradiated metal scrap and spent nuclear fuel. Our document, bearing the same title as this paper, will soon be available in the Hanford document system [Plys, et al., 2000]. This paper explains general content of our topical reference and provides examples useful throughout the DOE complex. Moreover, the methods described here can be applied to analysis of potentially pyrophoric plutonium, metal, or metal hydride compounds provided that kinetic data are available. A key feature of this paper is a set of straightforward equations and values that are immediately applicable to safety analysis.

  1. Nonstationary Phenomena in the Heliosheath

    NASA Astrophysics Data System (ADS)

    Pogorelov, N. V.; Borovikov, S. N.; Ebert, R. W.; Heerikhuisen, J.; Kim, T. K.; Kryukov, I.; Richardson, J. D.; Suess, S. T.; Zank, G. P.

    2012-12-01

    As Voyagers (V1 and V2) are approaching the heliopause (HP), they keep delivering important information about the solar wind (SW) behavior which sometimes appears to be substantially different at V1 and V2 locations. We argue that the observed differences may be attributed to SW variations. In particular, negative values of the radial velocity component derived from V1 observations may be due to the presence of time-dependent magnetic barriers formed due to the slow/fast wind interactions in the vicinity of solar cycle minima. The inner heliosheath is the venue of wave interaction, MHD instabilities, and turbulence. We further investigate these phenomena in the HP vicinity using a new, based on the Ulysses observations, solar cycle model. We show that some puzzling observational data, such as the difference in the heliocentric distances at which V1 and V2 crossed the termination shock, may be attributed to time-dependent effects. We also use other time-dependent sets of observational boundary conditions, e.g., interplanetary scintillation and OMNI data. Phenomena affecting the stability and shape of the HP are also discussed in the context of our time-dependent simulations. The satisfaction of the 2-3 kHz radio emission criteria beyond the HP is analyzed. Numerical results are validated by their comparison with observational data.

  2. Natural phenomena hazards, Hanford Site, Washington

    SciTech Connect

    Conrads, T.J.

    1998-09-29

    This document presents the natural phenomena hazard loads for use in implementing DOE Order 5480.28, Natural Phenomena Hazards Mitigation, and supports development of double-shell tank systems specifications at the Hanford Site in south-central Washington State. The natural phenomena covered are seismic, flood, wind, volcanic ash, lightning, snow, temperature, solar radiation, suspended sediment, and relative humidity.

  3. Interpolating function and Stokes phenomena

    NASA Astrophysics Data System (ADS)

    Honda, Masazumi; Jatkar, Dileep P.

    2015-11-01

    When we have two expansions of physical quantity around two different points in parameter space, we can usually construct a family of functions, which interpolates the both expansions. In this paper we study analytic structures of such interpolating functions and discuss their physical implications. We propose that the analytic structures of the interpolating functions provide information on analytic property and Stokes phenomena of the physical quantity, which we approximate by the interpolating functions. We explicitly check our proposal for partition functions of zero-dimensional φ4 theory and Sine-Gordon model. In the zero dimensional Sine-Gordon model, we compare our result with a recent result from resurgence analysis. We also comment on construction of interpolating function in Borel plane.

  4. Emergent Phenomena at Oxide Interfaces

    SciTech Connect

    Hwang, H.Y.

    2012-02-16

    Transition metal oxides (TMOs) are an ideal arena for the study of electronic correlations because the s-electrons of the transition metal ions are removed and transferred to oxygen ions, and hence the strongly correlated d-electrons determine their physical properties such as electrical transport, magnetism, optical response, thermal conductivity, and superconductivity. These electron correlations prohibit the double occupancy of metal sites and induce a local entanglement of charge, spin, and orbital degrees of freedom. This gives rise to a variety of phenomena, e.g., Mott insulators, various charge/spin/orbital orderings, metal-insulator transitions, multiferroics, and superconductivity. In recent years, there has been a burst of activity to manipulate these phenomena, as well as create new ones, using oxide heterostructures. Most fundamental to understanding the physical properties of TMOs is the concept of symmetry of the order parameter. As Landau recognized, the essence of phase transitions is the change of the symmetry. For example, ferromagnetic ordering breaks the rotational symmetry in spin space, i.e., the ordered phase has lower symmetry than the Hamiltonian of the system. There are three most important symmetries to be considered here. (i) Spatial inversion (I), defined as r {yields} -r. In the case of an insulator, breaking this symmetry can lead to spontaneous electric polarization, i.e. ferroelectricity, or pyroelectricity once the point group belongs to polar group symmetry. (ii) Time-reversal symmetry (T) defined as t {yields} -t. In quantum mechanics, the time-evolution of the wave-function {Psi} is given by the phase factor e{sup -iEt/{h_bar}} with E being the energy, and hence time-reversal basically corresponds to taking the complex conjugate of the wave-function. Also the spin, which is induced by the 'spinning' of the particle, is reversed by time-reversal. Broken T-symmetry is most naturally associated with magnetism, since the spin

  5. Earthquake prediction with electromagnetic phenomena

    NASA Astrophysics Data System (ADS)

    Hayakawa, Masashi

    2016-02-01

    Short-term earthquake (EQ) prediction is defined as prospective prediction with the time scale of about one week, which is considered to be one of the most important and urgent topics for the human beings. If this short-term prediction is realized, casualty will be drastically reduced. Unlike the conventional seismic measurement, we proposed the use of electromagnetic phenomena as precursors to EQs in the prediction, and an extensive amount of progress has been achieved in the field of seismo-electromagnetics during the last two decades. This paper deals with the review on this short-term EQ prediction, including the impossibility myth of EQs prediction by seismometers, the reason why we are interested in electromagnetics, the history of seismo-electromagnetics, the ionospheric perturbation as the most promising candidate of EQ prediction, then the future of EQ predictology from two standpoints of a practical science and a pure science, and finally a brief summary.

  6. Entanglement and boundary critical phenomena

    SciTech Connect

    Zhou Huanqiang; Barthel, Thomas; Schollwoeck, Ulrich; Fjaerestad, John Ove

    2006-11-15

    We investigate boundary critical phenomena from a quantum-information perspective. Bipartite entanglement in the ground state of one-dimensional quantum systems is quantified using the Renyi entropy S{sub {alpha}}, which includes the von Neumann entropy ({alpha}{yields}1) and the single-copy entanglement ({alpha}{yields}{infinity}) as special cases. We identify the contribution of the boundaries to the Renyi entropy, and show that there is an entanglement loss along boundary renormalization group (RG) flows. This property, which is intimately related to the Affleck-Ludwig g theorem, is a consequence of majorization relations between the spectra of the reduced density matrix along the boundary RG flows. We also point out that the bulk contribution to the single-copy entanglement is half of that to the von Neumann entropy, whereas the boundary contribution is the same.

  7. Unidentified phenomena - Unusual plasma behavior?

    NASA Astrophysics Data System (ADS)

    Avakian, S. V.; Kovalenok, V. V.

    1992-06-01

    The paper describes observations of a phenomenon belonging to the UFO category and the possible causes of these events. Special attention is given to an event which occurred during the night of September 19-20, 1974, when a huge 'star' was observed over Pertrozavodsk (Russia), consisting of a bright-white luminous center, emitting beams of light, and a less bright light-blue shell. The star gradually formed a cometlike object with a tail consisting of beams of light and started to descend. It is suggested that this event was related to cosmic disturbances caused by an occurrence of unusually strong solar flares. Other examples are presented that relate unusual phenomena observed in space to the occurrence of strong magnetic turbulence events.

  8. Wetting phenomena on rough substrates

    NASA Astrophysics Data System (ADS)

    Li, Hao; Kardar, Mehran

    1990-10-01

    We consider wetting phenomena in the vicinity of rough substrates. The quenched random geometry of the substrate is assumed to be a self-affine fractal with a roughness exponent of ζS. Asymptotic critical properties on approaching complete and critical wetting transitions are studied by combining the replica method with scaling and renormalization-group arguments. We find new critical behavior, controlled by a zero-temperature fixed point, when ζS exceeds the thermal roughness exponent of the emerging wetting layer. The possibility of an effective dimensional reduction due to randomness is considered. In two dimensions a number of exact results are obtained by using a many-body transfer-matrix technique.

  9. Critical phenomena in magnetic nanowires.

    PubMed

    Kamalakar, M Venkata; Raychaudhuri, A K

    2009-09-01

    In this paper we report the first experimental study of critical phenomena in case of magnetic nanowires of nickel near the ferromagnetic-paramagnetic transition from the electrical transport properties. Nickel nanowire arrays, prepared by potentiostatic electrodeposition of nickel inside pores of nanoporous anodic alumina template were well characterized by X-ray Diffraction, Transmission electron microscopy and Energy dispersive Spectroscopy. Precise electrical resistance measurement of the nanowire arrays of wire diameter 20 nm have been done in the temperature range between 300 K to 700 K. We see a drop in the Curie temperature as observed from the resistivity anomaly. We analyzed the resistance data near the critical region and extracted the critical exponent alpha directly from the resistance. We observed a decrease in the critical part of the resistivity including a decrease in the magnitude of the critical exponent alpha and severe modification in the correction to scaling. PMID:19928208

  10. ORNL rod-bundle heat-transfer test data. Volume 3. Thermal-hydraulic test facility experimental data report for test 3. 06. 6B - transient film boiling in upflow. [PWR

    SciTech Connect

    Mullins, C.B.; Felde, D.K.; Sutton, A.G.; Gould, S.S.; Morris, D.G.; Robinson, J.J.

    1982-05-01

    Reduced instrument responses are presented for Thermal-Hyraulic Test Facility (THTF) Test 3.06.6B. This test was conducted by members of the Oak Ridge National Laboratory Pressurized-Water-Reactor (PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on August 29, 1980. The objective of the program was to investigate heat transfer phenomena believed to occur in PWR's during accidents, including small and large break loss-of-coolant accidents. Test 3.06.6B was conducted to obtain transient film boiling data in rod bundle geometry under reactor accident-type conditions. The primary purpose of this report is to make the reduced instrument responses for THTF Test 3.06.6B available. Included in the report are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers.

  11. PREFACE Integrability and nonlinear phenomena Integrability and nonlinear phenomena

    NASA Astrophysics Data System (ADS)

    Gómez-Ullate, David; Lombardo, Sara; Mañas, Manuel; Mazzocco, Marta; Nijhoff, Frank; Sommacal, Matteo

    2010-10-01

    Back in 1967, Clifford Gardner, John Greene, Martin Kruskal and Robert Miura published a seminal paper in Physical Review Letters which was to become a cornerstone in the theory of integrable systems. In 2006, the authors of this paper received the AMS Steele Prize. In this award the AMS pointed out that `In applications of mathematics, solitons and their descendants (kinks, anti-kinks, instantons, and breathers) have entered and changed such diverse fields as nonlinear optics, plasma physics, and ocean, atmospheric, and planetary sciences. Nonlinearity has undergone a revolution: from a nuisance to be eliminated, to a new tool to be exploited.' From this discovery the modern theory of integrability bloomed, leading scientists to a deep understanding of many nonlinear phenomena which is by no means reachable by perturbation methods or other previous tools from linear theories. Nonlinear phenomena appear everywhere in nature, their description and understanding is therefore of great interest both from the theoretical and applicative point of view. If a nonlinear phenomenon can be represented by an integrable system then we have at our disposal a variety of tools to achieve a better mathematical description of the phenomenon. This special issue is largely dedicated to investigations of nonlinear phenomena which are related to the concept of integrability, either involving integrable systems themselves or because they use techniques from the theory of integrability. The idea of this special issue originated during the 18th edition of the Nonlinear Evolution Equations and Dynamical Systems (NEEDS) workshop, held at Isola Rossa, Sardinia, Italy, 16-23 May 2009 (http://needs-conferences.net/2009/). The issue benefits from the occasion offered by the meeting, in particular by its mini-workshops programme, and contains invited review papers and contributed papers. It is worth pointing out that there was an open call for papers and all contributions were peer reviewed

  12. Electromechanical phenomena in semiconductor nanostructures

    NASA Astrophysics Data System (ADS)

    Lew Yan Voon, L. C.; Willatzen, M.

    2011-02-01

    Electromechanical phenomena in semiconductors are still poorly studied from a fundamental and an applied science perspective, even though significant strides have been made in the last decade or so. Indeed, most current electromechanical devices are based on ferroelectric oxides. Yet, the importance of the effect in certain semiconductors is being increasingly recognized. For instance, the magnitude of the electric field in an AlN/GaN nanostructure can reach 1-10 MV/cm. In fact, the basic functioning of an (0001) AlGaN/GaN high electron mobility transistor is due to the two-dimensional electron gas formed at the material interface by the polarization fields. The goal of this review is to inform the reader of some of the recent developments in the field for nanostructures and to point out still open questions. Examples of recent work that involves the piezoelectric and pyroelectric effects in semiconductors include: the study of the optoelectronic properties of III-nitrides quantum wells and dots, the current controversy regarding the importance of the nonlinear piezoelectric effect, energy harvesting using ZnO nanowires as a piezoelectric nanogenerator, the use of piezoelectric materials in surface acoustic wave devices, and the appropriateness of various models for analyzing electromechanical effects. Piezoelectric materials such as GaN and ZnO are gaining more and more importance for energy-related applications; examples include high-brightness light-emitting diodes for white lighting, high-electron mobility transistors, and nanogenerators. Indeed, it remains to be demonstrated whether these materials could be the ideal multifunctional materials. The solutions to these and other related problems will not only lead to a better understanding of the basic physics of these materials, but will validate new characterization tools, and advance the development of new and better devices. We will restrict ourselves to nanostructures in the current article even though the

  13. EDITORIAL: Quantum phenomena in Nanotechnology Quantum phenomena in Nanotechnology

    NASA Astrophysics Data System (ADS)

    Loss, Daniel

    2009-10-01

    Twenty years ago the Institute of Physics launched the journal Nanotechnology from its publishing house based in the home town of Paul Dirac, a legendary figure in the development of quantum mechanics at the turn of the last century. At the beginning of the 20th century, the adoption of quantum mechanical descriptions of events transformed the existing deterministic world view. But in many ways it also revolutionised the progress of research itself. For the first time since the 17th century when Francis Bacon established inductive reasoning as the means of advancing science from fact to axiom to law, theory was progressing ahead of experiments instead of providing explanations for observations that had already been made. Dirac's postulation of antimatter through purely theoretical investigation before its observation is the archetypal example of theory leading the way for experiment. The progress of nanotechnology and the development of tools and techniques that enabled the investigation of systems at the nanoscale brought with them many fascinating observations of phenomena that could only be explained through quantum mechanics, first theoretically deduced decades previously. At the nanoscale, quantum confinement effects dominate the electrical and optical properties of systems. They also render new opportunities for manipulating the response of systems. For example, a better understanding of these systems has enabled the rapid development of quantum dots with precisely determined properties, which can be exploited in a range of applications from medical imaging and photovoltaic solar cells to quantum computation, a radically new information technology being currently developed in many labs worldwide. As the first ever academic journal in nanotechnology, {\\it Nanotechnology} has been the forum for papers detailing progress of the science through extremely exciting times. In the early years of the journal, the investigation of electron spin led to the formulation

  14. Monitoring of Transient Lunar Phenomena

    NASA Astrophysics Data System (ADS)

    Barker, Timothy; Farber, Ryan; Ahrendts, Gary

    2014-06-01

    Transient Lunar Phenomena (TLP’s) are described as short-lived changes in the brightness of areas on the face of the Moon. TLP research is characterized by the inability to substantiate, reproduce, and verify findings. Our current research includes the analysis of lunar images taken with two Santa Barbara Instrument Group (SBIG) ST8-E CCD cameras mounted on two 0.36m Celestron telescopes. On one telescope, we are using a sodium filter, and on the other an H-alpha filter, imaging approximately one-third of the lunar surface. We are focusing on two regions: Hyginus and Ina. Ina is of particular interest because it shows evidence of recent activity (Schultz, P., Staid, M., Pieters, C. Nature, Volume 444, Issue 7116, pp. 184-186, 2006). A total of over 50,000 images have been obtained over approximately 35 nights and visually analyzed to search for changes. As of March, 2014, no evidence of TLPs has been found. We are currently developing a Matlab program to do image analysis to detect TLPs that might not be apparent by visual inspection alone.

  15. Modeling Defect-Induced Phenomena

    NASA Astrophysics Data System (ADS)

    Kuklja, Maija M.; Rashkeev, Sergey N.

    Elucidation of dissociation mechanisms, energy localization, and transfer phenomena in the course of explosive decomposition of energetic materials (EMs) are central for understanding, controlling, and enhancing the performance of these materials as fuels, propellants, and explosives. Quality of energetic materials is often judged using two main parameters: sensitivity to detonation and its performance. Low sensitivity is desired to make the material relatively stable to external stimuli, i.e., controllable and able of triggering rapid dissociation only when needed and not accidentally. Performance, on the other hand, is to be high to provide larger heat of the explosive reaction. These parameters do not necessarily correlate with each other and depend on many variables such as molecular and crystalline structures, history of samples, the particle size, crystal hardness and orientation, external stimuli, aging, storage conditions, and others. Mechanisms governing performance are fairly well understood whereas mechanisms of sensitivity are poorly known and need to be much more extensively studied. It is widely accepted though that the thermal decomposition reactions of the materials play a significant role in their sensitivity to mechanical stimuli and their explosive properties [1].

  16. Precursor films in wetting phenomena.

    PubMed

    Popescu, M N; Oshanin, G; Dietrich, S; Cazabat, A-M

    2012-06-20

    The spontaneous spreading of non-volatile liquid droplets on solid substrates poses a classic problem in the context of wetting phenomena. It is well known that the spreading of a macroscopic droplet is in many cases accompanied by a thin film of macroscopic lateral extent, the so-called precursor film, which emanates from the three-phase contact line region and spreads ahead of the latter with a much higher speed. Such films have been usually associated with liquid-on-solid systems, but in the last decade similar films have been reported to occur in solid-on-solid systems. While the situations in which the thickness of such films is of mesoscopic size are fairly well understood, an intriguing and yet to be fully understood aspect is the spreading of microscopic, i.e. molecularly thin, films. Here we review the available experimental observations of such films in various liquid-on-solid and solid-on-solid systems, as well as the corresponding theoretical models and studies aimed at understanding their formation and spreading dynamics. Recent developments and perspectives for future research are discussed. PMID:22627067

  17. Bleed Hole Flow Phenomena Studied

    NASA Technical Reports Server (NTRS)

    1997-01-01

    Boundary-layer bleed is an invaluable tool for controlling the airflow in supersonic aircraft engine inlets. Incoming air is decelerated to subsonic speeds prior to entering the compressor via a series of oblique shocks. The low momentum flow in the boundary layer interacts with these shocks, growing in thickness and, under some conditions, leading to flow separation. To remedy this, bleed holes are strategically located to remove mass from the boundary layer, reducing its thickness and helping to maintain uniform flow to the compressor. The bleed requirements for any inlet design are unique and must be validated by extensive wind tunnel testing to optimize performance and efficiency. To accelerate this process and reduce cost, researchers at the NASA Lewis Research Center initiated an experimental program to study the flow phenomena associated with bleed holes. Knowledge of these flow properties will be incorporated into computational fluid dynamics (CFD) models that will aid engine inlet designers in optimizing bleed configurations before any hardware is fabricated. This ongoing investigation is currently examining two hole geometries, 90 and 20 (both with 5-mm diameters), and various flow features.

  18. Review - Axial compressor stall phenomena

    NASA Technical Reports Server (NTRS)

    Greitzer, E. M.

    1980-01-01

    Stall in compressors can be associated with the initiation of several types of fluid dynamic instabilities. These instabilities and the different phenomena, surge and rotating stall, which result from them, are discussed in this paper. Assessment is made of the various methods of predicting the onset of compressor and/or compression system instability, such as empirical correlations, linearized stability analyses, and numerical unsteady flow calculation procedures. Factors which affect the compressor stall point, in particular inlet flow distortion, are reviewed, and the techniques which are used to predict the loss in stall margin due to these factors are described. The influence of rotor casing treatment (grooves) on increasing compressor flow range is examined. Compressor and compression system behavior subsequent to the onset of stall is surveyed, with particular reference to the problem of engine recovery from a stalled condition. The distinction between surge and rotating stall is emphasized because of the very different consequences on recoverability. The structure of the compressor flow field during rotating stall is examined, and the prediction of compressor performance in rotating stall, including stall/unstall hysteresis, is described.

  19. WESF natural phenomena hazards survey

    SciTech Connect

    Wagenblast, G.R., Westinghouse Hanford

    1996-07-01

    A team of engineers conducted a systematic natural hazards phenomena (NPH) survey for the 225-B Waste Encapsulation and Storage Facility (WESF). The survey is an assessment of the existing design documentation to serve as the structural design basis for WESF, and the Interim Safety Basis (ISB). The lateral force resisting systems for the 225-B building structures, and the anchorages for the WESF safety related systems were evaluated. The original seismic and other design analyses were technically reviewed. Engineering judgment assessments were made of the probability of NPH survival, including seismic, for the 225-B structures and WESF safety systems. The method for the survey is based on the experience of the investigating engineers,and documented earthquake experience (expected response) data.The survey uses knowledge on NPH performance and engineering experience to determine the WESF strengths for NPH resistance, and uncover possible weak links. The survey, in general, concludes that the 225-B structures and WESF safety systems are designed and constructed commensurate with the current Hanford Site design criteria.

  20. Electronic phenomena at high pressure

    SciTech Connect

    Drickamer, H.G.

    1981-01-01

    High pressure research is undertaken either to investigate intrinsically high pressure phenomena or in order to get a better understanding of the effect of the chemical environment on properties or processes at one atmosphere. Studies of electronic properties which fall in each area are presented. Many molecules and complexes can assume in the excited state different molecular arrangements and intermolecular forces depending on the medium. Their luminescence emission is then very different in a rigid or a fluid medium. With pressure one can vary the viscosity of the medium by a factor of 10/sup 7/ and thus control the distribution and rate of crossing between the excited state conformations. In rare earth chelates the efficiency of 4f-4f emission of the rare earth is controlled by the feeding from the singlet and triplet levels of the organic ligand. These ligand levels can be strongly shifted by pressure. A study of the effect of pressure on the emission efficiency permits one to understand the effect of ligand chemistry at one atmosphere. At high pressure electronic states can be sufficiently perturbed to provide new ground states. In EDA complexes these new ground states exhibit unusual chemical reactivity and new products.

  1. Intrinsic interfacial phenomena in manganite heterostructures

    NASA Astrophysics Data System (ADS)

    Vaz, C. A. F.; Walker, F. J.; Ahn, C. H.; Ismail-Beigi, S.

    2015-04-01

    We review recent advances in our understanding of interfacial phenomena that emerge when dissimilar materials are brought together at atomically sharp and coherent interfaces. In particular, we focus on phenomena that are intrinsic to the interface and review recent work carried out on perovskite manganites interfaces, a class of complex oxides whose rich electronic properties have proven to be a useful playground for the discovery and prediction of novel phenomena.

  2. Electromagnetic phenomena and hysteresis losses in superconductors

    NASA Astrophysics Data System (ADS)

    Matsushita, T.

    Hysteresis losses in superconductors are caused by irreversible motion of fluxoids. This motion is, in most cases, described by the critical state model. In this article, various electromagnetic phenomena due to flux pinning effects are reviewed and explanations of these phenomena are given using the critical state model. The phenomena which cannot be well described by the present model, such as reversible fluxoid motion and the longitudinal field effect, are also introduced.

  3. Nonepileptic motor phenomena in the neonate

    PubMed Central

    Huntsman, Richard James; Lowry, Noel John; Sankaran, Koravangattu

    2008-01-01

    The newborn infant is prone to clinical motor phenomena that are not epileptic in nature. These include tremors, jitteriness, various forms of myoclonus and brainstem release phenomena. They are frequently misdiagnosed as seizures, resulting in unnecessary investigations and treatment with anticonvulsants, which have potentially harmful side effects. Unfortunately, there is a paucity of literature about many of these phenomena in the newborn, and some of the major textbooks refer to these events as nonepileptic seizures, leading to further confusion for the practitioner. The present paper aims to review these phenomena with special emphasis on differentiating them from epileptic seizures, and offers information on treatment and prognosis wherever possible. PMID:19436521

  4. Observation of Celestial Phenomena in Ancient China

    NASA Astrophysics Data System (ADS)

    Sun, Xiaochun

    Because of the need for calendar-making and portent astrology, the Chinese were diligent and meticulous observers of celestial phenomena. China has maintained the longest continuous historical records of celestial phenomena in the world. Extraordinary or abnormal celestial events were particularly noted because of their astrological significance. The historical records cover various types of celestial phenomena, which include solar and lunar eclipses, sunspots, "guest stars" (novae or supernovae as we understand today), comets and meteors, and all kinds of planetary phenomena. These records provide valuable historical data for astronomical studies today.

  5. Structural materials for ITER in-vessel component design

    NASA Astrophysics Data System (ADS)

    Kalinin, G.; Gauster, W.; Matera, R.; Tavassoli, A.-A. F.; Rowcliffe, A.; Fabritsiev, S.; Kawamura, H.

    1996-10-01

    The materials proposed for ITER in-vessel components have to exhibit adequate performance for the operating lifetime of the reactor or for specified replacement intervals. Estimates show that maximum irradiation dose to be up to 5-7 dpa (for 1 MWa/m 2 in the basic performance phase (BPP)) within a temperature range from 20 to 300°C. Austenitic SS 316LN-ITER Grade was defined as a reference option for the vacuum vessel, blanket, primary wall, pipe lines and divertor body. Conventional technologies and mill products are proposed for blanket, back plate and manifold manufacturing. HIPing is proposed as a reference manufacturing method for the primary wall and blanket and as an option for the divertor body. The existing data show that mechanical properties of HIPed SS are no worse than those of forged 316LN SS. Irradiation will result in property changes. Minimum ductility has been observed after irradiation in an approximate temperature range between 250 and 350°C, for doses of 5-10 dpa. In spite of radiation-induced changes in tensile deformation behavior, the fracture remains ductile. Irradiation assisted corrosion cracking is a concern for high doses of irradiation and at high temperatures. Re-welding is one of the critical issues because of the need to replace failed components. It is also being considered for the replacement of shielding blanket modules by breeding modules after the BPP. Estimates of radiation damage at the locations for re-welding show that the dose will not exceed 0.05 dpa (with He generation of 1 appm) for the manifold and 0.01 dpa (with He generation 0.1 appm) for the back plate for the BPP of ITER operation. Existing experimental data show that these levels will not result in property changes for SS; however, neutron irradiation and He generation promote crack formation in the heat affected zone during welding. Cu based alloys, DS-Cu (Glidcop A125) and PHCu CuCrZr bronze) are proposed as a structural materials for high heat flux

  6. Phenomena resulting from hypergolic contact

    NASA Astrophysics Data System (ADS)

    Forness, Jordan M.

    Understanding hypergolic ignition is critical for the safe and successful operation of hypergolic engines. The complex coupling of physical and chemical processes during hypergolic ignition complicates analysis of the event. Presently, hypergolic ignition models cannot simulate liquid contact and mixing or liquid-phase chemical reactions, and rely on experimental results for validation. In some cases, chemical kinetics of hypergolic propellants and fluid dynamics of droplet collisions couple to produce unexpected phenomena. This research investigates contact between droplets and pools of liquid hypergolic propellants under various conditions in order to investigate these liquid-phase reactions and categorize the resulting interaction. During this experiment, 142 drop tests were performed to investigate phenomena associated with hypergolic contact of various propellants. A drop of fuel impacted a semi-ellipsoidal pool of oxidizer at varying impact velocities and impact geometries. The temperature, pressure, ambient atmosphere, and propellant quality were all controlled during the experiment, as these factors have been shown to influence hypergolic ignition delay. Three distinct types of impacts were identified: explosions, bounces, and splashes. The impact type was found to depend on the impact Weber number and impact angle. Splashes occurred above a critical Weber number of 250, regardless of impact angle. Explosions occurred for Weber numbers less than 250, and for impact angles less than seven degrees. If the impact angle was greater than seven degrees then the test resulted in a bounce. Literature related to explosions induced by hypergolic contact was reviewed. Explosions were observed to occur inconsistently, a feature that has never been addressed. Literature related to non-reactive splashing, bouncing, and coalescence was reviewed for insight into the explosion phenomenon. I propose that the dependence of impact angle on the transition between explosion and

  7. In-Vessel and Ex-Vessel Neutron Dosimetry Programs in Korea

    NASA Astrophysics Data System (ADS)

    Yoo, Choon Sung; Kim, Byoung Chul; Fero, Arnold H.; Anderson, Stanwood L.

    2016-02-01

    In Korea, 20 PWRs are operating and 4 more PWRs are under construction. The in-vessel neutron dosimetry programs have been designed and implemented since each plant began operation. In addition to the in-vessel dosimetry program, ex-vessel neutron dosimetry systems have been installed for 16 PWRs. The objective of this paper is to describe the in-vessel and ex-vessel neutron dosimetry program of the PWRs in Korea and to compare in-vessel and ex-vessel dosimetry evaluation results. For this purpose plant and cycle specific forward neutron transport calculations and dosimetry measurement evaluations were carried out according to Regulatory Guide 1.190. Comparisons between the calculations and measurements were also performed for the reaction rates of each dosimetry sensor and the results show good agreement.

  8. Fluctuation theory of critical phenomena in fluids

    NASA Astrophysics Data System (ADS)

    Martynov, G. A.

    2016-07-01

    It is assumed that critical phenomena are generated by density wave fluctuations carrying a certain kinetic energy. It is noted that all coupling equations for critical indices are obtained within the context of this hypothesis. Critical indices are evaluated for 15 liquids more accurately than when using the current theory of critical phenomena.

  9. PREFACE Integrability and nonlinear phenomena Integrability and nonlinear phenomena

    NASA Astrophysics Data System (ADS)

    Gómez-Ullate, David; Lombardo, Sara; Mañas, Manuel; Mazzocco, Marta; Nijhoff, Frank; Sommacal, Matteo

    2010-10-01

    Back in 1967, Clifford Gardner, John Greene, Martin Kruskal and Robert Miura published a seminal paper in Physical Review Letters which was to become a cornerstone in the theory of integrable systems. In 2006, the authors of this paper received the AMS Steele Prize. In this award the AMS pointed out that `In applications of mathematics, solitons and their descendants (kinks, anti-kinks, instantons, and breathers) have entered and changed such diverse fields as nonlinear optics, plasma physics, and ocean, atmospheric, and planetary sciences. Nonlinearity has undergone a revolution: from a nuisance to be eliminated, to a new tool to be exploited.' From this discovery the modern theory of integrability bloomed, leading scientists to a deep understanding of many nonlinear phenomena which is by no means reachable by perturbation methods or other previous tools from linear theories. Nonlinear phenomena appear everywhere in nature, their description and understanding is therefore of great interest both from the theoretical and applicative point of view. If a nonlinear phenomenon can be represented by an integrable system then we have at our disposal a variety of tools to achieve a better mathematical description of the phenomenon. This special issue is largely dedicated to investigations of nonlinear phenomena which are related to the concept of integrability, either involving integrable systems themselves or because they use techniques from the theory of integrability. The idea of this special issue originated during the 18th edition of the Nonlinear Evolution Equations and Dynamical Systems (NEEDS) workshop, held at Isola Rossa, Sardinia, Italy, 16-23 May 2009 (http://needs-conferences.net/2009/). The issue benefits from the occasion offered by the meeting, in particular by its mini-workshops programme, and contains invited review papers and contributed papers. It is worth pointing out that there was an open call for papers and all contributions were peer reviewed

  10. Lessons on in-vessel severe accidents from experiments at KfK and the INEL

    SciTech Connect

    Hagen, S.; Hofmann, P. ); Allison, C. )

    1993-01-01

    At the time of the Three Mile Island unit 2 accident, the severe fuel damage data base was very limited. However, as a result of international severe accident research programs, that data base has been greatly expanded to include experiments ranging from single rod tests in the NIELS facility at the Kernforschungszentrum, Karlsruhe, Germany (KfK), to the coupled reactor coolant system (RCS) thermal-hydraulics-assembly melting loss-of-fluid test (LOFT) LP-FP-2 test in the LOFT reactor at the Idaho National Engineering Laboratory (INEL). In addition to these tests at the extreme ends of the scaling spectrum, other experiments in the CORA facility at the KfK and the Power Burst Facility at the INEL have also made a substantial contribution to the severe fuel damage data base. The experiments performed at the KfK and the INEL have included different modes of heating from the electrically heated experiments in Germany to the fission and decay-heat-driven experiments in Idaho. The tests have included (a) different heating rates, peak bundle temperatures, and coolant conditions, (b) fresh, trace-irradiated, and previously irradiated (36,000 MWd/tonne U) fuel, and (c) different bundle designs, including fuel-only bundles, fuel bundles with Ag-In-Cd control rods, and fuel bundles with representative B[sub 4]C control blade/channel boxes. This paper discusses several key results from these experiments that are common across all the facilities.

  11. Measurement of Turbulent Flow Phenomena for the Lower Plenum of a Prismatic Gas-Cooled Reactor

    SciTech Connect

    Hugh M. McIlroy Jr.; Donald M. McEligot; Robert J. Pink; Keith G. Condie; Glenn E. McCreery

    2007-09-01

    Mean velocity field and turbulence data are presented for flow phenomena in a lower plenum of a typical prismatic gas-cooled reactor (GCR), such as in a Very High Temperature Reactor (VHTR) concept. In preparation for design, safety analyses and licensing, research has begun on readying the computational tools that will be needed to predict the thermal-hydraulics behavior of the reactor design. Fluid dynamics experiments have been designed and built to develop benchmark databases for the assessment of computational fluid dynamics (CFD) codes and their turbulence models for a typical VHTR plenum geometry in the limiting case of negligible buoyancy and constant fluid properties. This experiment has been proposed as a “Standard Problem” for assessing advanced reactor (CFD) analysis tools. Present results concentrate on the region of the plenum near its far reflector wall (away from the outlet duct). The flow in the lower plenum can locally be considered as multiple jets into a confined cross flow - with obstructions. A model of the lower plenum has been fabricated and scaled to the geometric dimensions of the Next Generation Nuclear Plant (NGNP) Point Design. The model consists of a row of full circular posts along its centerline with half-posts on the two parallel walls to induce flow features somewhat comparable to those expected from the staggered parallel rows of posts in the reactor design. Posts, side walls and end walls are fabricated from clear, fused quartz to match the refractive-index of the working fluid so that optical techniques may be employed for the measurements. The experiments were conducted in the Matched-Index-of-Refraction (MIR) Facility at the Idaho National Laboratory (INL). The benefit of the MIR technique is that it permits optical measurements to determine complex flow characteristics in passages and around objects to be obtained without locating a disturbing transducer in the flow field and without distortion of the optical paths. The

  12. Synchronization Phenomena and Epoch Filter of Electroencephalogram

    NASA Astrophysics Data System (ADS)

    Matani, Ayumu

    Nonlinear electrophysiological synchronization phenomena in the brain, such as event-related (de)synchronization, long distance synchronization, and phase-reset, have received much attention in neuroscience over the last decade. These phenomena contain more electrical than physiological keywords and actually require electrical techniques to capture with electroencephalography (EEG). For instance, epoch filters, which have just recently been proposed, allow us to investigate such phenomena. Moreover, epoch filters are still developing and would hopefully generate a new paradigm in neuroscience from an electrical engineering viewpoint. Consequently, electrical engineers could be interested in EEG once again or from now on.

  13. Perspective: Emergent magnetic phenomena at interfaces

    SciTech Connect

    Suzuki, Yuri

    2015-06-01

    The discovery of emergent magnetic phenomena is of fundamental and technological interest. This perspective highlights recent promising examples of emergent ferromagnetism at complex oxide interfaces in the context of spin based electronics.

  14. Canister storage building natural phenomena design loads

    SciTech Connect

    Tallman, A.M.

    1996-02-01

    This document presents natural phenomena hazard (NPH) loads for use in the design and construction of the Canister Storage Building (CSB), which will be located in the 200 East Area of the Hanford Site.

  15. Analysis of nuclear reactor instability phenomena

    SciTech Connect

    Lahey, R.T. Jr.

    1993-01-01

    The phenomena known as density-wave instability often occurs in phase change systems, such as boiling water nuclear reactors (BWRS). Our current understanding of density-wave oscillations is in fairly good shape for linear phenomena (eg, the onset of instabilities) but is not very advanced for non-linear phenomena [Lahey and Podowski, 1989]. In particular, limit cycle and chaotic instability modes are not well understood in boiling systems such as current and advanced generation BWRs (eg, SBWR). In particular, the SBWR relies on natural circulation and is thus inherently prone to problems with density-wave instabilities. The purpose of this research is to develop a quantitative understanding of nonlinear nuclear-coupled density-wave instability phenomena in BWRS. This research builds on the work of Achard et al [1985] and Clausse et al [1991] who showed, respectively, that Hopf bifurcations and chaotic oscillations may occur in boiling systems.

  16. The impact of microwave stray radiation to in-vessel diagnostic components

    SciTech Connect

    Hirsch, M.; Laqua, H. P.; Hathiramani, D.; Baldzuhn, J.; Biedermann, C.; Cardella, A.; Erckmann, V.; König, R.; Köppen, M.; Zhang, D.; Oosterbeek, J.; Brand, H. von der; Parquay, S.; Jimenez, R. [Centro de Investigationes Energeticas, Medioambientales y Technológicas, Association EURATOM Collaboration: W7-X Teasm

    2014-08-21

    Microwave stray radiation resulting from unabsorbed multiple reflected ECRH / ECCD beams may cause severe heating of microwave absorbing in-vessel components such as gaskets, bellows, windows, ceramics and cable insulations. In view of long-pulse operation of WENDELSTEIN-7X the MIcrowave STray RAdiation Launch facility, MISTRAL, allows to test in-vessel components in the environment of isotropic 140 GHz microwave radiation at power load of up to 50 kW/m{sup 2} over 30 min. The results show that both, sufficient microwave shielding measures and cooling of all components are mandatory. If shielding/cooling measures of in-vessel diagnostic components are not efficient enough, the level of stray radiation may be (locally) reduced by dedicated absorbing ceramic coatings on cooled structures.

  17. Multidimensional shielding analysis of the JASPER in-vessel fuel storage experiments

    SciTech Connect

    Bucholz, J.A.

    1993-03-01

    The In-Vessel Fuel Storage (IVFS) experiments analyzed in this report were conducted at the Oak Ridge National Laboratory`s Tower Shielding Reactor (TSR) as part of the Japanese-American Shielding Program for Experimental Research (JASPER). These IVFS experiments were designed to study source multiplication and three-dimensional effects related to in-vessel storage of spent fuel elements in liquid metal reactor (LMR) systems. The present report describes the 2-D and 3-D models, analyses, and calculated results corresponding to a limited subset of those IVFS experiments in which the US LMR program has a particular interest.

  18. Shielding analysis of the LMR in-vessel fuel storage experiments

    SciTech Connect

    Bucholz, J.A.

    1994-06-01

    The In-Vessel Fuel Storage (IVFS) experiments analyzed in this paper were conducted at the Oak Ridge National Laboratory`s Tower Shielding Reactor (TSR) as part of the Japanese-American Shielding Program for Experimental Research (JASPER). These IVFS experiments were designed to study source multiplication and three-dimensional effects related to in-vessel storage of spent fuel elements in liquid metal reactor (LMR) systems. The present paper describes the 2- and 3-D calculations and results corresponding to a limited subset of those IVFS experiments in which the US LMR program had a particular interest.

  19. FY-09 Report: Experimental Validation of Stratified Flow Phenomena, Graphite Oxidation, and Mitigation Strategies of Air Ingress Accidents

    SciTech Connect

    Chang H. Oh; Eung S. Kim

    2009-12-01

    The Idaho National Laboratory (INL), under the auspices of the U.S. Department of Energy, is performing research and development that focuses on key phenomena important during potential scenarios that may occur in the Next Generation Nuclear Plant (NGNP)/Gen-IV very high temperature reactor (VHTR). Phenomena Identification and Ranking Studies to date have identified that an air ingress event following on the heels of a VHTR depressurization is a very important incident. Consequently, the development of advanced air ingress-related models and verification and validation data are a very high priority for the NGNP Project. Following a loss of coolant and system depressurization incident, air will enter the core through the break, leading to oxidation of the in-core graphite structure and fuel. If this accident occurs, the oxidation will accelerate heat-up of the bottom reflector and the reactor core and will eventually cause the release of fission products. The potential collapse of the core bottom structures causing the release of CO and fission products is one of the concerns. Therefore, experimental validation with the analytical model and computational fluid dynamic (CFD) model developed in this study is very important. Estimating the proper safety margin will require experimental data and tools, including accurate multidimensional thermal-hydraulic and reactor physics models, a burn-off model, and a fracture model. It will also require effective strategies to mitigate the effects of oxidation. The results from this research will provide crucial inputs to the INL NGNP/VHTR Methods Research and Development project. The second year of this three-year project (FY-08 to FY-10) was focused on (a) the analytical, CFD, and experimental study of air ingress caused by density-driven, stratified, countercurrent flow; (b) advanced graphite oxidation experiments and modeling; (c) experimental study of burn-off in the core bottom structures, (d) implementation of advanced

  20. Anomalous Light Phenomena vs. Bioelectric Brain Activity

    NASA Astrophysics Data System (ADS)

    Teodorani, M.; Nobili, G.

    We present a research proposal concerning the instrumented investigation of anomalous light phenomena that are apparently correlated with particular mind states, such as prayer, meditation or psi. Previous research by these authors demonstrate that such light phenomena can be monitored and measured quite efficiently in areas of the world where they are reported in a recurrent way. Instruments such as optical equipment for photography and spectroscopy, VLF spectrometers, magnetometers, radar and IR viewers were deployed and used massively in several areas of the world. Results allowed us to develop physical models concerning the structural and time-variable behaviour of light phenomena, and their kinematics. Recent insights and witnesses have suggested to us that a sort of "synchronous connection" seems to exist between plasma-like phenomena and particular mind states of experiencers who seem to trigger a light manifestation which is very similar to the one previously investigated. The main goal of these authors is now aimed at the search for a concrete "entanglement-like effect" between the experiencer's mind and the light phenomena, in such a way that both aspects are intended to be monitored and measured simultaneously using appropriate instrumentation. The goal of this research project is twofold: a) to verify quantitatively the existence of one very particular kind of mind-matter interaction and to study in real time its physical and biophysical manifestations; b) to repeat the same kind of experiment using the same test-subject in different locations and under various conditions of geomagnetic activity.

  1. Investigating the students' understanding of surface phenomena

    NASA Astrophysics Data System (ADS)

    Hamed, Kastro Mohamad

    1999-11-01

    This study investigated students' understanding of surface phenomena. The main purpose for conducting this research endeavor was to understand how students think about a complex topic about which they have little direct or formal instruction. The motivation for focusing on surface phenomena stemmed from an interest in integrating research and education. Despite the importance of surfaces and interfaces in research laboratories, in technological applications, and in everyday experiences, no previous systematic effort was done on pedagogy related to surface phenomena. The design of this research project was qualitative, exploratory, based on a Piagetian semi-structured clinical piloted interview, focused on obtaining a longitudinal view of the intended sample. The sampling was purposeful and the sample consisted of forty-four undergraduate students at Kansas State University. The student participants were enrolled in physics classes that spanned a wide academic spectrum. The data were analyzed qualitatively. The main themes that emerged from the analysis were: (a) students used analogies when confronted with novel situations, (b) students mixed descriptions and explanations, (c) students used the same explanation for several phenomena, (d) students manifested difficulties transferring the meaning of vocabulary across discipline boundaries, (e) in addition to the introductory chemistry classes, students used everyday experiences and job-related experiences as sources of knowledge, and (f) students' inquisitiveness and eagerness to investigate and discuss novel phenomena seemed to peak about the time students were enrolled in second year physics classes.

  2. On The Problem Of In-vessel Mirrors For Diagnostic Systems Of ITER

    SciTech Connect

    Voitsenya, V. S.; Litnovsky, A.

    2008-03-12

    The present status of the investigations with ITER-candidate mirror materials and directed on solution of the in-vessel mirror problem, are presented in the paper. The current tasks in the R and D of diagnostic mirrors and outstanding questions are discussed.

  3. The making of extraordinary psychological phenomena.

    PubMed

    Lamont, Peter

    2012-01-01

    This article considers the extraordinary phenomena that have been central to unorthodox areas of psychological knowledge. It shows how even the agreed facts relating to mesmerism, spiritualism, psychical research, and parapsychology have been framed as evidence both for and against the reality of the phenomena. It argues that these disputes can be seen as a means through which beliefs have been formulated and maintained in the face of potentially challenging evidence. It also shows how these disputes appealed to different forms of expertise, and that both sides appealed to belief in various ways as part of the ongoing dispute about both the facts and expertise. Finally, it shows how, when a formal Psychology of paranormal belief emerged in the twentieth century, it took two different forms, each reflecting one side of the ongoing dispute about the reality of the phenomena. PMID:25363382

  4. Theories of dynamical phenomena in sunspots

    NASA Technical Reports Server (NTRS)

    Thomas, J. H.

    1981-01-01

    Attempts that have been made to understand and explain observed dynamical phenomena in sunspots within the framework of magnetohydrodynamic theory are surveyed. The qualitative aspects of the theory and physical arguments are emphasized, with mathematical details generally avoided. The dynamical phenomena in sunspots are divided into two categories: aperiodic (quasi-steady) and oscillatory. For each phenomenon discussed, the salient observational features that any theory should explain are summarized. The two contending theoretical models that can account for the fine structure of the Evershed motion, namely the convective roll model and the siphon flow model, are described. With regard to oscillatory phenomena, attention is given to overstability and oscillatory convection, umbral oscillations and flashes. penumbral waves, five-minute oscillations in sunspots, and the wave cooling of sunspots.

  5. Fundamental investigation of duct/ESP phenomena

    SciTech Connect

    Brown, C.A. ); Durham, M.D. ); Sowa, W.A. . Combustion Lab.); Himes, R.M. ); Mahaffey, W.A. )

    1991-10-21

    Radian Corporation was contracted to investigate duct injection and ESP phenomena in a 1.7 MW pilot plant constructed for this test program. This study was an attempt to resolve problems found in previous studies and answer remaining questions for the technology using an approach which concentrates on the fundamental mechanisms of the process. The goal of the study was to obtain a better understanding of the basic physical and chemical phenomena that control: (1) the desulfurization of flue gas by calcium-based reagent, and (2) the coupling of an existing ESP particulate collection device to the duct injection process. Process economics are being studied by others. (VC)

  6. Modeling of fundamental phenomena in welds

    SciTech Connect

    Zacharia, T.; Vitek, J.M.; Goldak, J.A.; DebRoy, T.A.; Rappaz, M.; Bhadeshia, H.K.D.H.

    1993-12-31

    Recent advances in the mathematical modeling of fundamental phenomena in welds are summarized. State-of-the-art mathematical models, advances in computational techniques, emerging high-performance computers, and experimental validation techniques have provided significant insight into the fundamental factors that control the development of the weldment. The current status and scientific issues in the areas of heat and fluid flow in welds, heat source metal interaction, solidification microstructure, and phase transformations are assessed. Future research areas of major importance for understanding the fundamental phenomena in weld behavior are identified.

  7. Incorporating interfacial phenomena in solidification models

    NASA Technical Reports Server (NTRS)

    Beckermann, Christoph; Wang, Chao Yang

    1994-01-01

    A general methodology is available for the incorporation of microscopic interfacial phenomena in macroscopic solidification models that include diffusion and convection. The method is derived from a formal averaging procedure and a multiphase approach, and relies on the presence of interfacial integrals in the macroscopic transport equations. In a wider engineering context, these techniques are not new, but their application in the analysis and modeling of solidification processes has largely been overlooked. This article describes the techniques and demonstrates their utility in two examples in which microscopic interfacial phenomena are of great importance.

  8. Teaching Thermal Hydraulics & Numerical Methods: An Introductory Control Volume Primer

    SciTech Connect

    Lucas, D.S.

    2004-10-03

    This paper covers the basics of the implementation of the control volume method in the context of the Homogeneous Equilibrium Model (HEM)(T/H) code using the conservation equations of mass, momentum, and energy. This primer uses the advection equation as a template. The discussion will cover the basic equations of the control volume portion of the course in the primer, which includes the advection equation, numerical methods, along with the implementation of the various equations via FORTRAN into computer programs and the final result for a three equation HEM code and its validation.

  9. Thermal hydraulic modeling of a natural circulation loop

    NASA Astrophysics Data System (ADS)

    Jiang, S. Y.; Wu, X. X.; Zhang, Y. J.; Jia, H. J.

    The experiment was carried out on the test loop HRTL-5, which simulates the geometry and system design of a 5MW nuclear heating reactor. The analysis was based on a one-dimensional two-phase flow drift model with conservation equations for mass, steam, energy and momentum. Clausius-Clapeyron equation was used for the calculation of flashing front in the riser. A set of ordinary equations, which describes the behavior of two-phase flow in the natural circulation system, was derived through integration of the above conservation equations for the subcooled boiling region, bulk boiling region in the heated section and for the riser. The method of time-domain was used for the calculation. Both static and dynamic results are presented. System pressure, inlet subcooling and heat flux are varied as input parameters. The results show that subcooled boiling in the heated section and void flashing in the riser have significant influence on the distribution of the void fraction, mass flow rate and flow instability of the system, especially at low pressure. The response of mass flow rate, after a small disturbance in the heat flux is shown, and based on it the instability map of the system is given through experiment and calculation. There exists three regions in the instability map of the investigated natural circulation system, namely, the stable two-phase flow region, the unstable bulk and subcooled boiling flow region and the stable subcooled boiling and single phase flow region. The mechanism of two-phase flow oscillation is interpreted.

  10. Thermal hydraulic modeling of the mock fuel facility

    NASA Astrophysics Data System (ADS)

    Gardner, Jacob

    The major focus of this thesis was to make improved three dimensional models of the Mock Fuel Facility. Three distinct experiment types run with the Mock Fuel Facility (MFF) were the main focus of this thesis. Two of the experiments were modeled and an in-depth analysis of the model results was performed to gain a better understanding of the Mock Fuel Facility. For the third experiment the process of creating a model was begun. There were multiple purposes for the work completed in this thesis. The work was done partially to gain a greater understanding of the UMass Lowell Research Reactor (UMLRR). There is minimal instrumentation within the UMLRR to measure localized temperatures within the UMLRR. It is hoped that the work done in this thesis will provide a basis for future modeling work which will give insight into the temperature profiles within the UMLRR. This work is also being done to gain insight into the capabilities of the COMSOL multiphysics modelling software and evaluate its potential for future modelling work. Finally this work is also being done for its potential as an educational tool. The MFF and COMSOL have potential to be used for experimental lab work by students to learn about computer modeling and validation.

  11. Thermal-hydraulic analysis of spent fuel storage systems

    SciTech Connect

    Rector, D.R.; Wheeler, C.L.; Lombardo, N.J.

    1987-01-01

    This paper describes the COBRA-SFS (Spent Fuel Storage) computer code, which is designed to predict flow and temperature distributions in spent nuclear fuel storage and transportation systems. The decay heat generated by spent fuel in a dry storage cask is removed through a combination of conduction, natural convection, and thermal radiation. One major advantage of COBRA-SFS is that fluid recirculation within the cask is computed directly by solving the mass and momentum conservation equations. In addition, thermal radiation heat transfer is modeled using detailed radiation exchange factors based on quarter-rod segments. The equations governing mass, momentum, and energy conservation for incompressible flows are presented, and the semi-implicit solution method is described. COBRA-SFS predictions are compared to temperature data from a spent fuel storage cask test and the effect of different fill media on the cladding temperature distribution is discussed. The effect of spent fuel consolidation on cask thermal performance is also investigated. 16 refs., 6 figs., 2 tabs.

  12. Optimal thermal-hydraulic performance for helium-cooled divertors

    SciTech Connect

    Izenson, M.G.; Martin, J.L.

    1996-07-01

    Normal flow heat exchanger (NFHX) technology offers the potential for cooling divertor panels with reduced pressure drops (<0.5% {Delta}p/p), reduced pumping power (<0.75% pumping/thermal power), and smaller duct sizes than conventional helium heat exchangers. Furthermore, the NFHX can easily be fabricated in the large sizes required for divertors in large tokamaks. Recent experimental and computational results from a program to develop NFHX technology for divertor coolings using porous metal heat transfer media are described. We have tested the thermal and flow characteristics of porous metals and identified the optimal heat transfer material for the divertor heat exchanger. Methods have been developed to create highly conductive thermal bonds between the porous material and a solid substrate. Computational fluid dynamics calculations of flow and heat transfer in the porous metal layer have shown the capability of high thermal effectiveness. An 18-kW NFHX, designed to meet specifications for the international Thermonuclear Experimental Reactor divertor, has been fabricated and tested for thermal and flow performance. Preliminary results confirm design and fabrication methods. 11 refs., 12 figs., 1 tab.

  13. A hybrid incremental projection method for thermal-hydraulics applications

    NASA Astrophysics Data System (ADS)

    Christon, Mark A.; Bakosi, Jozsef; Nadiga, Balasubramanya T.; Berndt, Markus; Francois, Marianne M.; Stagg, Alan K.; Xia, Yidong; Luo, Hong

    2016-07-01

    A new second-order accurate, hybrid, incremental projection method for time-dependent incompressible viscous flow is introduced in this paper. The hybrid finite-element/finite-volume discretization circumvents the well-known Ladyzhenskaya-Babuška-Brezzi conditions for stability, and does not require special treatment to filter pressure modes by either Rhie-Chow interpolation or by using a Petrov-Galerkin finite element formulation. The use of a co-velocity with a high-resolution advection method and a linearly consistent edge-based treatment of viscous/diffusive terms yields a robust algorithm for a broad spectrum of incompressible flows. The high-resolution advection method is shown to deliver second-order spatial convergence on mixed element topology meshes, and the implicit advective treatment significantly increases the stable time-step size. The algorithm is robust and extensible, permitting the incorporation of features such as porous media flow, RANS and LES turbulence models, and semi-/fully-implicit time stepping. A series of verification and validation problems are used to illustrate the convergence properties of the algorithm. The temporal stability properties are demonstrated on a range of problems with 2 ≤ CFL ≤ 100. The new flow solver is built using the Hydra multiphysics toolkit. The Hydra toolkit is written in C++ and provides a rich suite of extensible and fully-parallel components that permit rapid application development, supports multiple discretization techniques, provides I/O interfaces, dynamic run-time load balancing and data migration, and interfaces to scalable popular linear solvers, e.g., in open-source packages such as HYPRE, PETSc, and Trilinos.

  14. Thermal-hydraulic studies on molten core-concrete interactions

    SciTech Connect

    Greene, G.A.

    1986-10-01

    This report discusses studies carried out in connection with light water power reactor accidents. Recent assessments have indicated that the consequences of molten-core concrete interactions dominate the considerations of severe accidents. The two areas of interest that have been investigated are interlayer heat and mass transfer and liquid-liquid boiling. Interlayer heat and mass transfer refers to processes that occur within a core melt between the stratified, immiscible phases of core oxides and metals. Liquid-liquid boiling refers to processes that occur at the melt-concrete on melt-coolant interface. (JDH)

  15. AP600 design certification thermal hydraulics testing and analysis

    SciTech Connect

    Hochreiter, L.E.; Piplica, E.J.

    1995-09-01

    Westinghouse Electric Corporation, in conjunction with the Department of Energy and the Electric Power Research Institute, have been developing an advanced light water reactor design; the AP600. The AP600 is a 1940 Mwt, 600Mwe unit which is similar to a Westinghouse two-loop Pressurized Water Reactor. The accumulated knowledge on reactor design to reduce the capital costs, construction time, and the operational and maintenance cost of the unit once it begins to generate electrical power. The AP600 design goal is to maintain an overall cost advantage over fossil generated electrical power.

  16. A hybrid incremental projection method for thermal-hydraulics applications

    DOE PAGESBeta

    Christon, Mark A.; Bakosi, Jozsef; Nadiga, Balasubramanya T.; Berndt, Markus; Francois, Marianne M.; Stagg, Alan K.; Xia, Yidong; Luo, Hong

    2016-05-04

    In this paper, a new second-order accurate, hybrid, incremental projection method for time-dependent incompressible viscous flow is introduced in this paper. The hybrid finite-element/finite-volume discretization circumvents the well-known Ladyzhenskaya–Babuška–Brezzi conditions for stability, and does not require special treatment to filter pressure modes by either Rhie–Chow interpolation or by using a Petrov–Galerkin finite element formulation. The use of a co-velocity with a high-resolution advection method and a linearly consistent edge-based treatment of viscous/diffusive terms yields a robust algorithm for a broad spectrum of incompressible flows. The high-resolution advection method is shown to deliver second-order spatial convergence on mixed element topology meshes,more » and the implicit advective treatment significantly increases the stable time-step size. The algorithm is robust and extensible, permitting the incorporation of features such as porous media flow, RANS and LES turbulence models, and semi-/fully-implicit time stepping. A series of verification and validation problems are used to illustrate the convergence properties of the algorithm. The temporal stability properties are demonstrated on a range of problems with 2 ≤ CFL ≤ 100. The new flow solver is built using the Hydra multiphysics toolkit. The Hydra toolkit is written in C++ and provides a rich suite of extensible and fully-parallel components that permit rapid application development, supports multiple discretization techniques, provides I/O interfaces, dynamic run-time load balancing and data migration, and interfaces to scalable popular linear solvers, e.g., in open-source packages such as HYPRE, PETSc, and Trilinos.« less

  17. Applications of a general thermal/hydraulic simulation tool

    NASA Technical Reports Server (NTRS)

    Cullimore, B. A.

    1989-01-01

    The analytic techniques, sample applications, and development status of a general-purpose computer program called SINDA '85/FLUINT (for systems improved numerical differencing analyzer, 1985 version with fluid integrator), designed for simulating thermal structures and internal fluid systems, are described, with special attention given to the applications of the fluid system capabilities. The underlying assumptions, methodologies, and modeling capabilities of the system are discussed. Sample applications include component-level and system-level simulations. A system-level analysis of a cryogenic storage system is presented.

  18. Thermal hydraulic analysis of the FFTF core using SUPERENERGY-2

    SciTech Connect

    Cramer, E.R.; Basehore, K.L.

    1980-01-01

    SUPERENERGY-2 is the latest steady-state code in the ENERGY series, combining all of the desirable features of the previous ENERGY-I and SUPERENERGY versions in an optimized form. The result is an easily redimensionable, multiassembly code with many user-convenience features, such as automatic noding and a default constitutive package, that help minimize the effort and time associated with setting up large forced-convection problems. Improvements in physical modeling include generalized facial boundary conditions, duct wall gamma heating, and a model for double-ducted assemblies. The latter is used for modeling both multiduct test and absorber assemblies. SUPERENERGY-2 was used to calculate the temperature distribution in the first six rows of the FFTF core.

  19. Spin Circuit Representation for Spin Pumping Phenomena

    NASA Astrophysics Data System (ADS)

    Roy, Kuntal; Datta, Supriyo

    2015-03-01

    There has been enormous progress in the field of spintronics and nanomagnetics in recent years with the discovery of many new materials and phenomena and it remains a formidable challenge to integrate these phenomena into functional devices and evaluate their potential. To facilitate this process a modular approach has been proposed whereby different phenomena are represented by spin circuit components. Unlike ordinary circuit components, these spin circuit components are characterized by 4-component voltages and currents (one for charge and three for spin). In this talk we will (1) present a spin circuit representation for spin pumping phenomena, (2) combine it with a spin circuit representation for the spin Hall effect to show that it reproduces established results obtained earlier by other means, and finally (3) use it to propose a possible method for enhancing the spin pumping efficiency by an order of magnitude through the addition of a spin sink layer. This work was supported by FAME, one of six centers of STARnet, a Semiconductor Research Corporation program sponsored by MARCO and DARPA.

  20. Phylogeny of Aging and Related Phenoptotic Phenomena.

    PubMed

    Libertini, G

    2015-12-01

    The interpretation of aging as adaptive, i.e. as a phenomenon genetically determined and modulated, and with an evolutionary advantage, implies that aging, as any physiologic mechanism, must have phylogenetic connections with similar phenomena. This review tries to find the phylogenetic connections between vertebrate aging and some related phenomena in other species, especially within those phenomena defined as phenoptotic, i.e. involving the death of one or more individuals for the benefit of other individuals. In particular, the aim of the work is to highlight and analyze similarities and connections, in the mechanisms and in the evolutionary causes, between: (i) proapoptosis in prokaryotes and apoptosis in unicellular eukaryotes; (ii) apoptosis in unicellular and multicellular eukaryotes; (iii) aging in yeast and in vertebrates; and (iv) the critical importance of the DNA subtelomeric segment in unicellular and multicellular eukaryotes. In short, there is strong evidence that vertebrate aging has clear similarities and connections with phenomena present in organisms with simpler organization. These phylogenetic connections are a necessary element for the sustainability of the thesis of aging explained as an adaptive phenomenon, and, on the contrary, are incompatible with the opposite view of aging as being due to the accumulation of random damages of various kinds. PMID:26638678

  1. Simple Phenomena, Slow Motion, Surprising Physics

    ERIC Educational Resources Information Center

    Koupil, Jan; Vicha, Vladimir

    2011-01-01

    This article describes a few simple experiments that are worthwhile for slow motion recording and analysis either because of interesting phenomena that can be seen only when slowed down significantly or because of the ability to do precise time measurements. The experiments described in this article are quite commonly done in Czech schools. All…

  2. MIXING PHENOMENA IN INDUSTRIAL FUME AFTERBURNER SYSTEMS

    EPA Science Inventory

    The report reviews the physical-mixing phenomena involved in the reactions that occur in afterburners or fume incinerators. It considers mixing in after-burners from three points of view. It first covers typical designs of afterburner components that are involved in the mixing ph...

  3. Geophysical phenomena classification by artificial neural networks

    NASA Technical Reports Server (NTRS)

    Gough, M. P.; Bruckner, J. R.

    1995-01-01

    Space science information systems involve accessing vast data bases. There is a need for an automatic process by which properties of the whole data set can be assimilated and presented to the user. Where data are in the form of spectrograms, phenomena can be detected by pattern recognition techniques. Presented are the first results obtained by applying unsupervised Artificial Neural Networks (ANN's) to the classification of magnetospheric wave spectra. The networks used here were a simple unsupervised Hamming network run on a PC and a more sophisticated CALM network run on a Sparc workstation. The ANN's were compared in their geophysical data recognition performance. CALM networks offer such qualities as fast learning, superiority in generalizing, the ability to continuously adapt to changes in the pattern set, and the possibility to modularize the network to allow the inter-relation between phenomena and data sets. This work is the first step toward an information system interface being developed at Sussex, the Whole Information System Expert (WISE). Phenomena in the data are automatically identified and provided to the user in the form of a data occurrence morphology, the Whole Information System Data Occurrence Morphology (WISDOM), along with relationships to other parameters and phenomena.

  4. Geophysical phenomena classification by artificial neural networks

    SciTech Connect

    Gough, M.P.; Bruckner, J.R.

    1995-01-01

    Space science information systems involve accessing vast data bases. There is a need for an automatic process by which properties of the whole data set can be assimilated and presented to the user. Where data are in the form of spectrograms, phenomena can be detected by pattern recognition techniques. Presented are the first results obtained by applying unsupervised Artificial Neural Networks (ANN`s) to the classification of magnetospheric wave spectra. The networks used here were a simple unsupervised Hamming network run on a PC and a more sophisticated CALM network run on a Sparc workstation. The ANN`s were compared in their geophysical data recognition performance. CALM networks offer such qualities as fast learning, superiority in generalizing, the ability to continuously adapt to changes in the pattern set, and the possibility to modularize the network to allow the inter-relation between phenomena and data sets. This work is the first step toward an information system interface being developed at Sussex, the Whole Information System Expert (WISE). Phenomena in the data are automatically identified and provided to the user in the form of a data occurrence morphology, the Whole Information System Data Occurrence Morphology (WISDOM), along with relationships to other parameters and phenomena.

  5. Solar Phenomena Associated with "EIT Waves"

    NASA Technical Reports Server (NTRS)

    Biesecker, D. A.; Myers, D. C.; Thompson, B. J.; Hammer, D. M.; Vourlidas, A.

    2002-01-01

    In an effort to understand what an 'EIT wave' is and what its causes are, we have looked for correlations between the initiation of EIT waves and the occurrence of other solar phenomena. An EIT wave is a coronal disturbance, typically appearing as a diffuse brightening propagating across the Sun. A catalog of EIT waves, covering the period from 1997 March through 1998 June, was used in this study. For each EIT wave, the catalog gives the heliographic location and a rating for each wave, where the rating is determined by the reliability of the observations. Since EIT waves are transient, coronal phenomena, we have looked for correlations with other transient, coronal phenomena: X-ray flares, coronal mass ejections (CMEs), and metric type II radio bursts. An unambiguous correlation between EIT waves and CMEs has been found. The correlation of EIT waves with flares is significantly weaker, and EIT waves frequently are not accompanied by radio bursts. To search for trends in the data, proxies for each of these transient phenomena are examined. We also use the accumulated data to show the robustness of the catalog and to reveal biases that must be accounted for in this study.

  6. Wave Phenomena in an Acoustic Resonant Chamber

    ERIC Educational Resources Information Center

    Smith, Mary E.; And Others

    1974-01-01

    Discusses the design and operation of a high Q acoustical resonant chamber which can be used to demonstrate wave phenomena such as three-dimensional normal modes, Q values, densities of states, changes in the speed of sound, Fourier decomposition, damped harmonic oscillations, sound-absorbing properties, and perturbation and scattering problems.…

  7. Atmospheric phenomena before and during sunset

    NASA Astrophysics Data System (ADS)

    Menat, M.

    The atmospheric transmittance and the astronomical refraction for low-elevation trajectories are discussed and quantitatively developed. The results are used to describe and calculate some of the fascinating atmospheric phenomena occurring shortly before and during sunset, such as the diminishing apparent luminance of the sun, its shape during sunset, and the green flash.

  8. A 'Phenomena Laboratory' for Physics Students.

    ERIC Educational Resources Information Center

    Houlden, M. A.; And Others

    1983-01-01

    Describes a laboratory designed to give students practical experiences with experimental phenomena discussed in lectures, focusing on laboratory organization and typical experiment. In addition to a list of experiments, three exercises are discussed: fluorescence/laser, ferromagnetic domains, and thermal population (which uses PET computer…

  9. Temporal Phenomena in the Korean Conjunctive Constructions

    ERIC Educational Resources Information Center

    Kim, Dongmin

    2015-01-01

    The goal of this study is to characterize the temporal phenomena in the Korean conjunctive constructions. These constructions consist of three components: a verbal stem, a clause medial temporal suffix, and a clause terminal suffix. This study focuses on both the temporality of the terminal connective suffixes and the grammatical meanings of the…

  10. Intervention in Biological Phenomena via Feedback Linearization.

    PubMed

    Fnaiech, Mohamed Amine; Nounou, Hazem; Nounou, Mohamed; Datta, Aniruddha

    2012-01-01

    The problems of modeling and intervention of biological phenomena have captured the interest of many researchers in the past few decades. The aim of the therapeutic intervention strategies is to move an undesirable state of a diseased network towards a more desirable one. Such an objective can be achieved by the application of drugs to act on some genes/metabolites that experience the undesirable behavior. For the purpose of design and analysis of intervention strategies, mathematical models that can capture the complex dynamics of the biological systems are needed. S-systems, which offer a good compromise between accuracy and mathematical flexibility, are a promising framework for modeling the dynamical behavior of biological phenomena. Due to the complex nonlinear dynamics of the biological phenomena represented by S-systems, nonlinear intervention schemes are needed to cope with the complexity of the nonlinear S-system models. Here, we present an intervention technique based on feedback linearization for biological phenomena modeled by S-systems. This technique is based on perfect knowledge of the S-system model. The proposed intervention technique is applied to the glycolytic-glycogenolytic pathway, and simulation results presented demonstrate the effectiveness of the proposed technique. PMID:23209459

  11. Crystal Melting and Wall Crossing Phenomena

    NASA Astrophysics Data System (ADS)

    Yamazaki, Masahito

    This paper summarizes recent developments in the theory of Bogomol'nyi-Prasad-Sommerfield (BPS) state counting and the wall crossing phenomena, emphasizing in particular the role of the statistical mechanical model of crystal melting. This paper is divided into two parts, which are closely related to each other. In the first part, we discuss the statistical mechanical model of crystal melting counting BPS states. Each of the BPS states contributing to the BPS index is in one-to-one correspondence with a configuration of a molten crystal, and the statistical partition function of the melting crystal gives the BPS partition function. We also show that smooth geometry of the Calabi-Yau manifold emerges in the thermodynamic limit of the crystal. This suggests a remarkable interpretation that an atom in the crystal is a discretization of the classical geometry, giving an important clue as such to the geometry at the Planck scale. In the second part, we discuss the wall crossing phenomena. Wall crossing phenomena states that the BPS index depends on the value of the moduli of the Calabi-Yau manifold, and jumps along real codimension one subspaces in the moduli space. We show that by using type IIA/M-theory duality, we can provide a simple and an intuitive derivation of the wall crossing phenomena, furthermore clarifying the connection with the topological string theory. This derivation is consistent with another derivation from the wall crossing formula, motivated by multicentered BPS extremal black holes. We also explain the representation of the wall crossing phenomena in terms of crystal melting, and the generalization of the counting problem and the wall crossing to the open BPS invariants.

  12. Crystal Melting and Wall Crossing Phenomena

    NASA Astrophysics Data System (ADS)

    Yamazaki, Masahito

    2010-02-01

    This paper summarizes recent developments in the theory of Bogomol'nyi-Prasad-Sommerfield (BPS) state counting and the wall crossing phenomena, emphasizing in particular the role of the statistical mechanical model of crystal melting. This paper is divided into two parts, which are closely related to each other. In the first part, we discuss the statistical mechanical model of crystal melting counting BPS states. Each of the BPS state contributing to the BPS index is in one-to-one correspondence with a configuration of a molten crystal, and the statistical partition function of the melting crystal gives the BPS partition function. We also show that smooth geometry of the Calabi-Yau manifold emerges in the thermodynamic limit of the crystal. This suggests a remarkable interpretation that an atom in the crystal is a discretization of the classical geometry, giving an important clue as to the geometry at the Planck scale.In the second part we discuss the wall crossing phenomena. Wall crossing phenomena states that the BPS index depends on the value of the moduli of the Calabi-Yau manifold, and jumps along real codimension one subspaces in the moduli space. We show that by using type IIA/M-theory duality, we can provide a simple and an intuitive derivation of the wall crossing phenomena, furthermore clarifying the connection with the topological string theory. This derivation is consistent with another derivation from the wall crossing formula, motivated by multi-centered BPS extremal black holes. We also explain the representation of the wall crossing phenomena in terms of crystal melting, and the generalization of the counting problem and the wall crossing to the open BPS invariants.

  13. Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-3: High Temperature Gas Cooled Reactor Thermal-Hydraulics.

    ERIC Educational Resources Information Center

    Reihman, Thomas C.

    This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical high temperature gas-cooled reactor (HTGR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating its use with a simplified model. The heart of the module…

  14. In-Vessel Coil Material Failure Rate Estimates for ITER Design Use

    SciTech Connect

    L. C. Cadwallader

    2013-01-01

    The ITER international project design teams are working to produce an engineering design for construction of this large tokamak fusion experiment. One of the design issues is ensuring proper control of the fusion plasma. In-vessel magnet coils may be needed for plasma control, especially the control of edge localized modes (ELMs) and plasma vertical stabilization (VS). These coils will be lifetime components that reside inside the ITER vacuum vessel behind the blanket modules. As such, their reliability is an important design issue since access will be time consuming if any type of repair were necessary. The following chapters give the research results and estimates of failure rates for the coil conductor and jacket materials to be used for the in-vessel coils. Copper and CuCrZr conductors, and stainless steel and Inconel jackets are examined.

  15. Identification and evaluation of PWR in-vessel severe accident management strategies

    SciTech Connect

    Dukelow, J S; Harrison, D G; Morgenstern, M

    1992-03-01

    This reports documents work performed the NRC/RES Accident Management Guidance Program to evaluate possible strategies for mitigating the consequences of PWR severe accidents. The selection and evaluation of strategies was limited to the in-vessel phase of the severe accident, i.e., after the initiation of core degradation and prior to RPV failure. A parallel project at BNL has been considering strategies applicable to the ex-vessel phase of PWR severe accidents.

  16. Auroral Phenomena: Associated with auroras in complex ways are an extraordinary number of other physical phenomena.

    PubMed

    O'brien, B J

    1965-04-23

    The array of auroral phenomena involves all the basic types of physical phenomena: heat, light, sound, electricity and magnetism, atomic physics, and plasma physics. The uncontrollability, the unreproducibility, and the sheer enormity of the phenomena will keep experimentalists and theorists busy but unsatisfied for many years to come. The greatest challenge in this field of research is an adequate experimentally verifiable theory of the local energization of auroral particle fluxes. Once that is achieved, there is every likelihood that the multitude of correlations between auroral phenomena can be understood and appreciated. Until that time, however, such correlations are to be regarded like icebergs-the parts that can be seen are only a small fraction of the whole phenomenon, and it is the large unseen parts that can be dangerous to theorists and experimentalists alike. PMID:17842831

  17. Transport Phenomena During Equiaxed Solidification of Alloys

    NASA Technical Reports Server (NTRS)

    Beckermann, C.; deGroh, H. C., III

    1997-01-01

    Recent progress in modeling of transport phenomena during dendritic alloy solidification is reviewed. Starting from the basic theorems of volume averaging, a general multiphase modeling framework is outlined. This framework allows for the incorporation of a variety of microscale phenomena in the macroscopic transport equations. For the case of diffusion dominated solidification, a simplified set of model equations is examined in detail and validated through comparisons with numerous experimental data for both columnar and equiaxed dendritic growth. This provides a critical assessment of the various model assumptions. Models that include melt flow and solid phase transport are also discussed, although their validation is still at an early stage. Several numerical results are presented that illustrate some of the profound effects of convective transport on the final compositional and structural characteristics of a solidified part. Important issues that deserve continuing attention are identified.

  18. Coherence Phenomena in Coupled Optical Resonators

    NASA Technical Reports Server (NTRS)

    Smith, D. D.; Chang, H.

    2004-01-01

    We predict a variety of photonic coherence phenomena in passive and active coupled ring resonators. Specifically, the effective dispersive and absorptive steady-state response of coupled resonators is derived, and used to determine the conditions for coupled-resonator-induced transparency and absorption, lasing without gain, and cooperative cavity emission. These effects rely on coherent photon trapping, in direct analogy with coherent population trapping phenomena in atomic systems. We also demonstrate that the coupled-mode equations are formally identical to the two-level atom Schrodinger equation in the rotating-wave approximation, and use this result for the analysis of coupled-resonator photon dynamics. Notably, because these effects are predicted directly from coupled-mode theory, they are not unique to atoms, but rather are fundamental to systems of coherently coupled resonators.

  19. Study of non-equilibrium transport phenomena

    NASA Technical Reports Server (NTRS)

    Sharma, Surendra P.

    1987-01-01

    Nonequilibrium phenomena due to real gas effects are very important features of low density hypersonic flows. The shock shape and emitted nonequilibrium radiation are identified as the bulk flow behavior parameters which are very sensitive to the nonequilibrium phenomena. These parameters can be measured in shock tubes, shock tunnels, and ballistic ranges and used to test the accuracy of computational fluid dynamic (CFD) codes. Since the CDF codes, by necessity, are based on multi-temperature models, it is also desirable to measure various temperatures, most importantly, the vibrational temperature. The CFD codes would require high temperature rate constants, which are not available at present. Experiments conducted at the NASA Electric Arc-driven Shock Tube (EAST) facility reveal that radiation from steel contaminants overwhelm the radiation from the test gas. For the measurement of radiation and the chemical parameters, further investigation and then appropriate modifications of the EAST facility are required.

  20. Optimizing Laboratory Experiments for Dynamic Astrophysical Phenomena

    SciTech Connect

    Ryutov, D; Remington, B

    2005-09-13

    To make a laboratory experiment an efficient tool for the studying the dynamical astrophysical phenomena, it is desirable to perform them in such a way as to observe the scaling invariance with respect to the astrophysical system under study. Several examples are presented of such scalings in the area of magnetohydrodynamic phenomena, where a number of scaled experiments have been performed. A difficult issue of the effect of fine-scale dissipative structures on the global scale dissipation-free flow is discussed. The second part of the paper is concerned with much less developed area of the scalings relevant to the interaction of an ultra-intense laser pulse with a pre-formed plasma. The use of the symmetry arguments in such experiments is also considered.

  1. A review of impulsive phase phenomena

    NASA Technical Reports Server (NTRS)

    Dejager, C.

    1986-01-01

    A brief review is given of impulsive phase phenomena in support of the models used to compute the energies of the different components of the flares under study. The observational characteristics of the impulsive phase are discussed as well as the evidence for multi-thermal or non-thermal phenomena. The significance of time delays between hard X-rays and microwaves is discussed in terms of electron beams and Alfven waves, two-step acceleration, and secondary bursts at large distances from the primary source. Observations indicating the occurrence of chromospheric evaporation, coronal explosions, and thermal conduction fronts are reviewed briefly, followed by the gamma ray and neutron results. Finally, a preferred flare scenario and energy source are presented involving the interactions in a complex of magnetic loops with the consequent reconnection and electron acceleration.

  2. Kinetically controlled phenomena in dynamic combinatorial libraries.

    PubMed

    Ji, Qing; Lirag, Rio Carlo; Miljanić, Ognjen Š

    2014-03-21

    Dynamic combinatorial libraries (DCLs) are collections of structurally related compounds that can interconvert through reversible chemical reaction(s). Such reversibility endows DCLs with adaptability to external stimuli, as rapid interconversion allows quick expression of those DCL components which best respond to the disturbing stimulus. This Tutorial Review focuses on the kinetically controlled phenomena that occur within DCLs. Specifically, it will describe dynamic chiral resolution of DCLs, their self-sorting under the influence of irreversible chemical and physical stimuli, and the autocatalytic behaviours within DCLs which can result in self-replicating systems. A brief discussion of precipitation-induced phenomena will follow and the review will conclude with the presentation of covalent organic frameworks (COFs)-porous materials whose synthesis critically depends on the fine tuning of the crystal growth and error correction rates within large DCLs. PMID:24445841

  3. Thermal-Mechanical Analysis for in-Vessel Diagnostic Components in W7-X

    NASA Astrophysics Data System (ADS)

    Ye, M. Y.; Werner, A.; Hirsch, M.; Thomsen, H.; Weller, A.; König, R.

    2008-03-01

    For long pulse plasma operation of the W7-X stellarator, the most serious challenge for the design of in-vessel diagnostic systems is the thermo-mechanical problem. Thermal load from convective losses and from plasma radiation can be as high as 500 kW/m2 at some locations close to plasma. The typical thermal load from plasma radiation alone ranges from several 10 to 100 kW/m2 as derived from 3-D Monte-Carlo simulations. A finite element analysis (FEA-ANSYS) is conducted for a better understanding of thermo-mechanical effects on in-vessel diagnostic components and to guide the design of the diagnostic system for steady state operation. All in-vessel diagnostic components require active cooling. Besides for long-pulse plasma operation optical components must be optimized to minimize thermal deformations. In this paper, we present the thermo-mechanical analyses of the CO2-laser interferometer retro-reflectors, the diamagnetic loops and the soft X-ray multi camera tomography system (XMCTS).

  4. Natural phenomena hazards site characterization criteria

    SciTech Connect

    Not Available

    1994-03-01

    The criteria and recommendations in this standard shall apply to site characterization for the purpose of mitigating Natural Phenomena Hazards (wind, floods, landslide, earthquake, volcano, etc.) in all DOE facilities covered by DOE Order 5480.28. Criteria for site characterization not related to NPH are not included unless necessary for clarification. General and detailed site characterization requirements are provided in areas of meteorology, hydrology, geology, seismology, and geotechnical studies.

  5. Coronal Mass Ejections (CMEs) and Associated Phenomena

    NASA Astrophysics Data System (ADS)

    Manoharan, P. K.

    2008-10-01

    The Sun is the most powerful radio waves emitting object in the sky. The first documented recognition of the reception of radio waves from the Sun was made in 1942 by Hey.15 Since then solar radio observations, from ground-based and space-based instruments, have played a major role in understanding the physics of the Sun and fundamental physical processes of the solar radio emitting phenomena...

  6. Coherent topological phenomena in protein folding.

    PubMed

    Bohr, H; Brunak, S; Bohr, J

    1997-01-01

    A theory is presented for coherent topological phenomena in protein dynamics with implications for protein folding and stability. We discuss the relationship to the writhing number used in knot diagrams of DNA. The winding state defines a long-range order along the backbone of a protein with long-range excitations, 'wring' modes, that play an important role in protein denaturation and stability. Energy can be pumped into these excitations, either thermally or by an external force. PMID:9218961

  7. Particle Modelling of Fluid Phenomena in Three -

    NASA Astrophysics Data System (ADS)

    Mahmoudi, Mohsen

    A new, numerical approach is developed to simulate fluid phenomena by means of molecular type behavior. First, consider the large number of molecules to be approximated by a smaller number of aggregates called particles. Then, let the particles interact with each other according to a classical molecular type force vec F whose magnitude F is given (Hirchfelder, Curtiss and Bird (1954)) by: F = rm -{Gover r^{p}} + {Hover r^{q}}, in which G, H, p, q, are positive constants and r is the distance between two particles. The acceleration of each particle is related to the force by the Newtonian dynamical equations vec F = m vec a. Displacement, velocity, and acceleration of each particle are then approximated by the "Leap Frog" formulas. The CRAY X-MP/24 is used to solve numerically the resulting large system of nonlinear, ordinary differential equations. We then study fluid phenomena in the following order. Part 1. Generation of particle fluids in a cylindrical region. Part 2. Verification of basic fluid properties. Part 3. Simulation of surface motion. In this part, we simulate three phenomena, which can be observed physically, by dropping a small object into a container filled with liquid. First, there is a backdrop. Then, a wave will be generated and going outward from the point of entry of the object into the container. Last, a reaction which can be recorded only with a high speed camera (Trefethen (1972)) is that very small drops of the container fluid may actually pinch off from the backdrop. Part 4. Simulation of surface tension. This phenomena can be observed by performing the following experiment. A small needle placed gently upon a water surface will not be sunk but will be supported by the molecular forces in the liquid surface. The molecules in the surface are depressed slightly in the process.

  8. Nonlinear phenomena in plasma physics and hydrodynamics

    NASA Astrophysics Data System (ADS)

    Sagdeev, R. Z.

    Advances in the theory of nonlinear phenomena are discussed in individual chapters contributed by Soviet physicists. Topics examined include vortices in plasma and hydrodynamics, oscillations and bifurcations in reversible systems, regular and chaotic dynamics of particles in a magnetic field, and renormalization-group theory and Kolmogorov-Arnold-Moser theory. Consideration is given to nonlinear problems of the turbulent dynamo, strong turbulence and topological solitons, self-oscillations in chemical systems, and autowaves in biologically active media.

  9. Tunable caustic phenomena in electron wavefields.

    PubMed

    Tavabi, Amir Hossein; Migunov, Vadim; Dwyer, Christian; Dunin-Borkowski, Rafal E; Pozzi, Giulio

    2015-10-01

    Novel caustic phenomena, which contain fold, butterfly and elliptic umbilic catastrophes, are observed in defocused images of two approximately collinear oppositely biased metallic tips in a transmission electron microscope. The observed patterns depend sensitively on defocus, on the applied voltage between the tips and on their separation and lateral offset. Their main features are interpreted on the basis of a projected electrostatic potential model for the electron-optical phase shift. PMID:26069930

  10. Seismoelectric Phenomena in Fluid-Saturated Sediments

    SciTech Connect

    Block, G I; Harris, J G

    2005-04-22

    Seismoelectric phenomena in sediments arise from acoustic wave-induced fluid motion in the pore space, which perturbs the electrostatic equilibrium of the electric double layer on the grain surfaces. Experimental techniques and the apparatus built to study this electrokinetic (EK) effect are described and outcomes for studies of seismoelectric phenomena in loose glass microspheres and medium-grain sand are presented. By varying the NaCl concentration in the pore fluid, we measured the conductivity dependence of two kinds of EK behavior: (1) the electric fields generated within the samples by the passage of transmitted acoustic waves, and (2) the electromagnetic wave produced at the fluid-sediment interface by the incident acoustic wave. Both phenomena are caused by relative fluid motion in the sediment pores--this feature is characteristic of poroelastic (Biot) media, but not predicted by either viscoelastic fluid or solid models. A model of plane-wave reflection from a fluid-sediment interface using EK-Biot theory leads to theoretical predictions that compare well to the experimental data for both sand and glass microspheres.

  11. Physical mechanism of membrane osmotic phenomena

    SciTech Connect

    Guell, D.C.; Brenner, H.

    1996-09-01

    The microscale, physicomechanical cause of osmosis and osmotic pressure in systems involving permeable and semipermeable membranes is not well understood, and no fully satisfactory mechanism has been offered to explain these phenomena. A general theory, albeit limited to dilute systems of inert, noninteracting solute particles, is presented which demonstrates that short-range forces exerted by the membrane on the dispersed solute particles constitute the origin of osmotic phenomena. At equilibrium, the greater total force exerted by the membrane on those solute particles present in the reservoir containing the more concentrated of the two solutions bathing the membrane is balanced by a macroscopically observable pressure difference between the two reservoirs. The latter constitutes the so-called osmotic pressure difference. Under nonequilibrium conditions, the membrane-solute force is transmitted to the solvent, thus driving the convective flow of solvent observed macroscopically as osmosis. While elements of these ideas have been proposed previously in various forms, the general demonstration offered here of the physicomechanical source of osmotic phenomena is novel. Beyond the purely academic interest that exists in establishing a mechanical understanding of osmotic pressure, the analysis lays the foundation underlying a quantitative theory of osmosis in dilute, nonequilibrium systems outlined in a companion paper.

  12. Stability and restoration phenomena in competitive systems

    NASA Astrophysics Data System (ADS)

    Uechi, Lisa; Akutsu, Tatsuya

    2013-10-01

    A conservation law along with stability, recovering phenomena, and characteristic patterns of a nonlinear dynamical system have been studied and applied to physical, biological, and ecological systems. In our previous study, we proposed a system of symmetric 2n-dimensional conserved nonlinear differential equations. In this paper, competitive systems described by a 2-dimensional nonlinear dynamical (ND) model with external perturbations are applied to population cycles and recovering phenomena of systems from microbes to mammals. The famous 10-year cycle of population density of Canadian lynx and snowshoe hare is numerically analyzed. We find that a nonlinear dynamical system with a conservation law is stable and generates a characteristic rhythm (cycle) of population density, which we call the standard rhythm of a nonlinear dynamical system. The stability and restoration phenomena are strongly related to a conservation law and the balance of a system. The standard rhythm of population density is a manifestation of the survival of the fittest to the balance of a nonlinear dynamical system.

  13. An interpretation of passive containment cooling phenomena

    SciTech Connect

    Chung, Bum-Jin; Kang, Chang-Sun,

    1995-09-01

    A simplified interpretation model for the cooling capability of the Westinghouse type PCCS is proposed in this paper. The PCCS domain was phenomenologically divided into 3 regions; water entrance effect region, asymptotic region, and air entrance effect region. The phenomena in the asymptotic region is focused in this paper. Due to the very large height to thickness ratio of the water film, the length of the asymptotic region is estimated to be over 90% of the whole domain. Using the analogy between heat and mass transfer phenomena in a turbulent situation, a new dependent variable combining temperature and vapor mass fraction was defined. The similarity between the PCCS phenomena, which contains the sensible and latent heat transfer, and the buoyant air flow on a vertical heated plate is derived. The modified buoyant coefficient and thermal conductivity were defined. Using these newly defined variable and coefficients, the modified correlation for the interfacial heat fluxes and the ratios of latent heat transfer to sensible heat transfer is established. To verify the accuracy of the correlation, the results of this study were compared with the results of other numerical analyses performed for the same configuration and they are well within the range of 15% difference.

  14. Further investigations of oblique hypervelocity impact phenomena

    NASA Technical Reports Server (NTRS)

    Schonberg, William P.

    1988-01-01

    The results of a continuing investigation of the phenomena associated with the oblique hypervelocity impact of spherical projectiles onto multi-sheet aluminum structures are described. A series of equations that quantitatively describes these phenomena is obtained through a regression of experimental data. These equations characterize observed ricoshet and penetration damage phenomena in a multi-sheet structure as functions of the geometric parameters of the structure and the diameter, obliquity, and velocity of the impacting projectile. Crater damage observed on the ricochet witness plates is used to determine the sizes and speeds of the ricochet debris particles that caused the damage. It is shown that, in general, the most damaging ricochet debris particle is approximately 0.25 cm (0.10 in) in diameter and travels at the speed of approximately 2.1 km/sec (6,890 ft/sec). The equations necessary for the design of shielding panels that will protect external systems from such ricochet debris damage are also developed. The dimensions of these shielding panels are shown to be strongly dependent on their inclination and on their circumferential distribution around the spacecraft. It is concluded that obliquity effects of high-speed impacts must be considered in the design of any structure exposed to the meteoroid and space debris environment.

  15. Search for collective phenomena in hadron interactions

    SciTech Connect

    Kokoulina, E. S. Nikitin, V. A. Petukhov, Y. P.; Karpov, A. V. Kutov, A. Ya.

    2010-12-15

    New results of the search for collective phenomena have been obtained and analyzed in the present report. The experimental studies are carried out on U-70 accelerator of IHEP in Protvino. It is suggested that these phenomena can be discovered at the energy range of 50-70 GeV in the extreme multiplicity region since the high-density matter can form in this very region. The collective behavior of secondary particles is considered to manifest itself in the Bose-Einstein condensation of pions, Vavilov-Cherenkov gluon radiation, excess of soft-photon yield, and other unique phenomena. The perceptible peak in the angular distribution has been revealed. It was interpreted as the gluon radiation and so the parton matter refraction index was determined. The new software was designed for the track reconstruction based on Kalman Filter technique. This algorithm allows one to estimate more precisely the track parameters (especially momentum). The search for Bose-Einstein condensation can be continued by using the selected events with the multiplicity of more than eight charged particles. The gluon dominance model predictions have shown good agreement with the multiplicity distribution at high multiplicity and confirmed the guark-gluon medium formation under these conditions.

  16. Microgravity Transport Phenomena Experiment (MTPE) Overview

    NASA Technical Reports Server (NTRS)

    Mason, Larry W.

    1999-01-01

    The Microgravity Transport Phenomena Experiment (MTPE) is a fluids experiment supported by the Fundamentals in Biotechnology program in association with the Human Exploration and Development of Space (BEDS) initiative. The MTP Experiment will investigate fluid transport phenomena both in ground based experiments and in the microgravity environment. Many fluid transport processes are affected by gravity. Osmotic flux kinetics in planar membrane systems have been shown to be influenced by gravimetric orientation, either through convective mixing caused by unstably stratified fluid layers, or through a stable fluid boundary layer structure that forms in association with the membrane. Coupled transport phenomena also show gravity related effects. Coefficients associated with coupled transport processes are defined in terms of a steady state condition. Buoyancy (gravity) driven convection interferes with the attainment of steady state, and the measurement of coupled processes. The MTP Experiment measures the kinetics of molecular migration that occurs in fluids, in response to the application of various driving potentials. Three separate driving potentials may be applied to the MTP Experiment fluids, either singly or in combination. The driving potentials include chemical potential, thermal potential, and electrical potential. Two separate fluid arrangements are used to study membrane mediated and bulk fluid transport phenomena. Transport processes of interest in membrane mediated systems include diffusion, osmosis, and streaming potential. Bulk fluid processes of interest include coupled phenomena such as the Soret Effect, Dufour Effect, Donnan Effect, and thermal diffusion potential. MTP Experiments are performed in the Microgravity Transport Apparatus (MTA), an instrument that has been developed specifically for precision measurement of transport processes. Experiment fluids are contained within the MTA fluid cells, designed to create a one dimensional flow geometry

  17. Studies of Novel Quantum Phenomena in Ruthenates

    SciTech Connect

    Mao, Zhiqiang

    2011-04-08

    Strongly correlated oxides have been the subject of intense study in contemporary condensed matter physics, and perovskite ruthenates (Sr,Ca)n+1RunO3n+1 have become a new focus in this field. One of important characteristics of ruthenates is that both lattice and orbital degrees of freedom are active and are strongly coupled to charge and spin degrees of freedom. Such a complex interplay of multiple degrees of freedom causes the properties of ruthenates to exhibit a gigantic response to external stimuli under certain circumstances. Magnetic field, pressure, and chemical composition all have been demonstrated to be effective in inducing electronic/magnetic phase transitions in ruthenates. Therefore, ruthenates are ideal candidates for searching for novel quantum phenomena through controlling external parameters. The objective of this project is to search for novel quantum phenomena in ruthenate materials using high-quality single crystals grown by the floating-zone technique, and investigate the underlying physics. The following summarizes our accomplishments. We have focused on trilayered Sr4Ru3O10 and bilayered (Ca1-xSrx)3Ru2O7. We have succeeded in growing high-quality single crystals of these materials using the floating-zone technique and performed systematic studies on their electronic and magnetic properties through a variety of measurements, including resistivity, Hall coefficient, angle-resolved magnetoresistivity, Hall probe microscopy, and specific heat. We have also studied microscopic magnetic properties for some of these materials using neutron scattering in collaboration with Los Alamos National Laboratory. We have observed a number of unusual exotic quantum phenomena through these studies, such as an orbital selective metamagnetic transition, bulk spin valve effect, and a heavy-mass nearly ferromagnetic state with a surprisingly large Wilson ratio. Our work has also revealed underlying physics of these exotic phenomena. Exotic phenomena of correlated

  18. Rod Driven Frequency Entrainment and Resonance Phenomena.

    PubMed

    Salchow, Christina; Strohmeier, Daniel; Klee, Sascha; Jannek, Dunja; Schiecke, Karin; Witte, Herbert; Nehorai, Arye; Haueisen, Jens

    2016-01-01

    A controversy exists on photic driving in the human visual cortex evoked by intermittent photic stimulation. Frequency entrainment and resonance phenomena are reported for frequencies higher than 12 Hz in some studies while missing in others. We hypothesized that this might be due to different experimental conditions, since both high and low intensity light stimulation were used. However, most studies do not report radiometric measurements, which makes it impossible to categorize the stimulation according to photopic, mesopic, and scotopic vision. Low intensity light stimulation might lead to scotopic vision, where rod perception dominates. In this study, we investigated photic driving for rod-dominated visual input under scotopic conditions. Twelve healthy volunteers were stimulated with low intensity light flashes at 20 stimulation frequencies, leading to rod activation only. The frequencies were multiples of the individual alpha frequency (α) of each volunteer in the range from 0.40 to 2.30(∗)α. Three hundred and six-channel whole head magnetoencephalography recordings were analyzed in time, frequency, and spatiotemporal domains with the Topographic Matching Pursuit algorithm. We found resonance phenomena and frequency entrainment for stimulations at or close to the individual alpha frequency (0.90-1.10(∗)α) and half of the alpha frequency (0.40-0.55(∗)α). No signs of resonance and frequency entrainment phenomena were revealed around 2.00(∗)α. Instead, on-responses at the beginning and off-responses at the end of each stimulation train were observed for the first time in a photic driving experiment at frequencies of 1.30-2.30(∗)α, indicating that the flicker fusion threshold was reached. All results, the resonance and entrainment as well as the fusion effects, provide evidence for rod-dominated photic driving in the visual cortex. PMID:27588002

  19. Simulating physical phenomena with a quantum computer

    NASA Astrophysics Data System (ADS)

    Ortiz, Gerardo

    2003-03-01

    In a keynote speech at MIT in 1981 Richard Feynman raised some provocative questions in connection to the exact simulation of physical systems using a special device named a ``quantum computer'' (QC). At the time it was known that deterministic simulations of quantum phenomena in classical computers required a number of resources that scaled exponentially with the number of degrees of freedom, and also that the probabilistic simulation of certain quantum problems were limited by the so-called sign or phase problem, a problem believed to be of exponential complexity. Such a QC was intended to mimick physical processes exactly the same as Nature. Certainly, remarks coming from such an influential figure generated widespread interest in these ideas, and today after 21 years there are still some open questions. What kind of physical phenomena can be simulated with a QC?, How?, and What are its limitations? Addressing and attempting to answer these questions is what this talk is about. Definitively, the goal of physics simulation using controllable quantum systems (``physics imitation'') is to exploit quantum laws to advantage, and thus accomplish efficient imitation. Fundamental is the connection between a quantum computational model and a physical system by transformations of operator algebras. This concept is a necessary one because in Quantum Mechanics each physical system is naturally associated with a language of operators and thus can be considered as a possible model of quantum computation. The remarkable result is that an arbitrary physical system is naturally simulatable by another physical system (or QC) whenever a ``dictionary'' between the two operator algebras exists. I will explain these concepts and address some of Feynman's concerns regarding the simulation of fermionic systems. Finally, I will illustrate the main ideas by imitating simple physical phenomena borrowed from condensed matter physics using quantum algorithms, and present experimental

  20. BWR core melt progression phenomena: Experimental analyses

    SciTech Connect

    Ott, L.J.

    1992-06-01

    In the BWR Core Melt in Progression Phenomena Program, experimental results concerning severe fuel damage and core melt progression in BWR core geometry are used to evaluate existing models of the governing phenomena. These include control blade eutectic liquefaction and the subsequent relocation and attack on the channel box structure; oxidation heating and hydrogen generation; Zircaloy melting and relocation; and the continuing oxidation of zirconium with metallic blockage formation. Integral data have been obtained from the BWR DF-4 experiment in the ACRR and from BWR tests in the German CORA exreactor fuel-damage test facility. Additional integral data will be obtained from new CORA BWR test, the full-length FLHT-6 BWR test in the NRU test reactor, and the new program of exreactor experiments at Sandia National Laboratories (SNL) on metallic melt relocation and blockage formation. an essential part of this activity is interpretation and use of the results of the BWR tests. The Oak Ridge National Laboratory (ORNL) has developed experiment-specific models for analysis of the BWR experiments; to date, these models have permitted far more precise analyses of the conditions in these experiments than has previously been available. These analyses have provided a basis for more accurate interpretation of the phenomena that the experiments are intended to investigate. The results of posttest analyses of BWR experiments are discussed and significant findings from these analyses are explained. The ORNL control blade/canister models with materials interaction, relocation and blockage models are currently being implemented in SCDAP/RELAP5 as an optional structural component.

  1. BWR core melt progression phenomena: Experimental analyses

    SciTech Connect

    Ott, L.J.

    1992-01-01

    In the BWR Core Melt in Progression Phenomena Program, experimental results concerning severe fuel damage and core melt progression in BWR core geometry are used to evaluate existing models of the governing phenomena. These include control blade eutectic liquefaction and the subsequent relocation and attack on the channel box structure; oxidation heating and hydrogen generation; Zircaloy melting and relocation; and the continuing oxidation of zirconium with metallic blockage formation. Integral data have been obtained from the BWR DF-4 experiment in the ACRR and from BWR tests in the German CORA exreactor fuel-damage test facility. Additional integral data will be obtained from new CORA BWR test, the full-length FLHT-6 BWR test in the NRU test reactor, and the new program of exreactor experiments at Sandia National Laboratories (SNL) on metallic melt relocation and blockage formation. an essential part of this activity is interpretation and use of the results of the BWR tests. The Oak Ridge National Laboratory (ORNL) has developed experiment-specific models for analysis of the BWR experiments; to date, these models have permitted far more precise analyses of the conditions in these experiments than has previously been available. These analyses have provided a basis for more accurate interpretation of the phenomena that the experiments are intended to investigate. The results of posttest analyses of BWR experiments are discussed and significant findings from these analyses are explained. The ORNL control blade/canister models with materials interaction, relocation and blockage models are currently being implemented in SCDAP/RELAP5 as an optional structural component.

  2. Fast Particle Methods for Multiscale Phenomena Simulations

    NASA Technical Reports Server (NTRS)

    Koumoutsakos, P.; Wray, A.; Shariff, K.; Pohorille, Andrew

    2000-01-01

    We are developing particle methods oriented at improving computational modeling capabilities of multiscale physical phenomena in : (i) high Reynolds number unsteady vortical flows, (ii) particle laden and interfacial flows, (iii)molecular dynamics studies of nanoscale droplets and studies of the structure, functions, and evolution of the earliest living cell. The unifying computational approach involves particle methods implemented in parallel computer architectures. The inherent adaptivity, robustness and efficiency of particle methods makes them a multidisciplinary computational tool capable of bridging the gap of micro-scale and continuum flow simulations. Using efficient tree data structures, multipole expansion algorithms, and improved particle-grid interpolation, particle methods allow for simulations using millions of computational elements, making possible the resolution of a wide range of length and time scales of these important physical phenomena.The current challenges in these simulations are in : [i] the proper formulation of particle methods in the molecular and continuous level for the discretization of the governing equations [ii] the resolution of the wide range of time and length scales governing the phenomena under investigation. [iii] the minimization of numerical artifacts that may interfere with the physics of the systems under consideration. [iv] the parallelization of processes such as tree traversal and grid-particle interpolations We are conducting simulations using vortex methods, molecular dynamics and smooth particle hydrodynamics, exploiting their unifying concepts such as : the solution of the N-body problem in parallel computers, highly accurate particle-particle and grid-particle interpolations, parallel FFT's and the formulation of processes such as diffusion in the context of particle methods. This approach enables us to transcend among seemingly unrelated areas of research.

  3. Displacement phenomena in lectin affinity chromatography.

    PubMed

    Cho, Wonryeon

    2015-10-01

    The work described here examines displacement phenomena that play a role in lectin affinity chromatography and their potential to impact reproducibility. This was achieved using Lycopersicon esculentum lectin (LEL), a lectin widely used in monitoring cancer. Four small identical LEL columns were coupled in series to form a single affinity chromatography system with the last in the series connected to an absorbance detector. The serial affinity column set (SACS) was then loaded with human plasma proteins. At the completion of loading, the column set was disassembled, the four columns were eluted individually, the captured proteins were trypsin digested, the peptides were deglycosylated with PNGase F, and the parent proteins were identified through mass spectral analyses. Significantly different sets of glycoproteins were selected by each column, some proteins appearing to be exclusively bound to the first column while others were bound further along in the series. Clearly, sample displacement chromatography (SDC) occurs. Glycoproteins were bound at different places in the column train, identifying the presence of glycoforms with different affinity on a single glycoprotein. It is not possible to see these phenomena in the single column mode of chromatography. Moreover, low abundance proteins were enriched, which facilitates detection. The great advantage of this method is that it differentiates between glycoproteins on the basis of their binding affinity. Displacement phenomena are concluded to be a significant component of the separation mechanism in heavily loaded lectin affinity chromatography columns. This further suggests that care must be exercised in sample loading of lectin columns to prevent analyte displacement with nonretained proteins. PMID:26348026

  4. Rod Driven Frequency Entrainment and Resonance Phenomena

    PubMed Central

    Salchow, Christina; Strohmeier, Daniel; Klee, Sascha; Jannek, Dunja; Schiecke, Karin; Witte, Herbert; Nehorai, Arye; Haueisen, Jens

    2016-01-01

    A controversy exists on photic driving in the human visual cortex evoked by intermittent photic stimulation. Frequency entrainment and resonance phenomena are reported for frequencies higher than 12 Hz in some studies while missing in others. We hypothesized that this might be due to different experimental conditions, since both high and low intensity light stimulation were used. However, most studies do not report radiometric measurements, which makes it impossible to categorize the stimulation according to photopic, mesopic, and scotopic vision. Low intensity light stimulation might lead to scotopic vision, where rod perception dominates. In this study, we investigated photic driving for rod-dominated visual input under scotopic conditions. Twelve healthy volunteers were stimulated with low intensity light flashes at 20 stimulation frequencies, leading to rod activation only. The frequencies were multiples of the individual alpha frequency (α) of each volunteer in the range from 0.40 to 2.30∗α. Three hundred and six-channel whole head magnetoencephalography recordings were analyzed in time, frequency, and spatiotemporal domains with the Topographic Matching Pursuit algorithm. We found resonance phenomena and frequency entrainment for stimulations at or close to the individual alpha frequency (0.90–1.10∗α) and half of the alpha frequency (0.40–0.55∗α). No signs of resonance and frequency entrainment phenomena were revealed around 2.00∗α. Instead, on-responses at the beginning and off-responses at the end of each stimulation train were observed for the first time in a photic driving experiment at frequencies of 1.30–2.30∗α, indicating that the flicker fusion threshold was reached. All results, the resonance and entrainment as well as the fusion effects, provide evidence for rod-dominated photic driving in the visual cortex. PMID:27588002

  5. Electronic phenomena in adsorption and catalysis

    SciTech Connect

    Kiselev, V.F.; Krylov, O.V.

    1987-01-01

    This book is the second of a three-volume treatment prepared by a physicist and a chemist, who took a common standpoint in considering the close relationship between the electronic processes taking place on the semiconductor-dielectric interface on the one hand, and the adsorptive and catalytic phenomena on the other. This volume brings together, and generalizes, a vast bulk of knowledge on the nature of surface and interface states, on the mechanism of surface electronic processes in semiconductors, as well as considers ways of controlling these processes. In addition, the authors discuss plausible mechanisms of elementary acts in surface charging during adsorption and catalysis.

  6. Phenomena and Parameters Important to Burnup Credit

    SciTech Connect

    Parks, C.V.

    2001-01-10

    Since the mid-1980s, a significant number of studies have been directed at understanding the phenomena and parameters important to implementation of burnup credit in out-of-reactor applications involving pressurized-water-reactor (PWR) spent fuel. The efforts directed at burnup credit involving boiling-water-reactor (BWR) spent fuel have been more limited. This paper reviews the knowledge and experience gained from work performed in the US and other countries in the study of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation for transport and dry cask storage is given.

  7. Complex Synchronization Phenomena in Ecological Systems

    NASA Astrophysics Data System (ADS)

    Stone, Lewi; Olinky, Ronen; Blasius, Bernd; Huppert, Amit; Cazelles, Bernard

    2002-07-01

    Ecological and biological systems provide us with many striking examples of synchronization phenomena. Here we discuss a number of intriguing cases and attempt to explain them taking advantage of a modelling framework. One main focus will concern synchronized ecological end epidemiological cycles which have Uniform Phase growth associated with their regular recurrence, and Chaotic Amplitudes - a feature we term UPCA. Examples come from different areas and include decadal cycles of small mammals, recurrent viral epidemics such as childhood infections (eg., measles), and seasonally driven phytoplankton blooms observed in lakes and the oceans. A more detailed theoretical analysis of seasonally synchronized chaotic population cycles is presented.

  8. Paramagnetic Meissner effect and related dynamical phenomena

    NASA Astrophysics Data System (ADS)

    Li, Mai Suan

    2003-03-01

    The hallmark of superconductivity is the diamagnetic response to external magnetic field. In striking contrast to this behavior, a paramagnetic response or paramagnetic Meissner effect was observed in ceramic high- Tc and in conventional superconductors. The present review is given on this interesting effect and related phenomena. We begin with a detailed discussion of experimental results on the paramagnetic Meissner effect in both granular and conventional superconductors. There are two main mechanisms leading to the paramagnetic response: the so-called d-wave and the flux compression. In the first scenario, the Josephson critical current between two d-wave superconductors becomes negative or equivalently one has a π junction. The paramagnetic signal occurs due to the nonzero spontaneous supercurrent circulating in a loop consisting of odd number of π junctions. In addition to the d-wave mechanism we present the flux compression mechanism for the paramagnetic Meissner effect. The compression may be due to either an inhomogeneous superconducting transition or flux trap inside the giant vortex state. The flux trapping which acts like a total nonzero spontaneous magnetic moment causes the paramagnetic signal. The anisotropic pairing scenario is believed to be valid for granular materials while the flux trap one can be applied to both conventional and high- Tc superconductors. The study of different phenomena by a three-dimensional lattice model of randomly distributed π Josephson junctions with finite self-inductance occupies the main part of our review. By simulations one can show that the chiral glass phase in which chiralities are frozen in time and in space may occur in granular superconductors possessing d-wave pairing symmetry. Experimental attempts on the search for the chiral glass phase are analysed. Experiments on dynamical phenomena such as AC susceptibility, compensation effect, anomalous microwave absorption, aging effect, AC resistivity and

  9. General unifying features of controlled quantum phenomena

    SciTech Connect

    Pechen, Alexander; Brif, Constantin; Wu, Rebing; Chakrabarti, Raj; Rabitz, Herschel

    2010-09-15

    Many proposals have been put forth for controlling quantum phenomena, including open-loop, adaptive feedback, and real-time feedback control. Each of these approaches has been viewed as operationally, and even physically, distinct from the others. This work shows that all such scenarios inherently share the same fundamental control features residing in the topology of the landscape relating the target physical observable to the applied controls. This unified foundation may provide a basis for development of hybrid control schemes that would combine the advantages of the existing approaches to achieve the best overall performance.

  10. Observations of cometary plasma wave phenomena

    NASA Technical Reports Server (NTRS)

    Scarf, F. L.; Coroniti, F. V.; Kennel, C. F.; Gurnett, D. A.; Ip, W.-H.; Smith, E. J.

    1986-01-01

    The ICE plasma wave investigation utilized very long electric antennas (100 m tip-to-tip) and a very high sensitivity magnetic search coil to obtain significant local information on plasma physics phenomena occurring in the distant pickup regions of Comet Giacobini-Zinner and Comet Halley; and information on the processes that developed in the coma and tail of Giacobini-Zinner. The ICE plasma wave measurements associated with both comet encounters are summarized, and high sensitivity ICE observations are related to corresponding measurements from the other Halley spacecraft.

  11. Quenching phenomena in natural circulation loop

    SciTech Connect

    Umekawa, Hisashi; Ozawa, Mamoru; Ishida, Naoki

    1995-09-01

    Quenching phenomena has been investigated experimentally using circulation loop of liquid nitrogen. During the quenching under natural circulation, the heat transfer mode changes from film boiling to nucleate boiling, and at the same time flux changes with time depending on the vapor generation rate and related two-phase flow characteristics. Moreover, density wave oscillations occur under a certain operating condition, which is closely related to the dynamic behavior of the cooling curve. The experimental results indicates that the occurrence of the density wave oscillation induces the deterioration of effective cooling of the heat surface in the film and the transition boiling regions, which results in the decrease in the quenching velocity.

  12. On periodicity of solar wind phenomena

    NASA Technical Reports Server (NTRS)

    Verma, V. K.; Joshi, G. C.

    1995-01-01

    We have investigated the rate of occurrence of solar wind phenomena observed between 1972-1984 using power spectrum analysis. The data have been taken from the high speed solar wind (HSSW) streams catalogue published by Mavromichalaki et al. (1988). The power spectrum analysis of HSSW events indicate that HSSW stream events have a periodicity of 9 days. This periodicity of HSSW events is 1/3 of the 27 days period of coronal holes which are the major source of solar wind events. In our opinion the 9 days period may be the energy build up time to produce the HSSW stream events.

  13. Monte Carlo analysis of magnetic aftereffect phenomena

    NASA Astrophysics Data System (ADS)

    Andrei, Petru; Stancu, Alexandru

    2006-04-01

    Magnetic aftereffect phenomena are analyzed by using the Monte Carlo technique. This technique has the advantage that it can be applied to any model of hysteresis. It is shown that a log t-type dependence of the magnetization can be qualitatively predicted even in the framework of hysteresis models with local history, such as the Jiles-Atherton model. These models are computationally much more efficient than the models with global history such as the Preisach model. Numerical results related to the decay of the magnetization as of function of time, as well as to the viscosity coefficient, are presented.

  14. Quantum Phenomena Tested By Neutron Interferometry

    SciTech Connect

    Rauch, Helmut

    2005-02-15

    Entanglement of two photons, or atoms is a complementary situation to a double slit situation of a single photon, neutron or atom. With neutrons single particle interference phenomena can be observed and the 'entanglement of degrees of freedom', i.e. contextuality can be verified. In this respect, neutrons are proper tools for testing quantum mechanics because they are massive, they couple to electromagnetic fields due to their magnetic moment and they are subject to all basic interactions, and they are sensitive to topological effects, as well. Related experiments will be discussed. Deterministic and stochastic partial absorption experiments can be described by Bell-type inequalities. Recent neutron interferometry experiments based on postselection methods renewed the discussion about quantum nonlocality and the quantum measuring process. It has been shown that interference phenomena can be revived even when the overall interference pattern has lost its contrast. This indicates a persisting coupling in phase space even in cases of spatially separated Schroedinger cat-like situations. These states are extremely fragile and sensitive against any kind of fluctuations and other decoherence processes. More complete quantum experiments also show that a complete retrieval of quantum states behind an interaction volume becomes impossible in principle.

  15. Effects of electrostatic correlations on electrokinetic phenomena.

    PubMed

    Storey, Brian D; Bazant, Martin Z

    2012-11-01

    The classical theory of electrokinetic phenomena is based on the mean-field approximation that the electric field acting on an individual ion is self-consistently determined by the local mean charge density. This paper considers situations, such as concentrated electrolytes, multivalent electrolytes, or solvent-free ionic liquids, where the mean-field approximation breaks down. A fourth-order modified Poisson equation is developed that captures the essential features in a simple continuum framework. The model is derived as a gradient approximation for nonlocal electrostatics of interacting effective charges, where the permittivity becomes a differential operator, scaled by a correlation length. The theory is able to capture subtle aspects of molecular simulations and allows for simple calculations of electrokinetic flows in correlated ionic fluids. Charge-density oscillations tend to reduce electro-osmotic flow and streaming current, and overscreening of surface charge can lead to flow reversal. These effects also help to explain the suppression of induced-charge electrokinetic phenomena at high salt concentrations. PMID:23214872

  16. Auroral Phenomena in Brown Dwarf Atmospheres

    NASA Astrophysics Data System (ADS)

    Pineda, J. Sebastian; Hallinan, Gregg

    2016-01-01

    Since the unexpected discovery of radio emission from brown dwarfs some 15 years ago, investigations into the nature of this emission have revealed that, despite their cool and neutral atmospheres, brown dwarfs harbor strong kG magnetic fields, but unlike the warmer stellar objects, they generate highly circularly polarized auroral radio emission, like the giant planets of the Solar System. Our recent results from Keck LRIS monitoring of the brown dwarf LSR1835+32 definitively confirm this picture by connecting the auroral radio emission to spectroscopic variability at optical wavelengths as coherent manifestations of strong large-scale magnetospheric auroral current systems. I present some of the results of my dissertation work to understand the nature brown dwarf auroral phenomena. My efforts include a survey of Late L dwarfs and T dwarfs, looking for auroral Hα emission and a concurrent survey looking for the auroral emission of H3+ from brown dwarfs with radio pulse detections. I discuss the potential connection of this auroral activity to brown dwarf weather phenomena and how brown dwarf aurorae may differ from the analogous emission of the magnetized giant planets in the Solar System.

  17. Animal network phenomena: insights from triadic games

    NASA Astrophysics Data System (ADS)

    Mesterton-Gibbons, Mike; Sherratt, Tom N.

    Games of animal conflict in networks rely heavily on computer simulation because analysis is difficult, the degree of difficulty increasing sharply with the size of the network. For this reason, virtually the entire analytical literature on evolutionary game theory has assumed either dyadic interaction or a high degree of symmetry, or both. Yet we cannot rely exclusively on computer simulation in the study of any complex system. So the study of triadic interactions has an important role to play, because triads are both the simplest groups in which asymmetric network phenomena can be studied and the groups beyond dyads in which analysis of population games is most likely to be tractable, especially when allowing for intrinsic variation. Here we demonstrate how such analyses can illuminate a variety of behavioral phenomena within networks, including coalition formation, eavesdropping (the strategic observation of contests between neighbors) and victory displays (which are performed by the winners of contests but not by the losers). In particular, we show that eavesdropping acts to lower aggression thresholds compared to games without it, and that victory displays to bystanders will be most intense when there is little difference in payoff between dominating an opponent and not subordinating.

  18. Physical phenomena and the microgravity response

    NASA Technical Reports Server (NTRS)

    Todd, Paul

    1989-01-01

    The living biological cell is not a sack of Newtonian fluid containing systems of chemical reactions at equilibrium. It is a kinetically driven system, not a thermodynamically driven system. While the cell as a whole might be considered isothermal, at the scale of individual macromolecular events there is heat generated, and presumably sharp thermal gradients exist at the submicron level. Basic physical phenomena to be considered when exploring the cell's response to inertial acceleration include particle sedimentation, solutal convection, motility electrokinetics, cytoskeletal work, and hydrostatic pressure. Protein crystal growth experiments, for example, illustrate the profound effects of convection currents on macromolecular assembly. Reaction kinetics in the cell vary all the way from diffusion-limited to life-time limited. Transport processes vary from free diffusion, to facilitated and active transmembrane transport, to contractile-protein-driven motility, to crystalline immobilization. At least four physical states of matter exist in the cell: aqueous, non-aqueous, immiscible-aqueous, and solid. Levels of order vary from crystalline to free solution. The relative volumes of these states profoundly influence the cell's response to inertial acceleration. Such subcellular phenomena as stretch-receptor activation, microtubule re-assembly, synaptic junction formation, chemotactic receptor activation, and statolith sedimentation were studied recently with respect to both their basic mechanisms and their responsiveness to inertial acceleration. From such studies a widespread role of cytoskeletal organization is becoming apparent.

  19. WHC natural phenomena hazards mitigation implementation plan

    SciTech Connect

    Conrads, T.J.

    1996-09-11

    Natural phenomena hazards (NPH) are unexpected acts of nature which pose a threat or danger to workers, the public or to the environment. Earthquakes, extreme winds (hurricane and tornado),snow, flooding, volcanic ashfall, and lightning strike are examples of NPH at Hanford. It is the policy of U.S. Department of Energy (DOE) to design, construct and operate DOE facilitiesso that workers, the public and the environment are protected from NPH and other hazards. During 1993 DOE, Richland Operations Office (RL) transmitted DOE Order 5480.28, ``Natural Phenomena Hazards Mitigation,`` to Westinghouse Hanford COmpany (WHC) for compliance. The Order includes rigorous new NPH criteria for the design of new DOE facilities as well as for the evaluation and upgrade of existing DOE facilities. In 1995 DOE issued Order 420.1, ``Facility Safety`` which contains the same NPH requirements and invokes the same applicable standards as Order 5480.28. It will supersede Order 5480.28 when an in-force date for Order 420.1 is established through contract revision. Activities will be planned and accomplished in four phases: Mobilization; Prioritization; Evaluation; and Upgrade. The basis for the graded approach is the designation of facilities/structures into one of five performance categories based upon safety function, mission and cost. This Implementation Plan develops the program for the Prioritization Phase, as well as an overall strategy for the implemention of DOE Order 5480.2B.

  20. Wireless, in-vessel neutron monitor for initial core-loading of advanced breeder reactors

    NASA Technical Reports Server (NTRS)

    Delorenzo, J. T.; Kennedy, E. J.; Blalock, T. V.; Rochelle, J. M.; Chiles, M. M.; Valentine, K. H.

    1981-01-01

    An experimental wireless, in-vessel neutron monitor was developed to measure the reactivity of an advanced breeder reactor as the core is loaded for the first time to preclude an accidental critically incident. The environment is liquid sodium at a temperature of approx. 220 C, with negligible gamma or neutron radiation. With ultrasonic transmission of neutron data, no fundamental limitation was observed after tests at 230 C for 2000 h. The neutron sensitivity was approx. 1 count/s-nv, and the potential data transmission rate was approx. 10,000 counts/s.

  1. Mathematical methods of studying physical phenomena

    NASA Astrophysics Data System (ADS)

    Man'ko, Margarita A.

    2013-03-01

    In recent decades, substantial theoretical and experimental progress was achieved in understanding the quantum nature of physical phenomena that serves as the foundation of present and future quantum technologies. Quantum correlations like the entanglement of the states of composite systems, the phenomenon of quantum discord, which captures other aspects of quantum correlations, quantum contextuality and, connected with these phenomena, uncertainty relations for conjugate variables and entropies, like Shannon and Rényi entropies, and the inequalities for spin states, like Bell inequalities, reflect the recently understood quantum properties of micro and macro systems. The mathematical methods needed to describe all quantum phenomena mentioned above were also the subject of intense studies in the end of the last, and beginning of the new, century. In this section of CAMOP 'Mathematical Methods of Studying Physical Phenomena' new results and new trends in the rapidly developing domain of quantum (and classical) physics are presented. Among the particular topics under discussion there are some reviews on the problems of dynamical invariants and their relations with symmetries of the physical systems. In fact, this is a very old problem of both classical and quantum systems, e.g. the systems of parametric oscillators with time-dependent parameters, like Ermakov systems, which have specific constants of motion depending linearly or quadratically on the oscillator positions and momenta. Such dynamical invariants play an important role in studying the dynamical Casimir effect, the essence of the effect being the creation of photons from the vacuum in a cavity with moving boundaries due to the presence of purely quantum fluctuations of the electromagnetic field in the vacuum. It is remarkable that this effect was recently observed experimentally. The other new direction in developing the mathematical approach in physics is quantum tomography that provides a new vision of

  2. Microgravity Transport Phenomena Experiment (MTPE) Overview

    NASA Technical Reports Server (NTRS)

    Mason, Larry W.

    1999-01-01

    The Microgravity Transport Phenomena Experiment (MTPE) is a fluids experiment supported by the Fundamentals in Biotechnology program in association with the Human Exploration and Development of Space (BEDS) initiative. The MTP Experiment will investigate fluid transport phenomena both in ground based experiments and in the microgravity environment. Many fluid transport processes are affected by gravity. Osmotic flux kinetics in planar membrane systems have been shown to be influenced by gravimetric orientation, either through convective mixing caused by unstably stratified fluid layers, or through a stable fluid boundary layer structure that forms in association with the membrane. Coupled transport phenomena also show gravity related effects. Coefficients associated with coupled transport processes are defined in terms of a steady state condition. Buoyancy (gravity) driven convection interferes with the attainment of steady state, and the measurement of coupled processes. The MTP Experiment measures the kinetics of molecular migration that occurs in fluids, in response to the application of various driving potentials. Three separate driving potentials may be applied to the MTP Experiment fluids, either singly or in combination. The driving potentials include chemical potential, thermal potential, and electrical potential. Two separate fluid arrangements are used to study membrane mediated and bulk fluid transport phenomena. Transport processes of interest in membrane mediated systems include diffusion, osmosis, and streaming potential. Bulk fluid processes of interest include coupled phenomena such as the Soret Effect, Dufour Effect, Donnan Effect, and thermal diffusion potential. MTP Experiments are performed in the Microgravity Transport Apparatus (MTA), an instrument that has been developed specifically for precision measurement of transport processes. Experiment fluids are contained within the MTA fluid cells, designed to create a one dimensional flow geometry

  3. Numerical Evaluation of Fluid Mixing Phenomena in Boiling Water Reactor Using Advanced Interface Tracking Method

    NASA Astrophysics Data System (ADS)

    Yoshida, Hiroyuki; Takase, Kazuyuki

    Thermal-hydraulic design of the current boiling water reactor (BWR) is performed with the subchannel analysis codes which incorporated the correlations based on empirical results including actual-size tests. Then, for the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) core, an actual size test of an embodiment of its design is required to confirm or modify such correlations. In this situation, development of a method that enables the thermal-hydraulic design of nuclear reactors without these actual size tests is desired, because these tests take a long time and entail great cost. For this reason, we developed an advanced thermal-hydraulic design method for FLWRs using innovative two-phase flow simulation technology. In this study, a detailed Two-Phase Flow simulation code using advanced Interface Tracking method: TPFIT is developed to calculate the detailed information of the two-phase flow. In this paper, firstly, we tried to verify the TPFIT code by comparing it with the existing 2-channel air-water mixing experimental results. Secondary, the TPFIT code was applied to simulation of steam-water two-phase flow in a model of two subchannels of a current BWRs and FLWRs rod bundle. The fluid mixing was observed at a gap between the subchannels. The existing two-phase flow correlation for fluid mixing is evaluated using detailed numerical simulation data. This data indicates that pressure difference between fluid channels is responsible for the fluid mixing, and thus the effects of the time average pressure difference and fluctuations must be incorporated in the two-phase flow correlation for fluid mixing. When inlet quality ratio of subchannels is relatively large, it is understood that evaluation precision of the existing two-phase flow correlations for fluid mixing are relatively low.

  4. Assessment and selection of materials for ITER in-vessel components

    NASA Astrophysics Data System (ADS)

    Kalinin, G.; Barabash, V.; Cardella, A.; Dietz, J.; Ioki, K.; Matera, R.; Santoro, R. T.; Tivey, R.; ITER Home Teams

    2000-12-01

    During the international thermonuclear experimental reactor (ITER) engineering design activities (EDA) significant progress has been made in the selection of materials for the in-vessel components of the reactor. This progress is a result of the worldwide collaboration of material scientists and industries which focused their effort on the optimisation of material and component manufacturing and on the investigation of the most critical material properties. Austenitic stainless steels 316L(N)-IG and 316L, nickel-based alloys Inconel 718 and Inconel 625, Ti-6Al-4V alloy and two copper alloys, CuCrZr-IG and CuAl25-IG, have been proposed as reference structural materials, and ferritic steel 430, and austenitic steel 304B7 with the addition of boron have been selected for some specific parts of the ITER in-vessel components. Beryllium, tungsten and carbon fibre composites are considered as plasma facing armour materials. The data base on the properties of all these materials is critically assessed and briefly reviewed in this paper together with the justification of the material selection (e.g., effect of neutron irradiation on the mechanical properties of materials, effect of manufacturing cycle, etc.).

  5. Topological Spintronics: Materials, Phenomena and Devices

    NASA Astrophysics Data System (ADS)

    Samarth, Nitin

    2015-03-01

    The two-dimensional surface states of three-dimensional topological insulators such as Bi2Se3and(Bi,Sb)2Te3 possess a spin texture that can potentially be exploited for spintronics applications. We provide a perspective on the emergence of ``topological spintronics,'' demonstrating how this spin texture can be engineered using either quantum tunneling between surfaces or by breaking time-reversal symmetry. We then discuss recent experiments that show striking spintronic phenomena useful for proof-of-concept devices, including a spin-orbit torque of record efficiency at room temperature and an electrically-gated ``giant anisotropic magnetoresistance'' at low temperature. This work was carried out in collaboration with A. Richardella, S.-Y. Xu, M. Neupane, A. Mellnik, A. Kandala, J. S. Lee, D. M. Zhang, M. Z. Hasan and D. C. Ralph. We acknowledge funding from the DARPA Meso program, ONR and C-SPIN (under sponsorship of MARCO and DARPA).

  6. Transient Phenomena: Opportunities for New Discoveries

    NASA Astrophysics Data System (ADS)

    Lazio, T. Joseph W.

    Known classes of radio wavelength transients range from the nearby (stellar flares and radio pulsars) to the distant Universe (γ-ray burst afterglows). Hypothesized classes of radio transients include analogs of known objects, such as extrasolar planets emitting Jovian-like radio bursts and giant-pulse emitting pulsars in other galaxies, to the exotic, such as prompt emission from γ-ray bursts, evaporating black holes and transmitters from other civilizations. Time domain astronomy has been recognized internationally as a means of addressing key scientific questions in astronomy and physics, and pathfinders and Precursors to the Square Kilometre Array (SKA) are beginning to offer a combination of wider fields of view and more wavelength agility than has been possible in the past. These improvements will continue when the SKA itself becomes operational. I illustrate the range of transient phenomena and discuss how the detection and study of radio transients will improve immensely.

  7. Transient Phenomena: Opportunities for New Discoveries

    NASA Technical Reports Server (NTRS)

    Lazio, T. Joseph W.

    2010-01-01

    Known classes of radio wavelength transients range from the nearby (stellar flares and radio pulsars) to the distant Universe (gamma-ray burst afterglows). Hypothesized classes of radio transients include analogs of known objects, such as extrasolar planets emitting Jovian-like radio bursts and giant-pulse emitting pulsars in other galaxies, to the exotic, such as prompt emission from gamma-ray bursts, evaporating black holes and transmitters from other civilizations. Time domain astronomy has been recognized internationally as a means of addressing key scientific questions in astronomy and physics, and pathfinders and Precursors to the Square Kilometre Array (SKA) are beginning to offer a combination of wider fields of view and more wavelength agility than has been possible in the past. These improvements will continue when the SKA itself becomes operational. I illustrate the range of transient phenomena and discuss how the detection and study of radio transients will improve immensely.

  8. Pump instability phenomena generated by fluid forces

    NASA Technical Reports Server (NTRS)

    Gopalakrishnan, S.

    1985-01-01

    Rotor dynamic behavior of high energy centrifugal pumps is significantly affected by two types of fluid forces; one due to the hydraulic interaction of the impeller with the surrounding volute or diffuser and the other due to the effect of the wear rings. The available data on these forces is first reviewed. A simple one degree-of-freedom system containing these forces is analytically solved to exhibit the rotor dynamic effects. To illustrate the relative magnitude of these phenomena, an example of a multistage boiler feed pump is worked out. It is shown that the wear ring effects tend to suppress critical speed and postpone instability onset. But the volute-impeller forces tend to lower the critical speed and the instability onset speed. However, for typical boiler feed pumps under normal running clearances, the wear ring effects are much more significant than the destabilizing hydraulic interaction effects.

  9. Using Spatial Gradients to Model Localization Phenomena

    SciTech Connect

    D.J.Bammann; D.Mosher; D.A.Hughes; N.R.Moody; P.R.Dawson

    1999-07-01

    We present the final report on a Laboratory-Directed Research and Development project, Using Spatial Gradients to Model Localization Phenomena, performed during the fiscal years 1996 through 1998. The project focused on including spatial gradients in the temporal evolution equations of the state variables that describe hardening in metal plasticity models. The motivation was to investigate the numerical aspects associated with post-bifurcation mesh dependent finite element solutions in problems involving damage or crack propagation as well as problems in which strain Localizations occur. The addition of the spatial gradients introduces a mathematical length scale that eliminates the mesh dependency of the solution. In addition, new experimental techniques were developed to identify the physical mechanism associated with the numerical length scale.

  10. Lunar orbital photography of astronomical phenomena.

    NASA Technical Reports Server (NTRS)

    Mercer, R. D.; Dunkelman, L.; Ross, C. L.; Worden, A.

    1972-01-01

    This paper reports further progress on photography of faint astronomical and geophysical phenomena accomplished during the recent Apollo missions. Command module pilots have been able to photograph such astronomical objects as the solar corona, zodiacal light-corona transition region, lunar libration region, and portions of the Milky Way. The methods utilized for calibration of the film by adaptation of the High Altitude Observatory sensitometer are discussed. Kodak 2485 high-speed recording film was used in both 35-mm and 70-mm formats. The cameras used were Nikon f/1.2 55-mm focal length and Hasselblad f/2.8 80-mm focal length. Preflight and postflight calibration exposures were included on both the flight and control films, corresponding to luminances extending from the inner solar corona to as faint as 1/10 of the luminance of the light of the night sky. The photographs obtained from unique vantage points available during lunar orbit are discussed.

  11. Electron Acceleration by Transient Ion Foreshock Phenomena

    NASA Astrophysics Data System (ADS)

    Wilson, L. B., III; Turner, D. L.

    2015-12-01

    Particle acceleration is a topic of considerable interest in space, laboratory, and astrophysical plasmas as it is a fundamental physical process to all areas of physics. Recent THEMIS [e.g., Turner et al., 2014] and Wind [e.g., Wilson et al., 2013] observations have found evidence for strong particle acceleration at macro- and meso-scale structures and/or pulsations called transient ion foreshock phenomena (TIFP). Ion acceleration has been extensively studied, but electron acceleration has received less attention. Electron acceleration can arise from fundamentally different processes than those affecting ions due to differences in their gyroradii. Electron acceleration is ubiquitous, occurring in the solar corona (e.g., solar flares), magnetic reconnection, at shocks, astrophysical plasmas, etc. We present new results analyzing the dependencies of electron acceleration on the properties of TIFP observed by the THEMIS spacecraft.

  12. Teaching wave phenomena via biophysical applications

    NASA Astrophysics Data System (ADS)

    Reich, Daniel; Robbins, Mark; Leheny, Robert; Wonnell, Steven

    2014-03-01

    Over the past several years we have developed a two-semester second-year physics course sequence for students in the biosciences, tailored in part to the needs of undergraduate biophysics majors. One semester, ``Biological Physics,'' is based on the book of that name by P. Nelson. This talk will focus largely on the other semester, ``Wave Phenomena with Biophysical Applications,'' where we provide a novel introduction to the physics of waves, primarily through the study of experimental probes used in the biosciences that depend on the interaction of electromagnetic radiation with matter. Topic covered include: Fourier analysis, sound and hearing, diffraction - culminating in an analysis of x-ray fiber diffraction and its use in the determination of the structure of DNA - geometrical and physical optics, the physics of modern light microscopy, NMR and MRI. Laboratory exercises tailored to this course will also be described.

  13. Critical and resonance phenomena in neural networks

    NASA Astrophysics Data System (ADS)

    Goltsev, A. V.; Lopes, M. A.; Lee, K.-E.; Mendes, J. F. F.

    2013-01-01

    Brain rhythms contribute to every aspect of brain function. Here, we study critical and resonance phenomena that precede the emergence of brain rhythms. Using an analytical approach and simulations of a cortical circuit model of neural networks with stochastic neurons in the presence of noise, we show that spontaneous appearance of network oscillations occurs as a dynamical (non-equilibrium) phase transition at a critical point determined by the noise level, network structure, the balance between excitatory and inhibitory neurons, and other parameters. We find that the relaxation time of neural activity to a steady state, response to periodic stimuli at the frequency of the oscillations, amplitude of damped oscillations, and stochastic fluctuations of neural activity are dramatically increased when approaching the critical point of the transition.

  14. Equatorial phenomena in neutral thermospheric composition.

    NASA Technical Reports Server (NTRS)

    Reber, C. A.; Hedin, A. E.; Chandra, S.

    1973-01-01

    Several interesting phenomena relating to the equatorial ionosphere have been observed in the data from the OGO-6 mass spectrometer. The diurnal variations during equinox at an altitude of 450 km show the N2 and O densities peaking near 1500 hr while He peaks near 1000 hr. The latitudinal variation in N2 during the day is very similar to the F-region electron density exhibiting the well known features of the ionospheric anomaly. During periods of intense geomagnetic disturbance (e.g. the large storm of 8 March 1970), the low latitude thermospheric temperature increases on the order of 50-150 K, while at mid latitudes, increases of more than 1000 K are observed.

  15. Hadronic and nuclear phenomena in quantum chromodynamics

    SciTech Connect

    Brodsky, S.J.

    1987-06-01

    Many of the key issues in understanding quantum chromodynamics involves processes at intermediate energies. We discuss a range of hadronic and nuclear phenomena - exclusive processes, color transparency, hidden color degrees of freedom in nuclei, reduced nuclear amplitudes, jet coalescence, formation zone effects, hadron helicity selection rules, spin correlations, higher twist effects, and nuclear diffraction - as tools for probing hadron structure and the propagation of quark and gluon jets in nuclei. Many of these processes can be studied in electroproduction, utilizing internal targets in storage rings. We also review several areas where there has been significant theoretical progress in determining the form of hadron and nuclear wavefunctions, including QCD sum rules, lattice gauge theory, and discretized light-cone quantization. 98 refs., 40 figs., 2 tabs.

  16. Surfactant-based critical phenomena in microgravity

    NASA Technical Reports Server (NTRS)

    Kaler, Eric W.; Paulaitis, Michael E.

    1994-01-01

    The objective of this research project is to characterize by experiment and theoretically both the kinetics of phase separation and the metastable structures produced during phase separation in a microgravity environment. The particular systems we are currently studying are mixtures of water, nonionic surfactants, and compressible supercritical fluids at temperatures and pressures where the coexisting liquid phases have equal densities (isopycnic phases). In this report, we describe experiments to locate equilibrium isopycnic phases and to determine the 'local' phase behavior and critical phenomena at nearby conditions of temperature, pressure, and composition. In addition, we report the results of preliminary small angle neutron scattering (SANS) experiments to characterize microstructures that exist in these mixtures at different fluid densities.

  17. Analysis of oblique hypervelocity impact phenomena

    NASA Technical Reports Server (NTRS)

    Schonberg, William P.; Taylor, Roy A.

    1988-01-01

    This paper describes the results of an experimental investigation of phenomena associated with the oblique hypervelocity impact of spherical projectiles on multisheet aluminum structures. A model that can be employed in the design of meteoroid and space debris protection systems for space structures is developed. The model consists of equations that relate crater and perforation damage of a multisheet structure to parameters such as projectile size, impact velocity, and trajectory obliquity. The equations are obtained through a regression analysis of oblique hypervelocity impact test data. This data shows that the response of a multisheet structure to oblique impact is significantly different from its response to normal hypervelocity impact. It was found that obliquely incident projectiles produce ricochet debris that can severely damage panels or instrumentation located on the exterior of a space structure. Obliquity effects of high-speed impact must, therefore, be considered in the design of any structure exposed to a meteoroid or space debris environement.

  18. Coherence Phenomena in Coupled Optical Resonators

    NASA Technical Reports Server (NTRS)

    Smith, David D.

    2007-01-01

    Quantum coherence effects in atomic media such as electromagnetically-induced transparency and absorption, lasing without inversion, super-radiance and gain-assisted superluminality have become well-known in atomic physics. But these effects are not unique to atoms, nor are they uniquely quantum in nature, but rather are fundamental to systems of coherently coupled oscillators. In this talk I will review a variety of analogous photonic coherence phenomena that can occur in passive and active coupled optical resonators. Specifically, I will examine the evolution of the response that can occur upon the addition of a second resonator, to a single resonator that is side-coupled to a waveguide, as the coupling is increased, and discuss the conditions for slow and fast light propagation, coupled-resonator-induced transparency and absorption, lasing without gain, and gain-assisted superluminal pulse propagation. Finally, I will discuss the application of these systems to laser stabilization and gyroscopy.

  19. Threshold Phenomena in a Throbbing Complex Plasma

    SciTech Connect

    Mikikian, Maxime; Coueedel, Lenaiec; Cavarroc, Marjorie; Tessier, Yves; Boufendi, Laiefa

    2010-08-13

    In complex plasmas, the trapped dust particle cloud is often characterized by a central dust-free region ('void'). The void induces a spatial inhomogeneity of the dust particle distribution and is at the origin of many intricate unstable phenomena. One type of this kind of behavior is the so-called heartbeat instability consisting of successive contractions and expansions of the void. This instability is characterized by a strong nonlinear dynamics which can reveal the occurrence of incomplete sequences corresponding to failed contractions. Experimental results based on high-speed imaging are presented for the first time and underline this threshold effect in both the dust cloud motion and the evolution of the plasma light emission.

  20. Novel nuclear phenomena in quantum chromodynamics

    SciTech Connect

    Brodsky, S.J.

    1987-08-01

    Many of the key issues in understanding quantum chromodynamics involve processes in nuclear targets at intermediate energies. A range of hadronic and nuclear phenomena-exclusive processes, color transparency, hidden color degrees of freedom in nuclei, reduced nuclear amplitudes, jet coalescence, formation zone effects, hadron helicity selection rules, spin correlations, higher twist effects, and nuclear diffraction were discussed as tools for probing hadron structure and the propagation of quark and gluon jets in nuclei. Several areas were also reviewed where there has been significant theoretical progress determining the form of hadron and nuclear wave functions, including QCD sum rules, lattice gauge theory, and discretized light-cone quantization. A possible interpretation was also discussed of the large spin correlation A/sub NN/ in proton-proton scattering, and how relate this effect to an energy and angular dependence of color transparency in nuclei. 76 refs., 24 figs.

  1. Single event phenomena: Testing and prediction

    NASA Technical Reports Server (NTRS)

    Kinnison, James D.

    1992-01-01

    Highly integrated microelectronic devices are often used to increase the performance of satellite systems while reducing the system power dissipation, size, and weight. However, these devices are usually more susceptible to radiation than less integrated devices. In particular, the problem of sensitivity to single event upset and latchup is greatly increased as the integration level is increased. Therefore, a method for accurately evaluating the susceptibility of new devices to single event phenomena is critical to qualifying new components for use in space systems. This evaluation includes testing devices for upset or latchup and extrapolating the results of these tests to the orbital environment. Current methods for testing devices for single event effects are reviewed, and methods for upset rate prediction, including a new technique based on Monte Carlo simulation, are presented.

  2. Meteorological phenomena in Western classical orchestral music

    NASA Astrophysics Data System (ADS)

    Williams, P. D.; Aplin, K. L.

    2012-12-01

    The creative output of composers, writers, and artists is often influenced by their surroundings. To give a literary example, it has been claimed recently that some of the characters in Oliver Twist and A Christmas Carol were based on real-life people who lived near Charles Dickens in London. Of course, an important part of what we see and hear is not only the people with whom we interact, but also our geophysical surroundings. Of all the geophysical phenomena to influence us, the weather is arguably the most significant, because we are exposed to it directly and daily. The weather was a great source of inspiration for Monet, Constable, and Turner, who are known for their scientifically accurate paintings of the skies. But to what extent does weather inspire composers? The authors of this presentation, who are atmospheric scientists by day but amateur classical musicians by night, have been contemplating this question. We have built a systematic musical database, which has allowed us to catalogue and analyze the frequencies with which weather is depicted in a sample of classical orchestral music. The depictions vary from explicit mimicry using traditional and specialized orchestral instruments, through to subtle suggestions. We have found that composers are generally influenced by their own environment in the type of weather they choose to represent. As befits the national stereotype, British composers seem disproportionately keen to depict the UK's variable weather patterns and stormy coastline. Reference: Aplin KL and Williams PD (2011) Meteorological phenomena in Western classical orchestral music. Weather, 66(11), pp 300-306. doi:10.1002/wea.765

  3. TRANSIENT LUNAR PHENOMENA: REGULARITY AND REALITY

    SciTech Connect

    Crotts, Arlin P. S.

    2009-05-20

    Transient lunar phenomena (TLPs) have been reported for centuries, but their nature is largely unsettled, and even their existence as a coherent phenomenon is controversial. Nonetheless, TLP data show regularities in the observations; a key question is whether this structure is imposed by processes tied to the lunar surface, or by terrestrial atmospheric or human observer effects. I interrogate an extensive catalog of TLPs to gauge how human factors determine the distribution of TLP reports. The sample is grouped according to variables which should produce differing results if determining factors involve humans, and not reflecting phenomena tied to the lunar surface. Features dependent on human factors can then be excluded. Regardless of how the sample is split, the results are similar: {approx}50% of reports originate from near Aristarchus, {approx}16% from Plato, {approx}6% from recent, major impacts (Copernicus, Kepler, Tycho, and Aristarchus), plus several at Grimaldi. Mare Crisium produces a robust signal in some cases (however, Crisium is too large for a 'feature' as defined). TLP count consistency for these features indicates that {approx}80% of these may be real. Some commonly reported sites disappear from the robust averages, including Alphonsus, Ross D, and Gassendi. These reports begin almost exclusively after 1955, when TLPs became widely known and many more (and inexperienced) observers searched for TLPs. In a companion paper, we compare the spatial distribution of robust TLP sites to transient outgassing (seen by Apollo and Lunar Prospector instruments). To a high confidence, robust TLP sites and those of lunar outgassing correlate strongly, further arguing for the reality of TLPs.

  4. Astrophysical phenomena related to supermassive black holes

    NASA Astrophysics Data System (ADS)

    Pott, Jörg-Uwe

    2006-12-01

    The thesis contains the results of my recent projects in astrophysical research. All projects aim at pushing the limits of our knowledge about the interaction between a galaxy, the fundamental building block of today's universe, and a supermassive black hole (SMBH) at its center. Over the past years a lot of observational evidence has been gathered for the current understanding, that at least a major part of the galaxies with a stellar bulge contain central SMBHs. The typical extragalactic approach consists of searching for the spectroscopic pattern of Keplerian rotation, produced by stars and gas, when orbiting a central dark mass (Kormendy & Richstone 1995). It suggests that a significant fraction of large galaxies host in their very nucleus a SMBH of millions to billions of solar masses (Kormendy & Gebhardt 2001). In the closest case, the center of our Milky Way, the most central stars, which can be imaged, were shown to move on orbits with circulation times of a few decades only, evidencing a mass and compactness of the dark counter part of the Keplerian motion, which can only be explained by a SMBH (Eckart & Genzel 1996; Ghez et al. 2000; Schödel et al. 2002). Having acknowledged the widespread existence of SMBHs the obvious next step is investigating the interaction with their environment. Although the basic property of a SMBH, which is concentrating a huge amount of mass in a ludicrously small volume defined by the Schwarzschild radius, only creates a deep gravitational trough, its existence evokes much more phenomena than simply attracting the surrounding matter. It can trigger or exacerbate star formation via tidal forces (Morris 1993). It shapes the distribution of its surrounding matter to accretion discs, which themselves release gravitational potential energy as radiation, possibly due to magnetic friction (Blandford 1995). The radiation efficiency of such active galactic nuclei (AGN) can become roughly 100 times more efficient than atomic nuclear

  5. Black Holes Admitting Strong Resonant Phenomena

    NASA Astrophysics Data System (ADS)

    Stuchlík, Z.; Kotrlová, A.; G. Török

    2008-12-01

    High-frequency twin peak quasiperiodic oscillations (QPOs) are observed in four microquasars, i.e., Galactic black hole binary systems, with frequency ratio very close to 3:2. In the microquasar GRS 1915+105 the structure of QPOs exhibits additional frequencies and more than two frequencies are observed in the Galaxy nuclei Sgr A* or in some extragalactic sources (NGC 4051, MCG-6-30-15 and NGC 5408 X-1). The observed QPOs can be explained by a variety of the orbital resonance model versions assuming resonance of oscillations with the Keplerian frequency νK or the vertical epicyclic frequency νθ, and the radial epicyclic frequency νr, or some combinations of these frequencies. Generally, different resonances could arise at different radii of an accretion disk. However, we have shown that for special values of dimensionless black hole spin a strong resonant phenomena could occur when different resonances can be excited at the same radius, as cooperative phenomena between the resonances may work in such situations. The special values of a are determined for triple frequency ratio sets νK:νθ:νr=s:t:u with s,t,u being small integers. The most promising example of such a special situation arises for black holes with extraordinary resonant spin a=0.983 at the radius r=2.395 M, where νK:νθ:νr=3:2:1. We also predict that when combinations of the orbital frequencies are allowed, QPOs with four frequency ratio set 4:3:2:1 could be observed in the field of black holes with a=0.866,0.882 and 0.962. Assuming the extraordinary resonant spin a=0.983 in Sgr A*, its QPOs with observed frequency ratio ≍3:2:1 imply the black hole mass in the interval 4.3×106 Msolar< M< 5.4×106 Msolar, in agreement with estimates given by other, independent, observations.

  6. Pathways toward understanding Macroscopic Quantum Phenomena

    NASA Astrophysics Data System (ADS)

    Hu, B. L.; Subaşi, Y.

    2013-06-01

    Macroscopic quantum phenomena refer to quantum features in objects of 'large' sizes, systems with many components or degrees of freedom, organized in some ways where they can be identified as macroscopic objects. This emerging field is ushered in by several categories of definitive experiments in superconductivity, electromechanical systems, Bose-Einstein condensates and others. Yet this new field which is rich in open issues at the foundation of quantum and statistical physics remains little explored theoretically (with the important exception of the work of A J Leggett [1], while touched upon or implied by several groups of authors represented in this conference. Our attitude differs in that we believe in the full validity of quantum mechanics stretching from the testable micro to meso scales, with no need for the introduction of new laws of physics.) This talk summarizes our thoughts in attempting a systematic investigation into some key foundational issues of quantum macroscopic phenomena, with the goal of ultimately revealing or building a viable theoretical framework. Three major themes discussed in three intended essays are the large N expansion [2], the correlation hierarchy [3] and quantum entanglement [4]. We give a sketch of the first two themes and then discuss several key issues in the consideration of macro and quantum, namely, a) recognition that there exist many levels of structure in a composite body and only by judicious choice of an appropriate set of collective variables can one give the best description of the dynamics of a specific level of structure. Capturing the quantum features of a macroscopic object is greatly facilitated by the existence and functioning of these collective variables; b) quantum entanglement, an exclusively quantum feature [5], is known to persist to high temperatures [6] and large scales [7] under certain conditions, and may actually decrease with increased connectivity in a quantum network [8]. We use entanglement as a

  7. Living matter: the "lunar eclipse" phenomena.

    PubMed

    Korpan, Nikolai N

    2010-01-01

    The present investigations describe a unique phenomenon, namely the phenomenon of the "lunar eclipse", which has been observed and discovered by the author in living substance during the freeze-thawing processes in vivo using temperatures of various intensities and its cryosurgical response in animal experiment. Similar phenomena author has observed in nature, namely the total lunar eclipse and total solar eclipse. In this experimental study 76 animals (mongrel dogs) were investigated. A disc cryogenic probe was placed on the pancreas after the laparotomy. For cryosurgical exposure a temperature range of -40 degrees C, -80 degrees C, -120 degrees C and -180 degrees C was selected in contact with pancreas parenchyma. The freeze-thaw cycle was monitored by intraoperative ultrasound before, during and after cryosurgery. Each cryolesion was observed for one hour after thawing intraoperatively. Immediately after freezing, during the thawing process, the snow-white pancreas parenchyma, frozen hard to an ice block and resembling a full moon with a sharp demarcation line, gradually assumed a ruby-red shade and a hemispherical shape as it grew in size depend on reconstruction vascular circulation from the periphery to the center. This snow-white cryogenic lesion dissolved in the same manner in all animal tissues. The "lunar eclipse" phenomenon contributes to a fundamental understanding of the mechanisms of biological tissue damage during low temperature exposure in cryoscience and cryomedicine. Properties of the pancreas parenchyma response during the phenomenon of the "lunar eclipse" provide important insights into the mechanisms of damage and the formation of cryogenic lesion immediately after thawing in cryosurgery. Vascular changes and circulatory stagnation are commonly considered to be the main mechanism of biological tissue injury during low temperature exposure. The phenomenon of the "lunar eclipse" suggests that cryosurgery is the first surgical technique to use

  8. Polar Phenomena in Outer Planet Atmospheres

    NASA Astrophysics Data System (ADS)

    Orton, G.; Fletcher, L.; Yanamandra-Fisher, P.; Leyrat, C.; Greathouse, T.; Parrish, P.; Encrenaz, T.; Simon-Miller, A.

    2008-12-01

    Infrared observations of the polar regions of the outer planets have revealed similarities to the Earth's atmosphere and some new phenomena. The most dominant force which is apparent in time-dependent studies of the poles is seasonal radiative forcing, which was detected in Saturn's stratosphere as early as 1973. For Saturn, Uranus and Neptune, planets with substantial obliquities, the seasonally dependent changes are predictable and can be used to constrain abundances of optically active gases and the rate of restoration by stratospheric circulation. In the case of Neptune, recent evidence shows that the heating is sufficient to allow a "leak" from the reservoir of methane in the deep atmosphere into the polar stratosphere. New thermal images of Uranus show that the winter pole of Uranus which has only recently emerged fully from darkness is colder than when it was in the middle of winter when Voyager 2 visited, confirming the substantial seasonal phase delay associated with radiative heating and cooling models. Even Jupiter with its 3-degree obliquity shows clear evidence for seasonal forcing of temperatures in the upper troposphere and stratosphere. The second most prominent characteristic of the resolvable polar temperature fields in Jupiter and Saturn is the formation of polar vortices. Jupiter's polar vortices are cold, similar to those detected in the terrestrial planets; they have sharp equatorward boundaries which are characterized by Rossby waves which rotate at the speed of the local zonal wind flow and are coincident with the similarly irregular boundaries of a polar haze, also known as "polar hoods". The cold vortex at Saturn's northern winter pole is muted, but Saturn also has a unique "warm polar vortex" in the south (late summer) pole which shows no apparent wave structure. Saturn's warm polar vortex has no counterpart in the Earth's atmosphere, where summer radiative warming simply dissipates the cold winter vortex. Saturn also possesses

  9. In-vessel Zircaloy oxidation/hydrogen generation behavior during severe accidents

    SciTech Connect

    Cronenberg, A.W. )

    1990-09-01

    In-vessel Zircaloy oxidation and hydrogen generation data from various US Nuclear Regulatory Commission severe-fuel damage test programs are presented and compared, where the effects of Zircaloy melting, bundle reconfiguration, and bundle quenching by reflooding are assessed for common findings. The experiments evaluated include fuel bundles incorporating fresh and previously irradiated fuel rods, as well as control rods. Findings indicate that the extent of bundle oxidation is largely controlled by steam supply conditions and that high rates of hydrogen generation continued after melt formation and relocation. Likewise, no retardation of hydrogen generation was noted for experiments which incorporated control rods. Metallographic findings indicate extensive oxidation of once-molten Zircaloy bearing test debris. Such test results indicate no apparent limitations to Zircaloy oxidation for fuel bundles subjected to severe-accident coolant-boiloff conditions. 46 refs., 22 figs., 12 tabs.

  10. Measurements for the JASPER program In-Vessel Fuel Storage experiment

    SciTech Connect

    Muckenthaler, F.J.; Spencer, R.R.; Hunter, H.T.; Hull, J.L.; Shono, A.

    1992-01-01

    The In-Vessel-Fuel-Storage (IVFS) experiment was conducted at the Oak Ridge National Laboratory`s (ORNL) Tower Shielding Facility (TSF) during the first nine months of 1991 as part of the continuing series of eight experiments planned for the Japanese-American Shielding Program for Experimental Research (JASPER) that was started in 1986. This is the fourth in a series of eight experiments that were planned, all of which are intended to provide support in the development of current reactor shield designs proposed for liquid metal reactor (LMR) systems both in Japan and the United States. The program is a cooperative effort between the United States Department of Energy (US DOE) and the Japanese Power Reactor and Nuclear Development Corporation (PNC). This document provides a description of the instrumentation and experimental configuration, test data, and data analysis.

  11. Insights from Investigations of In-Vessel Retention for High Powered Reactors

    SciTech Connect

    Joy L. Rempe

    2005-10-01

    In a three-year U.S. - Korean International Nuclear Energy Research Initiative (INERI), state-of-the-art analytical tools and key U.S. and Korean experimental facilities were used to explore two options, enhanced ERVC performance and the use of internal core catchers, that have the potential to increase the margin for in-vessel retention (IVR) in high power reactors (up to 1500 MWe). This increased margin has the potential to improve plant economics (owing to reduced regulatory requirements) and increase public acceptance (owing to reduced plant risk). Although this program focused upon the Korean Advanced Power Reactor -- 1400 MWe (APR 1400) design, recommentations were developed so that they can easily be applied to a wide range of existing and advanced reactor designs. This paper summarizes new data gained for evaluating the margin associated with various options investigated in this program. Insights from analyses completed with this data are also highlighted.

  12. Design Analysis and Manufacturing Studies for ITER In-Vessel Coils

    SciTech Connect

    Kalish, M.; Heitzenroeder, P.; Neumeyer, C.; Titus, P.; Zhai, Y.; Zatz, I.; Messineo, M.; Gomez, M.; Hause, C.; Daly, E.; Martin, A.; Wu, Y.; Jin, J.; Long, F.; Song, Y.; Wang, Z.; Yun, Zan; Hsiao, J.; Pillsbury, J. R.; Bohm, T.; Sawan, M.; Jiang, NFN

    2014-07-01

    ITER is incorporating two types of In Vessel Coils (IVCs): ELM Coils to mitigate Edge Localized Modes and VS Coils to provide Vertical Stabilization of the plasma. Strong coupling with the plasma is required so that the ELM and VS Coils can meet their performance requirements. Accordingly, the IVCs are in close proximity to the plasma, mounted just behind the Blanket Shield Modules. This location results in a radiation and temperature environment that is severe necessitating new solutions for material selection as well as challenging analysis and design solutions. Fitting the coil systems in between the blanket shield modules and the vacuum vessel leads to difficult integration with diagnostic cabling and cooling water manifolds.

  13. Rheological Properties and Transfer Phenomena of Nanofluids

    NASA Astrophysics Data System (ADS)

    Jung, Kang-min; Kim, Sung Hyun

    2008-07-01

    This study focused on the synthesis of stable nanofluids and investigation of their rhelogical properties and transfer phenomena. Nanofluids of diamond/ethylene glycol, alumina/transformer oil and silica/water were made to use in this study. Rheological properties of diamond nanofluids were determined at constant temperature (25 °C) using a viscometer. For the convective heat transfer experiment, alumina nanofluid passed through the plate heat exchanger. CO2 absorption experiment was conducted in a bubble type absorber containing silica nanofluid. Diamond nanofluid showed non-Newtonian behaviors under a steady-shear flow except the case of very low concentration of solid nanoparticles. The heat transfer coefficient of alumina nanofluid was higher than that of base fluid. One possible reason is that concentration of nanoparticles at the wall side is higher than that of microparticles. Silica nanofluid showed that both average CO2 absorption rate and total absorption amount enhanced than those of base fluid. The stably suspended nanoparticles create a mesh-like structure. That structure arrangement cracks the gas bubble and increases the surface area.

  14. Transition phenomena in unstably stratified turbulent flows.

    PubMed

    Bukai, M; Eidelman, A; Elperin, T; Kleeorin, N; Rogachevskii, I; Sapir-Katiraie, I

    2011-03-01

    We study experimentally and theoretically the transition phenomena caused by external forcing from Rayleigh-Bénard convection with large-scale circulation (LSC) to the limiting regime of unstably stratified turbulent flow without LSC, where the temperature field behaves like a passive scalar. In the experiments we use the Rayleigh-Bénard apparatus with an additional source of turbulence produced by two oscillating grids located near the sidewalls of the chamber. When the frequency of the grid oscillations is larger than 2 Hz, the LSC in turbulent convection is destroyed, and the destruction of the LSC is accompanied by a strong change of the mean temperature distribution. However, in all regimes of the unstably stratified turbulent flow the ratio [(ℓ{x}∇{x}T)²+(ℓ{y}∇{y}T)² + (ℓ{z}∇{z}T)²]/<θ²> varies slightly (even in the range of parameters where the behavior of the temperature field is different from that of the passive scalar). Here ℓ{i} are the integral scales of turbulence along the x,y,z directions, and T and θ are the mean and fluctuating parts of the fluid temperature. At all frequencies of the grid oscillations we have detected long-term nonlinear oscillations of the mean temperature. The theoretical predictions based on the budget equations for turbulent kinetic energy, turbulent temperature fluctuations, and turbulent heat flux, are in agreement with the experimental results. PMID:21517582

  15. Efferent feedback can explain many hearing phenomena

    NASA Astrophysics Data System (ADS)

    Holmes, W. Harvey; Flax, Matthew R.

    2015-12-01

    The mixed mode cochlear amplifier (MMCA) model was presented at the last Mechanics of Hearing workshop [4]. The MMCA consists principally of a nonlinear feedback loop formed when an efferent-controlled outer hair cell (OHC) is combined with the cochlear mechanics and the rest of the relevant neurobiology. Essential elements of this model are efferent control of the OHC motility and a delay in the feedback to the OHC. The input to the MMCA is the passive travelling wave. In the MMCA amplification is localized where both the neural and tuned mechanical systems meet in the Organ of Corti (OoC). The simplest model based on this idea is a nonlinear delay line resonator (DLR), which is mathematically described by a nonlinear delay-differential equation (DDE). This model predicts possible Hopf bifurcations and exhibits its most interesting behaviour when operating near a bifurcation. This contribution presents some simulation results using the DLR model. These show that various observed hearing phenomena can be accounted for by this model, at least qualitatively, including compression effects, two-tone suppression and some forms of otoacoustic emissions (OAEs).

  16. The Monitoring of Transient Lunar Phenomena

    NASA Astrophysics Data System (ADS)

    Doorn, Jarrel; Eaton, M.; Ahrendts, G.; Barker, T.

    2011-05-01

    Transient Lunar Phenomena (TLP's) are described as short-lived changes in the brightness of areas on the face of the Moon. TLP activity has a higher number of reports, though unsubstantiated, in specific areas of the Moon such as the Aristarchus plateau. Our current research includes the division of lunar images taken through multiple filters using a Santa-Barbara Instrument Group (SBIG) ST8-E CCD camera mounted on a 0.45m Centurion telescope. We are also taking spectra of regions such as Aristarchus and the crater Ina, which shows evidence of recent activity (Schultz, P., Staid, M., Pieters, C. Nature, Volume 444, Issue 7116, pp. 184-186, 2006) using an SBIG DSS-7 spectrometer mounted on a 0.30m Schmidt-Cassegrain optical tube assembly on a Software Bisque Paramount drive system. Future research will include infrared imaging of the lunar surface. We are grateful for the support provided by the NASA Rhode Island Space Grant Consortium and the National Geographic Society.

  17. Viscous theory of surface noise interaction phenomena

    NASA Technical Reports Server (NTRS)

    Yates, J. E.

    1980-01-01

    A viscous linear surface noise interaction problem is formulated that includes noise production by an oscillating surface, turbulent or vortical interaction with a surface, and scattering of sound by a surface. The importance of viscosity in establishing uniqueness of solution and partitioning of energy into acoustic and vortical modes is discussed. The results of inviscid two dimensional airfoil theory are used to examine the interactive noise problem in the limit of high reduced frequency and small Helmholtz number. It is shown that in the case of vortex interaction with a surface, the noise produced with the full Kutta condition is 3 dB less than the no Kutta condition result. The results of a study of an airfoil oscillating in a medium at rest are discussed. It is concluded that viscosity can be a controlling factor in analyses and experiments of surface noise interaction phenomena and that the effect of edge bluntness as well as viscosity must be included in the problem formulation to correctly calculate the interactive noise.

  18. Two-Stage Modelling Of Random Phenomena

    NASA Astrophysics Data System (ADS)

    Barańska, Anna

    2015-12-01

    The main objective of this publication was to present a two-stage algorithm of modelling random phenomena, based on multidimensional function modelling, on the example of modelling the real estate market for the purpose of real estate valuation and estimation of model parameters of foundations vertical displacements. The first stage of the presented algorithm includes a selection of a suitable form of the function model. In the classical algorithms, based on function modelling, prediction of the dependent variable is its value obtained directly from the model. The better the model reflects a relationship between the independent variables and their effect on the dependent variable, the more reliable is the model value. In this paper, an algorithm has been proposed which comprises adjustment of the value obtained from the model with a random correction determined from the residuals of the model for these cases which, in a separate analysis, were considered to be the most similar to the object for which we want to model the dependent variable. The effect of applying the developed quantitative procedures for calculating the corrections and qualitative methods to assess the similarity on the final outcome of the prediction and its accuracy, was examined by statistical methods, mainly using appropriate parametric tests of significance. The idea of the presented algorithm has been designed so as to approximate the value of the dependent variable of the studied phenomenon to its value in reality and, at the same time, to have it "smoothed out" by a well fitted modelling function.

  19. Fingering phenomena during grain-grain displacement

    NASA Astrophysics Data System (ADS)

    Mello, Nathália M. P.; Paiva, Humberto A.; Combe, G.; Atman, A. P. F.

    2016-05-01

    Spontaneous formation of fingered patterns during the displacement of dense granular assemblies was experimentally reported few years ago, in a radial Hele-Shaw cell. Here, by means of discrete element simulations, we have recovered the experimental findings and extended the original study to explore the control parameters space. In particular, using assemblies of grains with different geometries (monodisperse, bidisperse, or polydisperse), we measured the macroscopic stress tensor in the samples in order to confirm some conjectures proposed in analogy with Saffman-Taylor viscous fingering phenomena for immiscible fluids. Considering an axial setup which allows to control the discharge of grains and to follow the trajectory and the pressure gradient along the displacing interface, we have applied the Darcy law for laminar flow in fluids in order to measure an "effective viscosity" for each assembly combination, in an attempt to mimic variation of the viscosity ratio between the injected/displaced fluids in the Saffman-Taylor experiment. The results corroborate the analogy with the viscous fluids displacement, with the bidisperse assembly corresponding to the less viscous geometry. But, differently to fluid case, granular fingers only develop for a specific combination of displaced/injected geometries, and we have demonstrated that it is always related with the formation of a force chain network along the finger direction.

  20. Phantom black holes and critical phenomena

    SciTech Connect

    Azreg-Aïnou, Mustapha; Marques, Glauber T.

    2014-07-01

    We consider the two classes cosh and sinh of normal and phantom black holes of Einstein-Maxwell-dilaton theory. The thermodynamics of these holes is characterized by heat capacities that may have both signs depending on the parameters of the theory. Leaving aside the normal Reissner-Nordström black hole, it is shown that only some phantom black holes of both classes exhibit critical phenomena. The two classes share a nonextremality, but special, critical point where the transition is continuous and the heat capacity, at constant charge, changes sign with an infinite discontinuity. This point yields a classification scheme for critical points. It is concluded that the two unstable and stable phases coexist on one side of the criticality state and disappear on the other side, that is, there is no configuration where only one phase exists. The sinh class has an extremality critical point where the entropy diverges. The transition from extremality to nonextremality with the charge held constant is accompanied by a loss of mass and an increase in the temperature. A special case of this transition is when the hole is isolated (microcanonical ensemble), it will evolve by emission of energy, which results in a decrease of its mass, to the final state of minimum mass and vanishing heat capacity. The Ehrenfest scheme of classification is inaccurate in this case but the generalized one due to Hilfer leads to conclude that the transition is of order less than unity. Fluctuations near criticality are also investigated.