WorldWideScience

Sample records for advanced mechanistic code

  1. Advanced reach tool (ART) : Development of the mechanistic model

    NARCIS (Netherlands)

    Fransman, W.; Tongeren, M. van; Cherrie, J.W.; Tischer, M.; Schneider, T.; Schinkel, J.; Kromhout, H.; Warren, N.; Goede, H.; Tielemans, E.

    2011-01-01

    This paper describes the development of the mechanistic model within a collaborative project, referred to as the Advanced REACH Tool (ART) project, to develop a tool to model inhalation exposure for workers sharing similar operational conditions across different industries and locations in Europe.

  2. ESCADRE and ICARE code systems

    International Nuclear Information System (INIS)

    Reocreux, M.; Gauvain, J.

    1992-01-01

    The French sever accident code development program is following two parallel approaches: the first one is dealing with ''integral codes'' which are designed for giving immediate engineer answers, the second one is following a more mechanistic way in order to have the capability of detailed analysis of experiments, in order to get a better understanding of the scaling problem and reach a better confidence in plant calculations. In the first approach a complete system has been developed and is being used for practical cases: this is the ESCADRE system. In the second approach, a set of codes dealing first with primary circuit is being developed: a mechanistic core degradation code, ICARE, has been issued and is being coupled with the advanced thermalhydraulic code CATHARE. Fission product codes have been also coupled to CATHARE. The ''integral'' ESCADRE system and the mechanistic ICARE and associated codes are described. Their main characteristics are reviewed and the status of their development and assessment given. Future studies are finally discussed. 36 refs, 4 figs, 1 tab

  3. Mechanistic modelling of gaseous fission product behaviour in UO2 fuel by Rtop code

    International Nuclear Information System (INIS)

    Kanukova, V.D.; Khoruzhii, O.V.; Kourtchatov, S.Y.; Likhanskii, V.V.; Matveew, L.V.

    2002-01-01

    The current status of a mechanistic modelling by the RTOP code of the fission product behaviour in polycrystalline UO 2 fuel is described. An outline of the code and implemented physical models is presented. The general approach to code validation is discussed. It is exemplified by the results of validation of the models of fuel oxidation and grain growth. The different models of intragranular and intergranular gas bubble behaviour have been tested and the sensitivity of the code in the framework of these models has been analysed. An analysis of available models of the resolution of grain face bubbles is also presented. The possibilities of the RTOP code are presented through the example of modelling behaviour of WWER fuel over the course of a comparative WWER-PWR experiment performed at Halden and by comparison with Yanagisawa experiments. (author)

  4. Evaluation of Advanced Models for PAFS Condensation Heat Transfer in SPACE Code

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Byoung-Uhn; Kim, Seok; Park, Yu-Sun; Kang, Kyung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ahn, Tae-Hwan; Yun, Byong-Jo [Pusan National University, Busan (Korea, Republic of)

    2015-10-15

    The PAFS (Passive Auxiliary Feedwater System) is operated by the natural circulation to remove the core decay heat through the PCHX (Passive Condensation Heat Exchanger) which is composed of the nearly horizontal tubes. For validation of the cooling and operational performance of the PAFS, PASCAL (PAFS Condensing Heat Removal Assessment Loop) facility was constructed and the condensation heat transfer and natural convection phenomena in the PAFS was experimentally investigated at KAERI (Korea Atomic Energy Research Institute). From the PASCAL experimental result, it was found that conventional system analysis code underestimated the condensation heat transfer. In this study, advanced condensation heat transfer models which can treat the heat transfer mechanisms with the different flow regimes in the nearly horizontal heat exchanger tube were analyzed. The models were implemented in a thermal hydraulic safety analysis code, SPACE (Safety and Performance Analysis Code for Nuclear Power Plant), and it was evaluated with the PASCAL experimental data. With an aim of enhancing the prediction capability for the condensation phenomenon inside the PCHX tube of the PAFS, advanced models for the condensation heat transfer were implemented into the wall condensation model of the SPACE code, so that the PASCAL experimental result was utilized to validate the condensation models. Calculation results showed that the improved model for the condensation heat transfer coefficient enhanced the prediction capability of the SPACE code. This result confirms that the mechanistic modeling for the film condensation in the steam phase and the convection in the condensate liquid contributed to enhance the prediction capability of the wall condensation model of the SPACE code and reduce conservatism in prediction of condensation heat transfer.

  5. Requirements on mechanistic NPP models used in CSS for diagnostics and predictions

    International Nuclear Information System (INIS)

    Juslin, K.

    1996-01-01

    Mechanistic models have for several years with good experience been used for operators' support in electric power dispatching centres. Some models of limited scope have already been in use at nuclear power plants. It is considered that also advanced mechanistic models in combination with present computer technology with preference could be used in Computerized Support Systems (CSS) for the assistance of Nuclear Power Plant (NPP) operators. Requirements with respect to accuracy, validity range, speed flexibility and level of detail on the models used for such purposes are discussed. Quality Assurance, Verification and Validation efforts are considered. A long term commitment in the field of mechanistic modelling and real time simulation is considered as the key to successful implementations. The Advanced PROcess Simulation (APROS) code system and simulation environment developed at the Technical Research Centre of Finland (VTT) is intended also for CSS applications in NPP control rooms. (author). 4 refs

  6. A strategy of implementation of the improved constitutive equations for the advanced subchannel code

    International Nuclear Information System (INIS)

    Shirai, Hiroshi; Hotta, Akitoshi; Ninokata, Hisashi

    2004-01-01

    To develop the advanced subchannel analysis code, the dominant factors that influence the boiling transitional process must be taken into account in the mechanistic constitutive equations based on the flow geometries and the fluid properties. The dominant factors that influence the boiling transitional processes are (1) the gas-liquid re-distribution by cross flow, (2) the liquid film dryout, (3) the two-phase flow regime transition, (4) the droplet deposition, and (5) the spacer-droplet interaction. At first, we indicated the strategy for the development of the constitutive equations for the five dominant factors based on the experimental database by the latest measurement technique and the latest computational fluid dynamics method. Then, the problems of the present constitutive equations and the improvement plan of the constitutive equations were indicated. Finally, the layered structure for the two-phase/three-field subchannel code including the new constitutive equations was designed. (author)

  7. Advanced hardware design for error correcting codes

    CERN Document Server

    Coussy, Philippe

    2015-01-01

    This book provides thorough coverage of error correcting techniques. It includes essential basic concepts and the latest advances on key topics in design, implementation, and optimization of hardware/software systems for error correction. The book’s chapters are written by internationally recognized experts in this field. Topics include evolution of error correction techniques, industrial user needs, architectures, and design approaches for the most advanced error correcting codes (Polar Codes, Non-Binary LDPC, Product Codes, etc). This book provides access to recent results, and is suitable for graduate students and researchers of mathematics, computer science, and engineering. • Examines how to optimize the architecture of hardware design for error correcting codes; • Presents error correction codes from theory to optimized architecture for the current and the next generation standards; • Provides coverage of industrial user needs advanced error correcting techniques.

  8. Regulatory Technology Development Plan - Sodium Fast Reactor: Mechanistic Source Term - Trial Calculation

    International Nuclear Information System (INIS)

    Grabaskas, David

    2016-01-01

    The potential release of radioactive material during a plant incident, referred to as the source term, is a vital design metric and will be a major focus of advanced reactor licensing. The U.S. Nuclear Regulatory Commission has stated an expectation for advanced reactor vendors to present a mechanistic assessment of the potential source term in their license applications. The mechanistic source term presents an opportunity for vendors to realistically assess the radiological consequences of an incident, and may allow reduced emergency planning zones and smaller plant sites. However, the development of a mechanistic source term for advanced reactors is not without challenges, as there are often numerous phenomena impacting the transportation and retention of radionuclides. This project sought to evaluate U.S. capabilities regarding the mechanistic assessment of radionuclide release from core damage incidents at metal fueled, pool-type sodium fast reactors (SFRs). The purpose of the analysis was to identify, and prioritize, any gaps regarding computational tools or data necessary for the modeling of radionuclide transport and retention phenomena. To accomplish this task, a parallel-path analysis approach was utilized. One path, led by Argonne and Sandia National Laboratories, sought to perform a mechanistic source term assessment using available codes, data, and models, with the goal to identify gaps in the current knowledge base. The second path, performed by an independent contractor, performed sensitivity analyses to determine the importance of particular radionuclides and transport phenomena in regards to offsite consequences. The results of the two pathways were combined to prioritize gaps in current capabilities.

  9. Regulatory Technology Development Plan - Sodium Fast Reactor: Mechanistic Source Term – Trial Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Bucknor, Matthew [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Jerden, James [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Brunett, Acacia J. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Denman, Matthew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Nuclear Engineering Division; Clark, Andrew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Nuclear Engineering Division; Denning, Richard S. [Consultant, Columbus, OH (United States)

    2016-10-01

    The potential release of radioactive material during a plant incident, referred to as the source term, is a vital design metric and will be a major focus of advanced reactor licensing. The U.S. Nuclear Regulatory Commission has stated an expectation for advanced reactor vendors to present a mechanistic assessment of the potential source term in their license applications. The mechanistic source term presents an opportunity for vendors to realistically assess the radiological consequences of an incident, and may allow reduced emergency planning zones and smaller plant sites. However, the development of a mechanistic source term for advanced reactors is not without challenges, as there are often numerous phenomena impacting the transportation and retention of radionuclides. This project sought to evaluate U.S. capabilities regarding the mechanistic assessment of radionuclide release from core damage incidents at metal fueled, pool-type sodium fast reactors (SFRs). The purpose of the analysis was to identify, and prioritize, any gaps regarding computational tools or data necessary for the modeling of radionuclide transport and retention phenomena. To accomplish this task, a parallel-path analysis approach was utilized. One path, led by Argonne and Sandia National Laboratories, sought to perform a mechanistic source term assessment using available codes, data, and models, with the goal to identify gaps in the current knowledge base. The second path, performed by an independent contractor, performed sensitivity analyses to determine the importance of particular radionuclides and transport phenomena in regards to offsite consequences. The results of the two pathways were combined to prioritize gaps in current capabilities.

  10. Zebra: An advanced PWR lattice code

    Energy Technology Data Exchange (ETDEWEB)

    Cao, L.; Wu, H.; Zheng, Y. [School of Nuclear Science and Technology, Xi' an Jiaotong Univ., No. 28, Xianning West Road, Xi' an, ShannXi, 710049 (China)

    2012-07-01

    This paper presents an overview of an advanced PWR lattice code ZEBRA developed at NECP laboratory in Xi'an Jiaotong Univ.. The multi-group cross-section library is generated from the ENDF/B-VII library by NJOY and the 361-group SHEM structure is employed. The resonance calculation module is developed based on sub-group method. The transport solver is Auto-MOC code, which is a self-developed code based on the Method of Characteristic and the customization of AutoCAD software. The whole code is well organized in a modular software structure. Some numerical results during the validation of the code demonstrate that this code has a good precision and a high efficiency. (authors)

  11. Zebra: An advanced PWR lattice code

    International Nuclear Information System (INIS)

    Cao, L.; Wu, H.; Zheng, Y.

    2012-01-01

    This paper presents an overview of an advanced PWR lattice code ZEBRA developed at NECP laboratory in Xi'an Jiaotong Univ.. The multi-group cross-section library is generated from the ENDF/B-VII library by NJOY and the 361-group SHEM structure is employed. The resonance calculation module is developed based on sub-group method. The transport solver is Auto-MOC code, which is a self-developed code based on the Method of Characteristic and the customization of AutoCAD software. The whole code is well organized in a modular software structure. Some numerical results during the validation of the code demonstrate that this code has a good precision and a high efficiency. (authors)

  12. Huffman coding in advanced audio coding standard

    Science.gov (United States)

    Brzuchalski, Grzegorz

    2012-05-01

    This article presents several hardware architectures of Advanced Audio Coding (AAC) Huffman noiseless encoder, its optimisations and working implementation. Much attention has been paid to optimise the demand of hardware resources especially memory size. The aim of design was to get as short binary stream as possible in this standard. The Huffman encoder with whole audio-video system has been implemented in FPGA devices.

  13. Benchmarking of the PHOENIX-P/ANC [Advanced Nodal Code] advanced nuclear design system

    International Nuclear Information System (INIS)

    Nguyen, T.Q.; Liu, Y.S.; Durston, C.; Casadei, A.L.

    1988-01-01

    At Westinghouse, an advanced neutronic methods program was designed to improve the quality of the predictions, enhance flexibility in designing advanced fuel and related products, and improve design lead time. Extensive benchmarking data is presented to demonstrate the accuracy of the Advanced Nodal Code (ANC) and the PHOENIX-P advanced lattice code. Qualification data to demonstrate the accuracy of ANC include comparison of key physics parameters against a fine-mesh diffusion theory code, TORTISE. Benchmarking data to demonstrate the validity of the PHOENIX-P methodologies include comparison of physics predictions against critical experiments, isotopics measurements and measured power distributions from spatial criticals. The accuracy of the PHOENIX-P/ANC Advanced Design System is demonstrated by comparing predictions of hot zero power physics parameters and hot full power core follow against measured data from operating reactors. The excellent performance of this system for a broad range of comparisons establishes the basis for implementation of these tools for core design, licensing and operational follow of PWR [pressurized water reactor] cores at Westinghouse

  14. Towards advanced code simulators

    International Nuclear Information System (INIS)

    Scriven, A.H.

    1990-01-01

    The Central Electricity Generating Board (CEGB) uses advanced thermohydraulic codes extensively to support PWR safety analyses. A system has been developed to allow fully interactive execution of any code with graphical simulation of the operator desk and mimic display. The system operates in a virtual machine environment, with the thermohydraulic code executing in one virtual machine, communicating via interrupts with any number of other virtual machines each running other programs and graphics drivers. The driver code itself does not have to be modified from its normal batch form. Shortly following the release of RELAP5 MOD1 in IBM compatible form in 1983, this code was used as the driver for this system. When RELAP5 MOD2 became available, it was adopted with no changes needed in the basic system. Overall the system has been used for some 5 years for the analysis of LOBI tests, full scale plant studies and for simple what-if studies. For gaining rapid understanding of system dependencies it has proved invaluable. The graphical mimic system, being independent of the driver code, has also been used with other codes to study core rewetting, to replay results obtained from batch jobs on a CRAY2 computer system and to display suitably processed experimental results from the LOBI facility to aid interpretation. For the above work real-time execution was not necessary. Current work now centers on implementing the RELAP 5 code on a true parallel architecture machine. Marconi Simulation have been contracted to investigate the feasibility of using upwards of 100 processors, each capable of a peak of 30 MIPS to run a highly detailed RELAP5 model in real time, complete with specially written 3D core neutronics and balance of plant models. This paper describes the experience of using RELAP5 as an analyzer/simulator, and outlines the proposed methods and problems associated with parallel execution of RELAP5

  15. Advanced thermionic reactor systems design code

    International Nuclear Information System (INIS)

    Lewis, B.R.; Pawlowski, R.A.; Greek, K.J.; Klein, A.C.

    1991-01-01

    An overall systems design code is under development to model an advanced in-core thermionic nuclear reactor system for space applications at power levels of 10 to 50 kWe. The design code is written in an object-oriented programming environment that allows the use of a series of design modules, each of which is responsible for the determination of specific system parameters. The code modules include a neutronics and core criticality module, a core thermal hydraulics module, a thermionic fuel element performance module, a radiation shielding module, a module for waste heat transfer and rejection, and modules for power conditioning and control. The neutronics and core criticality module determines critical core size, core lifetime, and shutdown margins using the criticality calculation capability of the Monte Carlo Neutron and Photon Transport Code System (MCNP). The remaining modules utilize results of the MCNP analysis along with FORTRAN programming to predict the overall system performance

  16. Recent advances in the Poisson/superfish codes

    International Nuclear Information System (INIS)

    Ryne, R.; Barts, T.; Chan, K.C.D.; Cooper, R.; Deaven, H.; Merson, J.; Rodenz, G.

    1992-01-01

    We report on advances in the POISSON/SUPERFISH family of codes used in the design and analysis of magnets and rf cavities. The codes include preprocessors for mesh generation and postprocessors for graphical display of output and calculation of auxiliary quantities. Release 3 became available in January 1992; it contains many code corrections and physics enhancements, and it also includes support for PostScript, DISSPLA, GKS and PLOT10 graphical output. Release 4 will be available in September 1992; it is free of all bit packing, making the codes more portable and able to treat very large numbers of mesh points. Release 4 includes the preprocessor FRONT and a new menu-driven graphical postprocessor that runs on workstations under X-Windows and that is capable of producing arrow plots. We will present examples that illustrate the new capabilities of the codes. (author). 6 refs., 3 figs

  17. Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC).

    Energy Technology Data Exchange (ETDEWEB)

    Schultz, Peter Andrew

    2011-12-01

    The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. Achieving the objective of modeling the performance of a disposal scenario requires describing processes involved in waste form degradation and radionuclide release at the subcontinuum scale, beginning with mechanistic descriptions of chemical reactions and chemical kinetics at the atomic scale, and upscaling into effective, validated constitutive models for input to high-fidelity continuum scale codes for coupled multiphysics simulations of release and transport. Verification and validation (V&V) is required throughout the system to establish evidence-based metrics for the level of confidence in M&S codes and capabilities, including at the subcontiunuum scale and the constitutive models they inform or generate. This Report outlines the nature of the V&V challenge at the subcontinuum scale, an approach to incorporate V&V concepts into subcontinuum scale modeling and simulation (M&S), and a plan to incrementally incorporate effective V&V into subcontinuum scale M&S destined for use in the NEAMS Waste IPSC work flow to meet requirements of quantitative confidence in the constitutive models informed by subcontinuum scale phenomena.

  18. Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC)

    International Nuclear Information System (INIS)

    Schultz, Peter Andrew

    2011-01-01

    The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M and S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. Achieving the objective of modeling the performance of a disposal scenario requires describing processes involved in waste form degradation and radionuclide release at the subcontinuum scale, beginning with mechanistic descriptions of chemical reactions and chemical kinetics at the atomic scale, and upscaling into effective, validated constitutive models for input to high-fidelity continuum scale codes for coupled multiphysics simulations of release and transport. Verification and validation (V and V) is required throughout the system to establish evidence-based metrics for the level of confidence in M and S codes and capabilities, including at the subcontiunuum scale and the constitutive models they inform or generate. This Report outlines the nature of the V and V challenge at the subcontinuum scale, an approach to incorporate V and V concepts into subcontinuum scale modeling and simulation (M and S), and a plan to incrementally incorporate effective V and V into subcontinuum scale M and S destined for use in the NEAMS Waste IPSC work flow to meet requirements of quantitative confidence in the constitutive models informed by subcontinuum scale phenomena.

  19. Code qualification of structural materials for AFCI advanced recycling reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Li, M.; Majumdar, S.; Nanstad, R.K.; Sham, T.-L. (Nuclear Engineering Division); (ORNL)

    2012-05-31

    This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Code Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP

  20. Foundational development of an advanced nuclear reactor integrated safety code

    International Nuclear Information System (INIS)

    Clarno, Kevin; Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth; Hooper, Russell Warren; Humphries, Larry LaRon

    2010-01-01

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

  1. Foundational development of an advanced nuclear reactor integrated safety code.

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin (Oak Ridge National Laboratory, Oak Ridge, TN); Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth (Ktech Corporation, Albuquerque, NM); Hooper, Russell Warren; Humphries, Larry LaRon

    2010-02-01

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

  2. Application of the DART Code for the Assessment of Advanced Fuel Behavior

    International Nuclear Information System (INIS)

    Rest, J.; Totev, T.

    2007-01-01

    The Dispersion Analysis Research Tool (DART) code is a dispersion fuel analysis code that contains mechanistically-based fuel and reaction-product swelling models, a one dimensional heat transfer analysis, and mechanical deformation models. DART has been used to simulate the irradiation behavior of uranium oxide, uranium silicide, and uranium molybdenum aluminum dispersion fuels, as well as their monolithic counterparts. The thermal-mechanical DART code has been validated against RERTR tests performed in the ATR for irradiation data on interaction thickness, fuel, matrix, and reaction product volume fractions, and plate thickness changes. The DART fission gas behavior model has been validated against UO 2 fission gas release data as well as measured fission gas-bubble size distributions. Here DART is utilized to analyze various aspects of the observed bubble growth in U-Mo/Al interaction product. (authors)

  3. Measurement and modeling of advanced coal conversion processes

    Energy Technology Data Exchange (ETDEWEB)

    Solomon, P.R.; Serio, M.A.; Hamblen, D.G.; Smoot, L.D.; Brewster, B.S. (Advanced Fuel Research, Inc., East Hartford, CT (United States) Brigham Young Univ., Provo, UT (United States))

    1991-01-01

    The overall objective of this program is the development of predictive capability for the design, scale up, simulation, control and feedstock evaluation in advanced coal conversion devices. This program will merge significant advances made in measuring and quantitatively describing the mechanisms in coal conversion behavior. Comprehensive computer codes for mechanistic modeling of entrained-bed gasification. Additional capabilities in predicting pollutant formation will be implemented and the technology will be expanded to fixed-bed reactors.

  4. ANDREA: Advanced nodal diffusion code for reactor analysis

    International Nuclear Information System (INIS)

    Belac, J.; Josek, R.; Klecka, L.; Stary, V.; Vocka, R.

    2005-01-01

    A new macro code is being developed at NRI which will allow coupling of the advanced thermal-hydraulics model with neutronics calculations as well as efficient use in core loading pattern optimization process. This paper describes the current stage of the macro code development. The core simulator is based on the nodal expansion method, Helios lattice code is used for few group libraries preparation. Standard features such as pin wise power reconstruction and feedback iterations on critical control rod position, boron concentration and reactor power are implemented. A special attention is paid to the system and code modularity in order to enable flexible and easy implementation of new features in future. Precision of the methods used in the macro code has been verified on available benchmarks. Testing against Temelin PWR operational data is under way (Authors)

  5. AECL's advanced code program

    Energy Technology Data Exchange (ETDEWEB)

    McGee, G.; Ball, J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    This paper discusses the advanced code project at AECL.Current suite of Analytical, Scientific and Design (ASD) computer codes in use by Canadian Nuclear Power Industry is mostly developed 20 or more years ago. It is increasingly difficult to develop and maintain. It consist of many independent tools and integrated analysis is difficult, time consuming and error-prone. The objectives of this project is to demonstrate that nuclear facility systems, structures and components meet their design objectives in terms of function, cost, and safety; demonstrate that the nuclear facility meets licensing requirements in terms of consequences of off-normal events; dose to public, workers, impact on environment and demonstrate that the nuclear facility meets operational requirements with respect to on-power fuelling and outage management.

  6. Advanced codes and methods supporting improved fuel cycle economics - 5493

    International Nuclear Information System (INIS)

    Curca-Tivig, F.; Maupin, K.; Thareau, S.

    2015-01-01

    AREVA's code development program was practically completed in 2014. The basic codes supporting a new generation of advanced methods are the followings. GALILEO is a state-of-the-art fuel rod performance code for PWR and BWR applications. Development is completed, implementation started in France and the U.S.A. ARCADIA-1 is a state-of-the-art neutronics/ thermal-hydraulics/ thermal-mechanics code system for PWR applications. Development is completed, implementation started in Europe and in the U.S.A. The system thermal-hydraulic codes S-RELAP5 and CATHARE-2 are not really new but still state-of-the-art in the domain. S-RELAP5 was completely restructured and re-coded such that its life cycle increases by further decades. CATHARE-2 will be replaced in the future by the new CATHARE-3. The new AREVA codes and methods are largely based on first principles modeling with an extremely broad international verification and validation data base. This enables AREVA and its customers to access more predictable licensing processes in a fast evolving regulatory environment (new safety criteria, requests for enlarged qualification databases, statistical applications, uncertainty propagation...). In this context, the advanced codes and methods and the associated verification and validation represent the key to avoiding penalties on products, on operational limits, or on methodologies themselves

  7. XSOR codes users manual

    International Nuclear Information System (INIS)

    Jow, Hong-Nian; Murfin, W.B.; Johnson, J.D.

    1993-11-01

    This report describes the source term estimation codes, XSORs. The codes are written for three pressurized water reactors (Surry, Sequoyah, and Zion) and two boiling water reactors (Peach Bottom and Grand Gulf). The ensemble of codes has been named ''XSOR''. The purpose of XSOR codes is to estimate the source terms which would be released to the atmosphere in severe accidents. A source term includes the release fractions of several radionuclide groups, the timing and duration of releases, the rates of energy release, and the elevation of releases. The codes have been developed by Sandia National Laboratories for the US Nuclear Regulatory Commission (NRC) in support of the NUREG-1150 program. The XSOR codes are fast running parametric codes and are used as surrogates for detailed mechanistic codes. The XSOR codes also provide the capability to explore the phenomena and their uncertainty which are not currently modeled by the mechanistic codes. The uncertainty distributions of input parameters may be used by an. XSOR code to estimate the uncertainty of source terms

  8. Recent advances in neutral particle transport methods and codes

    International Nuclear Information System (INIS)

    Azmy, Y.Y.

    1996-01-01

    An overview of ORNL's three-dimensional neutral particle transport code, TORT, is presented. Special features of the code that make it invaluable for large applications are summarized for the prospective user. Advanced capabilities currently under development and installation in the production release of TORT are discussed; they include: multitasking on Cray platforms running the UNICOS operating system; Adjacent cell Preconditioning acceleration scheme; and graphics codes for displaying computed quantities such as the flux. Further developments for TORT and its companion codes to enhance its present capabilities, as well as expand its range of applications are disucssed. Speculation on the next generation of neutron particle transport codes at ORNL, especially regarding unstructured grids and high order spatial approximations, are also mentioned

  9. Verification of the CONPAS (CONtainment Performance Analysis System) code package

    International Nuclear Information System (INIS)

    Kim, See Darl; Ahn, Kwang Il; Song, Yong Man; Choi, Young; Park, Soo Yong; Kim, Dong Ha; Jin, Young Ho.

    1997-09-01

    CONPAS is a computer code package to integrate the numerical, graphical, and results-oriented aspects of Level 2 probabilistic safety assessment (PSA) for nuclear power plants under a PC window environment automatically. For the integrated analysis of Level 2 PSA, the code utilizes four distinct, but closely related modules: (1) ET Editor, (2) Computer, (3) Text Editor, and (4) Mechanistic Code Plotter. Compared with other existing computer codes for Level 2 PSA, and CONPAS code provides several advanced features: computational aspects including systematic uncertainty analysis, importance analysis, sensitivity analysis and data interpretation, reporting aspects including tabling and graphic as well as user-friendly interface. The computational performance of CONPAS has been verified through a Level 2 PSA to a reference plant. The results of the CONPAS code was compared with an existing level 2 PSA code (NUCAP+) and the comparison proves that CONPAS is appropriate for Level 2 PSA. (author). 9 refs., 8 tabs., 14 figs

  10. Development of Advanced Suite of Deterministic Codes for VHTR Physics Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Seog; Cho, J. Y.; Lee, K. H. (and others)

    2007-07-15

    Advanced Suites of deterministic codes for VHTR physics analysis has been developed for detailed analysis of current and advanced reactor designs as part of a US-ROK collaborative I-NERI project. These code suites include the conventional 2-step procedure in which a few group constants are generated by a transport lattice calculation, and the reactor physics analysis is performed by a 3-dimensional diffusion calculation, and a whole core transport code that can model local heterogeneities directly at the core level. Particular modeling issues in physics analysis of the gas-cooled VHTRs were resolved, which include a double heterogeneity of the coated fuel particles, a neutron streaming in the coolant channels, a strong core-reflector interaction, and large spectrum shifts due to changes of the surrounding environment, temperature and burnup. And the geometry handling capability of the DeCART code were extended to deal with the hexagonal fuel elements of the VHTR core. The developed code suites were validated and verified by comparing the computational results with those of the Monte Carlo calculations for the benchmark problems.

  11. A 3-D CFD approach to the mechanistic prediction of forced convective critical heat flux at low quality

    International Nuclear Information System (INIS)

    Jean-Marie Le Corre; Cristina H Amon; Shi-Chune Yao

    2005-01-01

    Full text of publication follows: The prediction of the Critical Heat Flux (CHF) in a heat flux controlled boiling heat exchanger is important to assess the maximal thermal capability of the system. In the case of a nuclear reactor, CHF margin gain (using improved mixing vane grid design, for instance) can allow power up-rate and enhanced operating flexibility. In general, current nuclear core design procedures use quasi-1D approach to model the coolant thermal-hydraulic conditions within the fuel bundles coupled with fully empirical CHF prediction methods. In addition, several CHF mechanistic models have been developed in the past and coupled with 1D and quasi-1D thermal-hydraulic codes. These mechanistic models have demonstrated reasonable CHF prediction characteristics and, more remarkably, correct parametric trends over wide range of fluid conditions. However, since the phenomena leading to CHF are localized near the heater, models are needed to relate local quantities of interest to area-averaged quantities. As a consequence, large CHF prediction uncertainties may be introduced and 3D fluid characteristics (such as swirling flow) cannot be accounted properly. Therefore, a fully mechanistic approach to CHF prediction is, in general, not possible using the current approach. The development of CHF-enhanced fuel assembly designs requires the use of more advanced 3D coolant properties computations coupled with a CHF mechanistic modeling. In the present work, the commercial CFD code CFX-5 is used to compute 3D coolant conditions in a vertical heated tube with upward flow. Several CHF mechanistic models at low quality available in the literature are coupled with the CFD code by developing adequate models between local coolant properties and local parameters of interest to predict CHF. The prediction performances of these models are assessed using CHF databases available in the open literature and the 1995 CHF look-up table. Since CFD can reasonably capture 3D fluid

  12. Processes of code status transitions in hospitalized patients with advanced cancer.

    Science.gov (United States)

    El-Jawahri, Areej; Lau-Min, Kelsey; Nipp, Ryan D; Greer, Joseph A; Traeger, Lara N; Moran, Samantha M; D'Arpino, Sara M; Hochberg, Ephraim P; Jackson, Vicki A; Cashavelly, Barbara J; Martinson, Holly S; Ryan, David P; Temel, Jennifer S

    2017-12-15

    Although hospitalized patients with advanced cancer have a low chance of surviving cardiopulmonary resuscitation (CPR), the processes by which they change their code status from full code to do not resuscitate (DNR) are unknown. We conducted a mixed-methods study on a prospective cohort of hospitalized patients with advanced cancer. Two physicians used a consensus-driven medical record review to characterize processes that led to code status order transitions from full code to DNR. In total, 1047 hospitalizations were reviewed among 728 patients. Admitting clinicians did not address code status in 53% of hospitalizations, resulting in code status orders of "presumed full." In total, 275 patients (26.3%) transitioned from full code to DNR, and 48.7% (134 of 275 patients) of those had an order of "presumed full" at admission; however, upon further clarification, the patients expressed that they had wished to be DNR before the hospitalization. We identified 3 additional processes leading to order transition from full code to DNR acute clinical deterioration (15.3%), discontinuation of cancer-directed therapy (17.1%), and education about the potential harms/futility of CPR (15.3%). Compared with discontinuing therapy and education, transitions because of acute clinical deterioration were associated with less patient involvement (P = .002), a shorter time to death (P cancer were because of full code orders in patients who had a preference for DNR before hospitalization. Transitions due of acute clinical deterioration were associated with less patient engagement and a higher likelihood of inpatient death. Cancer 2017;123:4895-902. © 2017 American Cancer Society. © 2017 American Cancer Society.

  13. Computer code qualification program for the Advanced CANDU Reactor

    International Nuclear Information System (INIS)

    Popov, N.K.; Wren, D.J.; Snell, V.G.; White, A.J.; Boczar, P.G.

    2003-01-01

    Atomic Energy of Canada Ltd (AECL) has developed and implemented a Software Quality Assurance program (SQA) to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. This paper provides an overview of the computer programs used in Advanced CANDU Reactor (ACR) safety analysis, and assessment of their applicability in the safety analyses of the ACR design. An outline of the incremental validation program, and an overview of the experimental program in support of the code validation are also presented. An outline of the SQA program used to qualify these computer codes is also briefly presented. To provide context to the differences in the SQA with respect to current CANDUs, the paper also provides an overview of the ACR design features that have an impact on the computer code qualification. (author)

  14. Measurement and modeling of advanced coal conversion processes. Annual report, October 1990--September 1991

    Energy Technology Data Exchange (ETDEWEB)

    Solomon, P.R.; Serio, M.A.; Hamblen, D.G.; Smoot, L.D.; Brewster, B.S. [Advanced Fuel Research, Inc., East Hartford, CT (United States)]|[Brigham Young Univ., Provo, UT (United States)

    1991-12-31

    The overall objective of this program is the development of predictive capability for the design, scale up, simulation, control and feedstock evaluation in advanced coal conversion devices. This program will merge significant advances made in measuring and quantitatively describing the mechanisms in coal conversion behavior. Comprehensive computer codes for mechanistic modeling of entrained-bed gasification. Additional capabilities in predicting pollutant formation will be implemented and the technology will be expanded to fixed-bed reactors.

  15. Advanced video coding systems

    CERN Document Server

    Gao, Wen

    2015-01-01

    This comprehensive and accessible text/reference presents an overview of the state of the art in video coding technology. Specifically, the book introduces the tools of the AVS2 standard, describing how AVS2 can help to achieve a significant improvement in coding efficiency for future video networks and applications by incorporating smarter coding tools such as scene video coding. Topics and features: introduces the basic concepts in video coding, and presents a short history of video coding technology and standards; reviews the coding framework, main coding tools, and syntax structure of AV

  16. Application of advanced validation concepts to oxide fuel performance codes: LIFE-4 fast-reactor and FRAPCON thermal-reactor fuel performance codes

    Energy Technology Data Exchange (ETDEWEB)

    Unal, C., E-mail: cu@lanl.gov [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Williams, B.J. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Yacout, A. [Argonne National Laboratory, 9700 S. Cass Avenue, Lemont, IL 60439 (United States); Higdon, D.M. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)

    2013-10-15

    Highlights: ► The application of advanced validation techniques (sensitivity, calibration and prediction) to nuclear performance codes FRAPCON and LIFE-4 is the focus of the paper. ► A sensitivity ranking methodology narrows down the number of selected modeling parameters from 61 to 24 for FRAPCON and from 69 to 35 for LIFE-4. ► Fuel creep, fuel thermal conductivity, fission gas transport/release, crack/boundary, and fuel gap conductivity models of LIFE-4 are identified for improvements. ► FRAPCON sensitivity results indicated the importance of the fuel thermal conduction and the fission gas release models. -- Abstract: Evolving nuclear energy programs expect to use enhanced modeling and simulation (M and S) capabilities, using multiscale, multiphysics modeling approaches, to reduce both cost and time from the design through the licensing phases. Interest in the development of the multiscale, multiphysics approach has increased in the last decade because of the need for predictive tools for complex interacting processes as a means of eliminating the limited use of empirically based model development. Complex interacting processes cannot be predicted by analyzing each individual component in isolation. In most cases, the mathematical models of complex processes and their boundary conditions are nonlinear. As a result, the solutions of these mathematical models often require high-performance computing capabilities and resources. The use of multiscale, multiphysics (MS/MP) models in conjunction with high-performance computational software and hardware introduces challenges in validating these predictive tools—traditional methodologies will have to be modified to address these challenges. The advanced MS/MP codes for nuclear fuels and reactors are being developed within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program of the US Department of Energy (DOE) – Nuclear Energy (NE). This paper does not directly address challenges in calibration

  17. The integrated code system CASCADE-3D for advanced core design and safety analysis

    International Nuclear Information System (INIS)

    Neufert, A.; Van de Velde, A.

    1999-01-01

    The new program system CASCADE-3D (Core Analysis and Safety Codes for Advanced Design Evaluation) links some of Siemens advanced code packages for in-core fuel management and accident analysis: SAV95, PANBOX/COBRA and RELAP5. Consequently by using CASCADE-3D the potential of modern fuel assemblies and in-core fuel management strategies can be much better utilized because safety margins which had been reduced due to conservative methods are now predicted more accurately. By this innovative code system the customers can now take full advantage of the recent progress in fuel assembly design and in-core fuel management.(author)

  18. FASTDART - A fast, accurate and friendly version of DART code

    International Nuclear Information System (INIS)

    Rest, Jeffrey; Taboada, Horacio

    2000-01-01

    A new enhanced, visual version of DART code is presented. DART is a mechanistic model based code, developed for the performance calculation and assessment of aluminum dispersion fuel. Major issues of this new version are the development of a new, time saving calculation routine, able to be run on PC, a friendly visual input interface and a plotting facility. This version, available for silicide and U-Mo fuels, adds to the classical accuracy of DART models for fuel performance prediction, a faster execution and visual interfaces. It is part of a collaboration agreement between ANL and CNEA in the area of Low Enriched Uranium Advanced Fuels, held by the Implementation Arrangement for Technical Exchange and Cooperation in the Area of Peaceful Uses of Nuclear Energy. (author)

  19. FAST: An advanced code system for fast reactor transient analysis

    International Nuclear Information System (INIS)

    Mikityuk, Konstantin; Pelloni, Sandro; Coddington, Paul; Bubelis, Evaldas; Chawla, Rakesh

    2005-01-01

    One of the main goals of the FAST project at PSI is to establish a unique analytical code capability for the core and safety analysis of advanced critical (and sub-critical) fast-spectrum systems for a wide range of different coolants. Both static and transient core physics, as well as the behaviour and safety of the power plant as a whole, are studied. The paper discusses the structure of the code system, including the organisation of the interfaces and data exchange. Examples of validation and application of the individual programs, as well as of the complete code system, are provided using studies carried out within the context of designs for experimental accelerator-driven, fast-spectrum systems

  20. Advanced thermal-hydraulic and neutronic codes: current and future applications. Summary and conclusions

    International Nuclear Information System (INIS)

    2001-05-01

    An OECD Workshop on Advanced Thermal-Hydraulic and Neutronic Codes Applications was held from 10 to 13 April 2000, in Barcelona, Spain, sponsored by the Committee on the Safety of Nuclear Installations (CSNI) of the OECD Nuclear Energy Agency (NEA). It was organised in collaboration with the Spanish Nuclear Safety Council (CSN) and hosted by CSN and the Polytechnic University of Catalonia (UPC) in collaboration with the Spanish Electricity Association (UNESA). The objectives of the Workshop were to review the developments since the previous CSNI Workshop held in Annapolis [NEA/CSNI/ R(97)4; NUREG/CP-0159], to analyse the present status of maturity and remnant needs of thermal-hydraulic (TH) and neutronic system codes and methods, and finally to evaluate the role of these tools in the evolving regulatory environment. The Technical Sessions and Discussion Sessions covered the following topics: - Regulatory requirements for Best-Estimate (BE) code assessment; - Application of TH and neutronic codes for current safety issues; - Uncertainty analysis; - Needs for integral plant transient and accident analysis; - Simulators and fast running codes; - Advances in next generation TH and neutronic codes; - Future trends in physical modeling; - Long term plans for development of advanced codes. The focus of the Workshop was on system codes. An incursion was made, however, in the new field of applying Computational Fluid Dynamic (CFD) codes to nuclear safety analysis. As a general conclusion, the Barcelona Workshop can be considered representative of the progress towards the targets marked at Annapolis almost four years ago. The Annapolis Workshop had identified areas where further development and specific improvements were needed, among them: multi-field models, transport of interfacial area, 2D and 3D thermal-hydraulics, 3-D neutronics consistent with level of details of thermal-hydraulics. Recommendations issued at Annapolis included: developing small pilot/test codes for

  1. On-line application of the PANTHER advanced nodal code

    International Nuclear Information System (INIS)

    Hutt, P.K.; Knight, M.P.

    1992-01-01

    Over the last few years, Nuclear Electric has developed an integrated core performance code package for both light water reactors (LWRs) and advanced gas-cooled reactors (AGRs) that can perform a comprehensive range of calculations for fuel cycle design, safety analysis, and on-line operational support for such plants. The package consists of the following codes: WIMS for lattice physics, PANTHER whole reactor nodal flux and AGR thermal hydraulics, VIPRE for LWR thermal hydraulics, and ENIGMA for fuel performance. These codes are integrated within a UNIX-based interactive system called the Reactor Physics Workbench (RPW), which provides an interactive graphic user interface and quality assurance records/data management. The RPW can also control calculational sequences and data flows. The package has been designed to run both off-line and on-line accessing plant data through the RPW

  2. Iterative Systems Biology for Medicine – time for advancing from network signature to mechanistic equations

    KAUST Repository

    Gomez-Cabrero, David

    2017-05-09

    The rise and growth of Systems Biology following the sequencing of the human genome has been astounding. Early on, an iterative wet-dry methodology was formulated which turned out as a successful approach in deciphering biological complexity. Such type of analysis effectively identified and associated molecular network signatures operative in biological processes across different systems. Yet, it has proven difficult to distinguish between causes and consequences, thus making it challenging to attack medical questions where we require precise causative drug targets and disease mechanisms beyond a web of associated markers. Here we review principal advances with regard to identification of structure, dynamics, control, and design of biological systems, following the structure in the visionary review from 2002 by Dr. Kitano. Yet, here we find that the underlying challenge of finding the governing mechanistic system equations enabling precision medicine remains open thus rendering clinical translation of systems biology arduous. However, stunning advances in raw computational power, generation of high-precision multi-faceted biological data, combined with powerful algorithms hold promise to set the stage for data-driven identification of equations implicating a fundamental understanding of living systems during health and disease.

  3. High burnup models in computer code fair

    Energy Technology Data Exchange (ETDEWEB)

    Dutta, B K; Swami Prasad, P; Kushwaha, H S; Mahajan, S C; Kakodar, A [Bhabha Atomic Research Centre, Bombay (India)

    1997-08-01

    An advanced fuel analysis code FAIR has been developed for analyzing the behavior of fuel rods of water cooled reactors under severe power transients and high burnups. The code is capable of analyzing fuel pins of both collapsible clad, as in PHWR and free standing clad as in LWR. The main emphasis in the development of this code is on evaluating the fuel performance at extended burnups and modelling of the fuel rods for advanced fuel cycles. For this purpose, a number of suitable models have been incorporated in FAIR. For modelling the fission gas release, three different models are implemented, namely Physically based mechanistic model, the standard ANS 5.4 model and the Halden model. Similarly the pellet thermal conductivity can be modelled by the MATPRO equation, the SIMFUEL relation or the Halden equation. The flux distribution across the pellet is modelled by using the model RADAR. For modelling pellet clad interaction (PCMI)/ stress corrosion cracking (SCC) induced failure of sheath, necessary routines are provided in FAIR. The validation of the code FAIR is based on the analysis of fuel rods of EPRI project ``Light water reactor fuel rod modelling code evaluation`` and also the analytical simulation of threshold power ramp criteria of fuel rods of pressurized heavy water reactors. In the present work, a study is carried out by analysing three CRP-FUMEX rods to show the effect of various combinations of fission gas release models and pellet conductivity models, on the fuel analysis parameters. The satisfactory performance of FAIR may be concluded through these case studies. (author). 12 refs, 5 figs.

  4. High burnup models in computer code fair

    International Nuclear Information System (INIS)

    Dutta, B.K.; Swami Prasad, P.; Kushwaha, H.S.; Mahajan, S.C.; Kakodar, A.

    1997-01-01

    An advanced fuel analysis code FAIR has been developed for analyzing the behavior of fuel rods of water cooled reactors under severe power transients and high burnups. The code is capable of analyzing fuel pins of both collapsible clad, as in PHWR and free standing clad as in LWR. The main emphasis in the development of this code is on evaluating the fuel performance at extended burnups and modelling of the fuel rods for advanced fuel cycles. For this purpose, a number of suitable models have been incorporated in FAIR. For modelling the fission gas release, three different models are implemented, namely Physically based mechanistic model, the standard ANS 5.4 model and the Halden model. Similarly the pellet thermal conductivity can be modelled by the MATPRO equation, the SIMFUEL relation or the Halden equation. The flux distribution across the pellet is modelled by using the model RADAR. For modelling pellet clad interaction (PCMI)/ stress corrosion cracking (SCC) induced failure of sheath, necessary routines are provided in FAIR. The validation of the code FAIR is based on the analysis of fuel rods of EPRI project ''Light water reactor fuel rod modelling code evaluation'' and also the analytical simulation of threshold power ramp criteria of fuel rods of pressurized heavy water reactors. In the present work, a study is carried out by analysing three CRP-FUMEX rods to show the effect of various combinations of fission gas release models and pellet conductivity models, on the fuel analysis parameters. The satisfactory performance of FAIR may be concluded through these case studies. (author). 12 refs, 5 figs

  5. TANDA TANGAN DIGITAL MENGGUNAKAN QR CODE DENGAN METODE ADVANCED ENCRYPTION STANDARD

    Directory of Open Access Journals (Sweden)

    Abdul Gani Putra Suratma

    2017-04-01

    Full Text Available Tanda tangan digital (digital signature adalah sebuah skema matematis yang secara unik mengidentifikasikan seorang pengirim, sekaligus untuk membuktikan keaslian dari pemilik sebuah pesan atau dokumen digital, sehingga sebuah tanda tangan digital yang autentik (sah, sudah cukup menjadi alasan bagi penerima un- tuk percaya bahwa sebuah pesan atau dokumen yang diterima adalah berasal dari pengirim yang telah diketahui. Perkembangan teknologi memungkinkan adanya tanda tangan digital yang dapat digunakan untuk melakukan pembuktian secara matematis, sehingga informasi yang didapat oleh satu pihak dari pihak lain dapat diidentifikasi untuk memastikan keaslian informasi yang diterima. Tanda tangan digital merupakan mekanisme otentikasi yang memungkinkan pembuat pesan menambahkan sebuah kode yang bertindak sebagai tanda tangannya. Tujuan dari penelitian ini menerapkan QR Code atau yang dikenal dengan istilah QR (Quick Respon dan Algoritma yang akan ditambahkan yaitu AES (Advanced Encryption Standard sebagai tanda tangan digital sehingga hasil dari penelitian penerapan QR Code menggunakan algoritma Advanced Encryption Standard sebagai tanda tangan digital dapat berfungsi sebagai otentikasi tanda tangan pimpinan serta ve- rivikasi dokumen pengambilan barang yang sah. dari penelitian ini akurasi klasifi- kasi QR Code dengan menggunakan naïve bayes classifier sebesar 90% dengan precision positif sebesar 80% dan precision negatif sebesar 100%.

  6. Development of an advanced code system for fast-reactor transient analysis

    International Nuclear Information System (INIS)

    Konstantin Mikityuk; Sandro Pelloni; Paul Coddington

    2005-01-01

    FAST (Fast-spectrum Advanced Systems for power production and resource management) is a recently approved PSI activity in the area of fast spectrum core and safety analysis with emphasis on generic developments and Generation IV systems. In frames of the FAST project we will study both statics and transients core physics, reactor system behaviour and safety; related international experiments. The main current goal of the project is to develop unique analytical and code capability for core and safety analysis of critical (and sub-critical) fast spectrum systems with an initial emphasis on a gas cooled fast reactors. A structure of the code system is shown on Fig. 1. The main components of the FAST code system are 1) ERANOS code for preparation of basic x-sections and their partial derivatives; 2) PARCS transient nodal-method multi-group neutron diffusion code for simulation of spatial (3D) neutron kinetics in hexagonal and square geometries; 3) TRAC/AAA code for system thermal hydraulics; 4) FRED transient model for fuel thermal-mechanical behaviour; 5) PVM system as an interface between separate parts of the code system. The paper presents a structure of the code system (Fig. 1), organization of interfaces and data exchanges between main parts of the code system, examples of verification and application of separate codes and the system as a whole. (authors)

  7. Application and analysis of performance of dqpsk advanced modulation format in spectral amplitude coding ocdma

    International Nuclear Information System (INIS)

    Memon, A.

    2015-01-01

    SAC (Spectral Amplitude Coding) is a technique of OCDMA (Optical Code Division Multiple Access) to encode and decode data bits by utilizing spectral components of the broadband source. Usually OOK (ON-Off-Keying) modulation format is used in this encoding scheme. To make SAC OCDMA network spectrally efficient, advanced modulation format of DQPSK (Differential Quaternary Phase Shift Keying) is applied, simulated and analyzed, m-sequence code is encoded in the simulated setup. Performance regarding various lengths of m-sequence code is also analyzed and displayed in the pictorial form. The results of the simulation are evaluated with the help of electrical constellation diagram, eye diagram and bit error rate graph. All the graphs indicate better transmission quality in case of advanced modulation format of DQPSK used in SAC OCDMA network as compared with OOK. (author)

  8. Generative mechanistic explanation building in undergraduate molecular and cellular biology

    Science.gov (United States)

    Southard, Katelyn M.; Espindola, Melissa R.; Zaepfel, Samantha D.; Bolger, Molly S.

    2017-09-01

    When conducting scientific research, experts in molecular and cellular biology (MCB) use specific reasoning strategies to construct mechanistic explanations for the underlying causal features of molecular phenomena. We explored how undergraduate students applied this scientific practice in MCB. Drawing from studies of explanation building among scientists, we created and applied a theoretical framework to explore the strategies students use to construct explanations for 'novel' biological phenomena. Specifically, we explored how students navigated the multi-level nature of complex biological systems using generative mechanistic reasoning. Interviews were conducted with introductory and upper-division biology students at a large public university in the United States. Results of qualitative coding revealed key features of students' explanation building. Students used modular thinking to consider the functional subdivisions of the system, which they 'filled in' to varying degrees with mechanistic elements. They also hypothesised the involvement of mechanistic entities and instantiated abstract schema to adapt their explanations to unfamiliar biological contexts. Finally, we explored the flexible thinking that students used to hypothesise the impact of mutations on multi-leveled biological systems. Results revealed a number of ways that students drew mechanistic connections between molecules, functional modules (sets of molecules with an emergent function), cells, tissues, organisms and populations.

  9. Validation of physics and thermalhydraulic computer codes for advanced Candu reactor applications

    International Nuclear Information System (INIS)

    Wren, D.J.; Popov, N.; Snell, V.G.

    2004-01-01

    Atomic Energy of Canada Ltd. (AECL) is developing an Advanced Candu Reactor (ACR) that is an evolutionary advancement of the currently operating Candu 6 reactors. The ACR is being designed to produce electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The ACR retains the modular Candu concept of horizontal fuel channels surrounded by a heavy water moderator. However, ACR uses slightly enriched uranium fuel compared to the natural uranium used in Candu 6. This achieves the twin goals of improved economics (via large reductions in the heavy water moderator volume and replacement of the heavy water coolant with light water coolant) and improved safety. AECL has developed and implemented a software quality assurance program to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. Since the basic design of the ACR is equivalent to that of the Candu 6, most of the key phenomena associated with the safety analyses of ACR are common, and the Candu industry standard tool-set of safety analysis codes can be applied to the analysis of the ACR. A systematic assessment of computer code applicability addressing the unique features of the ACR design was performed covering the important aspects of the computer code structure, models, constitutive correlations, and validation database. Arising from this assessment, limited additional requirements for code modifications and extensions to the validation databases have been identified. This paper provides an outline of the AECL software quality assurance program process for the validation of computer codes used to perform physics and thermal-hydraulics safety analyses of the ACR. It describes the additional validation work that has been identified for these codes and the planned, and ongoing, experimental programs to extend the code validation as required to address specific ACR design

  10. Micromagnetic Code Development of Advanced Magnetic Structures Final Report CRADA No. TC-1561-98

    Energy Technology Data Exchange (ETDEWEB)

    Cerjan, Charles J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Shi, Xizeng [Read-Rite Corporation, Fremont, CA (United States)

    2017-11-09

    The specific goals of this project were to: Further develop the previously written micromagnetic code DADIMAG (DOE code release number 980017); Validate the code. The resulting code was expected to be more realistic and useful for simulations of magnetic structures of specific interest to Read-Rite programs. We also planned to further the code for use in internal LLNL programs. This project complemented LLNL CRADA TC-840-94 between LLNL and Read-Rite, which allowed for simulations of the advanced magnetic head development completed under the CRADA. TC-1561-98 was effective concurrently with LLNL non-exclusive copyright license (TL-1552-98) to Read-Rite for DADIMAG Version 2 executable code.

  11. Application and Analysis of Performance of DQPSK Advanced Modulation Format in Spectral Amplitude Coding OCDMA

    Directory of Open Access Journals (Sweden)

    Abdul Latif Memon

    2014-04-01

    Full Text Available SAC (Spectral Amplitude Coding is a technique of OCDMA (Optical Code Division Multiple Access to encode and decode data bits by utilizing spectral components of the broadband source. Usually OOK (ON-Off-Keying modulation format is used in this encoding scheme. To make SAC OCDMA network spectrally efficient, advanced modulation format of DQPSK (Differential Quaternary Phase Shift Keying is applied, simulated and analyzed. m-sequence code is encoded in the simulated setup. Performance regarding various lengths of m-sequence code is also analyzed and displayed in the pictorial form. The results of the simulation are evaluated with the help of electrical constellation diagram, eye diagram and bit error rate graph. All the graphs indicate better transmission quality in case of advanced modulation format of DQPSK used in SAC OCDMA network as compared with OOK

  12. Utilization of MCNP code in the research and design for China advanced research reactor

    International Nuclear Information System (INIS)

    Shen Feng

    2006-01-01

    MCNP, which is the internationalized neutronics code, is used for nuclear research and design in China Advanced Research Reactor (CARR). MCNP is an important neutronics code in the research and design for CARR since many calculation tasks could be undertaken by it. Many nuclear parameters on reactor core, the design and optimization research for many reactor utilizations, much verification for other nuclear calculation code and so on are conducted with help of MCNP. (author)

  13. Proceedings of the workshop on advanced thermal-hydraulic and neutronic codes: current and future applications

    International Nuclear Information System (INIS)

    2001-01-01

    An OECD Workshop on Advanced Thermal-Hydraulic and Neutronic Codes Applications was held from 10 to 13 April 2000, in Barcelona, Spain, sponsored by the Committee on the Safety of Nuclear Installations (CSNI) of the OECD Nuclear Energy Agency (NEA). It was organised in collaboration with the Spanish Nuclear Safety Council (CSN) and hosted by CSN and the Polytechnic University of Catalonia (UPC) in collaboration with the Spanish Electricity Association (UNESA). The objectives of the Workshop were to review the developments since the previous CSNI Workshop held in Annapolis [NEA/CSNI/ R(97)4; NUREG/CP-0159], to analyse the present status of maturity and remnant needs of thermal-hydraulic (TH) and neutronic system codes and methods, and finally to evaluate the role of these tools in the evolving regulatory environment. The Technical Sessions and Discussion Sessions covered the following topics: - Regulatory requirements for Best-Estimate (BE) code assessment; - Application of TH and neutronic codes for current safety issues; - Uncertainty analysis; - Needs for integral plant transient and accident analysis; - Simulators and fast running codes; - Advances in next generation TH and neutronic codes; - Future trends in physical modeling; - Long term plans for development of advanced codes. The focus of the Workshop was on system codes. An incursion was made, however, in the new field of applying Computational Fluid Dynamic (CFD) codes to nuclear safety analysis. As a general conclusion, the Barcelona Workshop can be considered representative of the progress towards the targets marked at Annapolis almost four years ago. The Annapolis Workshop had identified areas where further development and specific improvements were needed, among them: multi-field models, transport of interfacial area, 2D and 3D thermal-hydraulics, 3-D neutronics consistent with level of details of thermal-hydraulics. Recommendations issued at Annapolis included: developing small pilot/test codes for

  14. Development of an advanced fluid-dynamic analysis code: α-flow

    International Nuclear Information System (INIS)

    Akiyama, Mamoru

    1990-01-01

    A Project for development of large scale three-dimensional fluid-dynamic analysis code, α-FLOW, coping with the recent advancement of supercomputers and workstations, has been in progress. This project is called the α-Project, which has been promoted by the Association for Large Scale Fluid Dynamics Analysis Code comprising private companies and research institutions such as universities. The developmental period for the α-FLOW is four years, March 1989 to March 1992. To date, the major portions of basic design and program preparation have been completed and the project is in the stage of testing each module. In this paper, the present status of the α-Project, design policy and outline of α-FLOW are described. (author)

  15. ADVANCED ELECTRIC AND MAGNETIC MATERIAL MODELS FOR FDTD ELECTROMAGNETIC CODES

    Energy Technology Data Exchange (ETDEWEB)

    Poole, B R; Nelson, S D; Langdon, S

    2005-05-05

    The modeling of dielectric and magnetic materials in the time domain is required for pulse power applications, pulsed induction accelerators, and advanced transmission lines. For example, most induction accelerator modules require the use of magnetic materials to provide adequate Volt-sec during the acceleration pulse. These models require hysteresis and saturation to simulate the saturation wavefront in a multipulse environment. In high voltage transmission line applications such as shock or soliton lines the dielectric is operating in a highly nonlinear regime, which require nonlinear models. Simple 1-D models are developed for fast parameterization of transmission line structures. In the case of nonlinear dielectrics, a simple analytic model describing the permittivity in terms of electric field is used in a 3-D finite difference time domain code (FDTD). In the case of magnetic materials, both rate independent and rate dependent Hodgdon magnetic material models have been implemented into 3-D FDTD codes and 1-D codes.

  16. ADVANCED ELECTRIC AND MAGNETIC MATERIAL MODELS FOR FDTD ELECTROMAGNETIC CODES

    International Nuclear Information System (INIS)

    Poole, B R; Nelson, S D; Langdon, S

    2005-01-01

    The modeling of dielectric and magnetic materials in the time domain is required for pulse power applications, pulsed induction accelerators, and advanced transmission lines. For example, most induction accelerator modules require the use of magnetic materials to provide adequate Volt-sec during the acceleration pulse. These models require hysteresis and saturation to simulate the saturation wavefront in a multipulse environment. In high voltage transmission line applications such as shock or soliton lines the dielectric is operating in a highly nonlinear regime, which require nonlinear models. Simple 1-D models are developed for fast parameterization of transmission line structures. In the case of nonlinear dielectrics, a simple analytic model describing the permittivity in terms of electric field is used in a 3-D finite difference time domain code (FDTD). In the case of magnetic materials, both rate independent and rate dependent Hodgdon magnetic material models have been implemented into 3-D FDTD codes and 1-D codes

  17. Development and Application of Subchannel Analysis Code Technology for Advanced Reactor Systems

    International Nuclear Information System (INIS)

    Hwang, Dae Hyun; Seo, K. W.

    2006-01-01

    A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled reactor condition and supercritical pressure conditions with water and Freon. A subchannel analysis has been conducted for European ADS(Accelerator-Driven subcritical System) with Pb-Bi coolant through the international cooperation work between KAERI and FZK, Germany. In addition, the prediction capability of the subchannel code was evaluated for the subchannel void distribution data by participating an international code benchmark program which was organized by OECD/NRC

  18. Development and Application of Subchannel Analysis Code Technology for Advanced Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hyun; Seo, K. W

    2006-01-15

    A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled reactor condition and supercritical pressure conditions with water and Freon. A subchannel analysis has been conducted for European ADS(Accelerator-Driven subcritical System) with Pb-Bi coolant through the international cooperation work between KAERI and FZK, Germany. In addition, the prediction capability of the subchannel code was evaluated for the subchannel void distribution data by participating an international code benchmark program which was organized by OECD/NRC.

  19. The 419 codes as business unusual: the advance fee fraud online ...

    African Journals Online (AJOL)

    The 419 codes as business unusual: the advance fee fraud online discourse. A Adogame. Abstract. No Abstract. International Journal of Humanistic Studies Vol. 5 2006: pp. 54-72. AJOL African Journals Online. HOW TO USE AJOL... for Researchers · for Librarians · for Authors · FAQ's · More about AJOL · AJOL's Partners ...

  20. Beam Optics Analysis - An Advanced 3D Trajectory Code

    International Nuclear Information System (INIS)

    Ives, R. Lawrence; Bui, Thuc; Vogler, William; Neilson, Jeff; Read, Mike; Shephard, Mark; Bauer, Andrew; Datta, Dibyendu; Beal, Mark

    2006-01-01

    Calabazas Creek Research, Inc. has completed initial development of an advanced, 3D program for modeling electron trajectories in electromagnetic fields. The code is being used to design complex guns and collectors. Beam Optics Analysis (BOA) is a fully relativistic, charged particle code using adaptive, finite element meshing. Geometrical input is imported from CAD programs generating ACIS-formatted files. Parametric data is inputted using an intuitive, graphical user interface (GUI), which also provides control of convergence, accuracy, and post processing. The program includes a magnetic field solver, and magnetic information can be imported from Maxwell 2D/3D and other programs. The program supports thermionic emission and injected beams. Secondary electron emission is also supported, including multiple generations. Work on field emission is in progress as well as implementation of computer optimization of both the geometry and operating parameters. The principle features of the program and its capabilities are presented

  1. Grammar Coding in the "Oxford Advanced Learner's Dictionary of Current English."

    Science.gov (United States)

    Wekker, Herman

    1992-01-01

    Focuses on the revised system of grammar coding for verbs in the fourth edition of the "Oxford Advanced Learner's Dictionary of Current English" (OALD4), comparing it with two other similar dictionaries. It is shown that the OALD4 is found to be more favorable on many criteria than the other comparable dictionaries. (16 references) (VWL)

  2. Recent advances in mathematical modeling of developmental abnormalities using mechanistic information.

    Science.gov (United States)

    Kavlock, R J

    1997-01-01

    During the last several years, significant changes in the risk assessment process for developmental toxicity of environmental contaminants have begun to emerge. The first of these changes is the development and beginning use of statistically based dose-response models [the benchmark dose (BMD) approach] that better utilize data derived from existing testing approaches. Accompanying this change is the greater emphasis placed on understanding and using mechanistic information to yield more accurate, reliable, and less uncertain risk assessments. The next stage in the evolution of risk assessment will be the use of biologically based dose-response (BBDR) models that begin to build into the statistically based models factors related to the underlying kinetic, biochemical, and/or physiologic processes perturbed by a toxicant. Such models are now emerging from several research laboratories. The introduction of quantitative models and the incorporation of biologic information into them has pointed to the need for even more sophisticated modifications for which we offer the term embryologically based dose-response (EBDR) models. Because these models would be based upon the understanding of normal morphogenesis, they represent a quantum leap in our thinking, but their complexity presents daunting challenges both to the developmental biologist and the developmental toxicologist. Implementation of these models will require extensive communication between developmental toxicologists, molecular embryologists, and biomathematicians. The remarkable progress in the understanding of mammalian embryonic development at the molecular level that has occurred over the last decade combined with advances in computing power and computational models should eventually enable these as yet hypothetical models to be brought into use.

  3. Functions of Code-Switching among Iranian Advanced and Elementary Teachers and Students

    Science.gov (United States)

    Momenian, Mohammad; Samar, Reza Ghafar

    2011-01-01

    This paper reports on the findings of a study carried out on the advanced and elementary teachers' and students' functions and patterns of code-switching in Iranian English classrooms. This concept has not been adequately examined in L2 (second language) classroom contexts than in outdoor natural contexts. Therefore, besides reporting on the…

  4. A metabonomic approach for mechanistic exploration of pre-clinical toxicology.

    Science.gov (United States)

    Coen, Muireann

    2010-12-30

    Metabonomics involves the application of advanced analytical tools to profile the diverse metabolic complement of a given biofluid or tissue. Subsequent statistical modelling of the complex multivariate spectral profiles enables discrimination between phenotypes of interest and identifies panels of discriminatory metabolites that represent candidate biomarkers. This review article presents an overview of recent developments in the field of metabonomics with a focus on application to pre-clinical toxicology studies. Recent research investigations carried out as part of the international COMET 2 consortium project on the hepatotoxic action of the aminosugar, galactosamine (galN) are presented. The application of advanced, high-field NMR spectroscopy is demonstrated, together with complementary application of a targeted mass spectrometry platform coupled with ultra-performance liquid chromatography. Much novel mechanistic information has been gleaned on both the mechanism of galN hepatotoxicity in multiple biofluids and tissues, and on the protection afforded by co-administration of glycine and uridine. The simultaneous identification of both the metabolic fate of galN and its associated endogenous consequences in spectral profiles is demonstrated. Furthermore, metabonomic assessment of inter-animal variability in response to galN presents enhanced mechanistic insight on variable response phentoypes and is relevant to understanding wider aspects of individual variability in drug response. This exemplar highlights the analytical and statistical tools commonly applied in metabonomic studies and notably, the approach is applicable to the study of any toxin/drug or intervention of interest. The metabonomic approach holds considerable promise and potential to significantly advance our understanding of the mechanistic bases for adverse drug reactions. Copyright © 2010 Elsevier Ireland Ltd. All rights reserved.

  5. Specification of advanced safety modeling requirements (Rev. 0).

    Energy Technology Data Exchange (ETDEWEB)

    Fanning, T. H.; Tautges, T. J.

    2008-06-30

    The U.S. Department of Energy's Global Nuclear Energy Partnership has lead to renewed interest in liquid-metal-cooled fast reactors for the purpose of closing the nuclear fuel cycle and making more efficient use of future repository capacity. However, the U.S. has not designed or constructed a fast reactor in nearly 30 years. Accurate, high-fidelity, whole-plant dynamics safety simulations will play a crucial role by providing confidence that component and system designs will satisfy established design limits and safety margins under a wide variety of operational, design basis, and beyond design basis transient conditions. Current modeling capabilities for fast reactor safety analyses have resulted from several hundred person-years of code development effort supported by experimental validation. The broad spectrum of mechanistic and phenomenological models that have been developed represent an enormous amount of institutional knowledge that needs to be maintained. Complicating this, the existing code architectures for safety modeling evolved from programming practices of the 1970s. This has lead to monolithic applications with interdependent data models which require significant knowledge of the complexities of the entire code in order for each component to be maintained. In order to develop an advanced fast reactor safety modeling capability, the limitations of the existing code architecture must be overcome while preserving the capabilities that already exist. To accomplish this, a set of advanced safety modeling requirements is defined, based on modern programming practices, that focuses on modular development within a flexible coupling framework. An approach for integrating the existing capabilities of the SAS4A/SASSYS-1 fast reactor safety analysis code into the SHARP framework is provided in order to preserve existing capabilities while providing a smooth transition to advanced modeling capabilities. In doing this, the advanced fast reactor safety models

  6. Specification of advanced safety modeling requirements (Rev. 0)

    International Nuclear Information System (INIS)

    Fanning, T. H.; Tautges, T. J.

    2008-01-01

    The U.S. Department of Energy's Global Nuclear Energy Partnership has lead to renewed interest in liquid-metal-cooled fast reactors for the purpose of closing the nuclear fuel cycle and making more efficient use of future repository capacity. However, the U.S. has not designed or constructed a fast reactor in nearly 30 years. Accurate, high-fidelity, whole-plant dynamics safety simulations will play a crucial role by providing confidence that component and system designs will satisfy established design limits and safety margins under a wide variety of operational, design basis, and beyond design basis transient conditions. Current modeling capabilities for fast reactor safety analyses have resulted from several hundred person-years of code development effort supported by experimental validation. The broad spectrum of mechanistic and phenomenological models that have been developed represent an enormous amount of institutional knowledge that needs to be maintained. Complicating this, the existing code architectures for safety modeling evolved from programming practices of the 1970s. This has lead to monolithic applications with interdependent data models which require significant knowledge of the complexities of the entire code in order for each component to be maintained. In order to develop an advanced fast reactor safety modeling capability, the limitations of the existing code architecture must be overcome while preserving the capabilities that already exist. To accomplish this, a set of advanced safety modeling requirements is defined, based on modern programming practices, that focuses on modular development within a flexible coupling framework. An approach for integrating the existing capabilities of the SAS4A/SASSYS-1 fast reactor safety analysis code into the SHARP framework is provided in order to preserve existing capabilities while providing a smooth transition to advanced modeling capabilities. In doing this, the advanced fast reactor safety models will

  7. Channel coding/decoding alternatives for compressed TV data on advanced planetary missions.

    Science.gov (United States)

    Rice, R. F.

    1972-01-01

    The compatibility of channel coding/decoding schemes with a specific TV compressor developed for advanced planetary missions is considered. Under certain conditions, it is shown that compressed data can be transmitted at approximately the same rate as uncompressed data without any loss in quality. Thus, the full gains of data compression can be achieved in real-time transmission.

  8. Mechanistic species distribution modelling as a link between physiology and conservation.

    Science.gov (United States)

    Evans, Tyler G; Diamond, Sarah E; Kelly, Morgan W

    2015-01-01

    Climate change conservation planning relies heavily on correlative species distribution models that estimate future areas of occupancy based on environmental conditions encountered in present-day ranges. The approach benefits from rapid assessment of vulnerability over a large number of organisms, but can have poor predictive power when transposed to novel environments and reveals little in the way of causal mechanisms that define changes in species distribution or abundance. Having conservation planning rely largely on this single approach also increases the risk of policy failure. Mechanistic models that are parameterized with physiological information are expected to be more robust when extrapolating distributions to future environmental conditions and can identify physiological processes that set range boundaries. Implementation of mechanistic species distribution models requires knowledge of how environmental change influences physiological performance, and because this information is currently restricted to a comparatively small number of well-studied organisms, use of mechanistic modelling in the context of climate change conservation is limited. In this review, we propose that the need to develop mechanistic models that incorporate physiological data presents an opportunity for physiologists to contribute more directly to climate change conservation and advance the field of conservation physiology. We begin by describing the prevalence of species distribution modelling in climate change conservation, highlighting the benefits and drawbacks of both mechanistic and correlative approaches. Next, we emphasize the need to expand mechanistic models and discuss potential metrics of physiological performance suitable for integration into mechanistic models. We conclude by summarizing other factors, such as the need to consider demography, limiting broader application of mechanistic models in climate change conservation. Ideally, modellers, physiologists and

  9. MELMRK 2.0: A description of computer models and results of code testing

    International Nuclear Information System (INIS)

    Wittman, R.S.; Denny, V.; Mertol, A.

    1992-01-01

    An advanced version of the MELMRK computer code has been developed that provides detailed models for conservation of mass, momentum, and thermal energy within relocating streams of molten metallics during meltdown of Savannah River Site (SRS) reactor assemblies. In addition to a mechanistic treatment of transport phenomena within a relocating stream, MELMRK 2.0 retains the MOD1 capability for real-time coupling of the in-depth thermal response of participating assembly heat structure and, further, augments this capability with models for self-heating of relocating melt owing to steam oxidation of metallics and fission product decay power. As was the case for MELMRK 1.0, the MOD2 version offers state-of-the-art numerics for solving coupled sets of nonlinear differential equations. Principal features include application of multi-dimensional Newton-Raphson techniques to accelerate convergence behavior and direct matrix inversion to advance primitive variables from one iterate to the next. Additionally, MELMRK 2.0 provides logical event flags for managing the broad range of code options available for treating such features as (1) coexisting flow regimes, (2) dynamic transitions between flow regimes, and (3) linkages between heatup and relocation code modules. The purpose of this report is to provide a detailed description of the MELMRK 2.0 computer models for melt relocation. Also included are illustrative results for code testing, as well as an integrated calculation for meltdown of a Mark 31a assembly

  10. Advanced Electric and Magnetic Material Models for FDTD Electromagnetic Codes

    CERN Document Server

    Poole, Brian R; Nelson, Scott D

    2005-01-01

    The modeling of dielectric and magnetic materials in the time domain is required for pulse power applications, pulsed induction accelerators, and advanced transmission lines. For example, most induction accelerator modules require the use of magnetic materials to provide adequate Volt-sec during the acceleration pulse. These models require hysteresis and saturation to simulate the saturation wavefront in a multipulse environment. In high voltage transmission line applications such as shock or soliton lines the dielectric is operating in a highly nonlinear regime, which requires nonlinear models. Simple 1-D models are developed for fast parameterization of transmission line structures. In the case of nonlinear dielectrics, a simple analytic model describing the permittivity in terms of electric field is used in a 3-D finite difference time domain code (FDTD). In the case of magnetic materials, both rate independent and rate dependent Hodgdon magnetic material models have been implemented into 3-D FDTD codes an...

  11. Multi codes and multi-scale analysis for void fraction prediction in hot channel for VVER-1000/V392

    International Nuclear Information System (INIS)

    Hoang Minh Giang; Hoang Tan Hung; Nguyen Huu Tiep

    2015-01-01

    Recently, an approach of multi codes and multi-scale analysis is widely applied to study core thermal hydraulic behavior such as void fraction prediction. Better results are achieved by using multi codes or coupling codes such as PARCS and RELAP5. The advantage of multi-scale analysis is zooming of the interested part in the simulated domain for detail investigation. Therefore, in this study, the multi codes between MCNP5, RELAP5, CTF and also the multi-scale analysis based RELAP5 and CTF are applied to investigate void fraction in hot channel of VVER-1000/V392 reactor. Since VVER-1000/V392 reactor is a typical advanced reactor that can be considered as the base to develop later VVER-1200 reactor, then understanding core behavior in transient conditions is necessary in order to investigate VVER technology. It is shown that the item of near wall boiling, Γ w in RELAP5 proposed by Lahey mechanistic method may not give enough accuracy of void fraction prediction as smaller scale code as CTF. (author)

  12. Static benchmarking of the NESTLE advanced nodal code

    International Nuclear Information System (INIS)

    Mosteller, R.D.

    1997-01-01

    Results from the NESTLE advanced nodal code are presented for multidimensional numerical benchmarks representing four different types of reactors, and predictions from NESTLE are compared with measured data from pressurized water reactors (PWRs). The numerical benchmarks include cases representative of PWRs, boiling water reactors (BWRs), CANDU heavy water reactors (HWRs), and high-temperature gas-cooled reactors (HTGRs). The measured PWR data include critical soluble boron concentrations and isothermal temperature coefficients of reactivity. The results demonstrate that NESTLE correctly solves the multigroup diffusion equations for both Cartesian and hexagonal geometries, that it reliably calculates k eff and reactivity coefficients for PWRs, and that--subsequent to the incorporation of additional thermal-hydraulic models--it will be able to perform accurate calculations for the corresponding parameters in BWRs, HWRs, and HTGRs as well

  13. Recent developments in the CONTAIN-LMR code

    International Nuclear Information System (INIS)

    Murata, K.K.

    1990-01-01

    Through an international collaborative effort, a special version of the CONTAIN code is being developed for integrated mechanistic analysis of the conditions in liquid metal reactor (LMR) containments during severe accidents. The capabilities of the most recent code version, CONTAIN LMR/1B-Mod.1, are discussed. These include new models for the treatment of two condensables, sodium condensation on aerosols, chemical reactions, hygroscopic aerosols, and concrete outgassing. This code version also incorporates all of the previously released LMR model enhancements. The results of an integral demonstration calculation of a sever core-melt accident scenario are given to illustrate the features of this code version. 11 refs., 7 figs., 1 tab

  14. Modeling and validation of a mechanistic tool (MEFISTO) for the prediction of critical power in BWR fuel assemblies

    International Nuclear Information System (INIS)

    Adamsson, Carl; Le Corre, Jean-Marie

    2011-01-01

    Highlights: → The MEFISTO code efficiently and accurately predicts the dryout event in a BWR fuel bundle, using a mechanistic model. → A hybrid approach between a fast and robust sub-channel analysis and a three-field two-phase analysis is adopted. → MEFISTO modeling approach, calibration, CPU usage, sensitivity, trend analysis and performance evaluation are presented. → The calibration parameters and process were carefully selected to preserve the mechanistic nature of the code. → The code dryout prediction performance is near the level of fuel-specific empirical dryout correlations. - Abstract: Westinghouse is currently developing the MEFISTO code with the main goal to achieve fast, robust, practical and reliable prediction of steady-state dryout Critical Power in Boiling Water Reactor (BWR) fuel bundle based on a mechanistic approach. A computationally efficient simulation scheme was used to achieve this goal, where the code resolves all relevant field (drop, steam and multi-film) mass balance equations, within the annular flow region, at the sub-channel level while relying on a fast and robust two-phase (liquid/steam) sub-channel solution to provide the cross-flow information. The MEFISTO code can hence provide highly detailed solution of the multi-film flow in BWR fuel bundle while enhancing flexibility and reducing the computer time by an order of magnitude as compared to a standard three-field sub-channel analysis approach. Models for the numerical computation of the one-dimensional field flowrate distributions in an open channel (e.g. a sub-channel), including the numerical treatment of field cross-flows, part-length rods, spacers grids and post-dryout conditions are presented in this paper. The MEFISTO code is then applied to dryout prediction in BWR fuel bundle using VIPRE-W as a fast and robust two-phase sub-channel driver code. The dryout power is numerically predicted by iterating on the bundle power so that the minimum film flowrate in the

  15. Refining the accuracy of validated target identification through coding variant fine-mapping in type 2 diabetes

    DEFF Research Database (Denmark)

    Mahajan, Anubha

    2018-01-01

    , compelling evidence for coding variant causality was obtained for only 16 signals. At 13 others, the associated coding variants clearly represent 'false leads' with potential to generate erroneous mechanistic inference. Coding variant associations offer a direct route to biological insight for complex...

  16. Association of code status discussion with invasive procedures among advanced-stage cancer and noncancer patients

    Directory of Open Access Journals (Sweden)

    Sasaki A

    2017-07-01

    Full Text Available Akinori Sasaki,1 Eiji Hiraoka,1 Yosuke Homma,2 Osamu Takahashi,3 Yasuhiro Norisue,4 Koji Kawai,5 Shigeki Fujitani4 1Department of Internal Medicine, 2Department of Emergency Medicine, Tokyo Bay Urayasu Ichikawa Medical Center, Urayasu City, Chiba, 3Department of Internal Medicine, St. Luke’s International Hospital, Chuo-ku, Tokyo, 4Department of Critical Care Medicine, Tokyo Bay Urayasu Ichikawa Medical Center, Urayasu City, Chiba, 5Department of Gastroenterology, Ito Municipal Hospital, Ito City, Shizuoka, Japan Background: Code status discussion is associated with a decrease in invasive procedures among terminally ill cancer patients. We investigated the association between code status discussion on admission and incidence of invasive procedures, cardiopulmonary resuscitation (CPR, and opioid use among inpatients with advanced stages of cancer and noncancer diseases. Methods: We performed a retrospective cohort study in a single center, Ito Municipal Hospital, Japan. Participants were patients who were admitted to the Department of Internal Medicine between October 1, 2013 and August 30, 2015, with advanced-stage cancer and noncancer. We collected demographic data and inquired the presence or absence of code status discussion within 24 hours of admission and whether invasive procedures, including central venous catheter placement, intubation with mechanical ventilation, and CPR for cardiac arrest, and opioid treatment were performed. We investigated the factors associated with CPR events by using multivariate logistic regression analysis. Results: Among the total 232 patients, code status was discussed with 115 patients on admission, of which 114 (99.1% patients had do-not-resuscitate (DNR orders. The code status was not discussed with the remaining 117 patients on admission, of which 69 (59% patients had subsequent code status discussion with resultant DNR orders. Code status discussion on admission decreased the incidence of central venous

  17. Advances in the development of interaction between the codes MCNPX and ANSYS Fluent and their fusion applications

    International Nuclear Information System (INIS)

    Colomer, C.; Salellas, J.; Ahmed, R.; Fabbrio, M.; Aleman, A.

    2012-01-01

    The advances are presented in the project for the development of a code of interaction between MCNPX y el ANSYS Fluent. Following the flow of the work carried out during the development of the project will study of the most appropriate remeshing algorithms between both codes. In addition explain the selection and implementation of methods to verify internally the correct transmission of the variables involved between both nets. Finally the selection of cases for verification and validation of the interaction between both codes in each of the possible fields of application will be exposed.

  18. CSAU (code scaling, applicability and uncertainty), a tool to prioritize advanced reactor research

    International Nuclear Information System (INIS)

    Wilson, G.E.; Boyack, B.E.

    1990-01-01

    Best Estimate computer codes have been accepted by the US Nuclear Regulatory Commission as an optional tool for performing safety analysis related to the licensing and regulation of current nuclear reactors producing commercial electrical power, providing their uncertainty is quantified. In support of this policy change, the NRC and its contractors and consultants have developed and demonstrated an uncertainty quantification methodology called CSAU. At the process level, the method is generic to any application which relies on best estimate computer code simulations to determine safe operating margins. The primary use of the CSAU methodology is to quantify safety margins for existing designs; however, the methodology can also serve an equally important role in advanced reactor research for plants not yet built. Applied early, during the period when alternate designs are being evaluated, the methodology can identify the relative importance of the sources of uncertainty in the knowledge of each plant behavior and, thereby, help prioritize the research needed to bring the new designs to fruition. This paper describes the CSAU methodology, at the generic process level, and provides the general principles whereby it may be applied to evaluations of advanced reactor designs. 9 refs., 1 fig., 1 tab

  19. Advanced Imaging Optics Utilizing Wavefront Coding.

    Energy Technology Data Exchange (ETDEWEB)

    Scrymgeour, David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Boye, Robert [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Adelsberger, Kathleen [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-06-01

    Image processing offers a potential to simplify an optical system by shifting some of the imaging burden from lenses to the more cost effective electronics. Wavefront coding using a cubic phase plate combined with image processing can extend the system's depth of focus, reducing many of the focus-related aberrations as well as material related chromatic aberrations. However, the optimal design process and physical limitations of wavefront coding systems with respect to first-order optical parameters and noise are not well documented. We examined image quality of simulated and experimental wavefront coded images before and after reconstruction in the presence of noise. Challenges in the implementation of cubic phase in an optical system are discussed. In particular, we found that limitations must be placed on system noise, aperture, field of view and bandwidth to develop a robust wavefront coded system.

  20. Advanced thermohydraulic simulation code for transients in LMFBRs (SSC-L code)

    Energy Technology Data Exchange (ETDEWEB)

    Agrawal, A.K.

    1978-02-01

    Physical models for various processes that are encountered in preaccident and transient simulation of thermohydraulic transients in the entire liquid metal fast breeder reactor (LMFBR) plant are described in this report. A computer code, SSC-L, was written as a part of the Super System Code (SSC) development project for the ''loop''-type designs of LMFBRs. This code has the self-starting capability, i.e., preaccident or steady-state calculations are performed internally. These results then serve as the starting point for the transient simulation.

  1. Advanced thermohydraulic simulation code for transients in LMFBRs (SSC-L code)

    International Nuclear Information System (INIS)

    Agrawal, A.K.

    1978-02-01

    Physical models for various processes that are encountered in preaccident and transient simulation of thermohydraulic transients in the entire liquid metal fast breeder reactor (LMFBR) plant are described in this report. A computer code, SSC-L, was written as a part of the Super System Code (SSC) development project for the ''loop''-type designs of LMFBRs. This code has the self-starting capability, i.e., preaccident or steady-state calculations are performed internally. These results then serve as the starting point for the transient simulation

  2. Integration of CFD codes and advanced combustion models for quantitative burnout determination

    Energy Technology Data Exchange (ETDEWEB)

    Javier Pallares; Inmaculada Arauzo; Alan Williams [University of Zaragoza, Zaragoza (Spain). Centre of Research for Energy Resources and Consumption (CIRCE)

    2007-10-15

    CFD codes and advanced kinetics combustion models are extensively used to predict coal burnout in large utility boilers. Modelling approaches based on CFD codes can accurately solve the fluid dynamics equations involved in the problem but this is usually achieved by including simple combustion models. On the other hand, advanced kinetics combustion models can give a detailed description of the coal combustion behaviour by using a simplified description of the flow field, this usually being obtained from a zone-method approach. Both approximations describe correctly general trends on coal burnout, but fail to predict quantitative values. In this paper a new methodology which takes advantage of both approximations is described. In the first instance CFD solutions were obtained of the combustion conditions in the furnace in the Lamarmora power plant (ASM Brescia, Italy) for a number of different conditions and for three coals. Then, these furnace conditions were used as inputs for a more detailed chemical combustion model to predict coal burnout. In this, devolatilization was modelled using a commercial macromolecular network pyrolysis model (FG-DVC). For char oxidation an intrinsic reactivity approach including thermal annealing, ash inhibition and maceral effects, was used. Results from the simulations were compared against plant experimental values, showing a reasonable agreement in trends and quantitative values. 28 refs., 4 figs., 4 tabs.

  3. Development of boiling transition analysis code TCAPE-INS/B based on mechanistic methods for BWR fuel bundles. Models and validations with boiling transition experimental data

    International Nuclear Information System (INIS)

    Ishida, Naoyuki; Utsuno, Hideaki; Kasahara, Fumio

    2003-01-01

    The Boiling Transition (BT) analysis code TCAPE-INS/B based on the mechanistic methods coupled with subchannel analysis has been developed for the evaluation of the integrity of Boiling Water Reactor (BWR) fuel rod bundles under abnormal operations. Objective of the development is the evaluation of the BT without using empirical BT and rewetting correlations needed for different bundle designs in the current analysis methods. TCAPE-INS/B consisted mainly of the drift-flux model, the film flow model, the cross-flow model, the thermal conductivity model and the heat transfer correlations. These models were validated systematically with the experimental data. The accuracy of the prediction for the steady-state Critical Heat Flux (CHF) and the transient temperature of the fuel rod surface after the occurrence of BT were evaluated on the validations. The calculations for the experiments with the single tube and bundles were carried out for the validations of the models incorporated in the code. The results showed that the steady-state CHF was predicted within about 6% average error. In the transient calculations, BT timing and temperature of the fuel rod surface gradient agreed well with experimental results, but rewetting was predicted lately. So, modeling of heat transfer phenomena during post-BT is under modification. (author)

  4. THEHYCO-3DT: Thermal hydrodynamic code for the 3 dimensional transient calculation of advanced LMFBR core

    Energy Technology Data Exchange (ETDEWEB)

    Vitruk, S.G.; Korsun, A.S. [Moscow Engineering Physics Institute (Russian Federation); Ushakov, P.A. [Institute of Physics and Power Engineering, Obninsk (R)] [and others

    1995-09-01

    The multilevel mathematical model of neutron thermal hydrodynamic processes in a passive safety core without assemblies duct walls and appropriate computer code SKETCH, consisted of thermal hydrodynamic module THEHYCO-3DT and neutron one, are described. A new effective discretization technique for energy, momentum and mass conservation equations is applied in hexagonal - z geometry. The model adequacy and applicability are presented. The results of the calculations show that the model and the computer code could be used in conceptual design of advanced reactors.

  5. THEHYCO-3DT: Thermal hydrodynamic code for the 3 dimensional transient calculation of advanced LMFBR core

    International Nuclear Information System (INIS)

    Vitruk, S.G.; Korsun, A.S.; Ushakov, P.A.

    1995-01-01

    The multilevel mathematical model of neutron thermal hydrodynamic processes in a passive safety core without assemblies duct walls and appropriate computer code SKETCH, consisted of thermal hydrodynamic module THEHYCO-3DT and neutron one, are described. A new effective discretization technique for energy, momentum and mass conservation equations is applied in hexagonal - z geometry. The model adequacy and applicability are presented. The results of the calculations show that the model and the computer code could be used in conceptual design of advanced reactors

  6. Tri-Lab Co-Design Milestone: In-Depth Performance Portability Analysis of Improved Integrated Codes on Advanced Architecture.

    Energy Technology Data Exchange (ETDEWEB)

    Hoekstra, Robert J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hammond, Simon David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Richards, David [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Bergen, Ben [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-09-01

    This milestone is a tri-lab deliverable supporting ongoing Co-Design efforts impacting applications in the Integrated Codes (IC) program element Advanced Technology Development and Mitigation (ATDM) program element. In FY14, the trilabs looked at porting proxy application to technologies of interest for ATS procurements. In FY15, a milestone was completed evaluating proxy applications in multiple programming models and in FY16, a milestone was completed focusing on the migration of lessons learned back into production code development. This year, the co-design milestone focuses on extracting the knowledge gained and/or code revisions back into production applications.

  7. Strategies for developing subchannel capability in an advanced system thermalhydraulic code: a literature review

    International Nuclear Information System (INIS)

    Cheng, J.; Rao, Y.F.

    2015-01-01

    In the framework of developing next generation safety analysis tools, Canadian Nuclear Laboratories (CNL) has planned to incorporate subchannel analysis capability into its advanced system thermalhydraulic code CATHENA 4. This paper provides a literature review and an assessment of current subchannel codes. It also evaluates three code-development methods: (i) static coupling of CATHENA 4 with the subchannel code ASSERT-PV, (ii) dynamic coupling of the two codes, and (iii) fully implicit implementation for a new, standalone CATHENA 4 version with subchannel capability. Results of the review and assessment suggest that the current ASSERT-PV modules can be used as the base for the fully implicit implementation of subchannel capability in CATHENA 4, and that this option may be the most cost-effective in the long run, resulting in savings in user application and maintenance costs. In addition, improved versatility of the tool could be accomplished by the addition of new features that could be added as part of its development. The new features would improve the capabilities of the existing subchannel code in handling low, reverse, and stagnant flows often encountered in system thermalhydraulic analysis. Therefore, the method of fully implicit implementation is preliminarily recommended for further exploration. A feasibility study will be performed in an attempt to extend the present work into a preliminary development plan. (author)

  8. A Life-Cycle Risk-Informed Systems Structured Nuclear Code

    International Nuclear Information System (INIS)

    Hill, Ralph S. III

    2002-01-01

    Current American Society of Mechanical Engineers (ASME) nuclear codes and standards rely primarily on deterministic and mechanistic approaches to design. The design code is a separate volume from the code for inservice inspections and both are separate from the standards for operations and maintenance. The ASME code for inservice inspections and code for nuclear plant operations and maintenance have adopted risk-informed methodologies for inservice inspection, preventive maintenance, and repair and replacement decisions. The American Institute of Steel Construction and the American Concrete Institute have incorporated risk-informed probabilistic methodologies into their design codes. It is proposed that the ASME nuclear code should undergo a planned evolution that integrates the various nuclear codes and standards and adopts a risk-informed approach across a facility life-cycle - encompassing design, construction, operation, maintenance and closure. (author)

  9. Algorithm for advanced canonical coding of planar chemical structures that considers stereochemical and symmetric information.

    Science.gov (United States)

    Koichi, Shungo; Iwata, Satoru; Uno, Takeaki; Koshino, Hiroyuki; Satoh, Hiroko

    2007-01-01

    We describe a rigorous and fast algorithm for advanced canonical coding of planar chemical structures based on the algorithm of Faulon et al. (J. Chem. Inf. Comput. Sci. 2004, 44, 427-436). Our algorithm works well even for highly symmetric structures; moreover, an advantage of our algorithm includes providing a rigorous canonical numbering of atoms with a consideration of stereochemistry and recognizing symmetric moieties. The planar structural line notation with the canonical numbering is also fit for use with stereochemical line notation. These capabilities are usable for general purposes in chemical structural coding and are particularly essential for detecting equivalent atoms in NMR studies. This algorithm was implemented on a 13C NMR chemical shift prediction system CAST/CNMR. Applications of the algorithm to several organic compounds demonstrate the practical efficiency of the rigorous coding.

  10. Fuel swelling importance in PCI mechanistic modelling

    International Nuclear Information System (INIS)

    Arimescu, V.I.

    2005-01-01

    Under certain conditions, fuel pellet swelling is the most important factor in determining the intensity of the pellet-to-cladding mechanical interaction (PCMI). This is especially true during power ramps, which lead to a temperature increase to a higher terminal plateau that is maintained for hours. The time-dependent gaseous swelling is proportional to temperature and is also enhanced by the increased gas atom migration to the grain boundary during the power ramp. On the other hand, gaseous swelling is inhibited by a compressive hydrostatic stress in the pellet. Therefore, PCMI is the net result of combining gaseous swelling and pellet thermal expansion with the opposing feedback from the cladding mechanical reaction. The coupling of the thermal and mechanical processes, mentioned above, with various feedback loops is best simulated by a mechanistic fuel code. This paper discusses a mechanistic swelling model that is coupled with a fission gas release model as well as a mechanical model of the fuel pellet. The role of fuel swelling is demonstrated for typical power ramps at different burn-ups. Also, fuel swelling plays a significant role in avoiding the thermal instability for larger gap fuel rods, by limiting the potentially exponentially increasing gap due to the positive feedback loop effect of increasing fission gas release and the associated over-pressure inside the cladding. (author)

  11. Evaluation of mechanistic DNB models using HCLWR CHF data

    International Nuclear Information System (INIS)

    Iwamura, Takamichi; Watanabe, Hironori; Okubo, Tsutomu; Araya, Fumimasa; Murao, Yoshio.

    1992-03-01

    An onset of departure from nucleate boiling (DNB) in light water reactor (LWR) has been generally predicted with empirical correlations. Since these correlations have less physical bases and contain adjustable empirical constants determined by best fitting of test data, applicable geometries and flow conditions are limited within the original experiment ranges. In order to obtain more universal prediction method, several mechanistic DNB models based on physical approaches have been proposed in recent years. However, the predictive capabilities of mechanistic DNB models have not been verified successfully especially for advanced LWR design purposes. In this report, typical DNB mechanistic models are reviewed and compared with critical heat flux (CHF) data for high conversion light water reactor (HCLWR). The experiments were performed using triangular 7-rods array with non-uniform axial heat flux distribution. Test pressure was 16 MPa, mass velocities ranged from 800 t0 3100 kg/s·m 2 and exit qualities from -0.07 to 0.19. The evaluated models are: 1) Wisman-Pei, 2) Chang-Lee, 3) Lee-Mudawwar, 4) Lin-Lee-Pei, and 5) Katto. The first two models are based on near-wall bubble crowding model and the other three models on sublayer dryout model. The comparison with experimental data indicated that the Weisman-Pei model agreed relatively well with the CHF data. Effects of empirical constants in each model on CHF calculation were clarified by sensitivity studies. It was also found that the magnitudes of physical quantities obtained in the course of calculation were significantly different for each model. Therefore, microscopic observation of the onset of DNB on heated surface is essential to clarify the DNB mechanism and establish a general DNB mechanistic model based on physical phenomenon. (author)

  12. Numerical analysis of reflood simulation based on a mechanistic, best-estimate, approach by KREWET code

    International Nuclear Information System (INIS)

    Chun, Moon-Hyun; Jeong, Eun-Soo

    1983-01-01

    A new computer code entitled KREWET has been developed in an effort to improve the accuracy and applicability of the existing reflood heat transfer simulation computer code. Sample calculations for temperature histories and heat transfer coefficient are made using KREWET code and the results are compared with the predictions of REFLUX, QUEN1D, and the PWR-FLECHT data for various conditions. These show favourable agreement in terms of clad temperature versus time. For high flooding rates (5-15cm/sec) and high pressure (∼413 Kpa), reflood predictions are reasonably well predicted by KREWET code as well as with other codes. For low flooding rates (less than ∼4cm/sec) and low pressure (∼138Kpa), predictions show considerable error in evaluating the rewet position versus time. This observation is common to all the codes examined in the present work

  13. Numerical analysis for reflood simulation based on a mechanistic, best-estimate, approach by KREWET code

    International Nuclear Information System (INIS)

    Chun, M.-H.; Jeong, E.-S.

    1983-01-01

    A new computer code entitled KREWET has been developed in an effort to improve the accuracy and applicability of the existing reflood heat transfer simulation computer code. Sample calculations for temperature histories and heat transfer coefficient are made using KREWET code and the results are compared with the predictions of REFLUX, QUENID, and the PWR-FLECHT data for various conditions. These show favorable agreement in terms of clad temperature versus time. For high flooding rates (5-15cm/sec) and high pressure (approx. =413 Kpa), reflood predictions are reasonably well predicted by KREWET code as well as with other codes. For low flooding rates (less than approx. =4cm/sec) and low pressure (approx. =138 Kpa), predictions show considerable error in evaluating the rewet position versus time. This observation is common to all the codes examined in the present work

  14. Fuel performance analysis code 'FAIR'

    International Nuclear Information System (INIS)

    Swami Prasad, P.; Dutta, B.K.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

    1994-01-01

    For modelling nuclear reactor fuel rod behaviour of water cooled reactors under severe power maneuvering and high burnups, a mechanistic fuel performance analysis code FAIR has been developed. The code incorporates finite element based thermomechanical module, physically based fission gas release module and relevant models for modelling fuel related phenomena, such as, pellet cracking, densification and swelling, radial flux redistribution across the pellet due to the build up of plutonium near the pellet surface, pellet clad mechanical interaction/stress corrosion cracking (PCMI/SSC) failure of sheath etc. The code follows the established principles of fuel rod analysis programmes, such as coupling of thermal and mechanical solutions along with the fission gas release calculations, analysing different axial segments of fuel rod simultaneously, providing means for performing local analysis such as clad ridging analysis etc. The modular nature of the code offers flexibility in affecting modifications easily to the code for modelling MOX fuels and thorium based fuels. For performing analysis of fuel rods subjected to very long power histories within a reasonable amount of time, the code has been parallelised and is commissioned on the ANUPAM parallel processing system developed at Bhabha Atomic Research Centre (BARC). (author). 37 refs

  15. Assessment of the GOTHIC code for prediction of hydrogen flame propagation in small scale experiments

    International Nuclear Information System (INIS)

    Lee, Jin-Yong . E-mail jinyong1@fnctech.com; Lee, Jung-Jae; Park, Goon-Cherl . E-mail parkgc@snu.ac.kr

    2006-01-01

    With the rising concerns regarding the time and space dependent hydrogen behavior in severe accidents, the calculation for local hydrogen combustion in compartment has been attempted using CFD codes like GOTHIC. In particular, the space resolved hydrogen combustion analysis is essential to address certain safety issues such as the safety components survivability, and to determine proper positions for hydrogen control devices as e.q. recombiners or igniters. In the GOTHIC 6.1b code, there are many advanced features associated with the hydrogen burn models to enhance its calculation capability. In this study, we performed premixed hydrogen/air combustion experiments with an upright, rectangular shaped, combustion chamber of dimensions 1 m x 0.024 m x 1 m. The GOTHIC 6.1b code was used to simulate the hydrogen/air combustion experiments, and its prediction capability was assessed by comparing the experimental with multidimensional calculational results. Especially, the prediction capability of the GOTHIC 6.1b code for local hydrogen flame propagation phenomena was examined. For some cases, comparisons are also presented for lumped modeling of hydrogen combustion. By evaluating the effect of parametric simulations, we present some instructions for local hydrogen combustion analysis using the GOTHIC 6.1b code. From the analyses results, it is concluded that the modeling parameter of GOTHIC 6.1b code should be modified when applying the mechanistic burn model for hydrogen propagation analysis in small geometry

  16. Mechanistic and Economical Characteristics of Asphalt Rubber Mixtures

    Directory of Open Access Journals (Sweden)

    Mena I. Souliman

    2016-01-01

    Full Text Available Load associated fatigue cracking is one of the major distress types occurring in flexible pavement systems. Flexural bending beam fatigue laboratory test has been used for several decades and is considered to be an integral part of the new superpave advanced characterization procedure. One of the most significant solutions to prolong the fatigue life for an asphaltic mixture is to utilize flexible materials as rubber. A laboratory testing program was performed on a conventional and Asphalt Rubber- (AR- gap-graded mixtures to investigate the impact of added rubber on the mechanical, mechanistic, and economical attributes of asphaltic mixtures. Strain controlled fatigue tests were conducted according to American Association of State Highway and Transportation Officials (AASHTO procedures. The results from the beam fatigue tests indicated that the AR-gap-graded mixtures would have much longer fatigue life compared with the reference (conventional mixtures. In addition, a mechanistic analysis using 3D-Move software coupled with a cost analysis study based on the fatigue performance on the two mixtures was performed. Overall, analysis showed that AR modified asphalt mixtures exhibited significantly lower cost of pavement per 1000 cycles of fatigue life per mile compared to conventional HMA mixture.

  17. Preliminary investigation study of code of developed country for developing Korean fuel cycle code

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Ko, Won Il; Lee, Ho Hee; Cho, Dong Keun; Park, Chang Je

    2012-01-01

    In order to develop Korean fuel cycle code, the analyses has been performed with the fuel cycle codes which are used in advanced country. Also, recommendations were proposed for future development. The fuel cycle codes are AS FLOOWS: VISTA which has been developed by IAEA, DANESS code which developed by ANL and LISTO, and VISION developed by INL for the Advanced Fuel Cycle Initiative (AFCI) system analysis. The recommended items were proposed for software, program scheme, material flow model, isotope decay model, environmental impact analysis model, and economics analysis model. The described things will be used for development of Korean nuclear fuel cycle code in future

  18. Supporting Mechanistic Reasoning in Domain-Specific Contexts

    Science.gov (United States)

    Weinberg, Paul J.

    2017-01-01

    Mechanistic reasoning is an epistemic practice central within science, technology, engineering, and mathematics disciplines. Although there has been some work on mechanistic reasoning in the research literature and standards documents, much of this work targets domain-general characterizations of mechanistic reasoning; this study provides…

  19. Mechanistic Systems Modeling to Improve Understanding and Prediction of Cardiotoxicity Caused by Targeted Cancer Therapeutics

    Directory of Open Access Journals (Sweden)

    Jaehee V. Shim

    2017-09-01

    Full Text Available Tyrosine kinase inhibitors (TKIs are highly potent cancer therapeutics that have been linked with serious cardiotoxicity, including left ventricular dysfunction, heart failure, and QT prolongation. TKI-induced cardiotoxicity is thought to result from interference with tyrosine kinase activity in cardiomyocytes, where these signaling pathways help to control critical processes such as survival signaling, energy homeostasis, and excitation–contraction coupling. However, mechanistic understanding is limited at present due to the complexities of tyrosine kinase signaling, and the wide range of targets inhibited by TKIs. Here, we review the use of TKIs in cancer and the cardiotoxicities that have been reported, discuss potential mechanisms underlying cardiotoxicity, and describe recent progress in achieving a more systematic understanding of cardiotoxicity via the use of mechanistic models. In particular, we argue that future advances are likely to be enabled by studies that combine large-scale experimental measurements with Quantitative Systems Pharmacology (QSP models describing biological mechanisms and dynamics. As such approaches have proven extremely valuable for understanding and predicting other drug toxicities, it is likely that QSP modeling can be successfully applied to cardiotoxicity induced by TKIs. We conclude by discussing a potential strategy for integrating genome-wide expression measurements with models, illustrate initial advances in applying this approach to cardiotoxicity, and describe challenges that must be overcome to truly develop a mechanistic and systematic understanding of cardiotoxicity caused by TKIs.

  20. RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - models and correlations

    Energy Technology Data Exchange (ETDEWEB)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

    1998-03-01

    This document describes the major modifications and improvements made to the modeling of the RAMONA-3B/MOD0 code since 1981, when the code description and assessment report was completed. The new version of the code is RAMONA-4B. RAMONA-4B is a systems transient code for application to different versions of Boiling Water Reactors (BWR) such as the current BWR, the Advanced Boiling Water Reactor (ABWR), and the Simplified Boiling Water Reactor (SBWR). This code uses a three-dimensional neutron kinetics model coupled with a multichannel, non-equilibrium, drift-flux, two-phase flow formulation of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients and instability issues. Chapter 1 is an overview of the code`s capabilities and limitations; Chapter 2 discusses the neutron kinetics modeling and the implementation of reactivity edits. Chapter 3 is an overview of the heat conduction calculations. Chapter 4 presents modifications to the thermal-hydraulics model of the vessel, recirculation loop, steam separators, boron transport, and SBWR specific components. Chapter 5 describes modeling of the plant control and safety systems. Chapter 6 presents and modeling of Balance of Plant (BOP). Chapter 7 describes the mechanistic containment model in the code. The content of this report is complementary to the RAMONA-3B code description and assessment document. 53 refs., 81 figs., 13 tabs.

  1. Mechanistic assessment of hillslope transpiration controls of diel subsurface flow: a steady-state irrigation approach

    Science.gov (United States)

    H.R. Barnard; C.B. Graham; W.J. van Verseveld; J.R. Brooks; B.J. Bond; J.J. McDonnell

    2010-01-01

    Mechanistic assessment of how transpiration influences subsurface flow is necessary to advance understanding of catchment hydrology. We conducted a 24-day, steady-state irrigation experiment to quantify the relationships among soil moisture, transpiration and hillslope subsurface flow. Our objectives were to: (1) examine the time lag between maximum transpiration and...

  2. Study of experimental validation for combustion analysis of GOTHIC code

    International Nuclear Information System (INIS)

    Lee, J. Y.; Yang, S. Y.; Park, K. C.; Jeong, S. H.

    2001-01-01

    In this study, present lumped and subdivided GOTHIC6 code analyses of the premixed hydrogen combustion experiment at the Seoul National University and comparison with the experiment results. The experimental facility has 16367 cc free volume and rectangular shape. And the test was performed with unit equivalence ratio of the hydrogen and air, and with various location of igniter position. Using the lumped and mechanistic combustion model in GOTHIC6 code, the experiments were simulated with the same conditions. In the comparison between experiment and calculated results, the GOTHIC6 prediction of the combustion response does not compare well with the experiment results. In the point of combustion time, the lumped combustion model of GOTHIC6 code does not simulate the physical phenomena of combustion appropriately. In the case of mechanistic combustion model, the combustion time is predicted well, but the induction time of calculation data is longer than the experiment data remarkably. Also, the laminar combustion model of GOTHIC6 has deficiency to simulate combustion phenomena unless control the user defined value appropriately. And the pressure is not a proper variable that characterize the three dimensional effect of combustion

  3. Modeling of fission product release in integral codes

    International Nuclear Information System (INIS)

    Obaidurrahman, K.; Raman, Rupak K.; Gaikwad, Avinash J.

    2014-01-01

    The Great Tohoku earthquake and tsunami that stroke the Fukushima-Daiichi nuclear power station in March 11, 2011 has intensified the needs of detailed nuclear safety research and with this objective all streams associated with severe accident phenomenology are being revisited thoroughly. The present paper would cover an overview of state of art FP release models being used, the important phenomenon considered in semi-mechanistic models and knowledge gaps in present FP release modeling. Capability of FP release module, ELSA of ASTEC integral code in appropriate prediction of FP release under several diversified core degraded conditions will also be demonstrated. Use of semi-mechanistic fission product release models at AERB in source-term estimation shall be briefed. (author)

  4. Trellis and turbo coding iterative and graph-based error control coding

    CERN Document Server

    Schlegel, Christian B

    2015-01-01

    This new edition has been extensively revised to reflect the progress in error control coding over the past few years. Over 60% of the material has been completely reworked, and 30% of the material is original. Convolutional, turbo, and low density parity-check (LDPC) coding and polar codes in a unified framework. Advanced research-related developments such as spatial coupling. A focus on algorithmic and implementation aspects of error control coding.

  5. A comparison of two nodal codes : Advanced nodal code (ANC) and analytic function expansion nodal (AFEN) code

    International Nuclear Information System (INIS)

    Chung, S.K.; Hah, C.J.; Lee, H.C.; Kim, Y.H.; Cho, N.Z.

    1996-01-01

    Modern nodal methods usually employs the transverse integration technique in order to reduce a multi-dimensional diffusion equation to one-dimensional diffusion equations. The use of the transverse integration technique requires two major approximations such as a transverse leakage approximation and a one-dimensional flux approximation. Both the transverse leakage and the one-dimensional flux are approximated by polynomials. ANC (Advanced Nodal Code) developed by Westinghouse employs a modern nodal expansion method for the flux calculation, the equivalence theory for the homogenization error reduction and a group theory for pin power recovery. Unlike the conventional modern nodal methods, AFEN (Analytic Function Expansion Nodal) method expands homogeneous flux distributions within a node into non-separable analytic basis functions, which eliminate two major approximations of the modern nodal methods. A comparison study of AFEN with ANC has been performed to see the applicability of AFEN to commercial PWR and different types of reactors such as MOX fueled reactor. The qualification comparison results demonstrate that AFEN methodology is accurate enough to apply for commercial PWR analysis. The results show that AFEN provides very accurate results (core multiplication factor and assembly power distribution) for cores that exhibit strong flux gradients as in a MOX loaded core. (author)

  6. Assessing uncertainty in mechanistic models

    Science.gov (United States)

    Edwin J. Green; David W. MacFarlane; Harry T. Valentine

    2000-01-01

    Concern over potential global change has led to increased interest in the use of mechanistic models for predicting forest growth. The rationale for this interest is that empirical models may be of limited usefulness if environmental conditions change. Intuitively, we expect that mechanistic models, grounded as far as possible in an understanding of the biology of tree...

  7. Fundamentals of information theory and coding design

    CERN Document Server

    Togneri, Roberto

    2003-01-01

    In a clear, concise, and modular format, this book introduces the fundamental concepts and mathematics of information and coding theory. The authors emphasize how a code is designed and discuss the main properties and characteristics of different coding algorithms along with strategies for selecting the appropriate codes to meet specific requirements. They provide comprehensive coverage of source and channel coding, address arithmetic, BCH, and Reed-Solomon codes and explore some more advanced topics such as PPM compression and turbo codes. Worked examples and sets of basic and advanced exercises in each chapter reinforce the text's clear explanations of all concepts and methodologies.

  8. Severe accident analysis code Sampson for impact project

    International Nuclear Information System (INIS)

    Hiroshi, Ujita; Takashi, Ikeda; Masanori, Naitoh

    2001-01-01

    Four years of the IMPACT project Phase 1 (1994-1997) had been completed with financial sponsorship from the Japanese government's Ministry of Economy, Trade and Industry. At the end of the phase, demonstration simulations by combinations of up to 11 analysis modules developed for severe accident analysis in the SAMPSON Code were performed and physical models in the code were verified. The SAMPSON prototype was validated by TMI-2 and Phebus-FP test analyses. Many of empirical correlation and conventional models have been replaced by mechanistic models during Phase 2 (1998-2000). New models for Accident Management evaluation have been also developed. (author)

  9. Mechanistic effect modeling for ecological risk assessment: where to go from here?

    Science.gov (United States)

    Grimm, Volker; Martin, Benjamin T

    2013-07-01

    Mechanistic effect models (MEMs) consider the mechanisms of how chemicals affect individuals and ecological systems such as populations and communities. There is an increasing awareness that MEMs have high potential to make risk assessment of chemicals more ecologically relevant than current standard practice. Here we discuss what kinds of MEMs are needed to improve scientific and regulatory aspects of risk assessment. To make valid predictions for a wide range of environmental conditions, MEMs need to include a sufficient amount of emergence, for example, population dynamics emerging from what individual organisms do. We present 1 example where the life cycle of individuals is described using Dynamic Energy Budget theory. The resulting individual-based population model is thus parameterized at the individual level but correctly predicts multiple patterns at the population level. This is the case for both control and treated populations. We conclude that the state-of-the-art in mechanistic effect modeling has reached a level where MEMs are robust and predictive enough to be used in regulatory risk assessment. Mechanistic effect models will thus be used to advance the scientific basis of current standard practice and will, if their development follows Good Modeling Practice, be included in a standardized way in future regulatory risk assessments. Copyright © 2013 SETAC.

  10. Numerical simulation in steam injection wellbores by mechanistic approach; Simulacao numerica do escoamento de vapor em pocos por uma abordagem mecanicista

    Energy Technology Data Exchange (ETDEWEB)

    Souza Junior, J.C. de; Campos, W.; Lopes, D.; Moura, L.S.S. [PETROBRAS, Rio de Janeiro, RJ (Brazil); Thomas, A. Clecio F. [Universidade Estadual do Ceara (UECE), CE (Brazil)

    2008-07-01

    This work addresses to the development of a hydrodynamic and heat transfer mechanistic model for steam flow in injection wellbores. The problem of two-phase steam flow in wellbores has been solved recently by using available empirical correlations from petroleum industry (Lopes, 1986) and nuclear industry (Moura, 1991).The good performance achieved by mechanistic models developed by Ansari (1994), Hasan (1995), Gomez (2000) and Kaya (2001) supports the importance of the mechanistic approach for the steam flow problem in injection wellbores. In this study, the methodology to solve the problem consists in the application of a numerical method to the governing equations of steam flow and a marching algorithm to determine the distribution of the pressure and temperature along the wellbore. So, a computer code has been formulated to get numerical results, which provides a comparative study to the main models found in the literature. Finally, when compared to available field data, the mechanistic model for downward vertical steam flow in wellbores gave better results than the empirical correlations. (author)

  11. Chirality-Controlled Growth of Single-Wall Carbon Nanotubes Using Vapor Phase Epitaxy: Mechanistic Understanding and Scalable Production

    Science.gov (United States)

    2016-09-15

    AFRL-AFOSR-VA-TR-2016-0319 Chirality -Controlled Growth of Single-Wall Carbon Nanotubes Using Vapor Phase Epitaxy: Mechanistic Understanding and...TELEPHONE NUMBER (Include area code) DISTRIBUTION A: Distribution approved for public release. 15-06-2016 final Jun 2014 - Jun 2016 Chirality ...for Public Release; Distribution is Unlimited. In this report, we present our efforts in establishing a novel and effective approach for chirality

  12. Study on natural circulation flow under reactor cavity flooding condition in advanced PWRs

    International Nuclear Information System (INIS)

    Tao Jun; Yang Jiang; Cao Jianhua; Lu Xianghui; Guo Dingqing

    2015-01-01

    Cavity flooding is an important severe accident management measure for the in-vessel retention of a degraded core by external reactor vessel cooling in advanced PWRs. A code simulation study on the natural circulation flow in the gap between the reactor vessel wall and insulation material under cavity flooding condition is performed by using a detailed mechanistic thermal-hydraulic code package RELAP 5. By simulating of an experiment carried out for studying the natural circulation flow for APR1400 shows that the code is applicable for analyzing the circulation flow under this condition. The analysis results show that heat removal capacity of the natural circulation flow in AP1000 is sufficient to prevent thermal failure of the reactor vessel under bounding heat load. Several conclusions can be drawn from the sensitivity analysis. Larger coolant inlet area induced larger natural circulation flow rate. The outlet should be large enough and should not be submerged by the cavity water to vent the steam-water mixture. In the implementation of cavity flooding, the flooding water level should be high enough to provide sufficient natural circulation driven force. (authors)

  13. Discrete rod burnup analysis capability in the Westinghouse advanced nodal code

    International Nuclear Information System (INIS)

    Buechel, R.J.; Fetterman, R.J.; Petrunyak, M.A.

    1992-01-01

    Core design analysis in the last several years has evolved toward the adoption of nodal-based methods to replace traditional fine-mesh models as the standard neutronic tool for first core and reload design applications throughout the nuclear industry. The accuracy, speed, and reduction in computation requirements associated with the nodal methods have made three-dimensional modeling the preferred approach to obtain the most realistic core model. These methods incorporate detailed rod power reconstruction as well. Certain design applications such as confirmation of fuel rod design limits and fuel reconstitution considerations, for example, require knowledge of the rodwise burnup distribution to avoid unnecessary conservatism in design analyses. The Westinghouse Advanced Nodal Code (ANC) incorporates the capability to generate the intra-assembly pin burnup distribution using an efficient algorithm

  14. Advanced REACH Tool : Development and application of the substance emission potential modifying factor

    NARCIS (Netherlands)

    Tongeren, M. van; Fransman, W.; Spankie, S.; Tischer, M.; Brouwer, D.; Schinkel, J.; Cherrie, J.W.; Tielemans, E.

    2011-01-01

    The Advanced REACH Tool (ART) is an exposure assessment tool that combines mechanistically modelled inhalation exposure estimates with available exposure data using a Bayesian approach. The mechanistic model is based on nine independent principal modifying factors (MF). One of these MF is the

  15. Validation of the COBRA code for dry out power calculation in CANDU type advanced fuels

    International Nuclear Information System (INIS)

    Daverio, Hernando J.

    2003-01-01

    Stern Laboratories perform a full scale CHF testing of the CANFLEX bundle under AECL request. This experiment is modeled with the COBRA IV HW code to verify it's capacity for the dry out power calculation . Good results were obtained: errors below 10 % with respect to all data measured and 1 % for standard operating conditions in CANDU reactors range . This calculations were repeated for the CNEA advanced fuel CARA obtaining the same performance as the CANFLEX fuel. (author)

  16. Performance evaluations of advanced massively parallel platforms based on gyrokinetic toroidal five-dimensional Eulerian code GT5D

    International Nuclear Information System (INIS)

    Idomura, Yasuhiro; Jolliet, Sebastien

    2010-01-01

    A gyrokinetic toroidal five dimensional Eulerian code GT5D is ported on six advanced massively parallel platforms and comprehensive benchmark tests are performed. A parallelisation technique based on physical properties of the gyrokinetic equation is presented. By extending the parallelisation technique with a hybrid parallel model, the scalability of the code is improved on platforms with multi-core processors. In the benchmark tests, a good salability is confirmed up to several thousands cores on every platforms, and the maximum sustained performance of ∼18.6 Tflops is achieved using 16384 cores of BX900. (author)

  17. Appropriateness of mechanistic and non-mechanistic models for the application of ultrafiltration to mixed waste

    International Nuclear Information System (INIS)

    Foust, Henry; Ghosehajra, Malay

    2007-01-01

    This study asks two questions: (1) How appropriate is the use of a basic filtration equation to the application of ultrafiltration of mixed waste, and (2) How appropriate are non-parametric models for permeate rates (volumes)? To answer these questions, mechanistic and non-mechanistic approaches are developed for permeate rates and volumes associated with an ultrafiltration/mixed waste system in dia-filtration mode. The mechanistic approach is based on a filtration equation which states that t/V vs. V is a linear relationship. The coefficients associated with this linear regression are composed of physical/chemical parameters of the system and based the mass balance equation associated with the membrane and associated developing cake layer. For several sets of data, a high correlation is shown that supports the assertion that t/V vs. V is a linear relationship. It is also shown that non-mechanistic approaches, i.e., the use of regression models to are not appropriate. One models considered is Q(p) = a*ln(Cb)+b. Regression models are inappropriate because the scale-up from a bench scale (pilot scale) study to full-scale for permeate rates (volumes) is not simply the ratio of the two membrane surface areas. (authors)

  18. Oxide fuel pin transient performance analysis and design with the TEMECH code

    International Nuclear Information System (INIS)

    Bard, F.E.; Dutt, S.P.; Hinman, C.A.; Hunter, C.W.; Pitner, A.L.

    1986-01-01

    The TEMECH code is a fast-running, thermal-mechanical-hydraulic, analytical program used to evaluate the transient performance of LMR oxide fuel pins. The code calculates pin deformation and failure probability due to fuel-cladding differential thermal expansion, expansion of fuel upon melting, and fission gas pressurization. The mechanistic fuel model in the code accounts for fuel cracking, crack closure, porosity decrease, and the temperature dependence of fuel creep through the course of the transient. Modeling emphasis has been placed on results obtained from Fuel Cladding Transient Test (FCTT) testing, Transient Fuel Deformation (TFD) tests and TREAT integral fuel pin experiments

  19. Probabilistic fuel rod analyses using the TRANSURANUS code

    Energy Technology Data Exchange (ETDEWEB)

    Lassmann, K; O` Carroll, C; Laar, J Van De [CEC Joint Research Centre, Karlsruhe (Germany)

    1997-08-01

    After more than 25 years of fuel rod modelling research, the basic concepts are well established and the limitations of the specific approaches are known. However, the widely used mechanistic approach leads in many cases to discrepancies between theoretical predictions and experimental evidence indicating that models are not exact and that some of the physical processes encountered are of stochastic nature. To better understand uncertainties and their consequences, the mechanistic approach must therefore be augmented by statistical analyses. In the present paper the basic probabilistic methods are briefly discussed. Two such probabilistic approaches are included in the fuel rod performance code TRANSURANUS: the Monte Carlo method and the Numerical Noise Analysis. These two techniques are compared and their capabilities are demonstrated. (author). 12 refs, 4 figs, 2 tabs.

  20. Basic research and industrialization of CANDU advanced fuel - A research for the improvement of RFSP code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Hyo; Jang, Chang Sun; Han, Tae Young [Seoul National University, Seoul (Korea)

    2000-03-01

    The objective of this project is to improve the RFSP code by adopting three dimensional two neutron energy group model and accelerated iterative solution scheme (FDM3D) to 2 group diffusion equations as well. The major contents of this research are the derivation of the finite difference equation to three dimensional two neutron energy group diffusion equation, application of accelerated iterative solution scheme to the finite difference diffusion equation and validation of the improved RFSP code (FDM3D) through benchmark tests. We have shown that SOR/Chebyshev two parameter method and BICG-STAB/Wielandt method are more effective than that of RFSP in terms of computing speed. SOR/Chebyshev two parameter method shows better efficiency than BICG-STAB/Wielandt method. Because calculation efficiency of the latter depends on the right choice of pre-conditioner, however, it is considered that more studies are necessary to improve the efficiency of this latter method and to validate it. We have incorporated the new efficient method into the existing RFSP so that the resulting RFSP becomes much faster and more accurate. RFSP currently uses POWDERPUFS code as main lattice code, which is adequate to the neutron energy group model of RFSP. Because of this, we can not make the full advantage of advanced RFSP without adopting lattice code WIMS-AECL which can generate exact two neutron energy group constants. Therefore, we suggest developing a new CANDU design and analysis code which incorporate WIMS-AECL into FDM3D. 16 refs., 10 figs., 23 tabs. (Author)

  1. Algebraic and stochastic coding theory

    CERN Document Server

    Kythe, Dave K

    2012-01-01

    Using a simple yet rigorous approach, Algebraic and Stochastic Coding Theory makes the subject of coding theory easy to understand for readers with a thorough knowledge of digital arithmetic, Boolean and modern algebra, and probability theory. It explains the underlying principles of coding theory and offers a clear, detailed description of each code. More advanced readers will appreciate its coverage of recent developments in coding theory and stochastic processes. After a brief review of coding history and Boolean algebra, the book introduces linear codes, including Hamming and Golay codes.

  2. Development of essential system technologies for advanced reactor - Development of natural circulation analysis code for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Goon Cherl; Park, Ik Gyu; Kim, Jae Hak; Lee, Sang Min; Kim, Tae Wan [Seoul National University, Seoul (Korea)

    1999-04-01

    The objective of this study is to understand the natural circulation characteristics of integral type reactors and to develope the natural circulation analysis code for integral type reactors. This study is focused on the asymmetric 3-dimensional flow during natural circulation such as 1/4 steam generator section isolation and the inclination of the reactor systems. Natural circulation experiments were done using small-scale facilities of integral reactor SMART (System-Integrated Modular Advanced ReacTor). CFX4 code was used to investigate the flow patterns and thermal mixing phenomena in upper pressure header and downcomer. Differences between normal operation of all steam generators and the 1/4 section isolation conditions were observed and the results were used as the data 1/4 section isolation conditions were observed and the results were used as the data for RETRAN-03/INT code validation. RETRAN-03 code was modified for the development of natural circulation analysis code for integral type reactors, which was development of natural circulation analysis code for integral type reactors, which was named as RETRAN-03/INT. 3-dimensional analysis models for asymmetric flow in integral type reactors were developed using vector momentum equations in RETRAN-03. Analysis results using RETRAN-03/INT were compared with experimental and CFX4 analysis results and showed good agreements. The natural circulation characteristics obtained in this study will provide the important and fundamental design features for the future small and medium integral reactors. (author). 29 refs., 75 figs., 18 tabs.

  3. Telemetry advances in data compression and channel coding

    Science.gov (United States)

    Miller, Warner H.; Morakis, James C.; Yeh, Pen-Shu

    1990-01-01

    Addressed in this paper is the dependence of telecommunication channel, forward error correcting coding and source data compression coding on integrated circuit technology. Emphasis is placed on real time high speed Reed Solomon (RS) decoding using full custom VLSI technology. Performance curves of NASA's standard channel coder and a proposed standard lossless data compression coder are presented.

  4. Multiple component codes based generalized LDPC codes for high-speed optical transport.

    Science.gov (United States)

    Djordjevic, Ivan B; Wang, Ting

    2014-07-14

    A class of generalized low-density parity-check (GLDPC) codes suitable for optical communications is proposed, which consists of multiple local codes. It is shown that Hamming, BCH, and Reed-Muller codes can be used as local codes, and that the maximum a posteriori probability (MAP) decoding of these local codes by Ashikhmin-Lytsin algorithm is feasible in terms of complexity and performance. We demonstrate that record coding gains can be obtained from properly designed GLDPC codes, derived from multiple component codes. We then show that several recently proposed classes of LDPC codes such as convolutional and spatially-coupled codes can be described using the concept of GLDPC coding, which indicates that the GLDPC coding can be used as a unified platform for advanced FEC enabling ultra-high speed optical transport. The proposed class of GLDPC codes is also suitable for code-rate adaption, to adjust the error correction strength depending on the optical channel conditions.

  5. Mechanistic systems modeling to guide drug discovery and development.

    Science.gov (United States)

    Schmidt, Brian J; Papin, Jason A; Musante, Cynthia J

    2013-02-01

    A crucial question that must be addressed in the drug development process is whether the proposed therapeutic target will yield the desired effect in the clinical population. Pharmaceutical and biotechnology companies place a large investment on research and development, long before confirmatory data are available from human trials. Basic science has greatly expanded the computable knowledge of disease processes, both through the generation of large omics data sets and a compendium of studies assessing cellular and systemic responses to physiologic and pathophysiologic stimuli. Given inherent uncertainties in drug development, mechanistic systems models can better inform target selection and the decision process for advancing compounds through preclinical and clinical research. Copyright © 2012 Elsevier Ltd. All rights reserved.

  6. Recent advances in coding theory for near error-free communications

    Science.gov (United States)

    Cheung, K.-M.; Deutsch, L. J.; Dolinar, S. J.; Mceliece, R. J.; Pollara, F.; Shahshahani, M.; Swanson, L.

    1991-01-01

    Channel and source coding theories are discussed. The following subject areas are covered: large constraint length convolutional codes (the Galileo code); decoder design (the big Viterbi decoder); Voyager's and Galileo's data compression scheme; current research in data compression for images; neural networks for soft decoding; neural networks for source decoding; finite-state codes; and fractals for data compression.

  7. Progress on DART code optimization

    International Nuclear Information System (INIS)

    Taboada, Horacio; Solis, Diego; Rest, Jeffrey

    1999-01-01

    This work consists about the progress made on the design and development of a new optimized version of DART code (DART-P), a mechanistic computer model for the performance calculation and assessment of aluminum dispersion fuel. It is part of a collaboration agreement between CNEA and ANL in the area of Low Enriched Uranium Advanced Fuels. It is held by the Implementation Arrangement for Technical Exchange and Cooperation in the Area of Peaceful Uses of Nuclear Energy, signed on October 16, 1997 between US DOE and the National Atomic Energy Commission of the Argentine Republic. DART optimization is a biannual program; it is operative since February 8, 1999 and has the following goals: 1. Design and develop a new DART calculation kernel for implementation within a parallel processing architecture. 2. Design and develop new user-friendly I/O routines to be resident on Personal Computer (PC)/WorkStation (WS) platform. 2.1. The new input interface will be designed and developed by means of a Visual interface, able to guide the user in the construction of the problem to be analyzed with the aid of a new database (described in item 3, below). The new I/O interface will include input data check controls in order to avoid corrupted input data. 2.2. The new output interface will be designed and developed by means of graphical tools, able to translate numeric data output into 'on line' graphic information. 3. Design and develop a new irradiated materials database, to be resident on PC/WS platform, so as to facilitate the analysis of the behavior of different fuel and meat compositions with DART-P. Currently, a different version of DART is used for oxide, silicide, and advanced alloy fuels. 4. Develop rigorous general inspection algorithms in order to provide valuable DART-P benchmarks. 5. Design and develop new models, such as superplasticity, elastoplastic feedback, improved models for the calculation of fuel deformation and the evolution of the fuel microstructure for

  8. Advanced Technology and Mitigation (ATDM) SPARC Re-Entry Code Fiscal Year 2017 Progress and Accomplishments for ECP.

    Energy Technology Data Exchange (ETDEWEB)

    Crozier, Paul [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Howard, Micah [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Rider, William J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Freno, Brian Andrew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bova, Steven W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Carnes, Brian [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-09-01

    The SPARC (Sandia Parallel Aerodynamics and Reentry Code) will provide nuclear weapon qualification evidence for the random vibration and thermal environments created by re-entry of a warhead into the earth’s atmosphere. SPARC incorporates the innovative approaches of ATDM projects on several fronts including: effective harnessing of heterogeneous compute nodes using Kokkos, exascale-ready parallel scalability through asynchronous multi-tasking, uncertainty quantification through Sacado integration, implementation of state-of-the-art reentry physics and multiscale models, use of advanced verification and validation methods, and enabling of improved workflows for users. SPARC is being developed primarily for the Department of Energy nuclear weapon program, with additional development and use of the code is being supported by the Department of Defense for conventional weapons programs.

  9. Advanced Chemical Propulsion Study

    Science.gov (United States)

    Woodcock, Gordon; Byers, Dave; Alexander, Leslie A.; Krebsbach, Al

    2004-01-01

    A study was performed of advanced chemical propulsion technology application to space science (Code S) missions. The purpose was to begin the process of selecting chemical propulsion technology advancement activities that would provide greatest benefits to Code S missions. Several missions were selected from Code S planning data, and a range of advanced chemical propulsion options was analyzed to assess capabilities and benefits re these missions. Selected beneficial applications were found for higher-performing bipropellants, gelled propellants, and cryogenic propellants. Technology advancement recommendations included cryocoolers and small turbopump engines for cryogenic propellants; space storable propellants such as LOX-hydrazine; and advanced monopropellants. It was noted that fluorine-bearing oxidizers offer performance gains over more benign oxidizers. Potential benefits were observed for gelled propellants that could be allowed to freeze, then thawed for use.

  10. Development of steady thermal-hydraulic analysis code for China advanced research reactor

    International Nuclear Information System (INIS)

    Tian Wenxi; Qiu Suizheng; Guo Yun; Su Guanghui; Jia Dounan; Liu Tiancai; Zhang Jianwei

    2006-01-01

    A multi-channel model steady-state thermal-hydraulic analysis code was developed for China Advanced Research Reactor (CARR). By simulating the whole reactor core, the detailed flow distribution in the core was obtained. The result shows that the structure size plays the most important role in flow distribution and the influence of core power could be neglected under single-phase flow. The temperature field of fuel element under unsymmetrical cooling condition was also obtained, which is necessary for the further study such as stress analysis etc. of the fuel element. At the same time, considering the hot channel effect including engineering factor and nuclear factor, calculation of hot channel was carried out and it is proved that all thermal-hydraulic parameters accord with the Safety Regulation of CARR. (authors)

  11. Reactor lattice codes

    International Nuclear Information System (INIS)

    Kulikowska, T.

    1999-01-01

    The present lecture has a main goal to show how the transport lattice calculations are realised in a standard computer code. This is illustrated on the example of the WIMSD code, belonging to the most popular tools for reactor calculations. Most of the approaches discussed here can be easily modified to any other lattice code. The description of the code assumes the basic knowledge of reactor lattice, on the level given in the lecture on 'Reactor lattice transport calculations'. For more advanced explanation of the WIMSD code the reader is directed to the detailed descriptions of the code cited in References. The discussion of the methods and models included in the code is followed by the generally used homogenisation procedure and several numerical examples of discrepancies in calculated multiplication factors based on different sources of library data. (author)

  12. Rational and Mechanistic Perspectives on Reinforcement Learning

    Science.gov (United States)

    Chater, Nick

    2009-01-01

    This special issue describes important recent developments in applying reinforcement learning models to capture neural and cognitive function. But reinforcement learning, as a theoretical framework, can apply at two very different levels of description: "mechanistic" and "rational." Reinforcement learning is often viewed in mechanistic terms--as…

  13. Experimental validation for combustion analysis of GOTHIC code in 2-dimensional combustion chamber

    International Nuclear Information System (INIS)

    Lee, J. W.; Yang, S. Y.; Park, K. C.; Jung, S. H.

    2002-01-01

    In this study, the prediction capability of GOTHIC code for hydrogen combustion phenomena was validated with the results of two-dimensional premixed hydrogen combustion experiment executed by Seoul National University. The experimental chamber has about 24 liter free volume (1x0.024x1 m 3 ) and 2-dimensional rectangular shape. The test were preformed with 10% hydrogen/air gas mixture and conducted with combination of two igniter positions (top center, top corner) and two boundary conditions (bottom full open, bottom right half open). Using the lumped parameter and mechanistic combustion model in GOTHIC code, the SNU experiments were simulated under the same conditions. The GOTHIC code prediction of the hydrogen combustion phenomena did not compare well with the experimental results. In case of lumped parameter simulation, the combustion time was predicted appropriately. But any other local information related combustion phenomena could not be obtained. In case of mechanistic combustion analysis, the physical combustion phenomena of gas mixture were not matched experimental ones. In boundary open cases, the GOTHIC predicted very long combustion time and the flame front propagation could not simulate appropriately. Though GOTHIC showed flame propagation phenomenon in adiabatic calculation, the induction time of combustion was still very long compare with experimental results. Also, it was found that the combustion model of GOTHIC code had some weak points in low concentration of hydrogen combustion simulation

  14. Parallel processing of structural integrity analysis codes

    International Nuclear Information System (INIS)

    Swami Prasad, P.; Dutta, B.K.; Kushwaha, H.S.

    1996-01-01

    Structural integrity analysis forms an important role in assessing and demonstrating the safety of nuclear reactor components. This analysis is performed using analytical tools such as Finite Element Method (FEM) with the help of digital computers. The complexity of the problems involved in nuclear engineering demands high speed computation facilities to obtain solutions in reasonable amount of time. Parallel processing systems such as ANUPAM provide an efficient platform for realising the high speed computation. The development and implementation of software on parallel processing systems is an interesting and challenging task. The data and algorithm structure of the codes plays an important role in exploiting the parallel processing system capabilities. Structural analysis codes based on FEM can be divided into two categories with respect to their implementation on parallel processing systems. The first category codes such as those used for harmonic analysis, mechanistic fuel performance codes need not require the parallelisation of individual modules of the codes. The second category of codes such as conventional FEM codes require parallelisation of individual modules. In this category, parallelisation of equation solution module poses major difficulties. Different solution schemes such as domain decomposition method (DDM), parallel active column solver and substructuring method are currently used on parallel processing systems. Two codes, FAIR and TABS belonging to each of these categories have been implemented on ANUPAM. The implementation details of these codes and the performance of different equation solvers are highlighted. (author). 5 refs., 12 figs., 1 tab

  15. Recent advances in multiview distributed video coding

    Science.gov (United States)

    Dufaux, Frederic; Ouaret, Mourad; Ebrahimi, Touradj

    2007-04-01

    We consider dense networks of surveillance cameras capturing overlapped images of the same scene from different viewing directions, such a scenario being referred to as multi-view. Data compression is paramount in such a system due to the large amount of captured data. In this paper, we propose a Multi-view Distributed Video Coding approach. It allows for low complexity / low power consumption at the encoder side, and the exploitation of inter-view correlation without communications among the cameras. We introduce a combination of temporal intra-view side information and homography inter-view side information. Simulation results show both the improvement of the side information, as well as a significant gain in terms of coding efficiency.

  16. Preface: Research advances in vadose zone hydrology through simulations with the TOUGH codes

    International Nuclear Information System (INIS)

    Finsterle, Stefan; Oldenburg, Curtis M.

    2004-01-01

    Numerical simulators are playing an increasingly important role in advancing our fundamental understanding of hydrological systems. They are indispensable tools for managing groundwater resources, analyzing proposed and actual remediation activities at contaminated sites, optimizing recovery of oil, gas, and geothermal energy, evaluating subsurface structures and mining activities, designing monitoring systems, assessing the long-term impacts of chemical and nuclear waste disposal, and devising improved irrigation and drainage practices in agricultural areas, among many other applications. The complexity of subsurface hydrology in the vadose zone calls for sophisticated modeling codes capable of handling the strong nonlinearities involved, the interactions of coupled physical, chemical and biological processes, and the multiscale heterogeneities inherent in such systems. The papers in this special section of ''Vadose Zone Journal'' are illustrative of the enormous potential of such numerical simulators as applied to the vadose zone. The papers describe recent developments and applications of one particular set of codes, the TOUGH family of codes, as applied to nonisothermal flow and transport in heterogeneous porous and fractured media (http://www-esd.lbl.gov/TOUGH2). The contributions were selected from presentations given at the TOUGH Symposium 2003, which brought together developers and users of the TOUGH codes at the Lawrence Berkeley National Laboratory (LBNL) in Berkeley, California, for three days of information exchange in May 2003 (http://www-esd.lbl.gov/TOUGHsymposium). The papers presented at the symposium covered a wide range of topics, including geothermal reservoir engineering, fracture flow and vadose zone hydrology, nuclear waste disposal, mining engineering, reactive chemical transport, environmental remediation, and gas transport. This Special Section of ''Vadose Zone Journal'' contains revised and expanded versions of selected papers from the

  17. Advancements in reactor physics modelling methodology of Monte Carlo Burnup Code MCB dedicated to higher simulation fidelity of HTR cores

    International Nuclear Information System (INIS)

    Cetnar, Jerzy

    2014-01-01

    The recent development of MCB - Monte Carlo Continuous Energy Burn-up code is directed towards advanced description of modern reactors, including double heterogeneity structures that exist in HTR-s. In this, we exploit the advantages of MCB methodology in integrated approach, where physics, neutronics, burnup, reprocessing, non-stationary process modeling (control rod operation) and refined spatial modeling are carried in a single flow. This approach allows for implementations of advanced statistical options like analysis of error propagation, perturbation in time domain, sensitivity and source convergence analyses. It includes statistical analysis of burnup process, emitted particle collection, thermal-hydraulic coupling, automatic power profile calculations, advanced procedures of burnup step normalization and enhanced post processing capabilities. (author)

  18. Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC): gap analysis for high fidelity and performance assessment code development

    International Nuclear Information System (INIS)

    Lee, Joon H.; Siegel, Malcolm Dean; Arguello, Jose Guadalupe Jr.; Webb, Stephen Walter; Dewers, Thomas A.; Mariner, Paul E.; Edwards, Harold Carter; Fuller, Timothy J.; Freeze, Geoffrey A.; Jove-Colon, Carlos F.; Wang, Yifeng

    2011-01-01

    This report describes a gap analysis performed in the process of developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with rigorous verification, validation, and software quality requirements. The gap analyses documented in this report were are performed during an initial gap analysis to identify candidate codes and tools to support the development and integration of the Waste IPSC, and during follow-on activities that delved into more detailed assessments of the various codes that were acquired, studied, and tested. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. The gap analysis indicates that significant capabilities may already exist in the existing THC codes although there is no single code able to fully account for all physical and chemical processes involved in a waste disposal system. Large gaps exist in modeling chemical processes and their couplings with other processes. The coupling of chemical processes with flow transport and mechanical deformation remains challenging. The data for extreme environments (e.g., for elevated temperature and high ionic strength media) that are

  19. Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC) : gap analysis for high fidelity and performance assessment code development.

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Joon H.; Siegel, Malcolm Dean; Arguello, Jose Guadalupe, Jr.; Webb, Stephen Walter; Dewers, Thomas A.; Mariner, Paul E.; Edwards, Harold Carter; Fuller, Timothy J.; Freeze, Geoffrey A.; Jove-Colon, Carlos F.; Wang, Yifeng

    2011-03-01

    This report describes a gap analysis performed in the process of developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with rigorous verification, validation, and software quality requirements. The gap analyses documented in this report were are performed during an initial gap analysis to identify candidate codes and tools to support the development and integration of the Waste IPSC, and during follow-on activities that delved into more detailed assessments of the various codes that were acquired, studied, and tested. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. The gap analysis indicates that significant capabilities may already exist in the existing THC codes although there is no single code able to fully account for all physical and chemical processes involved in a waste disposal system. Large gaps exist in modeling chemical processes and their couplings with other processes. The coupling of chemical processes with flow transport and mechanical deformation remains challenging. The data for extreme environments (e.g., for elevated temperature and high ionic strength media) that are

  20. Steam explosion simulation code JASMINE v.3 user's guide

    International Nuclear Information System (INIS)

    Moriyama, Kiyofumi; Maruyama, Yu; Nakamura, Hideo

    2008-07-01

    A steam explosion occurs when hot liquid contacts with cold volatile liquid. In this phenomenon, fine fragmentation of the hot liquid causes extremely rapid heat transfer from the hot liquid to the cold volatile liquid, and explosive vaporization, bringing shock waves and destructive forces. The steam explosion due to the contact of the molten core material and coolant water during severe accidents of light water reactors has been regarded as a potential threat to the integrity of the containment vessel. We developed a mechanistic steam explosion simulation code, JASMINE, that is applicable to plant scale assessment of the steam explosion loads. This document, as a manual for users of JASMINE code, describes the models, numerical solution methods, and also some verification and example calculations, as well as practical instructions for input preparation and usage of the code. (author)

  1. Coding for urologic office procedures.

    Science.gov (United States)

    Dowling, Robert A; Painter, Mark

    2013-11-01

    This article summarizes current best practices for documenting, coding, and billing common office-based urologic procedures. Topics covered include general principles, basic and advanced urologic coding, creation of medical records that support compliant coding practices, bundled codes and unbundling, global periods, modifiers for procedure codes, when to bill for evaluation and management services during the same visit, coding for supplies, and laboratory and radiology procedures pertinent to urology practice. Detailed information is included for the most common urology office procedures, and suggested resources and references are provided. This information is of value to physicians, office managers, and their coding staff. Copyright © 2013 Elsevier Inc. All rights reserved.

  2. Development of realistic thermal hydraulic system analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Chung, B. D; Kim, K. D. [and others

    2002-05-01

    The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others.

  3. Development of realistic thermal hydraulic system analysis code

    International Nuclear Information System (INIS)

    Lee, Won Jae; Chung, B. D; Kim, K. D.

    2002-05-01

    The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others

  4. CHF predictor derived from a 3D thermal-hydraulic code and an advanced statistical method

    International Nuclear Information System (INIS)

    Banner, D.; Aubry, S.

    2004-01-01

    A rod bundle CHF predictor has been determined by using a 3D code (THYC) to compute local thermal-hydraulic conditions at the boiling crisis location. These local parameters have been correlated to the critical heat flux by using an advanced statistical method based on spline functions. The main characteristics of the predictor are presented in conjunction with a detailed analysis of predictions (P/M ratio) in order to prove that the usual safety methodology can be applied with such a predictor. A thermal-hydraulic design criterion is obtained (1.13) and the predictor is compared with the WRB-1 correlation. (author)

  5. Development of Regulatory Audit Core Safety Code : COREDAX

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Chae Yong; Jo, Jong Chull; Roh, Byung Hwan [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Jae Jun; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2005-07-01

    Korea Institute of Nuclear Safety (KINS) has developed a core neutronics simulator, COREDAX code, for verifying core safety of SMART-P reactor, which is technically supported by Korea Advanced Institute of Science and Technology (KAIST). The COREDAX code would be used for regulatory audit calculations of 3- dimendional core neutronics. The COREDAX code solves the steady-state and timedependent multi-group neutron diffusion equation in hexagonal geometry as well as rectangular geometry by analytic function expansion nodal (AFEN) method. AFEN method was developed at KAIST, and it was internationally verified that its accuracy is excellent. The COREDAX code is originally programmed based on the AFEN method. Accuracy of the code on the AFEN method was excellent for the hexagonal 2-dimensional problems, but there was a need for improvement for hexagonal-z 3-dimensional problems. Hence, several solution routines of the AFEN method are improved, and finally the advanced AFEN method is created. COREDAX code is based on the advanced AFEN method . The initial version of COREDAX code is to complete a basic framework, performing eigenvalue calculations and kinetics calculations with thermal-hydraulic feedbacks, for audit calculations of steady-state core design and reactivity-induced accidents of SMART-P reactor. This study describes the COREDAX code for hexagonal geometry.

  6. New Advances in Photoionisation Codes: How and what for?

    International Nuclear Information System (INIS)

    Ercolano, Barbara

    2005-01-01

    The study of photoionised gas in planetary nebulae (PNe) has played a major role in the achievement, over the years, of a better understanding of a number of physical processes, pertinent to a broader range of fields than that of PNe studies, spanning from atomic physics to stellar evolution theories. Whilst empirical techniques are routinely employed for the analysis of the emission line spectra of these objects, the accurate interpretation of the observational data often requires the solution of a set of coupled equations, via the application of a photoionisation/plasma code. A number of large-scale codes have been developed since the late sixties, using various analytical or statistical techniques for the transfer of continuum radiation, mainly under the assumption of spherical symmetry and a few in 3D. These codes have been proved to be powerful and in many cases essential tools, but a clear idea of the underlying physical processes and assumptions is necessary in order to avoid reaching misleading conclusions.The development of the codes over the years has been driven by the observational constraints available, but also compromised by the available computer power. Modern codes are faster and more flexible, with the ultimate goal being the achievement of a description of the observations relying on the smallest number of parameters possible. In this light recent developments have been focused on the inclusion of new available atomic data, the inclusion of a realistic treatment for dust grains mixed in the ionised and photon dominated regions (PDRs) and the expansion of some codes to PDRs with the inclusion of chemical reaction networks. Furthermore the last few years have seen the development of fully 3D photoionisation codes based on the Monte Carlo method.A brief review of the field of photoionisation today is given here, with emphasis on the recent developments, including the expansion of the models to the 3D domain. Attention is given to the identification

  7. Mechanistic curiosity will not kill the Bayesian cat

    NARCIS (Netherlands)

    Borsboom, D.; Wagenmakers, E.-J.; Romeijn, J.-W.

    2011-01-01

    Jones & Love (J&L) suggest that Bayesian approaches to the explanation of human behavior should be constrained by mechanistic theories. We argue that their proposal misconstrues the relation between process models, such as the Bayesian model, and mechanisms. While mechanistic theories can answer

  8. Mechanistic curiosity will not kill the Bayesian cat

    NARCIS (Netherlands)

    Borsboom, Denny; Wagenmakers, Eric-Jan; Romeijn, Jan-Willem

    Jones & Love (J&L) suggest that Bayesian approaches to the explanation of human behavior should be constrained by mechanistic theories. We argue that their proposal misconstrues the relation between process models, such as the Bayesian model, and mechanisms. While mechanistic theories can answer

  9. SCDAP/RELAP5/MOD3 code development

    International Nuclear Information System (INIS)

    Allison, C.M.; Siefken, J.L.; Coryell, E.W.

    1992-01-01

    The SCOAP/RELAP5/MOD3 computer code is designed to describe the overall reactor coolant system (RCS) thermal-hydraulic response, core damage progression, and fission product release and transport during severe accidents. The code is being developed at the Idaho National Engineering Laboratory (INEL) under the primary sponsorship of the Office of Nuclear Regulatory Research of the US Nuclear Regulatory Commission (NRC). Code development activities are currently focused on three main areas - (a) code usability, (b) early phase melt progression model improvements, and (c) advanced reactor thermal-hydraulic model extensions. This paper describes the first two activities. A companion paper describes the advanced reactor model improvements being performed under RELAP5/MOD3 funding

  10. Advanced Design of Dumbbell-shaped Genetic Minimal Vectors Improves Non-coding and Coding RNA Expression.

    Science.gov (United States)

    Jiang, Xiaoou; Yu, Han; Teo, Cui Rong; Tan, Genim Siu Xian; Goh, Sok Chin; Patel, Parasvi; Chua, Yiqiang Kevin; Hameed, Nasirah Banu Sahul; Bertoletti, Antonio; Patzel, Volker

    2016-09-01

    Dumbbell-shaped DNA minimal vectors lacking nontherapeutic genes and bacterial sequences are considered a stable, safe alternative to viral, nonviral, and naked plasmid-based gene-transfer systems. We investigated novel molecular features of dumbbell vectors aiming to reduce vector size and to improve the expression of noncoding or coding RNA. We minimized small hairpin RNA (shRNA) or microRNA (miRNA) expressing dumbbell vectors in size down to 130 bp generating the smallest genetic expression vectors reported. This was achieved by using a minimal H1 promoter with integrated transcriptional terminator transcribing the RNA hairpin structure around the dumbbell loop. Such vectors were generated with high conversion yields using a novel protocol. Minimized shRNA-expressing dumbbells showed accelerated kinetics of delivery and transcription leading to enhanced gene silencing in human tissue culture cells. In primary human T cells, minimized miRNA-expressing dumbbells revealed higher stability and triggered stronger target gene suppression as compared with plasmids and miRNA mimics. Dumbbell-driven gene expression was enhanced up to 56- or 160-fold by implementation of an intron and the SV40 enhancer compared with control dumbbells or plasmids. Advanced dumbbell vectors may represent one option to close the gap between durable expression that is achievable with integrating viral vectors and short-term effects triggered by naked RNA.

  11. Development of a computer code for dynamic analysis of the primary circuit of advanced reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rocha, Jussie Soares da; Lira, Carlos A.B.O.; Magalhaes, Mardson A. de Sa, E-mail: cabol@ufpe.b [Universidade Federal de Pernambuco (DEN/UFPE), Recife, PE (Brazil). Dept. de Energia Nuclear

    2011-07-01

    Currently, advanced reactors are being developed, seeking for enhanced safety, better performance and low environmental impacts. Reactor designs must follow several steps and numerous tests before a conceptual project could be certified. In this sense, computational tools become indispensable in the preparation of such projects. Thus, this study aimed at the development of a computational tool for thermal-hydraulic analysis by coupling two computer codes to evaluate the influence of transients caused by pressure variations and flow surges in the region of the primary circuit of IRIS reactor between the core and the pressurizer. For the simulation, it was used a situation of 'insurge', characterized by the entry of water in the pressurizer, due to the expansion of the refrigerant in the primary circuit. This expansion was represented by a pressure disturbance in step form, through the block 'step' of SIMULINK, thus enabling the transient startup. The results showed that the dynamic tool, obtained through the coupling of the codes, generated very satisfactory responses within model limitations, preserving the most important phenomena in the process. (author)

  12. Development of a computer code for dynamic analysis of the primary circuit of advanced reactors

    International Nuclear Information System (INIS)

    Rocha, Jussie Soares da; Lira, Carlos A.B.O.; Magalhaes, Mardson A. de Sa

    2011-01-01

    Currently, advanced reactors are being developed, seeking for enhanced safety, better performance and low environmental impacts. Reactor designs must follow several steps and numerous tests before a conceptual project could be certified. In this sense, computational tools become indispensable in the preparation of such projects. Thus, this study aimed at the development of a computational tool for thermal-hydraulic analysis by coupling two computer codes to evaluate the influence of transients caused by pressure variations and flow surges in the region of the primary circuit of IRIS reactor between the core and the pressurizer. For the simulation, it was used a situation of 'insurge', characterized by the entry of water in the pressurizer, due to the expansion of the refrigerant in the primary circuit. This expansion was represented by a pressure disturbance in step form, through the block 'step' of SIMULINK, thus enabling the transient startup. The results showed that the dynamic tool, obtained through the coupling of the codes, generated very satisfactory responses within model limitations, preserving the most important phenomena in the process. (author)

  13. Further development of the computer code ATHLET-CD

    International Nuclear Information System (INIS)

    Weber, Sebastian; Austregesilo, Henrique; Bals, Christine; Band, Sebastian; Hollands, Thorsten; Koellein, Carsten; Lovasz, Liviusz; Pandazis, Peter; Schubert, Johann-Dietrich; Sonnenkalb, Martin

    2016-10-01

    In the framework of the reactor safety research program sponsored by the German Federal Ministry for Economic Affairs and Energy (BMWi), the computer code system ATHLET/ATHLET-CD has been further developed as an analysis tool for the simulation of accidents in nuclear power plants with pressurized and boiling water reactors as well as for the evaluation of accident management procedures. The main objective was to provide a mechanistic analysis tool for best estimate calculations of transients, accidents, and severe accidents with core degradation in light water reactors. With the continued development, the capability of the code system has been largely improved, allowing best estimate calculations of design and beyond design base accidents, and the simulation of advanced core degradation with enhanced model extent in a reasonable calculation time. ATHLET comprises inter alia a 6-equation model, models for the simulation of non-condensable gases and tracking of boron concentration, as well as additional component and process models for the complete system simulation. Among numerous model improvements, the code application has been extended to super critical pressures. The mechanistic description of the dynamic development of flow regimes on the basis of a transport equation for the interface area has been further developed. This ATHLET version is completely integrated in ATHLET-CD. ATHLET-CD further comprises dedicated models for the simulation of fuel and control assembly degradation for both pressurized and boiling water reactors, debris bed with melting in the core region, as well as fission product and aerosol release and transport in the cooling system, inclusive of decay of nuclide inventories and of chemical reactions in the gas phase. The continued development also concerned the modelling of absorber material release, of melting, melt relocation and freezing, and the interaction with the wall of the reactor pressure vessel. The following models were newly

  14. The H.264/MPEG4 advanced video coding

    Science.gov (United States)

    Gromek, Artur

    2009-06-01

    H.264/MPEG4-AVC is the newest video coding standard recommended by International Telecommunication Union - Telecommunication Standardization Section (ITU-T) and the ISO/IEC Moving Picture Expert Group (MPEG). The H.264/MPEG4-AVC has recently become leading standard for generic audiovisual services, since deployment for digital television. Nowadays is commonly used in wide range of video application ranging like mobile services, videoconferencing, IPTV, HDTV, video storage and many more. In this article, author briefly describes the technology applied in the H.264/MPEG4-AVC video coding standard, the way of real-time implementation and the way of future development.

  15. Failed fuel diagnosis during WWER reactor operation using the RTOP-CA code

    International Nuclear Information System (INIS)

    Likhanskii, V.; Afanasieva, E.; Sorokin, A.; Evdokimov, I.; Kanukova, V.; Khromov, A.

    2006-01-01

    The mechanistic code RTOP-CA is developed for objectives of failed fuel diagnosis during WWER reactor operation. The RTOP-CA code enables to solve a direct problem: modelling the failed fuel behavior and prediction of primary coolant activity if characteristics of failures in the reactor core are known. Results of verification of the RTOP-CA code are presented. Separate physical models were verified on small-scale in-pile and out-of-pile experiments. Integral verification cases included data obtained at research reactors and at nuclear power plants. The RTOP-CA code is used for development of a neural-network approach to the inverse problem: detection of failure characteristics on the base of data on primary coolant activity during reactor operation. Preliminary results of application of the neural-network approach for evaluation of fuel failure characteristics are presented. (authors)

  16. Performance of code 'FAIR' in IAEA CRP on FUMEX

    International Nuclear Information System (INIS)

    Swami Prasad, P.; Dutta, B.K.; Kushwaha, H.S.; Kakodkar, A.

    1996-01-01

    A modern fuel performance analysis code FAIR has been developed for analysing high burnup fuel pins of water/heavy water cooled reactors. The code employs finite element method for modelling thermo mechanical behaviour of fuel pins and mechanistic models for modelling various physical and chemical phenomena affecting the behaviour of nuclear reactor fuel pins. High burnup affects such as pellet thermal conductivity degradation, enhanced fission gas release and radial flux redistribution are incorporated in the code FAIR. The code FAIR is capable of performing statistical analysis of fuel pins using Monte Carlo technique. The code is implemented on BARC parallel processing system ANUPAM. The code has recently participated in an International Atomic Energy Agency (IAEA) coordinated research program (CRP) on fuel modelling at extended burnups (FUMEX). Nineteen agencies from different countries participated in this exercise. In this CRP, spread over a period of three years, a number of high burnup fuel pins irradiated at Halden reactor are analysed. The first phase of the CRP is a blind code comparison exercise, where the computed results are compared with experimental results. The second phase consists of modifications to the code based on the experimental results of first phase and statistical analysis of fuel pins. The performance of the code FAIR in this CRP has been very good. The present report highlights the main features of code FAIR and its performance in the IAEA CRP on FUMEX. 14 refs., 5 tabs., ills

  17. The general theory of convolutional codes

    Science.gov (United States)

    Mceliece, R. J.; Stanley, R. P.

    1993-01-01

    This article presents a self-contained introduction to the algebraic theory of convolutional codes. This introduction is partly a tutorial, but at the same time contains a number of new results which will prove useful for designers of advanced telecommunication systems. Among the new concepts introduced here are the Hilbert series for a convolutional code and the class of compact codes.

  18. Testing mechanistic models of growth in insects.

    Science.gov (United States)

    Maino, James L; Kearney, Michael R

    2015-11-22

    Insects are typified by their small size, large numbers, impressive reproductive output and rapid growth. However, insect growth is not simply rapid; rather, insects follow a qualitatively distinct trajectory to many other animals. Here we present a mechanistic growth model for insects and show that increasing specific assimilation during the growth phase can explain the near-exponential growth trajectory of insects. The presented model is tested against growth data on 50 insects, and compared against other mechanistic growth models. Unlike the other mechanistic models, our growth model predicts energy reserves per biomass to increase with age, which implies a higher production efficiency and energy density of biomass in later instars. These predictions are tested against data compiled from the literature whereby it is confirmed that insects increase their production efficiency (by 24 percentage points) and energy density (by 4 J mg(-1)) between hatching and the attainment of full size. The model suggests that insects achieve greater production efficiencies and enhanced growth rates by increasing specific assimilation and increasing energy reserves per biomass, which are less costly to maintain than structural biomass. Our findings illustrate how the explanatory and predictive power of mechanistic growth models comes from their grounding in underlying biological processes. © 2015 The Author(s).

  19. Zebrafish Models of Human Leukemia: Technological Advances and Mechanistic Insights.

    Science.gov (United States)

    Harrison, Nicholas R; Laroche, Fabrice J F; Gutierrez, Alejandro; Feng, Hui

    2016-01-01

    Insights concerning leukemic pathophysiology have been acquired in various animal models and further efforts to understand the mechanisms underlying leukemic treatment resistance and disease relapse promise to improve therapeutic strategies. The zebrafish (Danio rerio) is a vertebrate organism with a conserved hematopoietic program and unique experimental strengths suiting it for the investigation of human leukemia. Recent technological advances in zebrafish research including efficient transgenesis, precise genome editing, and straightforward transplantation techniques have led to the generation of a number of leukemia models. The transparency of the zebrafish when coupled with improved lineage-tracing and imaging techniques has revealed exquisite details of leukemic initiation, progression, and regression. With these advantages, the zebrafish represents a unique experimental system for leukemic research and additionally, advances in zebrafish-based high-throughput drug screening promise to hasten the discovery of novel leukemia therapeutics. To date, investigators have accumulated knowledge of the genetic underpinnings critical to leukemic transformation and treatment resistance and without doubt, zebrafish are rapidly expanding our understanding of disease mechanisms and helping to shape therapeutic strategies for improved outcomes in leukemic patients.

  20. Code Development and Analysis Program: developmental checkout of the BEACON/MOD2A code

    International Nuclear Information System (INIS)

    Ramsthaler, J.A.; Lime, J.F.; Sahota, M.S.

    1978-12-01

    A best-estimate transient containment code, BEACON, is being developed by EG and G Idaho, Inc. for the Nuclear Regulatory Commission's reactor safety research program. This is an advanced, two-dimensional fluid flow code designed to predict temperatures and pressures in a dry PWR containment during a hypothetical loss-of-coolant accident. The most recent version of the code, MOD2A, is presently in the final stages of production prior to being released to the National Energy Software Center. As part of the final code checkout, seven sample problems were selected to be run with BEACON/MOD2A

  1. Code package {open_quotes}SVECHA{close_quotes}: Modeling of core degradation phenomena at severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Veshchunov, M.S.; Kisselev, A.E.; Palagin, A.V. [Nuclear Safety Institute, Moscow (Russian Federation)] [and others

    1995-09-01

    The code package SVECHA for the modeling of in-vessel core degradation (CD) phenomena in severe accidents is being developed in the Nuclear Safety Institute, Russian Academy of Science (NSI RAS). The code package presents a detailed mechanistic description of the phenomenology of severe accidents in a reactor core. The modules of the package were developed and validated on separate effect test data. These modules were then successfully implemented in the ICARE2 code and validated against a wide range of integral tests. Validation results have shown good agreement with separate effect tests data and with the integral tests CORA-W1/W2, CORA-13, PHEBUS-B9+.

  2. (U) Ristra Next Generation Code Report

    Energy Technology Data Exchange (ETDEWEB)

    Hungerford, Aimee L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Daniel, David John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-09-22

    LANL’s Weapons Physics management (ADX) and ASC program office have defined a strategy for exascale-class application codes that follows two supportive, and mutually risk-mitigating paths: evolution for established codes (with a strong pedigree within the user community) based upon existing programming paradigms (MPI+X); and Ristra (formerly known as NGC), a high-risk/high-reward push for a next-generation multi-physics, multi-scale simulation toolkit based on emerging advanced programming systems (with an initial focus on data-flow task-based models exemplified by Legion [5]). Development along these paths is supported by the ATDM, IC, and CSSE elements of the ASC program, with the resulting codes forming a common ecosystem, and with algorithm and code exchange between them anticipated. Furthermore, solution of some of the more challenging problems of the future will require a federation of codes working together, using established-pedigree codes in partnership with new capabilities as they come on line. The role of Ristra as the high-risk/high-reward path for LANL’s codes is fully consistent with its role in the Advanced Technology Development and Mitigation (ATDM) sub-program of ASC (see Appendix C), in particular its emphasis on evolving ASC capabilities through novel programming models and data management technologies.

  3. A graph model for opportunistic network coding

    KAUST Repository

    Sorour, Sameh; Aboutoraby, Neda; Al-Naffouri, Tareq Y.; Alouini, Mohamed-Slim

    2015-01-01

    © 2015 IEEE. Recent advancements in graph-based analysis and solutions of instantly decodable network coding (IDNC) trigger the interest to extend them to more complicated opportunistic network coding (ONC) scenarios, with limited increase

  4. Development and validation of the ENIGMA code for MOX fuel performance modelling

    International Nuclear Information System (INIS)

    Palmer, I.; Rossiter, G.; White, R.J.

    2000-01-01

    The ENIGMA fuel performance code has been under development in the UK since the mid-1980s with contributions made by both the fuel vendor (BNFL) and the utility (British Energy). In recent years it has become the principal code for UO 2 fuel licensing for both PWR and AGR reactor systems in the UK and has also been used by BNFL in support of overseas UO 2 and MOX fuel business. A significant new programme of work has recently been initiated by BNFL to further develop the code specifically for MOX fuel application. Model development is proceeding hand in hand with a major programme of MOX fuel testing and PIE studies, with the objective of producing a fuel modelling code suitable for mechanistic analysis, as well as for licensing applications. This paper gives an overview of the model developments being undertaken and of the experimental data being used to underpin and to validate the code. The paper provides a summary of the code development programme together with specific examples of new models produced. (author)

  5. Ideas for Advancing Code Sharing: A Different Kind of Hack Day

    Science.gov (United States)

    Teuben, P.; Allen, A.; Berriman, B.; DuPrie, K.; Hanisch, R. J.; Mink, J.; Nemiroff, R. J.; Shamir, L.; Shortridge, K.; Taylor, M. B.; Wallin, J. F.

    2014-05-01

    How do we as a community encourage the reuse of software for telescope operations, data processing, and ? How can we support making codes used in research available for others to examine? Continuing the discussion from last year Bring out your codes! BoF session, participants separated into groups to brainstorm ideas to mitigate factors which inhibit code sharing and nurture those which encourage code sharing. The BoF concluded with the sharing of ideas that arose from the brainstorming sessions and a brief summary by the moderator.

  6. Performance and Complexity Co-evaluation of the Advanced Video Coding Standard for Cost-Effective Multimedia Communications

    Directory of Open Access Journals (Sweden)

    Saponara Sergio

    2004-01-01

    Full Text Available The advanced video codec (AVC standard, recently defined by a joint video team (JVT of ITU-T and ISO/IEC, is introduced in this paper together with its performance and complexity co-evaluation. While the basic framework is similar to the motion-compensated hybrid scheme of previous video coding standards, additional tools improve the compression efficiency at the expense of an increased implementation cost. As a first step to bridge the gap between the algorithmic design of a complex multimedia system and its cost-effective realization, a high-level co-evaluation approach is proposed and applied to a real-life AVC design. An exhaustive analysis of the codec compression efficiency versus complexity (memory and computational costs project space is carried out at the early algorithmic design phase. If all new coding features are used, the improved AVC compression efficiency (up to 50% compared to current video coding technology comes with a complexity increase of a factor 2 for the decoder and larger than one order of magnitude for the encoder. This represents a challenge for resource-constrained multimedia systems such as wireless devices or high-volume consumer electronics. The analysis also highlights important properties of the AVC framework allowing for complexity reduction at the high system level: when combining the new coding features, the implementation complexity accumulates, while the global compression efficiency saturates. Thus, a proper use of the AVC tools maintains the same performance as the most complex configuration while considerably reducing complexity. The reported results provide inputs to assist the profile definition in the standard, highlight the AVC bottlenecks, and select optimal trade-offs between algorithmic performance and complexity.

  7. Causation at Different Levels: Tracking the Commitments of Mechanistic Explanations

    DEFF Research Database (Denmark)

    Fazekas, Peter; Kertész, Gergely

    2011-01-01

    connections transparent. These general commitments get confronted with two claims made by certain proponents of the mechanistic approach: William Bechtel often argues that within the mechanistic framework it is possible to balance between reducing higher levels and maintaining their autonomy at the same time...... their autonomy at the same time than standard reductive accounts are, and that what mechanistic explanations are able to do at best is showing that downward causation does not exist....

  8. ATHENA code manual. Volume 1. Code structure, system models, and solution methods

    International Nuclear Information System (INIS)

    Carlson, K.E.; Roth, P.A.; Ransom, V.H.

    1986-09-01

    The ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) code has been developed to perform transient simulation of the thermal hydraulic systems which may be found in fusion reactors, space reactors, and other advanced systems. A generic modeling approach is utilized which permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of a complete facility. Several working fluids are available to be used in one or more interacting loops. Different loops may have different fluids with thermal connections between loops. The modeling theory and associated numerical schemes are documented in Volume I in order to acquaint the user with the modeling base and thus aid effective use of the code. The second volume contains detailed instructions for input data preparation

  9. Turbo coding, turbo equalisation and space-time coding for transmission over fading channels

    CERN Document Server

    Hanzo, L; Yeap, B

    2002-01-01

    Against the backdrop of the emerging 3G wireless personal communications standards and broadband access network standard proposals, this volume covers a range of coding and transmission aspects for transmission over fading wireless channels. It presents the most important classic channel coding issues and also the exciting advances of the last decade, such as turbo coding, turbo equalisation and space-time coding. It endeavours to be the first book with explicit emphasis on channel coding for transmission over wireless channels. Divided into 4 parts: Part 1 - explains the necessary background for novices. It aims to be both an easy reading text book and a deep research monograph. Part 2 - provides detailed coverage of turbo conventional and turbo block coding considering the known decoding algorithms and their performance over Gaussian as well as narrowband and wideband fading channels. Part 3 - comprehensively discusses both space-time block and space-time trellis coding for the first time in literature. Par...

  10. Real depletion in nodal diffusion codes

    International Nuclear Information System (INIS)

    Petkov, P.T.

    2002-01-01

    The fuel depletion is described by more than one hundred fuel isotopes in the advanced lattice codes like HELIOS, but only a few fuel isotopes are accounted for even in the advanced steady-state diffusion codes. The general assumption that the number densities of the majority of the fuel isotopes depend only on the fuel burnup is seriously in error if high burnup is considered. The real depletion conditions in the reactor core differ from the asymptotic ones at the stage of lattice depletion calculations. This study reveals which fuel isotopes should be explicitly accounted for in the diffusion codes in order to predict adequately the real depletion effects in the core. A somewhat strange conclusion is that if the real number densities of the main fissionable isotopes are not explicitly accounted for in the diffusion code, then Sm-149 should not be accounted for either, because the net error in k-inf is smaller (Authors)

  11. Development of computer code in PNC, 3

    International Nuclear Information System (INIS)

    Ohtaki, Akira; Ohira, Hiroaki

    1990-01-01

    Super-COPD, a code which is integrated by calculation modules, has been developed in order to evaluate kinds of dynamics of LMFBR plant by improving COPD. The code involves all models and its advanced models of COPD in module structures. The code makes it possible to simulate the system dynamics of LMFBR plant of any configurations and components. (author)

  12. Development of a steady thermal-hydraulic analysis code for the China Advanced Research Reactor

    Institute of Scientific and Technical Information of China (English)

    TIAN Wenxi; QIU Suizheng; GUO Yun; SU Guanghui; JIA Dounan; LIU Tiancai; ZHANG Jianwei

    2007-01-01

    A multi-channel model steady-state thermalhydraulic analysis code was developed for the China Advanced Research Reactor (CARR). By simulating the whole reactor core, the detailed mass flow distribution in the core was obtained. The result shows that structure size plays the most important role in mass flow distribution, and the influence of core power could be neglected under singlephase flow. The temperature field of the fuel element under unsymmetrical cooling condition was also obtained, which is necessary for further study such as stress analysis, etc. Of the fuel element. At the same time, considering the hot channel effect including engineering factor and nuclear factor, calculation of the mean and hot channel was carried out and it is proved that all thermal-hydraulic parameters satisfy the "Safety design regulation of CARR".

  13. FLOOD 3 code conversion from Apollo to HP

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hae Cho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-01-01

    FLOOD3 is a Fortran program used to analyze LOCA Reflood and Post Reflood transients for purpose of mechanistically generating containment sizing mass/energy source terms. The reflood time frame starts when safety injection water first enters the bottom of the core, following the initial LOCA blowdown phase, and the ends when liquid level in the core is high enough to quench the core. The post reflood phase starts at the end of reflood phase and lasts until the end of LOCA transient period. Longterm core cooling is initiated following post reflood phase. FLOOD3 code dose not contain separate capability for longterm cooling analysis. This report firstly describes detailed work carried out for installation of FLOOD3 on Apollo DN10000 and code validation results after installation. Secondly, A series of work is also describes in relation to installation of FLOOD3 on HP 9000/700 series as well as relevant code validation results. Attached is a report on software verification and validation results. 7 refs. (Author) .new.

  14. FLOOD 3 code conversion from Apollo to HP

    International Nuclear Information System (INIS)

    Lee, Hae Cho

    1996-01-01

    FLOOD3 is a Fortran program used to analyze LOCA Reflood and Post Reflood transients for purpose of mechanistically generating containment sizing mass/energy source terms. The reflood time frame starts when safety injection water first enters the bottom of the core, following the initial LOCA blowdown phase, and the ends when liquid level in the core is high enough to quench the core. The post reflood phase starts at the end of reflood phase and lasts until the end of LOCA transient period. Longterm core cooling is initiated following post reflood phase. FLOOD3 code dose not contain separate capability for longterm cooling analysis. This report firstly describes detailed work carried out for installation of FLOOD3 on Apollo DN10000 and code validation results after installation. Secondly, A series of work is also describes in relation to installation of FLOOD3 on HP 9000/700 series as well as relevant code validation results. Attached is a report on software verification and validation results. 7 refs. (Author) .new

  15. Benchmark calculation of subchannel analysis codes

    International Nuclear Information System (INIS)

    1996-02-01

    In order to evaluate the analysis capabilities of various subchannel codes used in thermal-hydraulic design of light water reactors, benchmark calculations were performed. The selected benchmark problems and major findings obtained by the calculations were as follows: (1)As for single-phase flow mixing experiments between two channels, the calculated results of water temperature distribution along the flow direction were agreed with experimental results by tuning turbulent mixing coefficients properly. However, the effect of gap width observed in the experiments could not be predicted by the subchannel codes. (2)As for two-phase flow mixing experiments between two channels, in high water flow rate cases, the calculated distributions of air and water flows in each channel were well agreed with the experimental results. In low water flow cases, on the other hand, the air mixing rates were underestimated. (3)As for two-phase flow mixing experiments among multi-channels, the calculated mass velocities at channel exit under steady-state condition were agreed with experimental values within about 10%. However, the predictive errors of exit qualities were as high as 30%. (4)As for critical heat flux(CHF) experiments, two different results were obtained. A code indicated that the calculated CHF's using KfK or EPRI correlations were well agreed with the experimental results, while another code suggested that the CHF's were well predicted by using WSC-2 correlation or Weisman-Pei mechanistic model. (5)As for droplets entrainment and deposition experiments, it was indicated that the predictive capability was significantly increased by improving correlations. On the other hand, a remarkable discrepancy between codes was observed. That is, a code underestimated the droplet flow rate and overestimated the liquid film flow rate in high quality cases, while another code overestimated the droplet flow rate and underestimated the liquid film flow rate in low quality cases. (J.P.N.)

  16. Development of a Code for the Long Term Radiological Safety Assessment of Radioactive Wastes from Advanced Nuclear Fuel Cycle Facilities in Republic of Korea

    International Nuclear Information System (INIS)

    Hwang, Yong Soo

    2010-01-01

    For the purpose of evaluating annual individual doses from a potential repository disposing of radioactive wastes from the operation of the prospective advanced nuclear fuel cycle facilities in Korea, the new safety assessment code based on the Goldsim has been developed. It was designed to compare the environmental impacts from many fuel cycle options such as direct disposal, wet and dry recycling. The code based on the compartment theory can be applied to assess both normal and what if scenarios

  17. Sodium/water pool-deposit bed model of the CONACS code

    International Nuclear Information System (INIS)

    Peak, R.D.

    1983-01-01

    A new Pool-Bed model of the CONACS (Containment Analysis Code System) code represents a major advance over the pool models of other containment analysis code (NABE code of France, CEDAN code of Japan and CACECO and CONTAIN codes of the United States). This new model advances pool-bed modeling because of the number of significant materials and processes which are included with appropriate rigor. This CONACS pool-bed model maintains material balances for eight chemical species (C, H 2 O, Na, NaH, Na 2 O, Na 2 O 2 , Na 2 CO 3 and NaOH) that collect in the stationary liquid pool on the floor and in the desposit bed on the elevated shelf of the standard CONACS analysis cell

  18. Film grain noise modeling in advanced video coding

    Science.gov (United States)

    Oh, Byung Tae; Kuo, C.-C. Jay; Sun, Shijun; Lei, Shawmin

    2007-01-01

    A new technique for film grain noise extraction, modeling and synthesis is proposed and applied to the coding of high definition video in this work. The film grain noise is viewed as a part of artistic presentation by people in the movie industry. On one hand, since the film grain noise can boost the natural appearance of pictures in high definition video, it should be preserved in high-fidelity video processing systems. On the other hand, video coding with film grain noise is expensive. It is desirable to extract film grain noise from the input video as a pre-processing step at the encoder and re-synthesize the film grain noise and add it back to the decoded video as a post-processing step at the decoder. Under this framework, the coding gain of the denoised video is higher while the quality of the final reconstructed video can still be well preserved. Following this idea, we present a method to remove film grain noise from image/video without distorting its original content. Besides, we describe a parametric model containing a small set of parameters to represent the extracted film grain noise. The proposed model generates the film grain noise that is close to the real one in terms of power spectral density and cross-channel spectral correlation. Experimental results are shown to demonstrate the efficiency of the proposed scheme.

  19. Changes in the Coding and Non-coding Transcriptome and DNA Methylome that Define the Schwann Cell Repair Phenotype after Nerve Injury.

    Science.gov (United States)

    Arthur-Farraj, Peter J; Morgan, Claire C; Adamowicz, Martyna; Gomez-Sanchez, Jose A; Fazal, Shaline V; Beucher, Anthony; Razzaghi, Bonnie; Mirsky, Rhona; Jessen, Kristjan R; Aitman, Timothy J

    2017-09-12

    Repair Schwann cells play a critical role in orchestrating nerve repair after injury, but the cellular and molecular processes that generate them are poorly understood. Here, we perform a combined whole-genome, coding and non-coding RNA and CpG methylation study following nerve injury. We show that genes involved in the epithelial-mesenchymal transition are enriched in repair cells, and we identify several long non-coding RNAs in Schwann cells. We demonstrate that the AP-1 transcription factor C-JUN regulates the expression of certain micro RNAs in repair Schwann cells, in particular miR-21 and miR-34. Surprisingly, unlike during development, changes in CpG methylation are limited in injury, restricted to specific locations, such as enhancer regions of Schwann cell-specific genes (e.g., Nedd4l), and close to local enrichment of AP-1 motifs. These genetic and epigenomic changes broaden our mechanistic understanding of the formation of repair Schwann cell during peripheral nervous system tissue repair. Copyright © 2017 The Authors. Published by Elsevier Inc. All rights reserved.

  20. Application of mechanistic models to fermentation and biocatalysis for next-generation processes

    DEFF Research Database (Denmark)

    Gernaey, Krist; Eliasson Lantz, Anna; Tufvesson, Pär

    2010-01-01

    of variables required for measurement, control and process design. In the near future, mechanistic models with a higher degree of detail will play key roles in the development of efficient next-generation fermentation and biocatalytic processes. Moreover, mechanistic models will be used increasingly......Mechanistic models are based on deterministic principles, and recently, interest in them has grown substantially. Herein we present an overview of mechanistic models and their applications in biotechnology, including future perspectives. Model utility is highlighted with respect to selection...

  1. ASME Code Efforts Supporting HTGRs

    Energy Technology Data Exchange (ETDEWEB)

    D.K. Morton

    2012-09-01

    In 1999, an international collaborative initiative for the development of advanced (Generation IV) reactors was started. The idea behind this effort was to bring nuclear energy closer to the needs of sustainability, to increase proliferation resistance, and to support concepts able to produce energy (both electricity and process heat) at competitive costs. The U.S. Department of Energy has supported this effort by pursuing the development of the Next Generation Nuclear Plant, a high temperature gas-cooled reactor. This support has included research and development of pertinent data, initial regulatory discussions, and engineering support of various codes and standards development. This report discusses the various applicable American Society of Mechanical Engineers (ASME) codes and standards that are being developed to support these high temperature gascooled reactors during construction and operation. ASME is aggressively pursuing these codes and standards to support an international effort to build the next generation of advanced reactors so that all can benefit.

  2. ASME Code Efforts Supporting HTGRs

    Energy Technology Data Exchange (ETDEWEB)

    D.K. Morton

    2011-09-01

    In 1999, an international collaborative initiative for the development of advanced (Generation IV) reactors was started. The idea behind this effort was to bring nuclear energy closer to the needs of sustainability, to increase proliferation resistance, and to support concepts able to produce energy (both electricity and process heat) at competitive costs. The U.S. Department of Energy has supported this effort by pursuing the development of the Next Generation Nuclear Plant, a high temperature gas-cooled reactor. This support has included research and development of pertinent data, initial regulatory discussions, and engineering support of various codes and standards development. This report discusses the various applicable American Society of Mechanical Engineers (ASME) codes and standards that are being developed to support these high temperature gascooled reactors during construction and operation. ASME is aggressively pursuing these codes and standards to support an international effort to build the next generation of advanced reactors so that all can benefit.

  3. Advancing Kohlberg through Codes: Using Professional Codes To Reach the Moral Reasoning Objective in Undergraduate Ethics Courses.

    Science.gov (United States)

    Whitehouse, Ginny; Ingram, Michael T.

    The development of moral reasoning as a key course objective in undergraduate communication ethics classes can be accomplished by the critical and deliberate introduction of professional codes of ethics and the internalization of values found in those codes. Notably, "fostering moral reasoning skills" and "surveying current ethical…

  4. Gap Conductance model Validation in the TASS/SMR-S code using MARS code

    International Nuclear Information System (INIS)

    Ahn, Sang Jun; Yang, Soo Hyung; Chung, Young Jong; Lee, Won Jae

    2010-01-01

    Korea Atomic Energy Research Institute (KAERI) has been developing the TASS/SMR-S (Transient and Setpoint Simulation/Small and Medium Reactor) code, which is a thermal hydraulic code for the safety analysis of the advanced integral reactor. An appropriate work to validate the applicability of the thermal hydraulic models within the code should be demanded. Among the models, the gap conductance model which is describes the thermal gap conductivity between fuel and cladding was validated through the comparison with MARS code. The validation of the gap conductance model was performed by evaluating the variation of the gap temperature and gap width as the changed with the power fraction. In this paper, a brief description of the gap conductance model in the TASS/SMR-S code is presented. In addition, calculated results to validate the gap conductance model are demonstrated by comparing with the results of the MARS code with the test case

  5. Evaluation of Advanced Thermohydraulic System Codes for Design and Safety Analysis of Integral Type Reactors

    International Nuclear Information System (INIS)

    2014-02-01

    The integral pressurized water reactor (PWR) concept, which incorporates the nuclear steam supply systems within the reactor vessel, is one of the innovative reactor types with high potential for near term deployment. An International Collaborative Standard Problem (ICSP) on Integral PWR Design, Natural Circulation Flow Stability and Thermohydraulic Coupling of Primary System and Containment during Accidents was established in 2010. Oregon State University, which made available the use of its experimental facility built to demonstrate the feasibility of the Multi-application Small Light Water Reactor (MASLWR) design, and sixteen institutes from seven Member States participated in this ICSP. The objective of the ICSP is to assess computer codes for reactor system design and safety analysis. This objective is achieved through the production of experimental data and computer code simulation of experiments. A loss of feedwater transient with subsequent automatic depressurization system blowdown and long term cooling was selected as the reference event since many different modes of natural circulation phenomena, including the coupling of primary system, high pressure containment and cooling pool are expected to occur during this transient. The power maneuvering transient is also tested to examine the stability of natural circulation during the single and two phase conditions. The ICSP was conducted in three phases: pre-test (with designed initial and boundary conditions established before the experiment was conducted), blind (with real initial and boundary conditions after the experiment was conducted) and open simulation (after the observation of real experimental data). Most advanced thermohydraulic system analysis codes such as TRACE, RELAPS and MARS have been assessed against experiments conducted at the MASLWR test facility. The ICSP has provided all participants with the opportunity to evaluate the strengths and weaknesses of their system codes in the transient

  6. Numerical simulation code for combustion of sodium liquid droplet and its verification

    International Nuclear Information System (INIS)

    Okano, Yasushi

    1997-11-01

    The computer programs for sodium leak and burning phenomena had been developed based on mechanistic approach. Direct numerical simulation code for sodium liquid droplet burning had been developed for numerical analysis of droplet combustion in forced convection air flow. Distributions of heat generation and temperature and reaction rate of chemical productions, such as sodium oxide and hydroxide, are calculated and evaluated with using this numerical code. Extended MAC method coupled with a higher-order upwind scheme had been used for combustion simulation of methane-air mixture. In the numerical simulation code for combustion of sodium liquid droplet, chemical reaction model of sodium was connected with the extended MAC method. Combustion of single sodium liquid droplet was simulated in this report for the verification of developed numerical simulation code. The changes of burning rate and reaction product with droplet diameter and inlet wind velocity were investigated. These calculation results were qualitatively and quantitatively conformed to the experimental and calculation observations in combustion engineering. It was confirmed that the numerical simulation code was available for the calculation of sodium liquid droplet burning. (author)

  7. A Mode Propagation Database Suitable for Code Validation Utilizing the NASA Glenn Advanced Noise Control Fan and Artificial Sources

    Science.gov (United States)

    Sutliff, Daniel L.

    2014-01-01

    The NASA Glenn Research Center's Advanced Noise Control Fan (ANCF) was developed in the early 1990s to provide a convenient test bed to measure and understand fan-generated acoustics, duct propagation, and radiation to the farfield. A series of tests were performed primarily for the use of code validation and tool validation. Rotating Rake mode measurements were acquired for parametric sets of: (i) mode blockage, (ii) liner insertion loss, (iii) short ducts, and (iv) mode reflection.

  8. Validation of thermal hydraulic computer codes for advanced light water reactor

    International Nuclear Information System (INIS)

    Macek, J.

    2001-01-01

    The Czech Republic operates 4 WWER-440 units, two WWER-1000 units are being finalised (one of them is undergoing commissioning). Thermal-hydraulics Department of the Nuclear Research Institute Rez performs accident analyses for these plants using a number of computer codes. To model the primary and secondary circuits behaviour the system codes ATHLET, CATHARE, RELAP, TRAC are applied. Containment and pressure-suppressure system are modelled with RALOC and MELCOR codes, the reactor power calculations (point and space-neutron kinetics) are made with DYN3D, NESTLE and CDF codes (FLUENT, TRIO) are used for some specific problems. An integral part of the current Czech project 'New Energy Sources' is selection of a new nuclear source. Within this and the preceding projects financed by the Czech Ministry of Industry and Trade and the EU PHARE, the Department carries and has carried out the systematic validation of thermal-hydraulic and reactor physics computer codes applying data obtained on several experimental facilities as well as the real operational data. The paper provides a concise information on these activities of the NRI and its Thermal-hydraulics Department. A detailed example of the system code validation and the consequent utilisation of the results for a real NPP purposes is included. (author)

  9. Advanced Code for Photocathode Design

    Energy Technology Data Exchange (ETDEWEB)

    Ives, Robert Lawrence [Calabazas Creek Research, Inc., San Mateo, CA (United States); Jensen, Kevin [Naval Research Lab. (NRL), Washington, DC (United States); Montgomery, Eric [Univ. of Maryland, College Park, MD (United States); Bui, Thuc [Calabazas Creek Research, Inc., San Mateo, CA (United States)

    2015-12-15

    The Phase I activity demonstrated that PhotoQE could be upgraded and modified to allow input using a graphical user interface. Specific calls to platform-dependent (e.g. IMSL) function calls were removed, and Fortran77 components were rewritten for Fortran95 compliance. The subroutines, specifically the common block structures and shared data parameters, were reworked to allow the GUI to update material parameter data, and the system was targeted for desktop personal computer operation. The new structures overcomes the previous rigid and unmodifiable library structures by implementing new, materials library data sets and repositioning the library values to external files. Material data may originate from published literature or experimental measurements. Further optimization and restructuring would allow custom and specific emission models for beam codes that rely on parameterized photoemission algorithms. These would be based on simplified and parametric representations updated and extended from previous versions (e.g., Modified Fowler-Dubridge, Modified Three-Step, etc.).

  10. SARNET, a success story. Survey of major achievements on severe accidents and of knowledge capitalization within the ASTEC code

    International Nuclear Information System (INIS)

    Albiol, T.; Van Dorsselaere, J.P.; Reinke, N.

    2013-01-01

    51 organizations from Europe and Canada cooperated within SARNET (Severe Accident Research Network of Excellence) joining their capacities of research in order to resolve the most important pending issues for enhancing, in regard to Severe Accidents (SA), the safety of existing and future Nuclear Power Plants (NPPs). SARNET defines common research programmes and develops common computer codes and methodologies for safety assessment. The ASTEC integral code, jointly developed by IRSN (France) and GRS (Germany) for Light Water Reactor (LWR) source term SA evaluation, Probabilistic Safety Assessment (PSA) level-2 studies and SA management evaluation, is the main integrating component of SARNET. The scientific knowledge generated in the Corium, Source Term and Containment Topics has been integrated into the code through improved or new physical models. ASTEC constitutes now the reference European SA integral code. During the 4 and half years of SARNET, 30 partners have assessed the successive versions of the ASTEC V1 code through validation. More than 60 scientists have been trained on the code use. Validation tasks on about 65 experiments were performed to cover all physical phenomena occurring in a severe accident: circuit thermalhydraulic, core degradation, fission products (FP) release and transport, Molten-Corium-Concrete-Interaction (MCCI), and in the containment, thermalhydraulic, aerosol and iodine as well as hydrogen behaviour. The overall status of validation can be considered as good, with results often close to results of mechanistic codes. Some reach the limits of present knowledge, for instance on MCCI, and, like in most codes, an adequate model for reflooding of a degraded core is still missing. IRSN and GRS are currently preparing the new series of ASTEC V2 versions that will account for most of the needs of evolution expressed by the SARNET partners. The first version V2.0, planned for March 09, will be applicable to EPR and will include the ICARE2

  11. International benchmark study of advanced thermal hydraulic safety analysis codes against measurements on IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hainoun, A., E-mail: pscientific2@aec.org.sy [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Doval, A. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400 S.C de Bariloche, Rio Negro (Argentina); Umbehaun, P. [Centro de Engenharia Nuclear – CEN, IPEN-CNEN/SP, Av. Lineu Prestes 2242-Cidade Universitaria, CEP-05508-000 São Paulo, SP (Brazil); Chatzidakis, S. [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States); Ghazi, N. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Park, S. [Research Reactor Design and Engineering Division, Basic Science Project Operation Dept., Korea Atomic Energy Research Institute (Korea, Republic of); Mladin, M. [Institute for Nuclear Research, Campului Street No. 1, P.O. Box 78, 115400 Mioveni, Arges (Romania); Shokr, A. [Division of Nuclear Installation Safety, Research Reactor Safety Section, International Atomic Energy Agency, A-1400 Vienna (Austria)

    2014-12-15

    Highlights: • A set of advanced system thermal hydraulic codes are benchmarked against IFA of IEA-R1. • Comparative safety analysis of IEA-R1 reactor during LOFA by 7 working teams. • This work covers both experimental and calculation effort and presents new out findings on TH of RR that have not been reported before. • LOFA results discrepancies from 7% to 20% for coolant and peak clad temperatures are predicted conservatively. - Abstract: In the framework of the IAEA Coordination Research Project on “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal hydraulic computational methods and tools for operation and safety analysis of research reactors” the Brazilian research reactor IEA-R1 has been selected as reference facility to perform benchmark calculations for a set of thermal hydraulic codes being widely used by international teams in the field of research reactor (RR) deterministic safety analysis. The goal of the conducted benchmark is to demonstrate the application of innovative reactor analysis tools in the research reactor community, validation of the applied codes and application of the validated codes to perform comprehensive safety analysis of RR. The IEA-R1 is equipped with an Instrumented Fuel Assembly (IFA) which provided measurements for normal operation and loss of flow transient. The measurements comprised coolant and cladding temperatures, reactor power and flow rate. Temperatures are measured at three different radial and axial positions of IFA summing up to 12 measuring points in addition to the coolant inlet and outlet temperatures. The considered benchmark deals with the loss of reactor flow and the subsequent flow reversal from downward forced to upward natural circulation and presents therefore relevant phenomena for the RR safety analysis. The benchmark calculations were performed independently by the participating teams using different thermal hydraulic and safety

  12. Analysis of selected Halden overpressure tests using the FALCON code

    Energy Technology Data Exchange (ETDEWEB)

    Khvostov, G., E-mail: grigori.khvostov@psi.ch [Paul Scherrer Institut, CH 5232 Villigen PSI (Switzerland); Wiesenack, W. [Institute for Energy Technology – OECD Halden Reactor Project, P.O. Box 173, N-1751 Halden (Norway)

    2016-12-15

    Highlights: • We analyse four Halden overpressure tests. • We determine a critical overpressure value for lift-off in a BWR fuel sample. • We show the role of bonding in over-pressurized rod behaviour. • We analytically quantify the degree of bonding via its impact on cladding elongation. • We hypothesize on an effect of circumferential cracks on thermal fuel response to overpressure. • We estimate a thermal effect of circumferential cracks based on interpretation of the data. - Abstract: Four Halden overpressure (lift-off) tests using samples with uranium dioxide fuel pre-irradiated in power reactors to a burnup of 60 MWd/kgU are analyzed. The FALCON code coupled to a mechanistic model, GRSW-A for fission gas release and gaseous-bubble swelling is used for the calculation. The advanced version of the FALCON code is shown to be applicable to best-estimate predictive analysis of overpressure tests using rods without, or weak pellet-cladding bonding, as well as scoping analysis of tests with fuels where stronger pellet-cladding bonding occurs. Significant effects of bonding and fuel cracking/relocation on the thermal and mechanical behaviour of highly over-pressurized rods are shown. The effect of bonding is particularly pronounced in the tests with the PWR samples. The present findings are basically consistent with an earlier analysis based on a direct interpretation of the experimental data. Additionally, in this paper, the specific effects are quantified based on the comparison of the data with the results of calculation. It is concluded that the identified effects are largely beyond the current traditional fuel-rod licensing analysis methods.

  13. Speech and audio processing for coding, enhancement and recognition

    CERN Document Server

    Togneri, Roberto; Narasimha, Madihally

    2015-01-01

    This book describes the basic principles underlying the generation, coding, transmission and enhancement of speech and audio signals, including advanced statistical and machine learning techniques for speech and speaker recognition with an overview of the key innovations in these areas. Key research undertaken in speech coding, speech enhancement, speech recognition, emotion recognition and speaker diarization are also presented, along with recent advances and new paradigms in these areas. ·         Offers readers a single-source reference on the significant applications of speech and audio processing to speech coding, speech enhancement and speech/speaker recognition. Enables readers involved in algorithm development and implementation issues for speech coding to understand the historical development and future challenges in speech coding research; ·         Discusses speech coding methods yielding bit-streams that are multi-rate and scalable for Voice-over-IP (VoIP) Networks; ·     �...

  14. NESTLE: A nodal kinetics code

    International Nuclear Information System (INIS)

    Al-Chalabi, R.M.; Turinsky, P.J.; Faure, F.-X.; Sarsour, H.N.; Engrand, P.R.

    1993-01-01

    The NESTLE nodal kinetics code has been developed for utilization as a stand-alone code for steady-state and transient reactor neutronic analysis and for incorporation into system transient codes, such as TRAC and RELAP. The latter is desirable to increase the simulation fidelity over that obtained from currently employed zero- and one-dimensional neutronic models and now feasible due to advances in computer performance and efficiency of nodal methods. As a stand-alone code, requirements are that it operate on a range of computing platforms from memory-limited personal computers (PCs) to supercomputers with vector processors. This paper summarizes the features of NESTLE that reflect the utilization and requirements just noted

  15. On the development of LWR fuel analysis code (1). Analysis of the FEMAXI code and proposal of a new model

    International Nuclear Information System (INIS)

    Lemehov, Sergei; Suzuki, Motoe

    2000-01-01

    This report summarizes the review on the modeling features of FEMAXI code and proposal of a new theoretical equation model of clad creep on the basis of irradiation-induced microstructure change. It was pointed out that plutonium build-up in fuel matrix and non-uniform radial power profile at high burn-up affect significantly fuel behavior through the interconnected effects with such phenomena as clad irradiation-induced creep, fission gas release, fuel thermal conductivity degradation, rim porous band formation and associated fuel swelling. Therefore, these combined effects should be properly incorporated into the models of the FEMAXI code so that the code can carry out numerical analysis at the level of accuracy and elaboration that modern experimental data obtained in test reactors have. Also, the proposed new mechanistic clad creep model has a general formalism which allows the model to be flexibly applied for clad behavior analysis under normal operation conditions and power transients as well for Zr-based clad materials by the use of established out-of-pile mechanical properties. The model has been tested against experimental data, while further verification is needed with specific emphasis on power ramps and transients. (author)

  16. MHD code using multi graphical processing units: SMAUG+

    Science.gov (United States)

    Gyenge, N.; Griffiths, M. K.; Erdélyi, R.

    2018-01-01

    This paper introduces the Sheffield Magnetohydrodynamics Algorithm Using GPUs (SMAUG+), an advanced numerical code for solving magnetohydrodynamic (MHD) problems, using multi-GPU systems. Multi-GPU systems facilitate the development of accelerated codes and enable us to investigate larger model sizes and/or more detailed computational domain resolutions. This is a significant advancement over the parent single-GPU MHD code, SMAUG (Griffiths et al., 2015). Here, we demonstrate the validity of the SMAUG + code, describe the parallelisation techniques and investigate performance benchmarks. The initial configuration of the Orszag-Tang vortex simulations are distributed among 4, 16, 64 and 100 GPUs. Furthermore, different simulation box resolutions are applied: 1000 × 1000, 2044 × 2044, 4000 × 4000 and 8000 × 8000 . We also tested the code with the Brio-Wu shock tube simulations with model size of 800 employing up to 10 GPUs. Based on the test results, we observed speed ups and slow downs, depending on the granularity and the communication overhead of certain parallel tasks. The main aim of the code development is to provide massively parallel code without the memory limitation of a single GPU. By using our code, the applied model size could be significantly increased. We demonstrate that we are able to successfully compute numerically valid and large 2D MHD problems.

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

    International Nuclear Information System (INIS)

    Baratta, A.J.

    1997-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Baratta, A.J.

    1997-07-01

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

  19. Mechanistic Insight into the Degradation of Nitrosamines via Aqueous-Phase UV Photolysis or a UV-Based Advanced Oxidation Process: Quantum Mechanical Calculations.

    Science.gov (United States)

    Minakata, Daisuke; Coscarelli, Erica

    2018-02-28

    Nitrosamines are a group of carcinogenic chemicals that are present in aquatic environments that result from byproducts of industrial processes and disinfection products. As indirect and direct potable reuse increase, the presence of trace nitrosamines presents challenges to water infrastructures that incorporate effluent from wastewater treatment. Ultraviolet (UV) photolysis or UV-based advanced oxidation processes that produce highly reactive hydroxyl radicals are promising technologies to remove nitrosamines from water. However, complex reaction mechanisms involving radicals limit our understandings of the elementary reaction pathways embedded in the overall reactions identified experimentally. In this study, we perform quantum mechanical calculations to identify the hydroxyl radical-induced initial elementary reactions with N -nitrosodimethylamine (NDMA), N -nitrosomethylethylamine, and N -nitrosomethylbutylamine. We also investigate the UV-induced NDMA degradation mechanisms. Our calculations reveal that the alkyl side chains of nitrosamine affect the reaction mechanism of hydroxyl radicals with each nitrosamine investigated in this study. Nitrosamines with one- or two-carbon alkyl chains caused the delocalization of the electron density, leading to slower subsequent degradation. Additionally, three major initial elementary reactions and the subsequent radical-involved reaction pathways are identified in the UV-induced NDMA degradation process. This study provides mechanistic insight into the elementary reaction pathways, and a future study will combine these results with the kinetic information to predict the time-dependent concentration profiles of nitrosamines and their transformation products.

  20. Mechanistic Insight into the Degradation of Nitrosamines via Aqueous-Phase UV Photolysis or a UV-Based Advanced Oxidation Process: Quantum Mechanical Calculations

    Directory of Open Access Journals (Sweden)

    Daisuke Minakata

    2018-02-01

    Full Text Available Nitrosamines are a group of carcinogenic chemicals that are present in aquatic environments that result from byproducts of industrial processes and disinfection products. As indirect and direct potable reuse increase, the presence of trace nitrosamines presents challenges to water infrastructures that incorporate effluent from wastewater treatment. Ultraviolet (UV photolysis or UV-based advanced oxidation processes that produce highly reactive hydroxyl radicals are promising technologies to remove nitrosamines from water. However, complex reaction mechanisms involving radicals limit our understandings of the elementary reaction pathways embedded in the overall reactions identified experimentally. In this study, we perform quantum mechanical calculations to identify the hydroxyl radical-induced initial elementary reactions with N-nitrosodimethylamine (NDMA, N-nitrosomethylethylamine, and N-nitrosomethylbutylamine. We also investigate the UV-induced NDMA degradation mechanisms. Our calculations reveal that the alkyl side chains of nitrosamine affect the reaction mechanism of hydroxyl radicals with each nitrosamine investigated in this study. Nitrosamines with one- or two-carbon alkyl chains caused the delocalization of the electron density, leading to slower subsequent degradation. Additionally, three major initial elementary reactions and the subsequent radical-involved reaction pathways are identified in the UV-induced NDMA degradation process. This study provides mechanistic insight into the elementary reaction pathways, and a future study will combine these results with the kinetic information to predict the time-dependent concentration profiles of nitrosamines and their transformation products.

  1. Advanced Wall Boiling Model with Wide Range Applicability for the Subcooled Boiling Flow and its Application into the CFD Code

    International Nuclear Information System (INIS)

    Yun, B. J.; Song, C. H.; Splawski, A.; Lo, S.

    2010-01-01

    Subcooled boiling is one of the crucial phenomena for the design, operation and safety analysis of a nuclear power plant. It occurs due to the thermally nonequilibrium state in the two-phase heat transfer system. Many complicated phenomena such as a bubble generation, a bubble departure, a bubble growth, and a bubble condensation are created by this thermally nonequilibrium condition in the subcooled boiling flow. However, it has been revealed that most of the existing best estimate safety analysis codes have a weakness in the prediction of the subcooled boiling phenomena in which multi-dimensional flow behavior is dominant. In recent years, many investigators are trying to apply CFD (Computational Fluid Dynamics) codes for an accurate prediction of the subcooled boiling flow. In the CFD codes, evaporation heat flux from heated wall is one of the key parameters to be modeled for an accurate prediction of the subcooled boiling flow. The evaporate heat flux for the CFD codes is expressed typically as follows, q' e = πD 3 d /6 ρ g h fg fN' where, D d , f ,N' are bubble departure size, bubble departure frequency and active nucleation site density, respectively. In the most of the commercial CFD codes, Tolubinsky bubble departure size model, Kurul and Podowski active nucleation site density model and Ceumem-Lindenstjerna bubble departure frequency model are adopted as a basic wall boiling model. However, these models do not consider their dependency on the flow, pressure and fluid type. In this paper, an advanced wall boiling model was proposed in order to improve subcooled boiling model for the CFD codes

  2. Toward a Rational and Mechanistic Account of Mental Effort.

    Science.gov (United States)

    Shenhav, Amitai; Musslick, Sebastian; Lieder, Falk; Kool, Wouter; Griffiths, Thomas L; Cohen, Jonathan D; Botvinick, Matthew M

    2017-07-25

    In spite of its familiar phenomenology, the mechanistic basis for mental effort remains poorly understood. Although most researchers agree that mental effort is aversive and stems from limitations in our capacity to exercise cognitive control, it is unclear what gives rise to those limitations and why they result in an experience of control as costly. The presence of these control costs also raises further questions regarding how best to allocate mental effort to minimize those costs and maximize the attendant benefits. This review explores recent advances in computational modeling and empirical research aimed at addressing these questions at the level of psychological process and neural mechanism, examining both the limitations to mental effort exertion and how we manage those limited cognitive resources. We conclude by identifying remaining challenges for theoretical accounts of mental effort as well as possible applications of the available findings to understanding the causes of and potential solutions for apparent failures to exert the mental effort required of us.

  3. SITE-94. Adaptation of mechanistic sorption models for performance assessment calculations

    International Nuclear Information System (INIS)

    Arthur, R.C.

    1996-10-01

    Sorption is considered in most predictive models of radionuclide transport in geologic systems. Most models simulate the effects of sorption in terms of empirical parameters, which however can be criticized because the data are only strictly valid under the experimental conditions at which they were measured. An alternative is to adopt a more mechanistic modeling framework based on recent advances in understanding the electrical properties of oxide mineral-water interfaces. It has recently been proposed that these 'surface-complexation' models may be directly applicable to natural systems. A possible approach for adapting mechanistic sorption models for use in performance assessments, using this 'surface-film' concept, is described in this report. Surface-acidity parameters in the Generalized Two-Layer surface complexation model are combined with surface-complexation constants for Np(V) sorption ob hydrous ferric oxide to derive an analytical model enabling direct calculation of corresponding intrinsic distribution coefficients as a function of pH, and Ca 2+ , Cl - , and HCO 3 - concentrations. The surface film concept is then used to calculate whole-rock distribution coefficients for Np(V) sorption by altered granitic rocks coexisting with a hypothetical, oxidized Aespoe groundwater. The calculated results suggest that the distribution coefficients for Np adsorption on these rocks could range from 10 to 100 ml/g. Independent estimates of K d for Np sorption in similar systems, based on an extensive review of experimental data, are consistent, though slightly conservative, with respect to the calculated values. 31 refs

  4. Toward a Mechanistic Source Term in Advanced Reactors: Characterization of Radionuclide Transport and Retention in a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Brunett, Acacia J.; Bucknor, Matthew; Grabaskas, David

    2016-04-17

    A vital component of the U.S. reactor licensing process is an integrated safety analysis in which a source term representing the release of radionuclides during normal operation and accident sequences is analyzed. Historically, source term analyses have utilized bounding, deterministic assumptions regarding radionuclide release. However, advancements in technical capabilities and the knowledge state have enabled the development of more realistic and best-estimate retention and release models such that a mechanistic source term assessment can be expected to be a required component of future licensing of advanced reactors. Recently, as part of a Regulatory Technology Development Plan effort for sodium cooled fast reactors (SFRs), Argonne National Laboratory has investigated the current state of knowledge of potential source terms in an SFR via an extensive review of previous domestic experiments, accidents, and operation. As part of this work, the significant sources and transport processes of radionuclides in an SFR have been identified and characterized. This effort examines all stages of release and source term evolution, beginning with release from the fuel pin and ending with retention in containment. Radionuclide sources considered in this effort include releases originating both in-vessel (e.g. in-core fuel, primary sodium, cover gas cleanup system, etc.) and ex-vessel (e.g. spent fuel storage, handling, and movement). Releases resulting from a primary sodium fire are also considered as a potential source. For each release group, dominant transport phenomena are identified and qualitatively discussed. The key product of this effort was the development of concise, inclusive diagrams that illustrate the release and retention mechanisms at a high level, where unique schematics have been developed for in-vessel, ex-vessel and sodium fire releases. This review effort has also found that despite the substantial range of phenomena affecting radionuclide release, the

  5. Crucial steps to life: From chemical reactions to code using agents.

    Science.gov (United States)

    Witzany, Guenther

    2016-02-01

    The concepts of the origin of the genetic code and the definitions of life changed dramatically after the RNA world hypothesis. Main narratives in molecular biology and genetics such as the "central dogma," "one gene one protein" and "non-coding DNA is junk" were falsified meanwhile. RNA moved from the transition intermediate molecule into centre stage. Additionally the abundance of empirical data concerning non-random genetic change operators such as the variety of mobile genetic elements, persistent viruses and defectives do not fit with the dominant narrative of error replication events (mutations) as being the main driving forces creating genetic novelty and diversity. The reductionistic and mechanistic views on physico-chemical properties of the genetic code are no longer convincing as appropriate descriptions of the abundance of non-random genetic content operators which are active in natural genetic engineering and natural genome editing. Copyright © 2015 Elsevier Ireland Ltd. All rights reserved.

  6. LASSIM-A network inference toolbox for genome-wide mechanistic modeling.

    Directory of Open Access Journals (Sweden)

    Rasmus Magnusson

    2017-06-01

    Full Text Available Recent technological advancements have made time-resolved, quantitative, multi-omics data available for many model systems, which could be integrated for systems pharmacokinetic use. Here, we present large-scale simulation modeling (LASSIM, which is a novel mathematical tool for performing large-scale inference using mechanistically defined ordinary differential equations (ODE for gene regulatory networks (GRNs. LASSIM integrates structural knowledge about regulatory interactions and non-linear equations with multiple steady state and dynamic response expression datasets. The rationale behind LASSIM is that biological GRNs can be simplified using a limited subset of core genes that are assumed to regulate all other gene transcription events in the network. The LASSIM method is implemented as a general-purpose toolbox using the PyGMO Python package to make the most of multicore computers and high performance clusters, and is available at https://gitlab.com/Gustafsson-lab/lassim. As a method, LASSIM works in two steps, where it first infers a non-linear ODE system of the pre-specified core gene expression. Second, LASSIM in parallel optimizes the parameters that model the regulation of peripheral genes by core system genes. We showed the usefulness of this method by applying LASSIM to infer a large-scale non-linear model of naïve Th2 cell differentiation, made possible by integrating Th2 specific bindings, time-series together with six public and six novel siRNA-mediated knock-down experiments. ChIP-seq showed significant overlap for all tested transcription factors. Next, we performed novel time-series measurements of total T-cells during differentiation towards Th2 and verified that our LASSIM model could monitor those data significantly better than comparable models that used the same Th2 bindings. In summary, the LASSIM toolbox opens the door to a new type of model-based data analysis that combines the strengths of reliable mechanistic models

  7. Mechanistic model for Sr and Ba release from severely damaged fuel

    International Nuclear Information System (INIS)

    Rest, J.; Cronenberg, A.W.

    1985-11-01

    Among radionuclides associated with fission product release during severe accidents, the primary ones with health consequences are the volatile species of I, Te, and Cs, and the next most important are Sr, Ba, and Ru. Considerable progress has been made in the mechanistic understanding of I, Cs, Te, and noble gas release; however, no capability presently exists for estimating the release of Sr, Ba, and Ru. This paper presents a description of the primary physical/chemical models recently incorporated into the FASTGRASS-VFP (volatile fission product) code for the estimation of Sr and Ba release. FASTGRASS-VFP release predictions are compared with two data sets: (1) data from out-of-reactor induction-heating experiments on declad low-burnup (1000 and 4000 MWd/t) pellets, and (2) data from the more recent in-reactor PBF Severe Fuel Damage Tests, in which one-meter-long, trace-irradiated (89 MWd/t) and normally irradiated (approx.35,000 MWd/t) fuel rods were tested under accident conditions. 10 refs

  8. "Ratio via Machina": Three Standards of Mechanistic Explanation in Sociology

    Science.gov (United States)

    Aviles, Natalie B.; Reed, Isaac Ariail

    2017-01-01

    Recently, sociologists have expended much effort in attempts to define social mechanisms. We intervene in these debates by proposing that sociologists in fact have a choice to make between three standards of what constitutes a good mechanistic explanation: substantial, formal, and metaphorical mechanistic explanation. All three standards are…

  9. Toward mechanistic classification of enzyme functions.

    Science.gov (United States)

    Almonacid, Daniel E; Babbitt, Patricia C

    2011-06-01

    Classification of enzyme function should be quantitative, computationally accessible, and informed by sequences and structures to enable use of genomic information for functional inference and other applications. Large-scale studies have established that divergently evolved enzymes share conserved elements of structure and common mechanistic steps and that convergently evolved enzymes often converge to similar mechanisms too, suggesting that reaction mechanisms could be used to develop finer-grained functional descriptions than provided by the Enzyme Commission (EC) system currently in use. Here we describe how evolution informs these structure-function mappings and review the databases that store mechanisms of enzyme reactions along with recent developments to measure ligand and mechanistic similarities. Together, these provide a foundation for new classifications of enzyme function. Copyright © 2011 Elsevier Ltd. All rights reserved.

  10. European Validation of the Integral Code ASTEC (EVITA)

    International Nuclear Information System (INIS)

    Allelein, H.-J.; Neu, K.; Dorsselaere, J.P. Van

    2005-01-01

    The main objective of the European Validation of the Integral Code ASTEC (EVITA) project is to distribute the severe accident integral code ASTEC to European partners in order to apply the validation strategy issued from the VASA project (4th EC FWP). Partners evaluate the code capability through validation on reference experiments and plant applications accounting for severe accident management measures, and compare results with reference codes. The basis version V0 of ASTEC (Accident Source Term Evaluation Code)-commonly developed and basically validated by GRS and IRSN-was made available in late 2000 for the EVITA partners on their individual platforms. Users' training was performed by IRSN and GRS. The code portability on different computers was checked to be correct. A 'hot line' assistance was installed continuously available for EVITA code users. The actual version V1 has been released to the EVITA partners end of June 2002. It allows to simulate the front-end phase by two new modules:- for reactor coolant system 2-phase simplified thermal hydraulics (5-equation approach) during both front-end and core degradation phases; - for core degradation, based on structure and main models of ICARE2 (IRSN) reference mechanistic code for core degradation and on other simplified models. Next priorities are clearly identified: code consolidation in order to increase the robustness, extension of all plant applications beyond the vessel lower head failure and coupling with fission product modules, and continuous improvements of users' tools. As EVITA has very successfully made the first step into the intention to provide end-users (like utilities, vendors and licensing authorities) with a well validated European integral code for the simulation of severe accidents in NPPs, the EVITA partners strongly recommend to continue validation, benchmarking and application of ASTEC. This work will continue in Severe Accident Research Network (SARNET) in the 6th Framework Programme

  11. Improving the International Agency for Research on Cancer's consideration of mechanistic evidence

    International Nuclear Information System (INIS)

    Goodman, Julie; Lynch, Heather

    2017-01-01

    Background: The International Agency for Research on Cancer (IARC) recently developed a framework for evaluating mechanistic evidence that includes a list of 10 key characteristics of carcinogens. This framework is useful for identifying and organizing large bodies of literature on carcinogenic mechanisms, but it lacks sufficient guidance for conducting evaluations that fully integrate mechanistic evidence into hazard assessments. Objectives: We summarize the framework, and suggest approaches to strengthen the evaluation of mechanistic evidence using this framework. Discussion: While the framework is useful for organizing mechanistic evidence, its lack of guidance for implementation limits its utility for understanding human carcinogenic potential. Specifically, it does not include explicit guidance for evaluating the biological significance of mechanistic endpoints, inter- and intra-individual variability, or study quality and relevance. It also does not explicitly address how mechanistic evidence should be integrated with other realms of evidence. Because mechanistic evidence is critical to understanding human cancer hazards, we recommend that IARC develop transparent and systematic guidelines for the use of this framework so that mechanistic evidence will be evaluated and integrated in a robust manner, and concurrently with other realms of evidence, to reach a final human cancer hazard conclusion. Conclusions: IARC does not currently provide a standardized approach to evaluating mechanistic evidence. Incorporating the recommendations discussed here will make IARC analyses of mechanistic evidence more transparent, and lead to assessments of cancer hazards that reflect the weight of the scientific evidence and allow for scientifically defensible decision-making. - Highlights: • IARC has a revised framework for evaluating literature on carcinogenic mechanisms. • The framework is based on 10 key characteristics of carcinogens. • IARC should develop transparent

  12. Improving the International Agency for Research on Cancer's consideration of mechanistic evidence

    Energy Technology Data Exchange (ETDEWEB)

    Goodman, Julie, E-mail: jgoodman@gradientcorp.com; Lynch, Heather

    2017-03-15

    Background: The International Agency for Research on Cancer (IARC) recently developed a framework for evaluating mechanistic evidence that includes a list of 10 key characteristics of carcinogens. This framework is useful for identifying and organizing large bodies of literature on carcinogenic mechanisms, but it lacks sufficient guidance for conducting evaluations that fully integrate mechanistic evidence into hazard assessments. Objectives: We summarize the framework, and suggest approaches to strengthen the evaluation of mechanistic evidence using this framework. Discussion: While the framework is useful for organizing mechanistic evidence, its lack of guidance for implementation limits its utility for understanding human carcinogenic potential. Specifically, it does not include explicit guidance for evaluating the biological significance of mechanistic endpoints, inter- and intra-individual variability, or study quality and relevance. It also does not explicitly address how mechanistic evidence should be integrated with other realms of evidence. Because mechanistic evidence is critical to understanding human cancer hazards, we recommend that IARC develop transparent and systematic guidelines for the use of this framework so that mechanistic evidence will be evaluated and integrated in a robust manner, and concurrently with other realms of evidence, to reach a final human cancer hazard conclusion. Conclusions: IARC does not currently provide a standardized approach to evaluating mechanistic evidence. Incorporating the recommendations discussed here will make IARC analyses of mechanistic evidence more transparent, and lead to assessments of cancer hazards that reflect the weight of the scientific evidence and allow for scientifically defensible decision-making. - Highlights: • IARC has a revised framework for evaluating literature on carcinogenic mechanisms. • The framework is based on 10 key characteristics of carcinogens. • IARC should develop transparent

  13. Predicting interactions from mechanistic information: Can omic data validate theories?

    International Nuclear Information System (INIS)

    Borgert, Christopher J.

    2007-01-01

    To address the most pressing and relevant issues for improving mixture risk assessment, researchers must first recognize that risk assessment is driven by both regulatory requirements and scientific research, and that regulatory concerns may expand beyond the purely scientific interests of researchers. Concepts of 'mode of action' and 'mechanism of action' are used in particular ways within the regulatory arena, depending on the specific assessment goals. The data requirements for delineating a mode of action and predicting interactive toxicity in mixtures are not well defined from a scientific standpoint due largely to inherent difficulties in testing certain underlying assumptions. Understanding the regulatory perspective on mechanistic concepts will be important for designing experiments that can be interpreted clearly and applied in risk assessments without undue reliance on extrapolation and assumption. In like fashion, regulators and risk assessors can be better equipped to apply mechanistic data if the concepts underlying mechanistic research and the limitations that must be placed on interpretation of mechanistic data are understood. This will be critically important for applying new technologies to risk assessment, such as functional genomics, proteomics, and metabolomics. It will be essential not only for risk assessors to become conversant with the language and concepts of mechanistic research, including new omic technologies, but also, for researchers to become more intimately familiar with the challenges and needs of risk assessment

  14. Explanation and inference: Mechanistic and functional explanations guide property generalization

    Directory of Open Access Journals (Sweden)

    Tania eLombrozo

    2014-09-01

    Full Text Available The ability to generalize from the known to the unknown is central to learning and inference. Two experiments explore the relationship between how a property is explained and how that property is generalized to novel species and artifacts. The experiments contrast the consequences of explaining a property mechanistically, by appeal to parts and processes, with the consequences of explaining the property functionally, by appeal to functions and goals. The findings suggest that properties that are explained functionally are more likely to be generalized on the basis of shared functions, with a weaker relationship between mechanistic explanations and generalization on the basis of shared parts and processes. The influence of explanation type on generalization holds even though all participants are provided with the same mechanistic and functional information, and whether an explanation type is freely generated (Experiment 1, experimentally provided (Experiment 2, or experimentally induced (Experiment 2. The experiments also demonstrate that explanations and generalizations of a particular type (mechanistic or functional can be experimentally induced by providing sample explanations of that type, with a comparable effect when the sample explanations come from the same domain or from a different domains. These results suggest that explanations serve as a guide to generalization, and contribute to a growing body of work supporting the value of distinguishing mechanistic and functional explanations.

  15. Explanation and inference: mechanistic and functional explanations guide property generalization.

    Science.gov (United States)

    Lombrozo, Tania; Gwynne, Nicholas Z

    2014-01-01

    The ability to generalize from the known to the unknown is central to learning and inference. Two experiments explore the relationship between how a property is explained and how that property is generalized to novel species and artifacts. The experiments contrast the consequences of explaining a property mechanistically, by appeal to parts and processes, with the consequences of explaining the property functionally, by appeal to functions and goals. The findings suggest that properties that are explained functionally are more likely to be generalized on the basis of shared functions, with a weaker relationship between mechanistic explanations and generalization on the basis of shared parts and processes. The influence of explanation type on generalization holds even though all participants are provided with the same mechanistic and functional information, and whether an explanation type is freely generated (Experiment 1), experimentally provided (Experiment 2), or experimentally induced (Experiment 2). The experiments also demonstrate that explanations and generalizations of a particular type (mechanistic or functional) can be experimentally induced by providing sample explanations of that type, with a comparable effect when the sample explanations come from the same domain or from a different domains. These results suggest that explanations serve as a guide to generalization, and contribute to a growing body of work supporting the value of distinguishing mechanistic and functional explanations.

  16. Soil pH controls the environmental availability of phosphorus: Experimental and mechanistic modelling approaches

    International Nuclear Information System (INIS)

    Devau, Nicolas; Cadre, Edith Le; Hinsinger, Philippe; Jaillard, Benoit; Gerard, Frederic

    2009-01-01

    Inorganic P is the least mobile major nutrient in most soils and is frequently the prime limiting factor for plant growth in terrestrial ecosystems. In this study, the extraction of soil inorganic P with CaCl 2 (P-CaCl 2 ) and geochemical modelling were combined in order to unravel the processes controlling the environmentally available P (EAP) of a soil over a range of pH values (pH ∼ 4-10). Mechanistic descriptions of the adsorption of cations and anions by the soil constituents were used (1-pK Triple Plane, ion-exchange and NICA-Donnan models). These models are implemented into the geochemical code Visual MINTEQ. An additive approach was used for their application to the surface horizon of a Cambisol. The geochemical code accurately reproduced the concentration of extracted P at the different soil pH values (R 2 = 0.9, RMSE = 0.03 mg kg -1 ). Model parameters were either directly found in the literature or estimated by fitting published experimental results in single mineral systems. The strong agreement between measurements and modelling results demonstrated that adsorption processes exerted a major control on the EAP of the soil over a large range of pH values. An influence of the precipitation of P-containing mineral is discounted based on thermodynamic calculations. Modelling results indicated that the variations in P-CaCl 2 with soil pH were controlled by the deprotonation/protonation of the surface hydroxyl groups, the distribution of P surface complexes, and the adsorption of Ca and Cl from the electrolyte background. Iron-oxides and gibbsite were found to be the major P-adsorbing soil constituents at acidic and alkaline pHs, whereas P was mainly adsorbed by clay minerals at intermediate pH values. This study demonstrates the efficacy of geochemical modelling to understand soil processes, and the applicability of mechanistic adsorption models to a 'real' soil, with its mineralogical complexity and the additional contribution of soil organic matter.

  17. Soil pH controls the environmental availability of phosphorus: Experimental and mechanistic modelling approaches

    Energy Technology Data Exchange (ETDEWEB)

    Devau, Nicolas [INRA, UMR 1222 Eco and Sols - Ecologie Fonctionnelle et Biogeochimie des Sols (INRA-IRD-SupAgro), Place Viala, F-34060 Montpellier (France); Cadre, Edith Le [Supagro, UMR 1222 Eco and Sols - Ecologie Fonctionnelle et Biogeochimie des Sols (INRA-IRD-SupAgro), Place Viala, F-34060 Montpellier (France); Hinsinger, Philippe; Jaillard, Benoit [INRA, UMR 1222 Eco and Sols - Ecologie Fonctionnelle et Biogeochimie des Sols (INRA-IRD-SupAgro), Place Viala, F-34060 Montpellier (France); Gerard, Frederic, E-mail: gerard@supagro.inra.fr [INRA, UMR 1222 Eco and Sols - Ecologie Fonctionnelle et Biogeochimie des Sols (INRA-IRD-SupAgro), Place Viala, F-34060 Montpellier (France)

    2009-11-15

    Inorganic P is the least mobile major nutrient in most soils and is frequently the prime limiting factor for plant growth in terrestrial ecosystems. In this study, the extraction of soil inorganic P with CaCl{sub 2} (P-CaCl{sub 2}) and geochemical modelling were combined in order to unravel the processes controlling the environmentally available P (EAP) of a soil over a range of pH values (pH {approx} 4-10). Mechanistic descriptions of the adsorption of cations and anions by the soil constituents were used (1-pK Triple Plane, ion-exchange and NICA-Donnan models). These models are implemented into the geochemical code Visual MINTEQ. An additive approach was used for their application to the surface horizon of a Cambisol. The geochemical code accurately reproduced the concentration of extracted P at the different soil pH values (R{sup 2} = 0.9, RMSE = 0.03 mg kg{sup -1}). Model parameters were either directly found in the literature or estimated by fitting published experimental results in single mineral systems. The strong agreement between measurements and modelling results demonstrated that adsorption processes exerted a major control on the EAP of the soil over a large range of pH values. An influence of the precipitation of P-containing mineral is discounted based on thermodynamic calculations. Modelling results indicated that the variations in P-CaCl{sub 2} with soil pH were controlled by the deprotonation/protonation of the surface hydroxyl groups, the distribution of P surface complexes, and the adsorption of Ca and Cl from the electrolyte background. Iron-oxides and gibbsite were found to be the major P-adsorbing soil constituents at acidic and alkaline pHs, whereas P was mainly adsorbed by clay minerals at intermediate pH values. This study demonstrates the efficacy of geochemical modelling to understand soil processes, and the applicability of mechanistic adsorption models to a 'real' soil, with its mineralogical complexity and the additional

  18. SPARC-90: A code for calculating fission product capture in suppression pools

    International Nuclear Information System (INIS)

    Owczarski, P.C.; Burk, K.W.

    1991-10-01

    This report describes the technical bases and use of two updated versions of a computer code initially developed to serve as a tool for calculating aerosol particle retention in boiling water reactor (BWR) pressure suppression pools during severe accidents, SPARC-87 and SPARC-90. The most recent version is SPARC-90. The initial or prototype version (Owczarski, Postma, and Schreck 1985) was improved to include the following: rigorous treatment of local particle deposition velocities on the surface of oblate spherical bubbles, new correlations for hydrodynamic behavior of bubble swarms, models for aerosol particle growth, both mechanistic and empirical models for vent exit region scrubbing, specific models for hydrodynamics of bubble breakup at various vent types, and models for capture of vapor iodine species. A complete user's guide is provided for SPARC-90 (along with SPARC-87). A code description, code operating instructions, partial code listing, examples of the use of SPARC-90, and summaries of experimental data comparison studies also support the use of SPARC-90. 29 refs., 4 figs., 11 tabs

  19. Bird Migration Under Climate Change - A Mechanistic Approach Using Remote Sensing

    Science.gov (United States)

    Smith, James A.; Blattner, Tim; Messmer, Peter

    2010-01-01

    The broad-scale reductions and shifts that may be expected under climate change in the availability and quality of stopover habitat for long-distance migrants is an area of increasing concern for conservation biologists. Researchers generally have taken two broad approaches to the modeling of migration behaviour to understand the impact of these changes on migratory bird populations. These include models based on causal processes and their response to environmental stimulation, "mechanistic models", or models that primarily are based on observed animal distribution patterns and the correlation of these patterns with environmental variables, i.e. "data driven" models. Investigators have applied the latter technique to forecast changes in migration patterns with changes in the environment, for example, as might be expected under climate change, by forecasting how the underlying environmental data layers upon which the relationships are built will change over time. The learned geostatstical correlations are then applied to the modified data layers.. However, this is problematic. Even if the projections of how the underlying data layers will change are correct, it is not evident that the statistical relationships will remain the same, i.e. that the animal organism may not adapt its' behaviour to the changing conditions. Mechanistic models that explicitly take into account the physical, biological, and behaviour responses of an organism as well as the underlying changes in the landscape offer an alternative to address these shortcomings. The availability of satellite remote sensing observations at multiple spatial and temporal scales, coupled with advances in climate modeling and information technologies enable the application of the mechanistic models to predict how continental bird migration patterns may change in response to environmental change. In earlier work, we simulated the impact of effects of wetland loss and inter-annual variability on the fitness of

  20. Best estimate LB LOCA approach based on advanced thermal-hydraulic codes

    International Nuclear Information System (INIS)

    Sauvage, J.Y.; Gandrille, J.L.; Gaurrand, M.; Rochwerger, D.; Thibaudeau, J.; Viloteau, E.

    2004-01-01

    Improvements achieved in thermal-hydraulics with development of Best Estimate computer codes, have led number of Safety Authorities to preconize realistic analyses instead of conservative calculations. The potentiality of a Best Estimate approach for the analysis of LOCAs urged FRAMATOME to early enter into the development with CEA and EDF of the 2nd generation code CATHARE, then of a LBLOCA BE methodology with BWNT following the Code Scaling Applicability and Uncertainty (CSAU) proceeding. CATHARE and TRAC are the basic tools for LOCA studies which will be performed by FRAMATOME according to either a deterministic better estimate (dbe) methodology or a Statistical Best Estimate (SBE) methodology. (author)

  1. Development and using computer codes for improvement of defect assembly detection on Russian WWER NPPs

    International Nuclear Information System (INIS)

    Likhanskii, V.; Evdokimov, I.; Zborovskii, V.; Kanukova, V.; Sorokin, A.; Taran, M.; Ugrumov, A.; Riabinin, Y.

    2009-01-01

    Diagnostic methods of fuel failure detection for improving the radiation safety and shortening of fuel reload time at Russian WWERs are currently in development . The works include creation new computer means for increase of effectiveness of fuel monitoring and reliability of leakage tests. Reliability of failure detection can be noticeably improved when we apply an integrated approach including the following methods. The first is fuel failure analysis under operating conditions. Analysis is performed with the pilot version of the expert system, which has been developed on the basis of the mechanistic code RTOP-CA. The second stage of failure monitoring is 'sipping' tests in the mast of the refueling machine. The leakage tests are the final stage of failure monitoring. A new technique with pressure cycling in the specialized casks was introduced to meet the requirements of higher reliability in detection/confirmation of the leakages. Measurements of the activity release kinetics during the pressure cycling and handling of the acquired data with the RTOP-LT code enable to evaluate a defect size in leaking fuel assembly. The mechanistic codes RTOP-CA and RTOP-LT were verified on a base of specialized experimental data and currently the code were certified by Russian authorities Rostechnadzor. Now the pressure cycling method in the specialized casks has official status and is utilized at the all Russian WWER units. Some results of application of the integrated approach to fuel failure monitoring at several Russian NPPs with WWER units are reported in the present paper. Predictions of the current version of the expert system are compared with the results of the leakage tests and with the estimations of the defect size by the pressure cycling technique. Using the RTOP-CA code the level of activity is assessed for the following fuel campaign if the leaking fuel assembly was decided to be reloaded into the core. A project of the automated computer system on the basis of

  2. DATING: A computer code for determining allowable temperatures for dry storage of spent fuel in inert and nitrogen gases

    International Nuclear Information System (INIS)

    Simonen, E.P.; Gilbert, E.R.

    1988-12-01

    The DATING (Determining Allowable Temperatures in Inert and Nitrogen Gases) code can be used to calculate allowable initial temperatures for dry storage of light-water-reactor spent fuel. The calculations are based on the life fraction rule using both measured data and mechanistic equations as reported by Chin et al. (1986). The code is written in FORTRAN and utilizes an efficient numerical integration method for rapid calculations on IBM-compatible personal computers. This report documents the technical basis for the DATING calculations, describes the computational method and code statements, and includes a user's guide with examples. The software for the DATING code is available through the National Energy Software Center operated by Argonne National Laboratory, Argonne, Illinois 60439. 5 refs., 8 figs., 5 tabs

  3. Multimedia signal coding and transmission

    CERN Document Server

    Ohm, Jens-Rainer

    2015-01-01

    This textbook covers the theoretical background of one- and multidimensional signal processing, statistical analysis and modelling, coding and information theory with regard to the principles and design of image, video and audio compression systems. The theoretical concepts are augmented by practical examples of algorithms for multimedia signal coding technology, and related transmission aspects. On this basis, principles behind multimedia coding standards, including most recent developments like High Efficiency Video Coding, can be well understood. Furthermore, potential advances in future development are pointed out. Numerous figures and examples help to illustrate the concepts covered. The book was developed on the basis of a graduate-level university course, and most chapters are supplemented by exercises. The book is also a self-contained introduction both for researchers and developers of multimedia compression systems in industry.

  4. MCNP code

    International Nuclear Information System (INIS)

    Cramer, S.N.

    1984-01-01

    The MCNP code is the major Monte Carlo coupled neutron-photon transport research tool at the Los Alamos National Laboratory, and it represents the most extensive Monte Carlo development program in the United States which is available in the public domain. The present code is the direct descendent of the original Monte Carlo work of Fermi, von Neumaum, and Ulam at Los Alamos in the 1940s. Development has continued uninterrupted since that time, and the current version of MCNP (or its predecessors) has always included state-of-the-art methods in the Monte Carlo simulation of radiation transport, basic cross section data, geometry capability, variance reduction, and estimation procedures. The authors of the present code have oriented its development toward general user application. The documentation, though extensive, is presented in a clear and simple manner with many examples, illustrations, and sample problems. In addition to providing the desired results, the output listings give a a wealth of detailed information (some optional) concerning each state of the calculation. The code system is continually updated to take advantage of advances in computer hardware and software, including interactive modes of operation, diagnostic interrupts and restarts, and a variety of graphical and video aids

  5. Development of the advanced CANDU technology

    International Nuclear Information System (INIS)

    Suk, Soo Dong; Min, Byung Joo; Na, Y. H.; Lee, S. Y.; Choi, J. H.; Lee, B. C.; Kim, S. N.; Jo, C. H.; Paik, J. S.; On, M. R.; Park, H. S.; Kim, S. R.

    1997-07-01

    The purpose of this study is to develop the advanced design technology to improve safety, operability and economy and to develop and advanced safety evaluation system. More realistic and reasonable methodology and modeling was employed to improve safety margin in containment analysis. Various efforts have been made to verify the CATHENA code which is the major safety analysis code for CANDU PHWR system. Fully computerized prototype ECCS was developed. The feasibility study and conceptual design of the distributed digital control system have been performed as well. The core characteristics of advanced fuel cycle, fuel management and power upgrade have been studied to determine the advanced core. (author). 77 refs., 51 tabs., 108 figs

  6. Development of the advanced CANDU technology

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Soo Dong; Min, Byung Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Na, Y H; Lee, S Y; Choi, J H; Lee, B C; Kim, S N; Jo, C H; Paik, J S; On, M R; Park, H S; Kim, S R [Korea Electric Power Co., Taejon (Korea, Republic of)

    1997-07-01

    The purpose of this study is to develop the advanced design technology to improve safety, operability and economy and to develop and advanced safety evaluation system. More realistic and reasonable methodology and modeling was employed to improve safety margin in containment analysis. Various efforts have been made to verify the CATHENA code which is the major safety analysis code for CANDU PHWR system. Fully computerized prototype ECCS was developed. The feasibility study and conceptual design of the distributed digital control system have been performed as well. The core characteristics of advanced fuel cycle, fuel management and power upgrade have been studied to determine the advanced core. (author). 77 refs., 51 tabs., 108 figs.

  7. Current Needs for the Experimental Investigation of the CHF Phenomenon Relevant to LWR Core Conditions

    International Nuclear Information System (INIS)

    Le Corre, J.M.

    2009-01-01

    The current achievements and needs toward the investigation, understanding and mechanistic prediction of the Critical Heat Flux (CHF) event, under PWR and BWR core conditions, are addressed in this paper. It is shown that, even when using advanced 3-D CFD simulation tools, the current approach to CHF mechanistic modeling has serious limitations. This is mainly due to the lack of information regarding the relevant two-phase flow pattern(s) (in particular near the heated wall) and associated mechanisms (at the meso and micro-scale) leading to the CHF event. Areas of current experimental needs are identified in order to address these shortcomings. In addition, the use of 1-D and 3-D numerical tools to mechanistically predict the CHF is discussed. It is shown that 3-D two-phase CFD codes may not be superior to 1-D codes without proper consideration of relevant constitutive relations. (author)

  8. Utility of QR codes in biological collections.

    Science.gov (United States)

    Diazgranados, Mauricio; Funk, Vicki A

    2013-01-01

    The popularity of QR codes for encoding information such as URIs has increased exponentially in step with the technological advances and availability of smartphones, digital tablets, and other electronic devices. We propose using QR codes on specimens in biological collections to facilitate linking vouchers' electronic information with their associated collections. QR codes can efficiently provide such links for connecting collections, photographs, maps, ecosystem notes, citations, and even GenBank sequences. QR codes have numerous advantages over barcodes, including their small size, superior security mechanisms, increased complexity and quantity of information, and low implementation cost. The scope of this paper is to initiate an academic discussion about using QR codes on specimens in biological collections.

  9. Utility of QR codes in biological collections

    Directory of Open Access Journals (Sweden)

    Mauricio Diazgranados

    2013-07-01

    Full Text Available The popularity of QR codes for encoding information such as URIs has increased exponentially in step with the technological advances and availability of smartphones, digital tablets, and other electronic devices. We propose using QR codes on specimens in biological collections to facilitate linking vouchers’ electronic information with their associated collections. QR codes can efficiently provide such links for connecting collections, photographs, maps, ecosystem notes, citations, and even GenBank sequences. QR codes have numerous advantages over barcodes, including their small size, superior security mechanisms, increased complexity and quantity of information, and low implementation cost. The scope of this paper is to initiate an academic discussion about using QR codes on specimens in biological collections.

  10. MARS code manual volume I: code structure, system models, and solution methods

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Kim, Kyung Doo; Bae, Sung Won; Jeong, Jae Jun; Lee, Seung Wook; Hwang, Moon Kyu; Yoon, Churl

    2010-02-01

    Korea Advanced Energy Research Institute (KAERI) conceived and started the development of MARS code with the main objective of producing a state-of-the-art realistic thermal hydraulic systems analysis code with multi-dimensional analysis capability. MARS achieves this objective by very tightly integrating the one dimensional RELAP5/MOD3 with the multi-dimensional COBRA-TF codes. The method of integration of the two codes is based on the dynamic link library techniques, and the system pressure equation matrices of both codes are implicitly integrated and solved simultaneously. In addition, the Equation-Of-State (EOS) for the light water was unified by replacing the EOS of COBRA-TF by that of the RELAP5. This theory manual provides a complete list of overall information of code structure and major function of MARS including code architecture, hydrodynamic model, heat structure, trip / control system and point reactor kinetics model. Therefore, this report would be very useful for the code users. The overall structure of the manual is modeled on the structure of the RELAP5 and as such the layout of the manual is very similar to that of the RELAP. This similitude to RELAP5 input is intentional as this input scheme will allow minimum modification between the inputs of RELAP5 and MARS3.1. MARS3.1 development team would like to express its appreciation to the RELAP5 Development Team and the USNRC for making this manual possible

  11. Deciphering the genetic regulatory code using an inverse error control coding framework.

    Energy Technology Data Exchange (ETDEWEB)

    Rintoul, Mark Daniel; May, Elebeoba Eni; Brown, William Michael; Johnston, Anna Marie; Watson, Jean-Paul

    2005-03-01

    We have found that developing a computational framework for reconstructing error control codes for engineered data and ultimately for deciphering genetic regulatory coding sequences is a challenging and uncharted area that will require advances in computational technology for exact solutions. Although exact solutions are desired, computational approaches that yield plausible solutions would be considered sufficient as a proof of concept to the feasibility of reverse engineering error control codes and the possibility of developing a quantitative model for understanding and engineering genetic regulation. Such evidence would help move the idea of reconstructing error control codes for engineered and biological systems from the high risk high payoff realm into the highly probable high payoff domain. Additionally this work will impact biological sensor development and the ability to model and ultimately develop defense mechanisms against bioagents that can be engineered to cause catastrophic damage. Understanding how biological organisms are able to communicate their genetic message efficiently in the presence of noise can improve our current communication protocols, a continuing research interest. Towards this end, project goals include: (1) Develop parameter estimation methods for n for block codes and for n, k, and m for convolutional codes. Use methods to determine error control (EC) code parameters for gene regulatory sequence. (2) Develop an evolutionary computing computational framework for near-optimal solutions to the algebraic code reconstruction problem. Method will be tested on engineered and biological sequences.

  12. A 3D-CFD code for accurate prediction of fluid flows and fluid forces in seals

    Science.gov (United States)

    Athavale, M. M.; Przekwas, A. J.; Hendricks, R. C.

    1994-01-01

    Current and future turbomachinery requires advanced seal configurations to control leakage, inhibit mixing of incompatible fluids and to control the rotodynamic response. In recognition of a deficiency in the existing predictive methodology for seals, a seven year effort was established in 1990 by NASA's Office of Aeronautics Exploration and Technology, under the Earth-to-Orbit Propulsion program, to develop validated Computational Fluid Dynamics (CFD) concepts, codes and analyses for seals. The effort will provide NASA and the U.S. Aerospace Industry with advanced CFD scientific codes and industrial codes for analyzing and designing turbomachinery seals. An advanced 3D CFD cylindrical seal code has been developed, incorporating state-of-the-art computational methodology for flow analysis in straight, tapered and stepped seals. Relevant computational features of the code include: stationary/rotating coordinates, cylindrical and general Body Fitted Coordinates (BFC) systems, high order differencing schemes, colocated variable arrangement, advanced turbulence models, incompressible/compressible flows, and moving grids. This paper presents the current status of code development, code demonstration for predicting rotordynamic coefficients, numerical parametric study of entrance loss coefficients for generic annular seals, and plans for code extensions to labyrinth, damping, and other seal configurations.

  13. Managing mechanistic and organic structure in health care organizations.

    Science.gov (United States)

    Olden, Peter C

    2012-01-01

    Managers at all levels in a health care organization must organize work to achieve the organization's mission and goals. This requires managers to decide the organization structure, which involves dividing the work among jobs and departments and then coordinating them all toward the common purpose. Organization structure, which is reflected in an organization chart, may range on a continuum from very mechanistic to very organic. Managers must decide how mechanistic versus how organic to make the entire organization and each of its departments. To do this, managers should carefully consider 5 factors for the organization and for each individual department: external environment, goals, work production, size, and culture. Some factors may push toward more mechanistic structure, whereas others may push in the opposite direction toward more organic structure. Practical advice can help managers at all levels design appropriate structure for their departments and organization.

  14. Validation of the THIRMAL-1 melt-water interaction code

    Energy Technology Data Exchange (ETDEWEB)

    Chu, C.C.; Sienicki, J.J.; Spencer, B.W. [Argonne National Lab., IL (United States)

    1995-09-01

    The THIRMAL-1 computer code has been used to calculate nonexplosive LWR melt-water interactions both in-vessel and ex-vessel. To support the application of the code and enhance its acceptability, THIRMAL-1 has been compared with available data from two of the ongoing FARO experiments at Ispra and two of the Corium Coolant Mixing (CCM) experiments performed at Argonne. THIRMAL-1 calculations for the FARO Scoping Test and Quenching Test 2 as well as the CCM-5 and -6 experiments were found to be in excellent agreement with the experiment results. This lends confidence to the modeling that has been incorporated in the code describing melt stream breakup due to the growth of both Kelvin-Helmholtz and large wave instabilities, the sizes of droplets formed, multiphase flow and heat transfer in the mixing zone surrounding and below the melt metallic phase. As part of the analysis of the FARO tests, a mechanistic model was developed to calculate the prefragmentation as it may have occurred when melt relocated from the release vessel to the water surface and the model was compared with the relevant data from FARO.

  15. Cognitive science as an interface between rational and mechanistic explanation.

    Science.gov (United States)

    Chater, Nick

    2014-04-01

    Cognitive science views thought as computation; and computation, by its very nature, can be understood in both rational and mechanistic terms. In rational terms, a computation solves some information processing problem (e.g., mapping sensory information into a description of the external world; parsing a sentence; selecting among a set of possible actions). In mechanistic terms, a computation corresponds to causal chain of events in a physical device (in engineering context, a silicon chip; in biological context, the nervous system). The discipline is thus at the interface between two very different styles of explanation--as the papers in the current special issue well illustrate, it explores the interplay of rational and mechanistic forces. Copyright © 2014 Cognitive Science Society, Inc.

  16. Introduction of thermal-hydraulic analysis code and system analysis code for HTGR

    International Nuclear Information System (INIS)

    Tanaka, Mitsuhiro; Izaki, Makoto; Koike, Hiroyuki; Tokumitsu, Masashi

    1984-01-01

    Kawasaki Heavy Industries Ltd. has advanced the development and systematization of analysis codes, aiming at lining up the analysis codes for heat transferring flow and control characteristics, taking up HTGR plants as the main object. In order to make the model of flow when shock waves propagate to heating tubes, SALE-3D which can analyze a complex system was developed, therefore, it is reported in this paper. Concerning the analysis code for control characteristics, the method of sensitivity analysis in a topological space including an example of application is reported. The flow analysis code SALE-3D is that for analyzing the flow of compressible viscous fluid in a three-dimensional system over the velocity range from incompressibility limit to supersonic velocity. The fundamental equations and fundamental algorithm of the SALE-3D, the calculation of cell volume, the plotting of perspective drawings and the analysis of the three-dimensional behavior of shock waves propagating in heating tubes after their rupture accident are described. The method of sensitivity analysis was added to the analysis code for control characteristics in a topological space, and blow-down phenomena was analyzed by its application. (Kako, I.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Ebert, D.

    1997-07-01

    This is a report on the CSNI Workshop on Transient Thermal-Hydraulic and Neutronic Codes Requirements held at Annapolis, Maryland, USA November 5-8, 1996. This experts` meeting consisted of 140 participants from 21 countries; 65 invited papers were presented. The meeting was divided into five areas: (1) current and prospective plans of thermal hydraulic codes development; (2) current and anticipated uses of thermal-hydraulic codes; (3) advances in modeling of thermal-hydraulic phenomena and associated additional experimental needs; (4) numerical methods in multi-phase flows; and (5) programming language, code architectures and user interfaces. The workshop consensus identified the following important action items to be addressed by the international community in order to maintain and improve the calculational capability: (a) preserve current code expertise and institutional memory, (b) preserve the ability to use the existing investment in plant transient analysis codes, (c) maintain essential experimental capabilities, (d) develop advanced measurement capabilities to support future code validation work, (e) integrate existing analytical capabilities so as to improve performance and reduce operating costs, (f) exploit the proven advances in code architecture, numerics, graphical user interfaces, and modularization in order to improve code performance and scrutibility, and (g) more effectively utilize user experience in modifying and improving the codes.

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

    International Nuclear Information System (INIS)

    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

  19. Bridging Mechanistic and Phenomenological Models of Complex Biological Systems.

    Science.gov (United States)

    Transtrum, Mark K; Qiu, Peng

    2016-05-01

    The inherent complexity of biological systems gives rise to complicated mechanistic models with a large number of parameters. On the other hand, the collective behavior of these systems can often be characterized by a relatively small number of phenomenological parameters. We use the Manifold Boundary Approximation Method (MBAM) as a tool for deriving simple phenomenological models from complicated mechanistic models. The resulting models are not black boxes, but remain expressed in terms of the microscopic parameters. In this way, we explicitly connect the macroscopic and microscopic descriptions, characterize the equivalence class of distinct systems exhibiting the same range of collective behavior, and identify the combinations of components that function as tunable control knobs for the behavior. We demonstrate the procedure for adaptation behavior exhibited by the EGFR pathway. From a 48 parameter mechanistic model, the system can be effectively described by a single adaptation parameter τ characterizing the ratio of time scales for the initial response and recovery time of the system which can in turn be expressed as a combination of microscopic reaction rates, Michaelis-Menten constants, and biochemical concentrations. The situation is not unlike modeling in physics in which microscopically complex processes can often be renormalized into simple phenomenological models with only a few effective parameters. The proposed method additionally provides a mechanistic explanation for non-universal features of the behavior.

  20. RELAP-7 Code Assessment Plan and Requirement Traceability Matrix

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Junsoo; Choi, Yong-joon; Smith, Curtis L.

    2016-10-01

    The RELAP-7, a safety analysis code for nuclear reactor system, is under development at Idaho National Laboratory (INL). Overall, the code development is directed towards leveraging the advancements in computer science technology, numerical solution methods and physical models over the last decades. Recently, INL has also been putting an effort to establish the code assessment plan, which aims to ensure an improved final product quality through the RELAP-7 development process. The ultimate goal of this plan is to propose a suitable way to systematically assess the wide range of software requirements for RELAP-7, including the software design, user interface, and technical requirements, etc. To this end, we first survey the literature (i.e., international/domestic reports, research articles) addressing the desirable features generally required for advanced nuclear system safety analysis codes. In addition, the V&V (verification and validation) efforts as well as the legacy issues of several recently-developed codes (e.g., RELAP5-3D, TRACE V5.0) are investigated. Lastly, this paper outlines the Requirement Traceability Matrix (RTM) for RELAP-7 which can be used to systematically evaluate and identify the code development process and its present capability.

  1. International Code Assessment and Applications Program: Summary of code assessment studies concerning RELAP5/MOD2, RELAP5/MOD3, and TRAC-B

    International Nuclear Information System (INIS)

    Schultz, R.R.

    1993-12-01

    Members of the International Code Assessment Program (ICAP) have assessed the US Nuclear Regulatory Commission (USNRC) advanced thermal-hydraulic codes over the past few years in a concerted effort to identify deficiencies, to define user guidelines, and to determine the state of each code. The results of sixty-two code assessment reviews, conducted at INEL, are summarized. Code deficiencies are discussed and user recommended nodalizations investigated during the course of conducting the assessment studies and reviews are listed. All the work that is summarized was done using the RELAP5/MOD2, RELAP5/MOD3, and TRAC-B codes

  2. Mechanistic Indicators of Childhood Asthma (MICA): piloting ...

    Science.gov (United States)

    Background: Modem methods in molecular biology and advanced computational tools show promise in elucidating complex interactions that occur between genes and environmental factors in diseases such as asthma; however appropriately designed studies are critical for these methods to reach their full potential. Objective: We used a case-control study to investigate whether genomic data (blood gene expression), viewed together with a spectrum of exposure effects and susceptibility markers (blood, urine and nail), can provide a mechanistic explanation for the increased susceptibility of asthmatics to ambient air pollutants. Methods: We studied 205 non-asthmatic and asthmatic children, (9-12 years of age) who participated in a clinical study in Detroit, Michigan. The study combines a traditional epidemiological design with an integrative approach to investigate the environmental exposure of children to indoor-outdoor air. The study includes measurements of internal dose (metals, allergen specific IgE, PAH and VOC metabolites) and clinical measures of health outcome (immunological, cardiovascular and respiratory). Results: Expected immunological indications of asthma have been obtained. In addition, initial results from our analyses point to the complex nature of childhood health and risk factors linked to metabolic syndrome (obesity, blood pressure and dyslipidemia). For example, 31% and 34% of the asthmatic MICA subjects were either overweight (BMI > 25) o

  3. Light water reactor fuel analysis code FEMAXI-7; model and structure

    International Nuclear Information System (INIS)

    Suzuki, Motoe; Udagawa, Yutaka; Saitou, Hiroaki

    2011-03-01

    A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in both normal conditions and anticipated transient conditions. This code is an advanced version which has been produced by incorporating the former version FEMAXI-6 with numerous functional improvements and extensions. In FEMAXI-7, many new models have been added and parameters have been clearly arranged. Also, to facilitate effective maintenance and accessibility of the code, modularization of subroutines and functions have been attained, and quality comment descriptions of variables or physical quantities have been incorporated in the source code. With these advancements, the FEMAXI-7 code has been upgraded to a versatile analytical tool for high burnup fuel behavior analyses. This report describes in detail the design, basic theory and structure, models and numerical method, and improvements and extensions. (author)

  4. Enzymatic Halogenation and Dehalogenation Reactions: Pervasive and Mechanistically Diverse.

    Science.gov (United States)

    Agarwal, Vinayak; Miles, Zachary D; Winter, Jaclyn M; Eustáquio, Alessandra S; El Gamal, Abrahim A; Moore, Bradley S

    2017-04-26

    Naturally produced halogenated compounds are ubiquitous across all domains of life where they perform a multitude of biological functions and adopt a diversity of chemical structures. Accordingly, a diverse collection of enzyme catalysts to install and remove halogens from organic scaffolds has evolved in nature. Accounting for the different chemical properties of the four halogen atoms (fluorine, chlorine, bromine, and iodine) and the diversity and chemical reactivity of their organic substrates, enzymes performing biosynthetic and degradative halogenation chemistry utilize numerous mechanistic strategies involving oxidation, reduction, and substitution. Biosynthetic halogenation reactions range from simple aromatic substitutions to stereoselective C-H functionalizations on remote carbon centers and can initiate the formation of simple to complex ring structures. Dehalogenating enzymes, on the other hand, are best known for removing halogen atoms from man-made organohalogens, yet also function naturally, albeit rarely, in metabolic pathways. This review details the scope and mechanism of nature's halogenation and dehalogenation enzymatic strategies, highlights gaps in our understanding, and posits where new advances in the field might arise in the near future.

  5. A new release of the S3M code

    International Nuclear Information System (INIS)

    Pavlovic, M.; Bokor, J.; Regodic, M.; Sagatova, A.

    2015-01-01

    This paper presents a new release of the code that contains some additional routines and advanced features of displaying the results. Special attention is paid to the processing of the SRIM range file, which was not included in the previous release of the code. Examples of distributions provided by the S 3 M code for implanted ions in thyroid and iron are presented. (authors)

  6. A novel method of generating and remembering international morse codes

    Digital Repository Service at National Institute of Oceanography (India)

    Charyulu, R.J.K.

    untethered communications have been advanced, despite as S.O.S International Morse Code will be at rescue as an emergency tool, when all other modes fail The details of hte method and actual codes have been enumerated....

  7. The release code package REVOLS/RENONS for fission product release from a liquid sodium pool into an inert gas atmosphere

    International Nuclear Information System (INIS)

    Starflinger, J.; Scholtyssek, W.; Unger, H.

    1994-12-01

    For aerosol source term considerations in the field of nuclear safety, the investigation of the release of volatile and non-volatile species from liquid surfaces into a gas atmosphere is important. In case of a hypothetical liquid metal fast breeder reactor accident with tank failure, primary coolant sodium with suspended or solved fuel particles and fission products may be released into the containment. The computer code package REVOLS/RENONS, based on a theoretical mechanistic model with a modular structure, has been developed for the prediction of sodium release as well as volatile and non-volatile radionuclide release from a liquid pool surface into the inert gas atmosphere of the inner containment. Hereby the release of sodium and volatile fission products, like cesium and sodium iodide, is calculated using a theoretical model in a mass transfer coefficient formulation. This model has been transposed into the code version REVOLS.MOD1.1, which is discussed here. It enables parameter analysis under highly variable user-defined boundary conditions. Whereas the evaporative release of the volatile components is governed by diffusive and convective transport processes, the release of the non-volatile ones may be governed by mechanical processes which lead to droplet entrainment from the wavy pool surface under conditions of natural or forced convection into the atmosphere. The mechanistic model calculates the liquid entrainment rate of the non-volatile species, like the fission product strontium oxide and the fuel (uranium dioxide) from a liquid pool surface into a parallel gas flow. The mechanistic model has been transposed into the computer code package REVOLS/RENONS, which is discussed here. Hereby the module REVOLS (RElease of VOLatile Species) calculates the evaporative release of the volatile species, while the module RENONS (RElease of NON-Volatile Species) computes the entrainment release of the non-volatile radionuclides. (orig./HP) [de

  8. Steady-State Calculation of the ATLAS Test Facility Using the SPACE Code

    International Nuclear Information System (INIS)

    Kim, Hyoung Tae; Choi, Ki Yong; Kim, Kyung Doo

    2011-01-01

    The Korean nuclear industry is developing a thermalhydraulic analysis code for safety analysis of pressurized water reactors (PWRs). The new code is called the Safety and Performance Analysis Code for Nuclear Power Plants (SPACE). Several research and industrial organizations including KAERI (Korea Atomic Energy Research Institute) are participating in the collaboration for the development of the SPACE code. One of the main tasks of KAERI is to carry out separate effect tests (SET) and integral effect tests (IET) for code verification and validation (V and V). The IET has been performed with ATLAS (Advanced Thermalhydraulic Test Loop for Accident Simulation) based on the design features of the APR1400 (Advanced Power Reactor of 1400MWe). In the present work the SPACE code input-deck for ATLAS is developed and used for simulation of the steady-state conditions of ATLAS as a preliminary work for IET V and V of the SPACE code

  9. A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karahan, Aydin, E-mail: karahan@mit.ed [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology (United States); Buongiorno, Jacopo [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology (United States)

    2010-01-31

    An engineering code to predict the irradiation behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and UO{sub 2}-PuO{sub 2} mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel-clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios. FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal-fuel version is called FEAST-METAL, and is described in this paper. The oxide-fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel-clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium

  10. A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

    International Nuclear Information System (INIS)

    Karahan, Aydin; Buongiorno, Jacopo

    2010-01-01

    An engineering code to predict the irradiation behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and UO 2 -PuO 2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel-clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios. FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal-fuel version is called FEAST-METAL, and is described in this paper. The oxide-fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel-clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium reactors

  11. Developments of HTGR thermofluid dynamic analysis codes and HTGR plant dynamic simulation code

    International Nuclear Information System (INIS)

    Tanaka, Mitsuhiro; Izaki, Makoto; Koike, Hiroyuki; Tokumitsu, Masashi

    1983-01-01

    In nuclear power plants as well as high temperature gas-cooled reactor plants, the design is mostly performed on the basis of the results after their characteristics have been grasped by carrying out the numerical simulation using the analysis code. Also in Kawasaki Heavy Industries Ltd., on the basis of the system engineering accumulated with gas-cooled reactors since several years ago, the preparation and systematization of analysis codes have been advanced, aiming at lining up the analysis codes for heat transferring flow and control characteristics, taking up HTGR plants as the main object. In this report, a part of the results is described. The example of the analysis applying the two-dimensional compressible flow analysis codes SOLA-VOF and SALE-2D, which were developed by Los Alamos National Laboratory in USA and modified for use in Kawasaki, to HTGR system is reported. Besides, Kawasaki has developed the control characteristics analyzing code DYSCO by which the change of system composition is easy and high versatility is available. The outline, fundamental equations, fundamental algorithms and examples of application of the SOLA-VOF and SALE-2D, the present status of system characteristic simulation codes and the outline of the DYSCO are described. (Kako, I.)

  12. Gap conductance model validation in the TASS/SMR-S code

    International Nuclear Information System (INIS)

    Ahn, Sang-Jun; Yang, Soo-Hyung; Chung, Young-Jong; Bae, Kyoo-Hwan; Lee, Won-Jae

    2011-01-01

    An advanced integral pressurized water reactor, SMART (System-Integrated Modular Advanced ReacTor) has been developed by KAERI (Korea Atomic Energy Research and Institute). The purposes of the SMART are sea water desalination and an electricity generation. For the safety evaluation and performance analysis of the SMART, TASS/SMR-S (Transient And Setpoint Simulation/System-integrated Modular Reactor) code, has been developed. In this paper, the gap conductance model for the calculation of gap conductance has been validated by using another system code, MARS code, and experimental results. In the validation, the behaviors of fuel temperature and gap width are selected as the major parameters. According to the evaluation results, the TASS/SMR-S code predicts well the behaviors of fuel temperatures and gap width variation, compared to the MARS calculation results and experimental data. (author)

  13. Monte Carlo advances for the Eolus Asci Project

    International Nuclear Information System (INIS)

    Hendrick, J. S.; McKinney, G. W.; Cox, L. J.

    2000-01-01

    The Eolus ASCI project includes parallel, 3-D transport simulation for various nuclear applications. The codes developed within this project provide neutral and charged particle transport, detailed interaction physics, numerous source and tally capabilities, and general geometry packages. One such code is MCNPW which is a general purpose, 3-dimensional, time-dependent, continuous-energy Monte Carlo fully-coupled N-Particle transport code. Significant advances are also being made in the areas of modern software engineering and parallel computing. These advances are described in detail

  14. Mechanistic Links Between PARP, NAD, and Brain Inflammation After TBI

    Science.gov (United States)

    2015-10-01

    1 AWARD NUMBER: W81XWH-13-2-0091 TITLE: Mechanistic Links Between PARP, NAD , and Brain Inflammation After TBI PRINCIPAL INVESTIGATOR...COVERED 25 Sep 2014 - 24 Sep 2015 4. TITLE AND SUBTITLE 5a. CONTRACT NUMBER Mechanistic Links Between PARP, NAD , and Brain Inflammation After TBI 5b. GRANT...efficacy of veliparib and NAD as agents for suppressing inflammation and improving outcomes after traumatic brain injury. The animal models include

  15. The sphere-PAC fuel code 'SPHERE-3'

    International Nuclear Information System (INIS)

    Wallin, H.

    2000-01-01

    Sphere-PAC fuel is an advanced nuclear fuel, in which the cladding tube is filled with small fuel spheres instead of the more usual fuel pellets. At PSI, the irradiation behaviour of sphere-PAC fuel is calculated using the computer code SPHERE-3. The paper describes the present status of the SPHERE-3 code, and some results of the qualification process against experimental data. (author)

  16. The sphere-pac fuel code 'SPHERE-3'

    International Nuclear Information System (INIS)

    Wallin, H.; Nordstroem, L.A.; Hellwig, C.

    2001-01-01

    Sphere-pac fuel is an advanced nuclear fuel, in which the cladding tube is filled with small fuel spheres instead of the more usual fuel pellets. At PSI, the irradiation behaviour of sphere-pac fuel is calculated using the computer code SPHERE-3. The paper describes the present status of the SPHERE-3 code, and some results of the qualification process against experimental data. (author)

  17. Improving Predictive Modeling in Pediatric Drug Development: Pharmacokinetics, Pharmacodynamics, and Mechanistic Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Slikker, William; Young, John F.; Corley, Rick A.; Dorman, David C.; Conolly, Rory B.; Knudsen, Thomas; Erstad, Brian L.; Luecke, Richard H.; Faustman, Elaine M.; Timchalk, Chuck; Mattison, Donald R.

    2005-07-26

    A workshop was conducted on November 18?19, 2004, to address the issue of improving predictive models for drug delivery to developing humans. Although considerable progress has been made for adult humans, large gaps remain for predicting pharmacokinetic/pharmacodynamic (PK/PD) outcome in children because most adult models have not been tested during development. The goals of the meeting included a description of when, during development, infants/children become adultlike in handling drugs. The issue of incorporating the most recent advances into the predictive models was also addressed: both the use of imaging approaches and genomic information were considered. Disease state, as exemplified by obesity, was addressed as a modifier of drug pharmacokinetics and pharmacodynamics during development. Issues addressed in this workshop should be considered in the development of new predictive and mechanistic models of drug kinetics and dynamics in the developing human.

  18. Short-Term Memory Coding in Children With Intellectual Disabilities

    OpenAIRE

    Henry, L.; Conners, F.

    2008-01-01

    To examine visual and verbal coding strategies, I asked children with intellectual disabilities and peers matched for MA and CA to perform picture memory span tasks with phonologically similar, visually similar, long, or nonsimilar named items. The CA group showed effects consistent with advanced verbal memory coding (phonological similarity and word length effects). Neither the intellectual disabilities nor MA groups showed evidence for memory coding strategies. However, children in these gr...

  19. The HELIOS-2 lattice physics code

    International Nuclear Information System (INIS)

    Wemple, C.A.; Gheorghiu, H-N.M.; Stamm'ler, R.J.J.; Villarino, E.A.

    2008-01-01

    Major advances have been made in the HELIOS code, resulting in the impending release of a new version, HELIOS-2. The new code includes a method of characteristics (MOC) transport solver to supplement the existing collision probabilities (CP) solver. A 177-group, ENDF/B-VII nuclear data library has been developed for inclusion with the new code package. Computational tests have been performed to verify the performance of the MOC solver against the CP solver, and validation testing against computational and measured benchmarks is underway. Results to-date of the verification and validation testing are presented, demonstrating the excellent performance of the new transport solver and nuclear data library. (Author)

  20. Light water reactor fuel analysis code FEMAXI-7. Model and structure

    International Nuclear Information System (INIS)

    Suzuki, Motoe; Udagawa, Yutaka; Nagase, Fumihisa; Saitou, Hiroaki

    2013-07-01

    A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in both normal conditions and anticipated transient conditions. This code is an advanced version which has been produced by incorporating the former version FEMAXI-6 with numerous functional improvements and extensions. In FEMAXI-7, many new models have been added and parameters have been clearly arranged. Also, to facilitate effective maintenance and accessibility of the code, modularization of subroutines and functions have been attained, and quality comment descriptions of variables or physical quantities have been incorporated in the source code. With these advancements, the FEMAXI-7 code has been upgraded to a versatile analytical tool for high burnup fuel behavior analyses. This report describes in detail the design, basic theory and structure, models and numerical method of FEMAXI-7, and its improvements and extensions. (author)

  1. Relationship between various pressure vessel and piping codes

    International Nuclear Information System (INIS)

    Canonico, D.A.

    1976-01-01

    Section VIII of the ASME Code provides stress allowable values for material specifications that are provided in Section II Parts A and B. Since the adoption of the ASME Code over 60 years ago the incidence of failure has been greatly reduced. The Codes are currently based on strength criteria and advancements in the technology of fracture toughness and fracture mechanics should permit an even greater degree of reliability and safety. This lecture discusses the various Sections of the Code. It describes the basis for the establishment of design stress allowables and promotes the idea of the use of fracture mechanics

  2. Proceedings of the international workshop on mechanistic understanding of radionuclide migration in compacted/intact systems

    International Nuclear Information System (INIS)

    Tachi, Yukio; Yui, Mikazu

    2010-03-01

    The international workshop on mechanistic understanding of radionuclide migration in compacted / intact systems was held at ENTRY, JAEA, Tokai on 21st - 23rd January, 2009. This workshop was hosted by Japan Atomic Energy Agency (JAEA) as part of the project on the mechanistic model/database development for radionuclide sorption and diffusion behavior in compacted / intact systems. The overall goal of the project is to develop the mechanistic model / database for a consistent understanding and prediction of migration parameters and its uncertainties for performance assessment of geological disposal of radioactive waste. The objective of the workshop is to integrate the state-of-the-art of mechanistic sorption and diffusion model in compacted / intact systems, especially in bentonite / clay systems, and discuss the JAEA's mechanistic approaches and future challenges, especially the following discussions points; 1) What's the status and difficulties for mechanistic model/database development? 2) What's the status and difficulties for applicability of mechanistic model to the compacted/intact system? 3) What's the status and difficulties for obtaining evidences for mechanistic model? 4) What's the status and difficulties for standardization of experimental methodology for batch sorption and diffusion? 5) What's the uncertainties of transport parameters in radionuclides migration analysis due to a lack of understanding/experimental methodologies, and how do we derive them? This report includes workshop program, overview and materials of each presentation, summary of discussions. (author)

  3. Code development and analyses within the area of transmutation and safety

    International Nuclear Information System (INIS)

    Maschek, W.

    2002-01-01

    A strong code development is going on to meet various demands resulting from the development of dedicated reactors for transmutation and incineration. Code development is concerned with safety codes and general codes needed for assessing scenarios and transmutation strategies. Analyses concentrate on various ADS systems with solid and liquid molten salt fuels. Analyses deal with ADS Demo Plant (5th FP EU) and transmuters with advanced fuels

  4. Void fraction prediction of NUPEC PSBT tests by CATHARE code

    International Nuclear Information System (INIS)

    Del Nevo, A.; Michelotti, L.; Moretti, F.; Rozzia, D.; D'Auria, F.

    2011-01-01

    The current generation of thermal-hydraulic system codes benefits of about sixty years of experiments and forty years of development and are considered mature tools to provide best estimate description of phenomena and detailed reactor system representations. However, there are continuous needs for checking the code capabilities in representing nuclear system, for drawing attention to their weak points, for identifying models which need to be refined for best-estimate calculations. Prediction of void fraction and Departure from Nucleate Boiling (DNB) in system thermal-hydraulics is currently based on empirical approaches. The database carried out by Nuclear Power Engineering Corporation (NUPEC), Japan addresses these issues. It is suitable for supporting the development of new computational tools based on more mechanistic approaches (i.e. three-field codes, two-phase CFD, etc.) as well as for validating current generation of thermal-hydraulic system codes. Selected experiments belonging to this database are used for the OECD/NRC PSBT benchmark. The paper reviews the activity carried out by CATHARE2 code on the basis of the subchannel (four test sections) and presents rod bundle (different axial power profile and test sections) experiments available in the database in steady state and transient conditions. The results demonstrate the accuracy of the code in predicting the void fraction in different thermal-hydraulic conditions. The tests are performed varying the pressure, coolant temperature, mass flow and power. Sensitivity analyses are carried out addressing nodalization effect and the influence of the initial and boundary conditions of the tests. (author)

  5. Analytical model for bottom reflooding heat transfer in light water reactors (the UCFLOOD code)

    International Nuclear Information System (INIS)

    Arrieta, L.; Yadigaroglu, G.

    1978-08-01

    The UCFLOOD code is based on mechanistic models developed to analyze bottom reflooding of a single flow channel and its associated fuel rod, or a tubular test section with internal flow. From the hydrodynamic point of view the flow channel is divided into a single-phase liquid region, a continuous-liquid two-phase region, and a dispersed-liquid region. The void fraction is obtained from drift flux models. For heat transfer calculations, the channel is divided into regions of single-phase-liquid heat transfer, nucleate boiling and forced-convection vaporization, inverted-annular film boiling, and dispersed-flow film boiling. The heat transfer coefficients are functions of the local flow conditions. Good agreement of calculated and experimental results has been obtained. A code user's manual is appended

  6. MELCOR code modeling for APR1400

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Young; Park, S. Y.; Kim, D. H.; Ahn, K. I.; Song, Y. M.; Kim, S. D.; Park, J. H

    2001-11-01

    The severe accident phenomena of nuclear power plant have large uncertainties. For the retention of the containment integrity and improvement of nuclear reactor safety against severe accident, it is essential to understand severe accident phenomena and be able to access the accident progression accurately using computer code. Furthermore, it is important to attain a capability for developing technique and assessment tools for an advanced nuclear reactor design as well as for the severe accident prevention and mitigation. The objective of this report is to establish technical bases for an application of the MELCOR code to the Korean Next Generation Reactor (APR1400) by modeling the plant and analyzing plant steady state. This report shows the data and the input preparation for MELCOR code as well as state-state assessment results using MELCOR code.

  7. Code Team Training: Demonstrating Adherence to AHA Guidelines During Pediatric Code Blue Activations.

    Science.gov (United States)

    Stewart, Claire; Shoemaker, Jamie; Keller-Smith, Rachel; Edmunds, Katherine; Davis, Andrew; Tegtmeyer, Ken

    2017-10-16

    Pediatric code blue activations are infrequent events with a high mortality rate despite the best effort of code teams. The best method for training these code teams is debatable; however, it is clear that training is needed to assure adherence to American Heart Association (AHA) Resuscitation Guidelines and to prevent the decay that invariably occurs after Pediatric Advanced Life Support training. The objectives of this project were to train a multidisciplinary, multidepartmental code team and to measure this team's adherence to AHA guidelines during code simulation. Multidisciplinary code team training sessions were held using high-fidelity, in situ simulation. Sessions were held several times per month. Each session was filmed and reviewed for adherence to 5 AHA guidelines: chest compression rate, ventilation rate, chest compression fraction, use of a backboard, and use of a team leader. After the first study period, modifications were made to the code team including implementation of just-in-time training and alteration of the compression team. Thirty-eight sessions were completed, with 31 eligible for video analysis. During the first study period, 1 session adhered to all AHA guidelines. During the second study period, after alteration of the code team and implementation of just-in-time training, no sessions adhered to all AHA guidelines; however, there was an improvement in percentage of sessions adhering to ventilation rate and chest compression rate and an improvement in median ventilation rate. We present a method for training a large code team drawn from multiple hospital departments and a method of assessing code team performance. Despite subjective improvement in code team positioning, communication, and role completion and some improvement in ventilation rate and chest compression rate, we failed to consistently demonstrate improvement in adherence to all guidelines.

  8. Short-Term Memory Coding in Children with Intellectual Disabilities

    Science.gov (United States)

    Henry, Lucy

    2008-01-01

    To examine visual and verbal coding strategies, I asked children with intellectual disabilities and peers matched for MA and CA to perform picture memory span tasks with phonologically similar, visually similar, long, or nonsimilar named items. The CA group showed effects consistent with advanced verbal memory coding (phonological similarity and…

  9. Use of advanced simulations in fuel performance codes

    International Nuclear Information System (INIS)

    Van Uffelen, P.

    2015-01-01

    The simulation of the cylindrical fuel rod behaviour in a reactor or a storage pool for spent fuel requires a fuel performance code. Such tool solves the equations for the heat transfer, the stresses and strains in fuel and cladding, the evolution of several isotopes and the behaviour of various fission products in the fuel rod. The main equations along with their limitations are briefly described. The current approaches adopted for overcoming these limitations and the perspectives are also outlined. (author)

  10. Specialists without spirit: limitations of the mechanistic biomedical model.

    Science.gov (United States)

    Hewa, S; Hetherington, R W

    1995-06-01

    This paper examines the origin and the development of the mechanistic model of the human body and health in terms of Max Weber's theory of rationalization. It is argued that the development of Western scientific medicine is a part of the broad process of rationalization that began in sixteenth century Europe as a result of the Reformation. The development of the mechanistic view of the human body in Western medicine is consistent with the ideas of calculability, predictability, and control-the major tenets of the process of rationalization as described by Weber. In recent years, however, the limitations of the mechanistic model have been the topic of many discussions. George Engel, a leading advocate of general systems theory, is one of the leading proponents of a new medical model which includes the general quality of life, clean environment, and psychological, or spiritual stability of life. The paper concludes with consideration of the potential of Engel's proposed new model in the context of the current state of rationalization in modern industrialized society.

  11. ASTEC V2 severe accident integral code main features, current V2.0 modelling status, perspectives

    International Nuclear Information System (INIS)

    Chatelard, P.; Reinke, N.; Arndt, S.; Belon, S.; Cantrel, L.; Carenini, L.; Chevalier-Jabet, K.; Cousin, F.; Eckel, J.; Jacq, F.; Marchetto, C.; Mun, C.; Piar, L.

    2014-01-01

    The severe accident integral code ASTEC, jointly developed since almost 20 years by IRSN and GRS, simulates the behaviour of a whole nuclear power plant under severe accident conditions, including severe accident management by engineering systems and procedures. Since 2004, the ASTEC code is progressively becoming the reference European severe accident integral code through in particular the intensification of research activities carried out in the frame of the SARNET European network of excellence. The first version of the new series ASTEC V2 was released in 2009 to about 30 organizations worldwide and in particular to SARNET partners. With respect to the previous V1 series, this new V2 series includes advanced core degradation models (issued from the ICARE2 IRSN mechanistic code) and necessary extensions to be applicable to Gen. III reactor designs, notably a description of the core catcher component to simulate severe accidents transients applied to the EPR reactor. Besides these two key-evolutions, most of the other physical modules have also been improved and ASTEC V2 is now coupled to the SUNSET statistical tool to make easier the uncertainty and sensitivity analyses. The ASTEC models are today at the state of the art (in particular fission product models with respect to source term evaluation), except for quenching of a severely damage core. Beyond the need to develop an adequate model for the reflooding of a degraded core, the main other mean-term objectives are to further progress on the on-going extension of the scope of application to BWR and CANDU reactors, to spent fuel pool accidents as well as to accidents in both the ITER Fusion facility and Gen. IV reactors (in priority on sodium-cooled fast reactors) while making ASTEC evolving towards a severe accident simulator constitutes the main long-term objective. This paper presents the status of the ASTEC V2 versions, focussing on the description of V2.0 models for water-cooled nuclear plants

  12. Conceptual models for waste tank mechanistic analysis

    International Nuclear Information System (INIS)

    Allemann, R.T.; Antoniak, Z.I.; Eyler, L.L.; Liljegren, L.M.; Roberts, J.S.

    1992-02-01

    Pacific Northwest Laboratory (PNL) is conducting a study for Westinghouse Hanford Company (Westinghouse Hanford), a contractor for the US Department of Energy (DOE). The purpose of the work is to study possible mechanisms and fluid dynamics contributing to the periodic release of gases from double-shell waste storage tanks at the Hanford Site in Richland, Washington. This interim report emphasizing the modeling work follows two other interim reports, Mechanistic Analysis of Double-Shell Tank Gas Release Progress Report -- November 1990 and Collection and Analysis of Existing Data for Waste Tank Mechanistic Analysis Progress Report -- December 1990, that emphasized data correlation and mechanisms. The approach in this study has been to assemble and compile data that are pertinent to the mechanisms, analyze the data, evaluate physical properties and parameters, evaluate hypothetical mechanisms, and develop mathematical models of mechanisms

  13. Nuclear Energy Advanced Modeling and Simulation (NEAMS) Waste Integrated Performance and Safety Codes (IPSC) : FY10 development and integration.

    Energy Technology Data Exchange (ETDEWEB)

    Criscenti, Louise Jacqueline; Sassani, David Carl; Arguello, Jose Guadalupe, Jr.; Dewers, Thomas A.; Bouchard, Julie F.; Edwards, Harold Carter; Freeze, Geoffrey A.; Wang, Yifeng; Schultz, Peter Andrew

    2011-02-01

    This report describes the progress in fiscal year 2010 in developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with robust verification, validation, and software quality requirements. Waste IPSC activities in fiscal year 2010 focused on specifying a challenge problem to demonstrate proof of concept, developing a verification and validation plan, and performing an initial gap analyses to identify candidate codes and tools to support the development and integration of the Waste IPSC. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. This year-end progress report documents the FY10 status of acquisition, development, and integration of thermal-hydrologic-chemical-mechanical (THCM) code capabilities, frameworks, and enabling tools and infrastructure.

  14. Channel coding in the space station data system network

    Science.gov (United States)

    Healy, T.

    1982-01-01

    A detailed discussion of the use of channel coding for error correction, privacy/secrecy, channel separation, and synchronization is presented. Channel coding, in one form or another, is an established and common element in data systems. No analysis and design of a major new system would fail to consider ways in which channel coding could make the system more effective. The presence of channel coding on TDRS, Shuttle, the Advanced Communication Technology Satellite Program system, the JSC-proposed Space Operations Center, and the proposed 30/20 GHz Satellite Communication System strongly support the requirement for the utilization of coding for the communications channel. The designers of the space station data system have to consider the use of channel coding.

  15. Why did Jacques Monod make the choice of mechanistic determinism?

    Science.gov (United States)

    Loison, Laurent

    2015-06-01

    The development of molecular biology placed in the foreground a mechanistic and deterministic conception of the functioning of macromolecules. In this article, I show that this conception was neither obvious, nor necessary. Taking Jacques Monod as a case study, I detail the way he gradually came loose from a statistical understanding of determinism to finally support a mechanistic understanding. The reasons of the choice made by Monod at the beginning of the 1950s can be understood only in the light of the general theoretical schema supported by the concept of mechanistic determinism. This schema articulates three fundamental notions for Monod, namely that of the rigidity of the sequence of the genetic program, that of the intrinsic stability of macromolecules (DNA and proteins), and that of the specificity of molecular interactions. Copyright © 2015 Académie des sciences. Published by Elsevier SAS. All rights reserved.

  16. Advancing methods for reliably assessing motivational interviewing fidelity using the motivational interviewing skills code.

    Science.gov (United States)

    Lord, Sarah Peregrine; Can, Doğan; Yi, Michael; Marin, Rebeca; Dunn, Christopher W; Imel, Zac E; Georgiou, Panayiotis; Narayanan, Shrikanth; Steyvers, Mark; Atkins, David C

    2015-02-01

    The current paper presents novel methods for collecting MISC data and accurately assessing reliability of behavior codes at the level of the utterance. The MISC 2.1 was used to rate MI interviews from five randomized trials targeting alcohol and drug use. Sessions were coded at the utterance-level. Utterance-based coding reliability was estimated using three methods and compared to traditional reliability estimates of session tallies. Session-level reliability was generally higher compared to reliability using utterance-based codes, suggesting that typical methods for MISC reliability may be biased. These novel methods in MI fidelity data collection and reliability assessment provided rich data for therapist feedback and further analyses. Beyond implications for fidelity coding, utterance-level coding schemes may elucidate important elements in the counselor-client interaction that could inform theories of change and the practice of MI. Copyright © 2015 Elsevier Inc. All rights reserved.

  17. A comparative analysis of reactor lower head debris cooling models employed in the existing severe accident analysis codes

    International Nuclear Information System (INIS)

    Ahn, K.I.; Kim, D.H.; Kim, S.B.; Kim, H.D.

    1998-08-01

    MELCOR and MAAP4 are the representative severe accident analysis codes which have been developed for the integral analysis of the phenomenological reactor lower head corium cooling behavior. Main objectives of the present study is to identify merits and disadvantages of each relevant model through the comparative analysis of the lower plenum corium cooling models employed in these two codes. The final results will be utilized for the development of LILAC phenomenological models and for the continuous improvement of the existing MELCOR reactor lower head models, which are currently being performed at the KAERI. For these purposes, first, nine reference models are selected featuring the lower head corium behavior based on the existing experimental evidences and related models. Then main features of the selected models have been critically analyzed, and finally merits and disadvantages of each corresponding model have been summarized in the view point of realistic corium behavior and reasonable modeling. Being on these evidences, summarized and presented the potential improvements for developing more advanced models. The present study has been focused on the qualitative comparison of each model and so more detailed quantitative analysis is strongly required to obtain the final conclusions for their merits and disadvantages. In addition, in order to compensate the limitations of the current model, required further studies relating closely the detailed mechanistic models with the molten material movement and heat transfer based on phase-change in the porous medium, to the existing simple models. (author). 36 refs

  18. Mechanistic Indicators of Childhood Asthma (MICA) Study

    Science.gov (United States)

    The Mechanistic Indicators of Childhood Asthma (MICA) Study has been designed to incorporate state-of-the-art technologies to examine the physiological and environmental factors that interact to increase the risk of asthmatic responses. MICA is primarily a clinically-bases obser...

  19. Static Code Analysis with Gitlab-CI

    CERN Document Server

    Datko, Szymon Tomasz

    2016-01-01

    Static Code Analysis is a simple but efficient way to ensure that application’s source code is free from known flaws and security vulnerabilities. Although such analysis tools are often coming with more advanced code editors, there are a lot of people who prefer less complicated environments. The easiest solution would involve education – where to get and how to use the aforementioned tools. However, counting on the manual usage of such tools still does not guarantee their actual usage. On the other hand, reducing the required effort, according to the idea “setup once, use anytime without sweat” seems like a more promising approach. In this paper, the approach to automate code scanning, within the existing CERN’s Gitlab installation, is described. For realization of that project, the Gitlab-CI service (the “CI” stands for "Continuous Integration"), with Docker assistance, was employed to provide a variety of static code analysers for different programming languages. This document covers the gene...

  20. SIMULATE-3 K coupled code applications

    Energy Technology Data Exchange (ETDEWEB)

    Joensson, Christian [Studsvik Scandpower AB, Vaesteraas (Sweden); Grandi, Gerardo; Judd, Jerry [Studsvik Scandpower Inc., Idaho Falls, ID (United States)

    2017-07-15

    This paper describes the coupled code system TRACE/SIMULATE-3 K/VIPRE and the application of this code system to the OECD PWR Main Steam Line Break. A short description is given for the application of the coupled system to analyze DNBR and the flexibility the system creates for the user. This includes the possibility to compare and evaluate the result with the TRACE/SIMULATE-3K (S3K) coupled code, the S3K standalone code (core calculation) as well as performing single-channel calculations with S3K and VIPRE. This is the typical separate-effect-analyses required for advanced calculations in order to develop methodologies to be used for safety analyses in general. The models and methods of the code systems are presented. The outline represents the analysis approach starting with the coupled code system, reactor and core model calculation (TRACE/S3K). This is followed by a more detailed core evaluation (S3K standalone) and finally a very detailed thermal-hydraulic investigation of the hot pin condition (VIPRE).

  1. Further development of the computer code ATHLET-CD; Weiterentwicklung des Rechenprogramms ATHLET-CD. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Weber, Sebastian; Austregesilo, Henrique; Bals, Christine; Band, Sebastian; Hollands, Thorsten; Koellein, Carsten; Lovasz, Liviusz; Pandazis, Peter; Schubert, Johann-Dietrich; Sonnenkalb, Martin

    2016-10-15

    In the framework of the reactor safety research program sponsored by the German Federal Ministry for Economic Affairs and Energy (BMWi), the computer code system ATHLET/ATHLET-CD has been further developed as an analysis tool for the simulation of accidents in nuclear power plants with pressurized and boiling water reactors as well as for the evaluation of accident management procedures. The main objective was to provide a mechanistic analysis tool for best estimate calculations of transients, accidents, and severe accidents with core degradation in light water reactors. With the continued development, the capability of the code system has been largely improved, allowing best estimate calculations of design and beyond design base accidents, and the simulation of advanced core degradation with enhanced model extent in a reasonable calculation time. ATHLET comprises inter alia a 6-equation model, models for the simulation of non-condensable gases and tracking of boron concentration, as well as additional component and process models for the complete system simulation. Among numerous model improvements, the code application has been extended to super critical pressures. The mechanistic description of the dynamic development of flow regimes on the basis of a transport equation for the interface area has been further developed. This ATHLET version is completely integrated in ATHLET-CD. ATHLET-CD further comprises dedicated models for the simulation of fuel and control assembly degradation for both pressurized and boiling water reactors, debris bed with melting in the core region, as well as fission product and aerosol release and transport in the cooling system, inclusive of decay of nuclide inventories and of chemical reactions in the gas phase. The continued development also concerned the modelling of absorber material release, of melting, melt relocation and freezing, and the interaction with the wall of the reactor pressure vessel. The following models were newly

  2. APOLLO-2: An advanced transport code for LWRs

    International Nuclear Information System (INIS)

    Mathonniere, G.

    1995-01-01

    APOLLO-2 is a fully modular code in which each module corresponds to a specific task: access to the cross-sections libraries, creation of isotopes medium or mixtures, geometry definition, self-shielding calculations, computation of multigroup collision probabilities, flux solver, depletion calculations, transport-transport or transport-diffusion equivalence process, SN calculations, etc... Modules communicate exclusively by ''objects'' containing structured data, these objects are identified and handled by user's given names. Among the major improvements offered by APOLLO-2 the modelization of the self-shielding: it is possible now to deal with a great precision, checked versus Montecarlo calculations, a fuel rod divided into several concentric rings. So the total production of Plutonium is quite better estimated than before and its radial distribution may be predicted also with a good accuracy. Thanks to the versatility of the code some reference calculations and routine ones may be compared easily because only one parameter is changed; for example the self-shielding approximations are modified, the libraries or the flux solver being exactly the same. Other interesting features have been introduced in APOLLO-2: the new isotopes JEF.2 are available in 99 and 172 energy groups libraries, the surface leakage model improves the calculation of the control rod efficiency, the flux-current method allows faster calculations, the possibility of an automatic convergence checking during the depletion calculations coupled with fully automatic corrections, heterogeneous diffusion coefficients used for voiding analysis. 17 refs, 1 tab

  3. Montana Advanced Biofuels Great Falls Approval

    Science.gov (United States)

    This November 20, 2015 letter from EPA approves the petition from Montana Advanced Biofuels, LLC, Great Falls facility, regarding ethanol produced through a dry mill process, qualifying under the Clean Air Act for advanced biofuel (D-code 5) and renewable

  4. Incorporation of lysosomal sequestration in the mechanistic model for prediction of tissue distribution of basic drugs.

    Science.gov (United States)

    Assmus, Frauke; Houston, J Brian; Galetin, Aleksandra

    2017-11-15

    cell types. Despite this extensive lysosomal sequestration in the individual cells types, the maximal change in the overall predicted tissue Kpu was model input parameters, in particular lysosomal pH and fraction of the cellular volume occupied by the lysosomes, only partially explained discrepancies between observed and predicted Kpu data in the lung. Improved understanding of the system properties, e.g., cell/organelle composition is required to support further development of mechanistic equations for the prediction of drug tissue distribution. Application of this revised mechanistic model is recommended for prediction of Kpu in lysosome-rich tissue to facilitate the advancement of physiologically-based prediction of volume of distribution and drug exposure in the tissues. Copyright © 2017 Elsevier B.V. All rights reserved.

  5. Electrical safety code manual a plan language guide to national electrical code, OSHA and NFPA 70E

    CERN Document Server

    Keller, Kimberley

    2010-01-01

    Safety in any workplace is extremely important. In the case of the electrical industry, safety is critical and the codes and regulations which determine safe practices are both diverse and complicated. Employers, electricians, electrical system designers, inspectors, engineers and architects must comply with safety standards listed in the National Electrical Code, OSHA and NFPA 70E. Unfortunately, the publications which list these safety requirements are written in very technically advanced terms and the average person has an extremely difficult time understanding exactly what they need to

  6. Recent advances in modeling nutrient utilization in ruminants.

    Science.gov (United States)

    Kebreab, E; Dijkstra, J; Bannink, A; France, J

    2009-04-01

    Mathematical modeling techniques have been applied to study various aspects of the ruminant, such as rumen function, postabsorptive metabolism, and product composition. This review focuses on advances made in modeling rumen fermentation and its associated rumen disorders, and energy and nutrient utilization and excretion with respect to environmental issues. Accurate prediction of fermentation stoichiometry has an impact on estimating the type of energy-yielding substrate available to the animal, and the ratio of lipogenic to glucogenic VFA is an important determinant of methanogenesis. Recent advances in modeling VFA stoichiometry offer ways for dietary manipulation to shift the fermentation in favor of glucogenic VFA. Increasing energy to the animal by supplementing with starch can lead to health problems such as subacute rumen acidosis caused by rumen pH depression. Mathematical models have been developed to describe changes in rumen pH and rumen fermentation. Models that relate rumen temperature to rumen pH have also been developed and have the potential to aid in the diagnosis of subacute rumen acidosis. The effect of pH has been studied mechanistically, and in such models, fractional passage rate has a large impact on substrate degradation and microbial efficiency in the rumen and should be an important theme in future studies. The efficiency with which energy is utilized by ruminants has been updated in recent studies. Mechanistic models of N utilization indicate that reducing dietary protein concentration, matching protein degradability to the microbial requirement, and increasing the energy status of the animal will reduce the output of N as waste. Recent mechanistic P models calculate the P requirement by taking into account P recycled through saliva and endogenous losses. Mechanistic P models suggest reducing current P amounts for lactating dairy cattle to at least 0.35% P in the diet, with a potential reduction of up to 1.3 kt/yr. A model that

  7. The ZPIC educational code suite

    Science.gov (United States)

    Calado, R.; Pardal, M.; Ninhos, P.; Helm, A.; Mori, W. B.; Decyk, V. K.; Vieira, J.; Silva, L. O.; Fonseca, R. A.

    2017-10-01

    Particle-in-Cell (PIC) codes are used in almost all areas of plasma physics, such as fusion energy research, plasma accelerators, space physics, ion propulsion, and plasma processing, and many other areas. In this work, we present the ZPIC educational code suite, a new initiative to foster training in plasma physics using computer simulations. Leveraging on our expertise and experience from the development and use of the OSIRIS PIC code, we have developed a suite of 1D/2D fully relativistic electromagnetic PIC codes, as well as 1D electrostatic. These codes are self-contained and require only a standard laptop/desktop computer with a C compiler to be run. The output files are written in a new file format called ZDF that can be easily read using the supplied routines in a number of languages, such as Python, and IDL. The code suite also includes a number of example problems that can be used to illustrate several textbook and advanced plasma mechanisms, including instructions for parameter space exploration. We also invite contributions to this repository of test problems that will be made freely available to the community provided the input files comply with the format defined by the ZPIC team. The code suite is freely available and hosted on GitHub at https://github.com/zambzamb/zpic. Work partially supported by PICKSC.

  8. Implementation of advanced finite element technology in structural analysis computer codes

    International Nuclear Information System (INIS)

    Kohli, T.D.; Wiley, J.W.; Koss, P.W.

    1975-01-01

    Advances in finite element technology over the last several years have been rapid and have largely outstripped the ability of general purpose programs in the public domain to assimilate them. As a result, it has become the burden of the structural analyst to incorporate these advances himself. This paper discusses the implementation and extension of specific technological advances in Bechtel structural analysis programs. In general these advances belong in two categories: (1) the finite elements themselves and (2) equation solution algorithms. Improvements in the finite elements involve increased accuracy of the elements and extension of their applicability to various specialized modelling situations. Improvements in solution algorithms have been almost exclusively aimed at expanding problem solving capacity. (Auth.)

  9. A preliminary study of mechanistic approach in pavement design to accommodate climate change effects

    Science.gov (United States)

    Harnaeni, S. R.; Pramesti, F. P.; Budiarto, A.; Setyawan, A.

    2018-03-01

    Road damage is caused by some factors, including climate changes, overload, and inappropriate procedure for material and development process. Meanwhile, climate change is a phenomenon which cannot be avoided. The effects observed include air temperature rise, sea level rise, rainfall changes, and the intensity of extreme weather phenomena. Previous studies had shown the impacts of climate changes on road damage. Therefore, several measures to anticipate the damage should be considered during the planning and construction in order to reduce the cost of road maintenance. There are three approaches generally applied in the design of flexible pavement thickness, namely mechanistic approach, mechanistic-empirical (ME) approach and empirical approach. The advantages of applying mechanistic approach or mechanistic-empirical (ME) approaches are its efficiency and reliability in the design of flexible pavement thickness as well as its capacity to accommodate climate changes in compared to empirical approach. However, generally, the design of flexible pavement thickness in Indonesia still applies empirical approach. This preliminary study aimed to emphasize the importance of the shifting towards a mechanistic approach in the design of flexible pavement thickness.

  10. An Applied Study of Implementation of the Advanced Decommissioning Costing Methodology for Intermediate Storage Facility for Spent Fuel in Studsvik, Sweden with special emphasis to the application of the Omega code

    Energy Technology Data Exchange (ETDEWEB)

    Kristofova, Kristina; Vasko, Marek; Daniska, Vladimir; Ondra, Frantisek; Bezak, Peter [DECOM Slovakia, spol. s.r.o., J. Bottu 2, SK-917 01 Trnava (Slovakia); Lindskog, Staffan [Swedish Nuclear Power Inspectorate, Stockholm (Sweden)

    2007-01-15

    The presented study is focused on an analysis of decommissioning costs for the Intermediate Storage Facility for Spent Fuel (FA) facility in Studsvik prepared by SVAFO and a proposal of the advanced decommissioning costing methodology application. Therefore, this applied study concentrates particularly in the following areas: 1. Analysis of FA facility cost estimates prepared by SVAFO including description of FA facility in Studsvik, summarised input data, applied cost estimates methodology and summarised results from SVAFO study. 2. Discussion of results of the SVAFO analysis, proposals for enhanced cost estimating methodology and upgraded structure of inputs/outputs for decommissioning study for FA facility. 3. Review of costing methodologies with the special emphasis on the advanced costing methodology and cost calculation code OMEGA. 4. Discussion on implementation of the advanced costing methodology for FA facility in Studsvik together with: - identification of areas of implementation; - analyses of local decommissioning infrastructure; - adaptation of the data for the calculation database; - inventory database; and - implementation of the style of work with the computer code OMEGA.

  11. An Applied Study of Implementation of the Advanced Decommissioning Costing Methodology for Intermediate Storage Facility for Spent Fuel in Studsvik, Sweden with special emphasis to the application of the Omega code

    International Nuclear Information System (INIS)

    Kristofova, Kristina; Vasko, Marek; Daniska, Vladimir; Ondra, Frantisek; Bezak, Peter; Lindskog, Staffan

    2007-01-01

    The presented study is focused on an analysis of decommissioning costs for the Intermediate Storage Facility for Spent Fuel (FA) facility in Studsvik prepared by SVAFO and a proposal of the advanced decommissioning costing methodology application. Therefore, this applied study concentrates particularly in the following areas: 1. Analysis of FA facility cost estimates prepared by SVAFO including description of FA facility in Studsvik, summarised input data, applied cost estimates methodology and summarised results from SVAFO study. 2. Discussion of results of the SVAFO analysis, proposals for enhanced cost estimating methodology and upgraded structure of inputs/outputs for decommissioning study for FA facility. 3. Review of costing methodologies with the special emphasis on the advanced costing methodology and cost calculation code OMEGA. 4. Discussion on implementation of the advanced costing methodology for FA facility in Studsvik together with: - identification of areas of implementation; - analyses of local decommissioning infrastructure; - adaptation of the data for the calculation database; - inventory database; and - implementation of the style of work with the computer code OMEGA

  12. Short-term memory coding in children with intellectual disabilities.

    Science.gov (United States)

    Henry, Lucy

    2008-05-01

    To examine visual and verbal coding strategies, I asked children with intellectual disabilities and peers matched for MA and CA to perform picture memory span tasks with phonologically similar, visually similar, long, or nonsimilar named items. The CA group showed effects consistent with advanced verbal memory coding (phonological similarity and word length effects). Neither the intellectual disabilities nor MA groups showed evidence for memory coding strategies. However, children in these groups with MAs above 6 years showed significant visual similarity and word length effects, broadly consistent with an intermediate stage of dual visual and verbal coding. These results suggest that developmental progressions in memory coding strategies are independent of intellectual disabilities status and consistent with MA.

  13. Esophageal function testing: Billing and coding update.

    Science.gov (United States)

    Khan, A; Massey, B; Rao, S; Pandolfino, J

    2018-01-01

    Esophageal function testing is being increasingly utilized in diagnosis and management of esophageal disorders. There have been several recent technological advances in the field to allow practitioners the ability to more accurately assess and treat such conditions, but there has been a relative lack of education in the literature regarding the associated Common Procedural Terminology (CPT) codes and methods of reimbursement. This review, commissioned and supported by the American Neurogastroenterology and Motility Society Council, aims to summarize each of the CPT codes for esophageal function testing and show the trends of associated reimbursement, as well as recommend coding methods in a practical context. We also aim to encourage many of these codes to be reviewed on a gastrointestinal (GI) societal level, by providing evidence of both discrepancies in coding definitions and inadequate reimbursement in this new era of esophageal function testing. © 2017 John Wiley & Sons Ltd.

  14. Mechanistic aspects of ionic reactions in flames

    DEFF Research Database (Denmark)

    Egsgaard, H.; Carlsen, L.

    1993-01-01

    Some fundamentals of the ion chemistry of flames are summarized. Mechanistic aspects of ionic reactions in flames have been studied using a VG PlasmaQuad, the ICP-system being substituted by a simple quartz burner. Simple hydrocarbon flames as well as sulfur-containing flames have been investigated...

  15. Joint ICTP-IAEA advanced workshop on model codes for spallation reactions

    International Nuclear Information System (INIS)

    Filges, D.; Leray, S.; Yariv, Y.; Mengoni, A.; Stanculescu, A.; Mank, G.

    2008-08-01

    The International Atomic Energy Agency (IAEA) and the Abdus Salam International Centre for Theoretical Physics (ICTP) organised an expert meeting at the ICTP from 4 to 8 February 2008 to discuss model codes for spallation reactions. These nuclear reactions play an important role in a wide domain of applications ranging from neutron sources for condensed matter and material studies, transmutation of nuclear waste and rare isotope production to astrophysics, simulation of detector set-ups in nuclear and particle physics experiments, and radiation protection near accelerators or in space. The simulation tools developed for these domains use nuclear model codes to compute the production yields and characteristics of all the particles and nuclei generated in these reactions. These codes are generally Monte-Carlo implementations of Intra-Nuclear Cascade (INC) or Quantum Molecular Dynamics (QMD) models, followed by de-excitation (principally evaporation/fission) models. Experts have discussed in depth the physics contained within the different models in order to understand their strengths and weaknesses. Such codes need to be validated against experimental data in order to determine their accuracy and reliability with respect to all forms of application. Agreement was reached during the course of the workshop to organise an international benchmark of the different models developed by different groups around the world. The specifications of the benchmark, including the set of selected experimental data to be compared to the models, were also defined during the workshop. The benchmark will be organised under the auspices of the IAEA in 2008, and the first results will be discussed at the next Accelerator Applications Conference (AccApp'09) to be held in Vienna in May 2009. (author)

  16. Energy Efficiency Program Administrators and Building Energy Codes

    Science.gov (United States)

    Explore how energy efficiency program administrators have helped advance building energy codes at federal, state, and local levels—using technical, institutional, financial, and other resources—and discusses potential next steps.

  17. Advanced training simulator models. Implementation and validation

    International Nuclear Information System (INIS)

    Borkowsky, Jeffrey; Judd, Jerry; Belblidia, Lotfi; O'farrell, David; Andersen, Peter

    2008-01-01

    Modern training simulators are required to replicate plant data for both thermal-hydraulic and neutronic response. Replication is required such that reactivity manipulation on the simulator properly trains the operator for reactivity manipulation at the plant. This paper discusses advanced models which perform this function in real-time using the coupled code system THOR/S3R. This code system models the all fluids systems in detail using an advanced, two-phase thermal-hydraulic a model. The nuclear core is modeled using an advanced, three-dimensional nodal method and also by using cycle-specific nuclear data. These models are configured to run interactively from a graphical instructor station or handware operation panels. The simulator models are theoretically rigorous and are expected to replicate the physics of the plant. However, to verify replication, the models must be independently assessed. Plant data is the preferred validation method, but plant data is often not available for many important training scenarios. In the absence of data, validation may be obtained by slower-than-real-time transient analysis. This analysis can be performed by coupling a safety analysis code and a core design code. Such a coupling exists between the codes RELAP5 and SIMULATE-3K (S3K). RELAP5/S3K is used to validate the real-time model for several postulated plant events. (author)

  18. REVA Advanced Fuel Design and Codes and Methods - Increasing Reliability, Operating Margin and Efficiency in Operation

    Energy Technology Data Exchange (ETDEWEB)

    Frichet, A.; Mollard, P.; Gentet, G.; Lippert, H. J.; Curva-Tivig, F.; Cole, S.; Garner, N.

    2014-07-01

    Since three decades, AREVA has been incrementally implementing upgrades in the BWR and PWR Fuel design and codes and methods leading to an ever greater fuel efficiency and easier licensing. For PWRs, AREVA is implementing upgraded versions of its HTP{sup T}M and AFA 3G technologies called HTP{sup T}M-I and AFA3G-I. These fuel assemblies feature improved robustness and dimensional stability through the ultimate optimization of their hold down system, the use of Q12, the AREVA advanced quaternary alloy for guide tube, the increase in their wall thickness and the stiffening of the spacer to guide tube connection. But an even bigger step forward has been achieved a s AREVA has successfully developed and introduces to the market the GAIA product which maintains the resistance to grid to rod fretting (GTRF) of the HTP{sup T}M product while providing addition al thermal-hydraulic margin and high resistance to Fuel Assembly bow. (Author)

  19. Breathing (and Coding?) a Bit Easier: Changes to International Classification of Disease Coding for Pulmonary Hypertension.

    Science.gov (United States)

    Mathai, Stephen C; Mathew, Sherin

    2018-04-20

    International Classification of Disease (ICD) coding system is broadly utilized by healthcare providers, hospitals, healthcare payers, and governments to track health trends and statistics at the global, national, and local levels and to provide a reimbursement framework for medical care based upon diagnosis and severity of illness. The current iteration of the ICD system, ICD-10, was implemented in 2015. While many changes to the prior ICD-9 system were included in the ICD-10 system, the newer revision failed to adequately reflect advances in the clinical classification of certain diseases such as pulmonary hypertension (PH). Recently, a proposal to modify the ICD-10 codes for PH was considered and ultimately adopted for inclusion as updates to ICD-10 coding system. While these revisions better reflect the current clinical classification of PH, in the future, further changes should be considered to improve the accuracy and ease of coding for all forms of PH. Copyright © 2018. Published by Elsevier Inc.

  20. The MELTSPREAD Code for Modeling of Ex-Vessel Core Debris Spreading Behavior, Code Manual – Version3-beta

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-09-01

    MELTSPREAD3 is a transient one-dimensional computer code that has been developed to predict the gravity-driven flow and freezing behavior of molten reactor core materials (corium) in containment geometries. Predictions can be made for corium flowing across surfaces under either dry or wet cavity conditions. The spreading surfaces that can be selected are steel, concrete, a user-specified material (e.g., a ceramic), or an arbitrary combination thereof. The corium can have a wide range of compositions of reactor core materials that includes distinct oxide phases (predominantly Zr, and steel oxides) plus metallic phases (predominantly Zr and steel). The code requires input that describes the containment geometry, melt “pour” conditions, and cavity atmospheric conditions (i.e., pressure, temperature, and cavity flooding information). For cases in which the cavity contains a preexisting water layer at the time of RPV failure, melt jet breakup and particle bed formation can be calculated mechanistically given the time-dependent melt pour conditions (input data) as well as the heatup and boiloff of water in the melt impingement zone (calculated). For core debris impacting either the containment floor or previously spread material, the code calculates the transient hydrodynamics and heat transfer which determine the spreading and freezing behavior of the melt. The code predicts conditions at the end of the spreading stage, including melt relocation distance, depth and material composition profiles, substrate ablation profile, and wall heatup. Code output can be used as input to other models such as CORQUENCH that evaluate long term core-concrete interaction behavior following the transient spreading stage. MELTSPREAD3 was originally developed to investigate BWR Mark I liner vulnerability, but has been substantially upgraded and applied to other reactor designs (e.g., the EPR), and more recently to the plant accidents at Fukushima Daiichi. The most recent round of

  1. Development of a domestically-made system code

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    According to lessons learned from the Fukushima-Daiichi NPP accidents, a new safety standard based on state-of-the-art findings has been established by the Japanese Nuclear Regulation Authority (NRA) and will soon come into force in Japan. In order to ensure a precise response to this movement from a technological point of view, it should be required for safety regulation to develop a new system code with much smaller uncertainty and reinforced simulation capability even in application to beyond-DBAs (BDBAs), as well as with the capability of close coupling to a newly developing severe accident code. Accordingly, development of a new domestically-made system code that incorporates 3-dimensional and 3 or more fluid thermal-hydraulics in tandem with a 3-dimensional neutronics has been started in 2012. In 2012, two branches of development activities, the development of 'main body' and advanced features have been started in parallel for development efficiency. The main body has been started from scratch and the following activities have therefore been performed: 1) development and determination of key principles and methodologies to realize a flexible, extensible and robust platform, 2) determination of requirements definition, 3) start of basic program design and coding and 4) start of a development of prototypical GUI-based pre-post processor. As for the advanced features, the following activities have been performed: 1) development of Phenomena Identification and Ranking Tables (PIRTs) and model capability matrix from normal operations to BDBAs in order to address requirements definition for advanced modeling, 2) development of detailed action plan for modification of field equations, numerical schemes and solvers and 3) start of the program development of field equations with an interfacial area concentration transport equation, a robust solver for condensation induced water hammer phenomena and a versatile Newton-Raphson solver. (author)

  2. Recent Advances in the Mechanistic Studies of Alkylaromatic Conversions over Zeolite Catalysts

    International Nuclear Information System (INIS)

    Min, Hyung-Ki; Hong, Suk Bong

    2013-01-01

    The transformation of alkylaromatic hydrocarbons using zeolite catalysts play big part in the current petrochemical industry. Here we review recent advances in the understanding of the reaction mechanisms of various alkylaromatic conversions with respect to the structural and physicochemical properties of zeolite catalysts employed. Indeed, the shape-selective nature of zeolite catalysts determines the type of reaction intermediates and hence the prevailing reaction mechanism together with the product distribution. The prospect of zeolite catalysis in the development of more efficient petrochemical processes is also described

  3. A New Monte Carlo Neutron Transport Code at UNIST

    International Nuclear Information System (INIS)

    Lee, Hyunsuk; Kong, Chidong; Lee, Deokjung

    2014-01-01

    Monte Carlo neutron transport code named MCS is under development at UNIST for the advanced reactor design and research purpose. This MC code can be used for fixed source calculation and criticality calculation. Continuous energy neutron cross section data and multi-group cross section data can be used for the MC calculation. This paper presents the overview of developed MC code and its calculation results. The real time fixed source calculation ability is also tested in this paper. The calculation results show good agreement with commercial code and experiment. A new Monte Carlo neutron transport code is being developed at UNIST. The MC codes are tested with several benchmark problems: ICSBEP, VENUS-2, and Hoogenboom-Martin benchmark. These benchmarks covers pin geometry to 3-dimensional whole core, and results shows good agreement with reference results

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

    Science.gov (United States)

    Blyth, Taylor S.

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Blyth, Taylor S. [Pennsylvania State Univ., University Park, PA (United States); Avramova, Maria [North Carolina State Univ., Raleigh, NC (United States)

    2017-04-01

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

  6. Mechanistic approach to the sodium leakage and fire analysis

    International Nuclear Information System (INIS)

    Yamaguchi, Akira; Muramatsu, Toshiharu; Ohira, Hiroaki; Ida, Masao

    1997-04-01

    In December 1995, a thermocouple well was broken and liquid sodium leaked out of the intermediate heat transport system of the prototype fast breeder reactor Monju. In the initiating process of the incident, liquid sodium flowed out through the hollow thermocouple well, nipple and connector. As a result, liquid sodium, following ignition and combustion, was dropping from the connector to colide with the duct and grating placed below. The collision may cause fragmentation and scattering of the sodium droplet that finally was piled up on the floor. This report deals with the development of computer programs for the phenomena based on mechanistics approach. Numerical analyses are also made for fundamental sodium leakage and combustion phenomenon, sodium combustion experiment, and Monju incident condition. The contents of this report is listed below: (1) Analysis of chemical reaction process based on molecular orbital method, (2) Thermalhy draulic analysis of the sodium combustion experiment II performed in 1996 at O-arai Engineering Center, PNC, (3) Thermalhy draulic analysis of room A-446 of Monju reactor when the sodium leakage took place, (4) Direct numerical simulation of sodium droplet, (5) Sodium leakage and scattering analysis using three dimensional particle method, (6) Multi-dimensional combustion analysis and multi-point approximation combustion analysis code. Subsequent to the development work of the programs, they are to be applied to the safety analysis of the Fast Breeder Reactor. (author)

  7. A Mechanistic Source Term Calculation for a Metal Fuel Sodium Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David; Bucknor, Matthew; Jerden, James

    2017-06-26

    A mechanistic source term (MST) calculation attempts to realistically assess the transport and release of radionuclides from a reactor system to the environment during a specific accident sequence. The U.S. Nuclear Regulatory Commission (NRC) has repeatedly stated its expectation that advanced reactor vendors will utilize an MST during the U.S. reactor licensing process. As part of a project to examine possible impediments to sodium fast reactor (SFR) licensing in the U.S., an analysis was conducted regarding the current capabilities to perform an MST for a metal fuel SFR. The purpose of the project was to identify and prioritize any gaps in current computational tools, and the associated database, for the accurate assessment of an MST. The results of the study demonstrate that an SFR MST is possible with current tools and data, but several gaps exist that may lead to possibly unacceptable levels of uncertainty, depending on the goals of the MST analysis.

  8. Advancing population ecology with integral projection models: a practical guide

    DEFF Research Database (Denmark)

    Merow, Cory; Dahlgren, Johan; Metcall, C. Jessica E.

    2014-01-01

    (e.g., environment). By combining regressions of vital rates, an IPM provides mechanistic insight into emergent ecological patterns such as population dynamics, species geographic distributions, or life history strategies. Here, we review important resources for building IPMs and provide...... a comprehensive guide, with extensive R code, for their construction. IPMs can be applied to any stage-structured population; here we illustrate IPMs for a series of plant life histories of increasing complexity and biological realism, highlighting the utility of various regression methods for capturing...

  9. Mechanistic Fermentation Models for Process Design, Monitoring, and Control

    DEFF Research Database (Denmark)

    Mears, Lisa; Stocks, Stuart M.; Albæk, Mads Orla

    2017-01-01

    Mechanistic models require a significant investment of time and resources, but their application to multiple stages of fermentation process development and operation can make this investment highly valuable. This Opinion article discusses how an established fermentation model may be adapted...... for application to different stages of fermentation process development: planning, process design, monitoring, and control. Although a longer development time is required for such modeling methods in comparison to purely data-based model techniques, the wide range of applications makes them a highly valuable tool...... for fermentation research and development. In addition, in a research environment, where collaboration is important, developing mechanistic models provides a platform for knowledge sharing and consolidation of existing process understanding....

  10. Advances in In Vitro and In Silico Tools for Toxicokinetic Dose Modeling and Predictive Toxicology (WC10)

    Science.gov (United States)

    Recent advances in vitro assays, in silico tools, and systems biology approaches provide opportunities for refined mechanistic understanding for chemical safety assessment that will ultimately lead to reduced reliance on animal-based methods. With the U.S. commercial chemical lan...

  11. Programming peptidomimetic syntheses by translating genetic codes designed de novo.

    Science.gov (United States)

    Forster, Anthony C; Tan, Zhongping; Nalam, Madhavi N L; Lin, Hening; Qu, Hui; Cornish, Virginia W; Blacklow, Stephen C

    2003-05-27

    Although the universal genetic code exhibits only minor variations in nature, Francis Crick proposed in 1955 that "the adaptor hypothesis allows one to construct, in theory, codes of bewildering variety." The existing code has been expanded to enable incorporation of a variety of unnatural amino acids at one or two nonadjacent sites within a protein by using nonsense or frameshift suppressor aminoacyl-tRNAs (aa-tRNAs) as adaptors. However, the suppressor strategy is inherently limited by compatibility with only a small subset of codons, by the ways such codons can be combined, and by variation in the efficiency of incorporation. Here, by preventing competing reactions with aa-tRNA synthetases, aa-tRNAs, and release factors during translation and by using nonsuppressor aa-tRNA substrates, we realize a potentially generalizable approach for template-encoded polymer synthesis that unmasks the substantially broader versatility of the core translation apparatus as a catalyst. We show that several adjacent, arbitrarily chosen sense codons can be completely reassigned to various unnatural amino acids according to de novo genetic codes by translating mRNAs into specific peptide analog polymers (peptidomimetics). Unnatural aa-tRNA substrates do not uniformly function as well as natural substrates, revealing important recognition elements for the translation apparatus. Genetic programming of peptidomimetic synthesis should facilitate mechanistic studies of translation and may ultimately enable the directed evolution of small molecules with desirable catalytic or pharmacological properties.

  12. NEAMS FPL M2 Milestone Report: Development of a UO₂ Grain Size Model using Multicale Modeling and Simulation

    Energy Technology Data Exchange (ETDEWEB)

    Tonks, Michael R [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Yongfeng [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bai, Xianming [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-06-01

    This report summarizes development work funded by the Nuclear Energy Advanced Modeling Simulation program's Fuels Product Line (FPL) to develop a mechanistic model for the average grain size in UO₂ fuel. The model is developed using a multiscale modeling and simulation approach involving atomistic simulations, as well as mesoscale simulations using INL's MARMOT code.

  13. Development of a nuclear power plant system analysis code

    International Nuclear Information System (INIS)

    Sim, Suk K.; Jeong, J. J.; Ha, K. S.; Moon, S. K.; Park, J. W.; Yang, S. K.; Song, C. H.; Chun, S. Y.; Kim, H. C.; Chung, B. D.; Lee, W. J.; Kwon, T. S.

    1997-07-01

    During the period of this study, TASS 1.0 code has been prepared for the non-LOCA licensing and reload safety analyses of the Westinghouse and the Korean Standard Nuclear Power Plants (KSNPP) type reactors operating in Korea. TASS-NPA also has been developed for a real time simulation of the Kori-3/4 transients using on-line graphical interactions. TASS 2.0 code has been further developed to timely apply the TASS 2.0 code for the design certification of the KNGR. The COBRA/RELAP5 code, a multi-dimensional best estimate system code, has been developed by integrating the realistic three-dimensional reactor vessel model with the RELAP5 /MOD3.2 code, a one-dimensional system code. Also, a 3D turbulent two-phase flow analysis code, FEMOTH-TF, has been developed using finite element technique to analyze local thermal hydraulic phenomena in support of the detailed design analysis for the development of the advanced reactors. (author). 84 refs., 27 tabs., 83 figs

  14. SSC-K code users manual (rev.1)

    International Nuclear Information System (INIS)

    Kwon, Y. M.; Lee, Y. B.; Chang, W. P.; Hahn, D.

    2002-01-01

    The Supper System Code of KAERI (SSC-K) is a best-estimate system code for analyzing a variety of off-normal or accidents in the heat transport system of a pool type LMR design. It is being developed at Korea Atomic Energy Research Institution (KAERI) on the basis of SSC-L, originally developed at BNL to analyze loop-type LMR transients. SSC-K can handle both designs of loop and pool type LMRs. SSC-K contains detailed mechanistic models of transient thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, coolant, fuel elements, and structures to accident conditions. This report provides a revised User's Manual (rev.1) of the SSC-K computer code, focusing on phenomenological model descriptions for new thermal, hydraulic, neutronic, and mechanical modules. A comprehensive description of the models for pool-type reactor is given in Chapters 2 and 3; the steady-state plant characterization, prior to the initiation of transient is described in Chapter 2 and their transient counterparts are discussed in Chapter 3. Discussions on the intermediate heat exchanger (IHX) and the electromagnetic (EM) pump are described in Chapter 4 and 5, respectively. A model of passive safety decay heat removal system (PSDRS) is discussed in Chapter 6, and models for various reactivity feedback effects are discussed in Chapter 7. In Chapter 8, constitutive laws and correlations required to execute the SSC-K are described. New models developed for SSC-K rev.1 are two dimensional hot pool model in Chapter 9, and long term cooling model in Chapter 10. Finally, a brief description of MINET code adopted to simulate BOP is presented in Chapter 11. Based on test runs for typical LMFBR accident analyses, it was found that the present version of SSC-K would be used for the safety analysis of KALIMER. However, the further validation of SSC-K is required for real applications. It is noted that the user's manual of SSC-K will be revised later with the

  15. Chemical kinetic mechanistic models to investigate cancer biology and impact cancer medicine

    International Nuclear Information System (INIS)

    Stites, Edward C

    2013-01-01

    Traditional experimental biology has provided a mechanistic understanding of cancer in which the malignancy develops through the acquisition of mutations that disrupt cellular processes. Several drugs developed to target such mutations have now demonstrated clinical value. These advances are unequivocal testaments to the value of traditional cellular and molecular biology. However, several features of cancer may limit the pace of progress that can be made with established experimental approaches alone. The mutated genes (and resultant mutant proteins) function within large biochemical networks. Biochemical networks typically have a large number of component molecules and are characterized by a large number of quantitative properties. Responses to a stimulus or perturbation are typically nonlinear and can display qualitative changes that depend upon the specific values of variable system properties. Features such as these can complicate the interpretation of experimental data and the formulation of logical hypotheses that drive further research. Mathematical models based upon the molecular reactions that define these networks combined with computational studies have the potential to deal with these obstacles and to enable currently available information to be more completely utilized. Many of the pressing problems in cancer biology and cancer medicine may benefit from a mathematical treatment. As work in this area advances, one can envision a future where such models may meaningfully contribute to the clinical management of cancer patients. (paper)

  16. International Accreditation of ASME Codes and Standards

    International Nuclear Information System (INIS)

    Green, Mervin R.

    1989-01-01

    ASME established a Boiler Code Committee to develop rules for the design, fabrication and inspection of boilers. This year we recognize 75 years of that Code and will publish a history of that 75 years. The first Code and subsequent editions provided for a Code Symbol Stamp or mark which could be affixed by a manufacturer to a newly constructed product to certify that the manufacturer had designed, fabricated and had inspected it in accordance with Code requirements. The purpose of the ASME Mark is to identify those boilers that meet ASME Boiler and Pressure Vessel Code requirements. Through thousands of updates over the years, the Code has been revised to reflect technological advances and changing safety needs. Its scope has been broadened from boilers to include pressure vessels, nuclear components and systems. Proposed revisions to the Code are published for public review and comment four times per year and revisions and interpretations are published annually; it's a living and constantly evolving Code. You and your organizations are a vital part of the feedback system that keeps the Code alive. Because of this dynamic Code, we no longer have columns in newspapers listing boiler explosions. Nevertheless, it has been argued recently that ASME should go further in internationalizing its Code. Specifically, representatives of several countries, have suggested that ASME delegate to them responsibility for Code implementation within their national boundaries. The question is, thus, posed: Has the time come to franchise responsibility for administration of ASME's Code accreditation programs to foreign entities or, perhaps, 'institutes.' And if so, how should this be accomplished?

  17. Review and assessment of thermodynamic and transport properties for the CONTAIN Code

    International Nuclear Information System (INIS)

    Valdez, G.D.

    1988-12-01

    A study was carried out to review available data and correlations on the thermodynamic and transport properties of materials applicable to the CONTAIN computer code. CONTAIN is the NRC's best-estimate, mechanistic computer code for modeling containment response to a severe accident. Where appropriate, recommendations have been made for suitable approximations for material properties of interests. Based on a modified Benedict-Webb-Rubin (BWR) equation of state, a procedure is introduced for calculating thermodynamic properties for common gases in the CONTAIN code. These gases are nitrogen, oxygen, hydrogen, carbon dioxide, carbon monoxide, steam, helium, and argon. The thermodynamic equations for density, currently represented in CONTAIN by relatively simple fits, were independently checked and are recommended to be replaced by the Lee-Kesler equation of state which substantially improves accuracy without too much sacrifice in computational efficiency. The accuracy of the calculated values have been found to be generally acceptable. Various correlations and models for single component gas transport properties, viscosity and thermal conductivity, were also assessed with available experimental data. When a suitable correlation or model was not available, transport properties were obtained by performing least-squares fit on experimental data. 50 refs., 126 figs., 3 tabs

  18. Long Non-Coding RNAs: The Key Players in Glioma Pathogenesis

    Energy Technology Data Exchange (ETDEWEB)

    Kiang, Karrie Mei-Yee; Zhang, Xiao-Qin; Leung, Gilberto Ka-Kit, E-mail: gilberto@hku.hk [Department of Surgery, Li Ka Shing Faculty of Medicine, The University of Hong Kong, Queen Mary Hospital, Hong Kong (China)

    2015-07-29

    Long non-coding RNAs (LncRNAs) represent a novel class of RNAs with no functional protein-coding ability, yet it has become increasingly clear that interactions between lncRNAs with other molecules are responsible for important gene regulatory functions in various contexts. Given their relatively high expressions in the brain, lncRNAs are now thought to play important roles in normal brain development as well as diverse disease processes including gliomagenesis. Intriguingly, certain lncRNAs are closely associated with the initiation, differentiation, progression, recurrence and stem-like characteristics in glioma, and may therefore be exploited for the purposes of sub-classification, diagnosis and prognosis. LncRNAs may also serve as potential therapeutic targets as well as a novel biomarkers in the treatment of glioma. In this article, the functional aspects of lncRNAs, particularly within the central nervous system (CNS), will be briefly discussed, followed by highlights of the important roles of lncRNAs in mediating critical steps during glioma development. In addition, the key lncRNA players and their possible mechanistic pathways associated with gliomagenesis will be addressed.

  19. Long Non-Coding RNAs: The Key Players in Glioma Pathogenesis

    International Nuclear Information System (INIS)

    Kiang, Karrie Mei-Yee; Zhang, Xiao-Qin; Leung, Gilberto Ka-Kit

    2015-01-01

    Long non-coding RNAs (LncRNAs) represent a novel class of RNAs with no functional protein-coding ability, yet it has become increasingly clear that interactions between lncRNAs with other molecules are responsible for important gene regulatory functions in various contexts. Given their relatively high expressions in the brain, lncRNAs are now thought to play important roles in normal brain development as well as diverse disease processes including gliomagenesis. Intriguingly, certain lncRNAs are closely associated with the initiation, differentiation, progression, recurrence and stem-like characteristics in glioma, and may therefore be exploited for the purposes of sub-classification, diagnosis and prognosis. LncRNAs may also serve as potential therapeutic targets as well as a novel biomarkers in the treatment of glioma. In this article, the functional aspects of lncRNAs, particularly within the central nervous system (CNS), will be briefly discussed, followed by highlights of the important roles of lncRNAs in mediating critical steps during glioma development. In addition, the key lncRNA players and their possible mechanistic pathways associated with gliomagenesis will be addressed

  20. CONTAIN calculations of direct containment heating in the Surry plant

    International Nuclear Information System (INIS)

    Williams, D.C.; Louie, D.L.Y.

    1988-01-01

    The draft NUREG-1150 risk analysis performed for the Surry plant identified direct containment heating (DCH) as a potentially dominant contributor to the total public risk associated with this plant. At that time, however, detailed mechanistic calculations of DCH loads were unavailable. Subsequently, a series of analyses of DCH scenarios using the CONTAIN-DCH code was performed in order to put the treatment of DCH on a firmer basis in the final draft of NUREG-1150. The present paper describes some of the results obtained for the Surry plant. A developmental model for DCH has been incorporated into CONTAIN code. This model includes mechanistic treatments of reasonably well-understood phenomena (e.g., heat and mass transfer), together with a parametric treatment of poorly understood phenomena for which mechanistic models are unavailable (e.g., debris de-entrainment from the gas stream due to debris-structure interactions). The DCH model was described in an earlier report, but the present version incorporates a number of advances, including treatment of the chemical equilibria involved in the iron-steam reaction

  1. COSY INFINITY, a new arbitrary order optics code

    International Nuclear Information System (INIS)

    Berz, M.

    1990-01-01

    The new arbitrary order particle optics and beam dynamics code COSY INFINITY is presented. The code is based on differential algebraic (DA) methods. COSY INFINITY has a full structured object oriented language environment. This provides a simple interface for the casual or novice user. At the same time, it offers the advanced user a very flexible and powerful tool for the utilization of DA. The power and generality of the environment is perhaps best demonstrated by the fact that the physics routines of COSY INFINITY are written in its own input language. The approach also facilitates the implementation of new features because special code generated by a user can be readily adopted to the source code. Besides being very compact in size, the code is also very fast, thanks to efficiently programmed elementary DA operations. For simple low order problems, which can be handled by conventional codes, the speed of COSY INFINITY is comparable and in certain cases even higher

  2. The sphere-PAC fuel code 'SPHERE-3'

    Energy Technology Data Exchange (ETDEWEB)

    Wallin, H

    2000-07-01

    Sphere-PAC fuel is an advanced nuclear fuel, in which the cladding tube is filled with small fuel spheres instead of the more usual fuel pellets. At PSI, the irradiation behaviour of sphere-PAC fuel is calculated using the computer code SPHERE-3. The paper describes the present status of the SPHERE-3 code, and some results of the qualification process against experimental data. (author)

  3. IFR code for secondary particle dynamics

    International Nuclear Information System (INIS)

    Teague, M.R.; Yu, S.S.

    1985-01-01

    A numerical simulation has been constructed to obtain a detailed, quantitative estimate of the electromagnetic fields and currents existing in the Advanced Test Accelerator under conditions of laser guiding. The code treats the secondary electrons by particle simulation and the beam dynamics by a time-dependent envelope model. The simulation gives a fully relativistic description of secondary electrons moving in self-consistent electromagnetic fields. The calculations are made using coordinates t, x, y, z for the electrons and t, ct-z, r for the axisymmetric electromagnetic fields and currents. Code results, showing in particular current enhancement effects, will be given

  4. Electromagnetic reprogrammable coding-metasurface holograms.

    Science.gov (United States)

    Li, Lianlin; Jun Cui, Tie; Ji, Wei; Liu, Shuo; Ding, Jun; Wan, Xiang; Bo Li, Yun; Jiang, Menghua; Qiu, Cheng-Wei; Zhang, Shuang

    2017-08-04

    Metasurfaces have enabled a plethora of emerging functions within an ultrathin dimension, paving way towards flat and highly integrated photonic devices. Despite the rapid progress in this area, simultaneous realization of reconfigurability, high efficiency, and full control over the phase and amplitude of scattered light is posing a great challenge. Here, we try to tackle this challenge by introducing the concept of a reprogrammable hologram based on 1-bit coding metasurfaces. The state of each unit cell of the coding metasurface can be switched between '1' and '0' by electrically controlling the loaded diodes. Our proof-of-concept experiments show that multiple desired holographic images can be realized in real time with only a single coding metasurface. The proposed reprogrammable hologram may be a key in enabling future intelligent devices with reconfigurable and programmable functionalities that may lead to advances in a variety of applications such as microscopy, display, security, data storage, and information processing.Realizing metasurfaces with reconfigurability, high efficiency, and control over phase and amplitude is a challenge. Here, Li et al. introduce a reprogrammable hologram based on a 1-bit coding metasurface, where the state of each unit cell of the coding metasurface can be switched electrically.

  5. Descriptive and mechanistic models of crop–weed competition

    NARCIS (Netherlands)

    Bastiaans, L.; Storkey, J.

    2017-01-01

    Crop-weed competitive relations are an important element of agroecosystems. Quantifying and understanding them helps to design appropriate weed management at operational, tactical and strategic level. This chapter presents and discusses simple descriptive and more mechanistic models for crop-weed

  6. Analysis of the three dimensional core kinetics NESTLE code coupling with the advanced thermo-hydraulic code systems, RELAP5/SCDAPSIM and its application to the Laguna Verde Central reactor

    International Nuclear Information System (INIS)

    Salazar C, J.H.; Nunez C, A.; Chavez M, C.

    2004-01-01

    The objective of the written present is to propose a methodology for the joining of the codes RELAP5/SCDAPSIM and NESTLE. The development of this joining will be carried out inside a doctoral program of Engineering in Energy with nuclear profile of the Ability of Engineering of the UNAM together with the National Commission of Nuclear Security and Safeguards (CNSNS). The general purpose of this type of developments, is to have tools that are implemented by multiple programs or codes such a that systems or models of the three-dimensional kinetics of the core can be simulated and those of the dynamics of the reactor (water heater-hydraulics). In the past, by limitations for the calculation of the complete answer of both systems, the developed models they were carried out for separate, putting a lot of emphasis in one but neglecting the other one. These methodologies, calls of better estimate, will be good to the nuclear industry to evaluate, with more high grades of detail, the designs of the nuclear power plant (for modifications to those already existent or for new concepts in the designs of advanced reactors), besides analysing events (transitory and have an accident), among other applications. The coupled system was applied to design studies and investigation of the Laguna Verde Nuclear power plant (CNLV). (Author)

  7. Mechanistic insight into neurotoxicity induced by developmental insults

    International Nuclear Information System (INIS)

    Tamm, Christoffer; Ceccatelli, Sandra

    2017-01-01

    Epidemiological and/or experimental studies have shown that unfavorable prenatal environmental factors, such as stress or exposure to certain neurotoxic environmental contaminants, may have adverse consequences for neurodevelopment. Alterations in neurogenesis can have harmful effects not only for the developing nervous system, but also for the adult brain where neurogenesis is believed to play a role in learning, memory, and even in depression. Many recent advances in the understanding of the complex process of nervous system development can be integrated into the field of neurotoxicology. In the past 15 years we have been using cultured neural stem or progenitor cells to investigate the effects of neurotoxic stimuli on cell survival, proliferation and differentiation, with special focus on heritable effects. This is an overview of the work performed by our group in the attempt to elucidate the mechanisms of developmental neurotoxicity and possibly provide relevant information for the understanding of the etiopathogenesis of complex brain disorders. - Highlights: • The developing nervous system is highly sensitive to toxic insults. • Neural stem cells are relevant models for mechanistic studies as well as for identifying heritable effects due to epigenetic changes. • Depending on the dose, the outcome of exposure to neurotoxicants ranges from altered proliferation and differentiation to cell death. • The elucidation of neurotoxicity mechanisms is relevant for understanding the etiopathogenesis of developmental and adult nervous system disorders.

  8. Precision and accuracy of mechanistic-empirical pavement design

    CSIR Research Space (South Africa)

    Theyse, HL

    2006-09-01

    Full Text Available are discussed in general. The effects of variability and error on the design accuracy and design risk are lastly illustrated at the hand of a simple mechanistic-empirical design problem, showing that the engineering models alone determine the accuracy...

  9. The spectral code Apollo2: from lattice to 2D core calculations

    International Nuclear Information System (INIS)

    Coste-Delclaux, M.; Santandrea, S.; Damian, F.; Blanc-Tranchant, P.; Zmijarevic, I.; Santamarina, A.

    2005-01-01

    Apollo2 is a powerful code dedicated to neutron transport, it is a highly qualified tool for a wide range of applications from research and development studies to industrial applications. Today Apollo2 is part of several advanced 3-dimensional nuclear code packages dedicated to reactor physics, fuel cycle, criticality and safety analysis. The presentations have been organized into 7 topics: -) an introduction to Apollo2, -) cross-sections, -) flux calculation, -) advanced applications, -) Apollo2 users, specialized packages, -) qualification program, and -) the future of Apollo2. This document gathers only the slides of the presentations

  10. The spectral code Apollo2: from lattice to 2D core calculations

    Energy Technology Data Exchange (ETDEWEB)

    Coste-Delclaux, M.; Santandrea, S.; Damian, F.; Blanc-Tranchant, P.; Zmijarevic, I. [CEA Saclay (DEN/DANS/SERMA), 91 - Gif-sur-Yvette (France); Santamarina, A. [CEA Cadarache (CEA/DEN/DER/SPRC), 13 - Saint Paul lez Durance (France)

    2005-07-01

    Apollo2 is a powerful code dedicated to neutron transport, it is a highly qualified tool for a wide range of applications from research and development studies to industrial applications. Today Apollo2 is part of several advanced 3-dimensional nuclear code packages dedicated to reactor physics, fuel cycle, criticality and safety analysis. The presentations have been organized into 7 topics: -) an introduction to Apollo2, -) cross-sections, -) flux calculation, -) advanced applications, -) Apollo2 users, specialized packages, -) qualification program, and -) the future of Apollo2. This document gathers only the slides of the presentations.

  11. Mechanistic Models for Process Development and Optimization of Fed-batch Fermentation Systems

    DEFF Research Database (Denmark)

    Mears, Lisa; Stocks, Stuart M.; Albæk, Mads O.

    2016-01-01

    This work discusses the application of mechanistic models to pilot scale filamentous fungal fermentation systems operated at Novozymes A/S. For on-line applications, a state estimator model is developed based on a stoichiometric balance in order to predict the biomass and product concentration....... This is based on on-line gas measurements and ammonia addition flow rate measurements. Additionally, a mechanistic model is applied offline as a tool for batch planning, based on definition of the process back pressure, aeration rate and stirrer speed. This allows the batch starting fill to be planned, taking...... into account the oxygen transfer conditions, as well as the evaporation rates of the system. Mechanistic models are valuable tools which are applicable for both process development and optimization. The state estimator described will be a valuable tool for future work as part of control strategy development...

  12. Digital color acquisition, perception, coding and rendering

    CERN Document Server

    Fernandez-Maloigne, Christine; Macaire, Ludovic

    2013-01-01

    In this book the authors identify the basic concepts and recent advances in the acquisition, perception, coding and rendering of color. The fundamental aspects related to the science of colorimetry in relation to physiology (the human visual system) are addressed, as are constancy and color appearance. It also addresses the more technical aspects related to sensors and the color management screen. Particular attention is paid to the notion of color rendering in computer graphics. Beyond color, the authors also look at coding, compression, protection and quality of color images and videos.

  13. Incorporation of advanced accident analysis methodology into safety analysis reports

    International Nuclear Information System (INIS)

    2003-05-01

    The IAEA Safety Guide on Safety Assessment and Verification defines that the aim of the safety analysis should be by means of appropriate analytical tools to establish and confirm the design basis for the items important to safety, and to ensure that the overall plant design is capable of meeting the prescribed and acceptable limits for radiation doses and releases for each plant condition category. Practical guidance on how to perform accident analyses of nuclear power plants (NPPs) is provided by the IAEA Safety Report on Accident Analysis for Nuclear Power Plants. The safety analyses are performed both in the form of deterministic and probabilistic analyses for NPPs. It is customary to refer to deterministic safety analyses as accident analyses. This report discusses the aspects of using the advanced accident analysis methods to carry out accident analyses in order to introduce them into the Safety Analysis Reports (SARs). In relation to the SAR, purposes of deterministic safety analysis can be further specified as (1) to demonstrate compliance with specific regulatory acceptance criteria; (2) to complement other analyses and evaluations in defining a complete set of design and operating requirements; (3) to identify and quantify limiting safety system set points and limiting conditions for operation to be used in the NPP limits and conditions; (4) to justify appropriateness of the technical solutions employed in the fulfillment of predetermined safety requirements. The essential parts of accident analyses are performed by applying sophisticated computer code packages, which have been specifically developed for this purpose. These code packages include mainly thermal-hydraulic system codes and reactor dynamics codes meant for the transient and accident analyses. There are also specific codes such as those for the containment thermal-hydraulics, for the radiological consequences and for severe accident analyses. In some cases, codes of a more general nature such

  14. Advanced GF(32) nonbinary LDPC coded modulation with non-uniform 9-QAM outperforming star 8-QAM.

    Science.gov (United States)

    Liu, Tao; Lin, Changyu; Djordjevic, Ivan B

    2016-06-27

    In this paper, we first describe a 9-symbol non-uniform signaling scheme based on Huffman code, in which different symbols are transmitted with different probabilities. By using the Huffman procedure, prefix code is designed to approach the optimal performance. Then, we introduce an algorithm to determine the optimal signal constellation sets for our proposed non-uniform scheme with the criterion of maximizing constellation figure of merit (CFM). The proposed nonuniform polarization multiplexed signaling 9-QAM scheme has the same spectral efficiency as the conventional 8-QAM. Additionally, we propose a specially designed GF(32) nonbinary quasi-cyclic LDPC code for the coded modulation system based on the 9-QAM non-uniform scheme. Further, we study the efficiency of our proposed non-uniform 9-QAM, combined with nonbinary LDPC coding, and demonstrate by Monte Carlo simulation that the proposed GF(23) nonbinary LDPC coded 9-QAM scheme outperforms nonbinary LDPC coded uniform 8-QAM by at least 0.8dB.

  15. Development and validation of a fuel performance analysis code

    International Nuclear Information System (INIS)

    Majalee, Aaditya V.; Chaturvedi, S.

    2015-01-01

    CAD has been developing a computer code 'FRAVIZ' for calculation of steady-state thermomechanical behaviour of nuclear reactor fuel rods. It contains four major modules viz., Thermal module, Fission Gas Release module, Material Properties module and Mechanical module. All these four modules are coupled to each other and feedback from each module is fed back to others to get a self-consistent evolution in time. The computer code has been checked against two FUMEX benchmarks. Modelling fuel performance in Advance Heavy Water Reactor would require additional inputs related to the fuel and some modification in the code.(author)

  16. ORMEC: a three-dimensional MHD spectral inverse equilibrium code

    International Nuclear Information System (INIS)

    Hirshman, S.P.; Hogan, J.T.

    1986-02-01

    The Oak Ridge Moments Equilibrium Code (ORMEC) is an efficient computer code that has been developed to calculate three-dimensional MHD equilibria using the inverse spectral method. The fixed boundary formulation, which is based on a variational principle for the spectral coefficients (moments) of the cylindrical coordinates R and Z, is described and compared with the finite difference code BETA developed by Bauer, Betancourt, and Garabedian. Calculations for the Heliotron, Wendelstein VIIA, and Advanced Toroidal Facility (ATF) configurations are performed to establish the accuracy and mesh convergence properties for the spectral method. 16 refs., 13 figs

  17. Development and Implementation of Mechanistic Terry Turbine Models in RELAP-7 to Simulate RCIC Normal Operation Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Haihua [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zou, Ling [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); O' Brien, James Edward [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    As part of the efforts to understand the unexpected “self-regulating” mode of the RCIC (Reactor Core Isolation Cooling) systems in Fukushima accidents and extend BWR RCIC and PWR AFW (Auxiliary Feed Water) operational range and flexibility, mechanistic models for the Terry turbine, based on Sandia’s original work [1], have been developed and implemented in the RELAP-7 code to simulate the RCIC system. In 2016, our effort has been focused on normal working conditions of the RCIC system. More complex off-design conditions will be pursued in later years when more data are available. In the Sandia model, the turbine stator inlet velocity is provided according to a reduced-order model which was obtained from a large number of CFD (computational fluid dynamics) simulations. In this work, we propose an alternative method, using an under-expanded jet model to obtain the velocity and thermodynamic conditions for the turbine stator inlet. The models include both an adiabatic expansion process inside the nozzle and a free expansion process outside of the nozzle to ambient pressure. The combined models are able to predict the steam mass flow rate and supersonic velocity to the Terry turbine bucket entrance, which are the necessary input information for the Terry turbine rotor model. The analytical models for the nozzle were validated with experimental data and benchmarked with CFD simulations. The analytical models generally agree well with the experimental data and CFD simulations. The analytical models are suitable for implementation into a reactor system analysis code or severe accident code as part of mechanistic and dynamical models to understand the RCIC behaviors. The newly developed nozzle models and modified turbine rotor model according to the Sandia’s original work have been implemented into RELAP-7, along with the original Sandia Terry turbine model. A new pump model has also been developed and implemented to couple with the Terry turbine model. An input

  18. Advanced C and C++ compiling

    CERN Document Server

    Stevanovic, Milan

    2014-01-01

    Learning how to write C/C++ code is only the first step. To be a serious programmer, you need to understand the structure and purpose of the binary files produced by the compiler: object files, static libraries, shared libraries, and, of course, executables.Advanced C and C++ Compiling explains the build process in detail and shows how to integrate code from other developers in the form of deployed libraries as well as how to resolve issues and potential mismatches between your own and external code trees.With the proliferation of open source, understanding these issues is increasingly the res

  19. Optimal codes as Tanner codes with cyclic component codes

    DEFF Research Database (Denmark)

    Høholdt, Tom; Pinero, Fernando; Zeng, Peng

    2014-01-01

    In this article we study a class of graph codes with cyclic code component codes as affine variety codes. Within this class of Tanner codes we find some optimal binary codes. We use a particular subgraph of the point-line incidence plane of A(2,q) as the Tanner graph, and we are able to describe ...

  20. Relevant thermal hydraulic aspects of advanced reactors design: status report

    International Nuclear Information System (INIS)

    1996-11-01

    This status report provides an overview on the relevant thermalhydraulic aspects of advanced reactor designs (e.g. ABWR, AP600, SBWR, EPR, ABB 80+, PIUS, etc.). Since all of the advanced reactor concepts are at the design stage, the information and data available in the open literature are still very limited. Some characteristics of advanced reactor designs are provided together with selected phenomena identification and ranking tables. Specific needs for thermalhydraulic codes together with the list of relevant and important thermalhydraulic phenomena for advanced reactor designs are summarized with the purpose of providing some guidance in development of research plans for considering further code development and assessment needs and for the planning of experimental programs

  1. Advances in the development of interaction between the codes MCNPX and ANSYS Fluent and their fusion applications; Avances en el desarrollo de la interaccion entre los codigos MCNPX y ANSYS Fluente y sus aplicaciones para fusion

    Energy Technology Data Exchange (ETDEWEB)

    Colomer, C.; Salellas, J.; Ahmed, R.; Fabbrio, M.; Aleman, A.

    2012-07-01

    The advances are presented in the project for the development of a code of interaction between MCNPX y el ANSYS Fluent. Following the flow of the work carried out during the development of the project will study of the most appropriate remeshing algorithms between both codes. In addition explain the selection and implementation of methods to verify internally the correct transmission of the variables involved between both nets. Finally the selection of cases for verification and validation of the interaction between both codes in each of the possible fields of application will be exposed.

  2. Development of realistic thermal-hydraulic system analysis codes ; development of thermal hydraulic test requirements for multidimensional flow modeling

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Kune Yull; Yoon, Sang Hyuk; Noh, Sang Woo; Lee, Il Suk [Seoul National University, Seoul (Korea)

    2002-03-01

    This study is concerned with developing a multidimensional flow model required for the system analysis code MARS to more mechanistically simulate a variety of thermal hydraulic phenomena in the nuclear stem supply system. The capability of the MARS code as a thermal hydraulic analysis tool for optimized system design can be expanded by improving the current calculational methods and adding new models. In this study the relevant literature was surveyed on the multidimensional flow models that may potentially be applied to the multidimensional analysis code. Research items were critically reviewed and suggested to better predict the multidimensional thermal hydraulic behavior and to identify test requirements. A small-scale preliminary test was performed in the downcomer formed by two vertical plates to analyze multidimensional flow pattern in a simple geometry. The experimental result may be applied to the code for analysis of the fluid impingement to the reactor downcomer wall. Also, data were collected to find out the controlling parameters for the one-dimensional and multidimensional flow behavior. 22 refs., 40 figs., 7 tabs. (Author)

  3. CONTAIN 2.0 code release and the transition to licensing

    International Nuclear Information System (INIS)

    Murata, K.K.; Griffith, R.O.; Bergeron, K.D.; Tills, J.

    1997-10-01

    CONTAIN is a reactor accident simulation code developed by Sandia National Laboratories under US Nuclear Regulatory Commission (USNRC) sponsorship to provide integrated analysis of containment phenomena, including those related to nuclear reactor containment loads and radiological source terms. The recently released CONTAIN 2.0 code version represents a significant advance in CONTAIN modeling capabilities over the last major code release (CONTAIN 1.12V). The new modeling capabilities are discussed here. The principal motivation for many of the recent model improvements has been to allow CONTAIN to model the special features in advanced light water reactor (ALWR) designs. The work done in this area is also summarized. In addition to the ALWR work, the USNRC is currently engaged in an effort to qualify CONTAIN for more general use in licensing, with the intent of supplementing or possibly replacing traditional licensing codes. To qualify the CONTAIN code for licensing applications, studies utilizing CONTAIN 2.0 are in progress. A number of results from this effort are presented in this paper to illustrate the code capabilities. In particular, CONTAIN calculations of the NUPEC M-8-1 and ISP-23 experiments and CVTR test number-sign 3 are presented to illustrate (1) the ability of CONTAIN to model non-uniform gas density and/or temperature distributions, and (2) the relationship between such gas distributions and containment loads. CONTAIN and CONTEMPT predictions for a large break loss of coolant accident scenario in the San Onofre plant are also compared

  4. Temporal Coding of Volumetric Imagery

    Science.gov (United States)

    Llull, Patrick Ryan

    'Image volumes' refer to realizations of images in other dimensions such as time, spectrum, and focus. Recent advances in scientific, medical, and consumer applications demand improvements in image volume capture. Though image volume acquisition continues to advance, it maintains the same sampling mechanisms that have been used for decades; every voxel must be scanned and is presumed independent of its neighbors. Under these conditions, improving performance comes at the cost of increased system complexity, data rates, and power consumption. This dissertation explores systems and methods capable of efficiently improving sensitivity and performance for image volume cameras, and specifically proposes several sampling strategies that utilize temporal coding to improve imaging system performance and enhance our awareness for a variety of dynamic applications. Video cameras and camcorders sample the video volume (x,y,t) at fixed intervals to gain understanding of the volume's temporal evolution. Conventionally, one must reduce the spatial resolution to increase the framerate of such cameras. Using temporal coding via physical translation of an optical element known as a coded aperture, the compressive temporal imaging (CACTI) camera emonstrates a method which which to embed the temporal dimension of the video volume into spatial (x,y) measurements, thereby greatly improving temporal resolution with minimal loss of spatial resolution. This technique, which is among a family of compressive sampling strategies developed at Duke University, temporally codes the exposure readout functions at the pixel level. Since video cameras nominally integrate the remaining image volume dimensions (e.g. spectrum and focus) at capture time, spectral (x,y,t,lambda) and focal (x,y,t,z) image volumes are traditionally captured via sequential changes to the spectral and focal state of the system, respectively. The CACTI camera's ability to embed video volumes into images leads to exploration

  5. Optics code development at Los Alamos

    International Nuclear Information System (INIS)

    Mottershead, C.T.; Lysenko, W.P.

    1988-01-01

    This paper is an overview of part of the beam optics code development effort in the Accelerator Technology Division at Los Alamos National Laboratory. The aim of this effort is to improve our capability to design advanced beam optics systems. The work reported is being carried out by a collaboration of permanent staff members, visiting consultants, and student research assistants. The main components of the effort are: building a new framework of common supporting utilities and software tools to facilitate further development; research and development on basic computational techniques in classical mechanics and electrodynamics; and evaluation and comparison of existing beam optics codes, and support for their continuing development. 17 refs

  6. Optics code development at Los Alamos

    International Nuclear Information System (INIS)

    Mottershead, C.T.; Lysenko, W.P.

    1988-01-01

    This paper is an overview of part of the beam optics code development effort in the Accelerator Technology Division at Los Alamos National Laboratory. The aim of this effort is to improve our capability to design advanced beam optics systems. The work reported is being carried out by a collaboration of permanent staff members, visiting consultants, and student research assistants. The main components of the effort are building a new framework of common supporting utilities and software tools to facilitate further development. research and development on basic computational techniques in classical mechanics and electrodynamics, and evaluation and comparison of existing beam optics codes, and support for their continuing development

  7. Issues affecting advanced passive light-water reactor safety analysis

    International Nuclear Information System (INIS)

    Beelman, R.J.; Fletcher, C.D.; Modro, S.M.

    1992-01-01

    Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented

  8. Mechanistic model for microbial growth on hydrocarbons

    Energy Technology Data Exchange (ETDEWEB)

    Mallee, F M; Blanch, H W

    1977-12-01

    Based on available information describing the transport and consumption of insoluble alkanes, a mechanistic model is proposed for microbial growth on hydrocarbons. The model describes the atypical growth kinetics observed, and has implications in the design of large scale equipment for single cell protein (SCP) manufacture from hydrocarbons. The model presents a framework for comparison of the previously published experimental kinetic data.

  9. The FORTRAN NALAP code adapted to a microcomputer compiler

    International Nuclear Information System (INIS)

    Lobo, Paulo David de Castro; Borges, Eduardo Madeira; Braz Filho, Francisco Antonio; Guimaraes, Lamartine Nogueira Frutuoso

    2010-01-01

    The Nuclear Energy Division of the Institute for Advanced Studies (IEAv) is conducting the TERRA project (TEcnologia de Reatores Rapidos Avancados), Technology for Advanced Fast Reactors project, aimed at a space reactor application. In this work, to attend the TERRA project, the NALAP code adapted to a microcomputer compiler called Compaq Visual Fortran (Version 6.6) is presented. This code, adapted from the light water reactor transient code RELAP 3B, simulates thermal-hydraulic responses for sodium cooled fast reactors. The strategy to run the code in a PC was divided in some steps mainly to remove unnecessary routines, to eliminate old statements, to introduce new ones and also to include extension precision mode. The source program was able to solve three sample cases under conditions of protected transients suggested in literature: the normal reactor shutdown, with a delay of 200 ms to start the control rod movement and a delay of 500 ms to stop the pumps; reactor scram after transient of loss of flow; and transients protected from overpower. Comparisons were made with results from the time when the NALAP code was acquired by the IEAv, back in the 80's. All the responses for these three simulations reproduced the calculations performed with the CDC compiler in 1985. Further modifications will include the usage of gas as coolant for the nuclear reactor to allow a Closed Brayton Cycle Loop - CBCL - to be used as a heat/electric converter. (author)

  10. The FORTRAN NALAP code adapted to a microcomputer compiler

    Energy Technology Data Exchange (ETDEWEB)

    Lobo, Paulo David de Castro; Borges, Eduardo Madeira; Braz Filho, Francisco Antonio; Guimaraes, Lamartine Nogueira Frutuoso, E-mail: plobo.a@uol.com.b, E-mail: eduardo@ieav.cta.b, E-mail: fbraz@ieav.cta.b, E-mail: guimarae@ieav.cta.b [Instituto de Estudos Avancados (IEAv/CTA), Sao Jose dos Campos, SP (Brazil)

    2010-07-01

    The Nuclear Energy Division of the Institute for Advanced Studies (IEAv) is conducting the TERRA project (TEcnologia de Reatores Rapidos Avancados), Technology for Advanced Fast Reactors project, aimed at a space reactor application. In this work, to attend the TERRA project, the NALAP code adapted to a microcomputer compiler called Compaq Visual Fortran (Version 6.6) is presented. This code, adapted from the light water reactor transient code RELAP 3B, simulates thermal-hydraulic responses for sodium cooled fast reactors. The strategy to run the code in a PC was divided in some steps mainly to remove unnecessary routines, to eliminate old statements, to introduce new ones and also to include extension precision mode. The source program was able to solve three sample cases under conditions of protected transients suggested in literature: the normal reactor shutdown, with a delay of 200 ms to start the control rod movement and a delay of 500 ms to stop the pumps; reactor scram after transient of loss of flow; and transients protected from overpower. Comparisons were made with results from the time when the NALAP code was acquired by the IEAv, back in the 80's. All the responses for these three simulations reproduced the calculations performed with the CDC compiler in 1985. Further modifications will include the usage of gas as coolant for the nuclear reactor to allow a Closed Brayton Cycle Loop - CBCL - to be used as a heat/electric converter. (author)

  11. Timing comparison of two-dimensional discrete-ordinates codes for criticality calculations

    International Nuclear Information System (INIS)

    Miller, W.F. Jr.; Alcouffe, R.E.; Bosler, G.E.; Brinkley, F.W. Jr.; O'dell, R.D.

    1979-01-01

    The authors compare two-dimensional discrete-ordinates neutron transport computer codes to solve reactor criticality problems. The fundamental interest is in determining which code requires the minimum Central Processing Unit (CPU) time for a given numerical model of a reasonably realistic fast reactor core and peripherals. The computer codes considered are the most advanced available and, in three cases, are not officially released. The conclusion, based on the study of four fast reactor core models, is that for this class of problems the diffusion synthetic accelerated version of TWOTRAN, labeled TWOTRAN-DA, is superior to the other codes in terms of CPU requirements

  12. Light water reactor fuel analysis code FEMAXI-7. Model and structure [Revised edition

    International Nuclear Information System (INIS)

    Suzuki, Motoe; Udagawa, Yutaka; Amaya, Masaki; Saitou, Hiroaki

    2014-03-01

    A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in both normal conditions and anticipated transient conditions. This code is an advanced version which has been produced by incorporating the former version FEMAXI-6 with numerous functional improvements and extensions. In FEMAXI-7, many new models have been added and parameters have been clearly arranged. Also, to facilitate effective maintenance and accessibility of the code, modularization of subroutines and functions have been attained, and quality comment descriptions of variables or physical quantities have been incorporated in the source code. With these advancements, the FEMAXI-7 code has been upgraded to a versatile analytical tool for high burnup fuel behavior analyses. This report is the revised edition of the first one which describes in detail the design, basic theory and structure, models and numerical method, and improvements and extensions. The first edition, JAEA-Data/Code 2010-035, was published in 2010. The first edition was extended by orderly addition and disposition of explanations of models and organized as the revised edition after three years interval. (author)

  13. Introduction and immigration of TRAC-PF1 code

    International Nuclear Information System (INIS)

    Yan Yuhua; Gao Zuying; Gao Cheng; Li Jincai

    1997-01-01

    TRAC-PF1 code performs best-estimate predictions of postulated accidents for pressurized light water reactors. It is one of the few system analyse codes which use two fluid model to treat two phase problems in nuclear system. In order to use this advanced software in China and make it possible to be run in different compute systems, IBM version of TRAC-PF1 code, imported from USA National Energy Software Center, is immigrated to CDC NOS/VE system and SUN workstation. The differences in computer languages from IBM 370 to CDC NOS/VE and to SUN workstation are modified properly. All the benchmark problems are calculated, and the results show that the immigration is successful

  14. Proceedings of the 8th topical meeting on nuclear code development

    International Nuclear Information System (INIS)

    1993-03-01

    The 8th Topical Meeting on Nuclear Code Development, organized by Committee on Reactor Physics and Nuclear Codes Committee of Japan Atomic Energy Research Institute (JAERI), was held at Tokai Research Establishment of JAERI, on 11th and 12th of November, 1992. In the meeting, 14 papers were presented on the topics of (1) the next generation nuclear reactor design system and (2) advances of the nuclear fuel reprocessing safety analysis codes. These papers are compiled in this proceedings. (author)

  15. Benchmarking Severe Accident Computer Codes for Heavy Water Reactor Applications

    International Nuclear Information System (INIS)

    2013-12-01

    Requests for severe accident investigations and assurance of mitigation measures have increased for operating nuclear power plants and the design of advanced nuclear power plants. Severe accident analysis investigations necessitate the analysis of the very complex physical phenomena that occur sequentially during various stages of accident progression. Computer codes are essential tools for understanding how the reactor and its containment might respond under severe accident conditions. The IAEA organizes coordinated research projects (CRPs) to facilitate technology development through international collaboration among Member States. The CRP on Benchmarking Severe Accident Computer Codes for HWR Applications was planned on the advice and with the support of the IAEA Nuclear Energy Department's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). This publication summarizes the results from the CRP participants. The CRP promoted international collaboration among Member States to improve the phenomenological understanding of severe core damage accidents and the capability to analyse them. The CRP scope included the identification and selection of a severe accident sequence, selection of appropriate geometrical and boundary conditions, conduct of benchmark analyses, comparison of the results of all code outputs, evaluation of the capabilities of computer codes to predict important severe accident phenomena, and the proposal of necessary code improvements and/or new experiments to reduce uncertainties. Seven institutes from five countries with HWRs participated in this CRP

  16. Numerical simulation in steam injection process by a mechanistic approach

    Energy Technology Data Exchange (ETDEWEB)

    De Souza, J.C.Jr.; Campos, W.; Lopes, D.; Moura, L.S.S. [Petrobras, Rio de Janeiro (Brazil)

    2008-10-15

    Steam injection is a common thermal recovery method used in very viscous oil reservoirs. The method involves the injection of heat to reduce viscosity and mobilize oil. A steam generation and injection system consists primarily of a steam source, distribution lines, injection wells and a discarding tank. In order to optimize injection and improve the oil recovery factor, one must determine the parameters of steam flow such as pressure, temperature and steam quality. This study focused on developing a unified mathematical model by means of a mechanistic approach for two-phase steam flow in pipelines and wells. The hydrodynamic and heat transfer mechanistic model was implemented in a computer simulator to model the parameters of steam injection while trying to avoid the use of empirical correlations. A marching algorithm was used to determine the distribution of pressure and temperature along the pipelines and wellbores. The mathematical model for steam flow in injection systems, developed by a mechanistic approach (VapMec) performed well when the simulated values of pressures and temperatures were compared with the values measured during field tests. The newly developed VapMec model was incorporated in the LinVap-3 simulator that constitutes an engineering supporting tool for steam injection wells operated by Petrobras. 23 refs., 7 tabs., 6 figs.

  17. Final Report for Project 13-4791: New Mechanistic Models of Creep-Fatigue Crack Growth Interactions for Advanced High Temperature Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Kruzic, Jamie J [Oregon State Univ., Corvallis, OR (United States); Siegmund, Thomas [Purdue Univ., West Lafayette, IN (United States); Tomar, Vikas [Purdue Univ., West Lafayette, IN (United States)

    2018-03-20

    This project developed and validated a novel, multi-scale, mechanism-based model to quantitatively predict creep-fatigue crack growth and failure for Ni-based Alloy 617 at 800°C. Alloy 617 is a target material for intermediate heat exchangers in Generation IV very high temperature reactor designs, and it is envisioned that this model will aid in the design of safe, long lasting nuclear power plants. The technical effectiveness of the model was shown by demonstrating that experimentally observed crack growth rates can be predicted under both steady state and overload crack growth conditions. Feasibility was considered by incorporating our model into a commercially available finite element method code, ABAQUS, that is commonly used by design engineers. While the focus of the project was specifically on an alloy targeted for Generation IV nuclear reactors, the benefits to the public are expected to be wide reaching. Indeed, creep-fatigue failure is a design consideration for a wide range of high temperature mechanical systems that rely on Ni-based alloys, including industrial gas power turbines, advanced ultra-super critical steam turbines, and aerospace turbine engines. It is envisioned that this new model can be adapted to a wide range of engineering applications.

  18. Identification and Analysis of Critical Gaps in Nuclear Fuel Cycle Codes Required by the SINEMA Program

    International Nuclear Information System (INIS)

    Miron, Adrian; Valentine, Joshua; Christenson, John; Hawwari, Majd; Bhatt, Santosh; Dunzik-Gougar, Mary Lou; Lineberry, Michael

    2009-01-01

    The current state of the art in nuclear fuel cycle (NFC) modeling is an eclectic mixture of codes with various levels of applicability, flexibility, and availability. In support of the advanced fuel cycle systems analyses, especially those by the Advanced Fuel Cycle Initiative (AFCI), University of Cincinnati in collaboration with Idaho State University carried out a detailed review of the existing codes describing various aspects of the nuclear fuel cycle and identified the research and development needs required for a comprehensive model of the global nuclear energy infrastructure and the associated nuclear fuel cycles. Relevant information obtained on the NFC codes was compiled into a relational database that allows easy access to various codes' properties. Additionally, the research analyzed the gaps in the NFC computer codes with respect to their potential integration into programs that perform comprehensive NFC analysis.

  19. Identification and Analysis of Critical Gaps in Nuclear Fuel Cycle Codes Required by the SINEMA Program

    Energy Technology Data Exchange (ETDEWEB)

    Adrian Miron; Joshua Valentine; John Christenson; Majd Hawwari; Santosh Bhatt; Mary Lou Dunzik-Gougar: Michael Lineberry

    2009-10-01

    The current state of the art in nuclear fuel cycle (NFC) modeling is an eclectic mixture of codes with various levels of applicability, flexibility, and availability. In support of the advanced fuel cycle systems analyses, especially those by the Advanced Fuel Cycle Initiative (AFCI), Unviery of Cincinnati in collaboration with Idaho State University carried out a detailed review of the existing codes describing various aspects of the nuclear fuel cycle and identified the research and development needs required for a comprehensive model of the global nuclear energy infrastructure and the associated nuclear fuel cycles. Relevant information obtained on the NFC codes was compiled into a relational database that allows easy access to various codes' properties. Additionally, the research analyzed the gaps in the NFC computer codes with respect to their potential integration into programs that perform comprehensive NFC analysis.

  20. Upgrade and benchmarking of the NIFS physics-engineering-cost code

    International Nuclear Information System (INIS)

    Dolan, T.J.; Yamazaki, K.

    2004-07-01

    The NIFS Physics-Engineering-Cost (PEC) code for helical and tokamak fusion reactors is upgraded by adding data from three blanket-shield designs, a new cost section based on the ARIES cost schedule, more recent unit costs, and improved algorithms for various computations. The PEC code is also benchmarked by modeling the ARIES-AT (advanced technology) tokamak and the ARIES-SPPS (stellarator power plant system). The PEC code succeeds in predicting many of the pertinent plasma parameters and reactor component masses within about 10%. There are cost differences greater than 10% for some fusion power core components, which may be attributed to differences of unit costs used by the codes. The COEs estimated by the PEC code differ from the COEs of the ARIES-AT and ARIES-SPPS studies by 5%. (author)

  1. SIMMER-III code-verification. Phase 1

    International Nuclear Information System (INIS)

    Maschek, W.

    1996-05-01

    SIMMER-III is a computer code to investigate core disruptive accidents in liquid metal fast reactors but should also be used to investigate safety related problems in other types of advanced reactors. The code is developed by PNC with cooperation of the European partners FZK, CEA and AEA-T. SIMMER-III is a two-dimensional, three-velocity-field, multiphase, multicomponent, Eulerian, fluid-dynamics code coupled with a space-, time-, and energy-dependent neutron dynamics model. In order to model complex flow situations in a postulated disrupting core, mass and energy conservation equations are solved for 27 density components and 16 energy components, respectively. Three velocity fields (two liquid and one vapor) are modeled to simulate the relative motion of different fluid components. An additional static field takes into account the structures available in a reactor (pins, hexans, vessel structures, internal structures etc.). The neutronics is based on the discrete ordinate method (S N method) coupled into a quasistatic dynamic model. The code assessment and verification of the fluid dynamic/thermohydraulic parts of the code is performed in several steps in a joint effort of all partners. The results of the FZK contributions to the first assessment and verification phase is reported. (orig.) [de

  2. Code of conduct for scientists (abstract)

    International Nuclear Information System (INIS)

    Khurshid, S.J.

    2011-01-01

    The emergence of advanced technologies in the last three decades and extraordinary progress in our knowledge on the basic Physical, Chemical and Biological properties of living matter has offered tremendous benefits to human beings but simultaneously highlighted the need of higher awareness and responsibility by the scientists of 21 century. Scientist is not born with ethics, nor science is ethically neutral, but there are ethical dimensions to scientific work. There is need to evolve an appropriate Code of Conduct for scientist particularly working in every field of Science. However, while considering the contents, promulgation and adaptation of Codes of Conduct for Scientists, a balance is needed to be maintained between freedom of scientists and at the same time some binding on them in the form of Code of Conducts. The use of good and safe laboratory procedures, whether, codified by law or by common practice must also be considered as part of the moral duties of scientists. It is internationally agreed that a general Code of Conduct can't be formulated for all the scientists universally, but there should be a set of 'building blocks' aimed at establishing the Code of Conduct for Scientists either as individual researcher or responsible for direction, evaluation, monitoring of scientific activities at the institutional or organizational level. (author)

  3. The mechanistic bases of the power-time relationship

    DEFF Research Database (Denmark)

    Vanhatalo, Anni; Black, Matthew I; DiMenna, Fred J

    2016-01-01

    .025) and inversely correlated with muscle type IIx fibre proportion (r = -0.76, P = 0.01). There was no relationship between W' (19.4 ± 6.3 kJ) and muscle fibre type. These data indicate a mechanistic link between the bioenergetic characteristics of different muscle fibre types and the power-duration relationship...

  4. Mechanistic studies of carbon monoxide reduction

    Energy Technology Data Exchange (ETDEWEB)

    Geoffroy, G.L.

    1990-06-12

    The progress made during the current grant period (1 January 1988--1 April 1990) in three different areas of research is summarized. The research areas are: (1) oxidatively-induced double carbonylation reactions to form {alpha}-ketoacyl complexes and studies of the reactivity of the resulting compounds, (2) mechanistic studies of the carbonylation of nitroaromatics to form isocyanates, carbamates, and ureas, and (3) studies of the formation and reactivity of unusual metallacycles and alkylidene ligands supported on binuclear iron carbonyl fragments. 18 refs., 5 figs., 1 tab.

  5. The Accurate Particle Tracer Code

    OpenAIRE

    Wang, Yulei; Liu, Jian; Qin, Hong; Yu, Zhi

    2016-01-01

    The Accurate Particle Tracer (APT) code is designed for large-scale particle simulations on dynamical systems. Based on a large variety of advanced geometric algorithms, APT possesses long-term numerical accuracy and stability, which are critical for solving multi-scale and non-linear problems. Under the well-designed integrated and modularized framework, APT serves as a universal platform for researchers from different fields, such as plasma physics, accelerator physics, space science, fusio...

  6. RELOS.MOD2: a code system for the determination of instationary fission product releases from molten pools

    International Nuclear Information System (INIS)

    Kortz, Ch.; Koch, M.K.; Unger, H.; Funke, F.

    1999-01-01

    For the assessment of molten corium pool source terms, a mechanistic model has been developed to describe the transport of fission products from liquid corium pool surfaces into a colder gas atmosphere. Modelling is based on an approach for diffusive and convective transport processes coupled with thermochemical equilibrium considerations enabling detailed speciation analyses of the fission products released. Both have been implemented into the code system RELOS.MOD2. RELOS.MOD2 sensitivity calculations on possible effects of anticipated uncertainties in the thermo-chemical data on the fission product release predictions are presented. (author)

  7. Development of in-vessel source term analysis code, tracer

    International Nuclear Information System (INIS)

    Miyagi, K.; Miyahara, S.

    1996-01-01

    Analyses of radionuclide transport in fuel failure accidents (generally referred to source terms) are considered to be important especially in the severe accident evaluation. The TRACER code has been developed to realistically predict the time dependent behavior of FPs and aerosols within the primary cooling system for wide range of fuel failure events. This paper presents the model description, results of validation study, the recent model advancement status of the code, and results of check out calculations under reactor conditions. (author)

  8. QR CODES IN EDUCATION AND COMMUNICATION

    Directory of Open Access Journals (Sweden)

    Gurhan DURAK

    2016-04-01

    Full Text Available Technological advances brought applications of innovations to education. Conventional education increasingly flourishes with new technologies accompanied by more learner active environments. In this continuum, there are learners preferring self-learning. Traditional learning materials yield attractive, motivating and technologically enhanced learning materials. The QR (Quick Response Codes are one of these innovations. The aim of this study is to redesign a lesson unit supported with QR Codes and to get the learner views about the redesigned material. For this purpose, the redesigned lesson unit was delivered to 15 learners in Balıkesir University in the academic year of 2013-2014. The learners were asked to study the material. The learners who had smart phones and Internet access were chosen for the study. To provide sectional diversity, three groups were created. The group learners were from Faculty of Education, Faculty of Science and Literature and Faculty of Engineering. After the semi-structured interviews were held, the learners were asked about their pre-knowledge about QR Codes, QR Codes’ contribution to learning, difficulties with using QR Codes about and design issues. Descriptive data analysis was used in the study. The findings were interpreted on the basis of Theory of Diffusion of Innovations and Theory of Uses and Gratifications. After the research, the themes found were awareness of QR Code, types of QR Codes and applications, contributions to learning, and proliferation of QR Codes. Generally, the learners participating in the study reported that they were aware of QR Codes; that they could use the QR Codes; and that using QR Codes in education was useful. They also expressed that such features as visual elements, attractiveness and direct routing had positive impact on learning. In addition, they generally mentioned that they did not have any difficulty using QR Codes; that they liked the design; and that the content should

  9. Benchmarking and tuning the MILC code on clusters and supercomputers

    International Nuclear Information System (INIS)

    Gottlieb, Steven

    2002-01-01

    Recently, we have benchmarked and tuned the MILC code on a number of architectures including Intel Itanium and Pentium IV (PIV), dual-CPU Athlon, and the latest Compaq Alpha nodes. Results will be presented for many of these, and we shall discuss some simple code changes that can result in a very dramatic speedup of the KS conjugate gradient on processors with more advanced memory systems such as PIV, IBM SP and Alpha

  10. Benchmarking and tuning the MILC code on clusters and supercomputers

    International Nuclear Information System (INIS)

    Steven A. Gottlieb

    2001-01-01

    Recently, we have benchmarked and tuned the MILC code on a number of architectures including Intel Itanium and Pentium IV (PIV), dual-CPU Athlon, and the latest Compaq Alpha nodes. Results will be presented for many of these, and we shall discuss some simple code changes that can result in a very dramatic speedup of the KS conjugate gradient on processors with more advanced memory systems such as PIV, IBM SP and Alpha

  11. Benchmarking and tuning the MILC code on clusters and supercomputers

    Science.gov (United States)

    Gottlieb, Steven

    2002-03-01

    Recently, we have benchmarked and tuned the MILC code on a number of architectures including Intel Itanium and Pentium IV (PIV), dual-CPU Athlon, and the latest Compaq Alpha nodes. Results will be presented for many of these, and we shall discuss some simple code changes that can result in a very dramatic speedup of the KS conjugate gradient on processors with more advanced memory systems such as PIV, IBM SP and Alpha.

  12. A new 3-D integral code for computation of accelerator magnets

    International Nuclear Information System (INIS)

    Turner, L.R.; Kettunen, L.

    1991-01-01

    For computing accelerator magnets, integral codes have several advantages over finite element codes; far-field boundaries are treated automatically, and computed field in the bore region satisfy Maxwell's equations exactly. A new integral code employing edge elements rather than nodal elements has overcome the difficulties associated with earlier integral codes. By the use of field integrals (potential differences) as solution variables, the number of unknowns is reduced to one less than the number of nodes. Two examples, a hollow iron sphere and the dipole magnet of Advanced Photon Source injector synchrotron, show the capability of the code. The CPU time requirements are comparable to those of three-dimensional (3-D) finite-element codes. Experiments show that in practice it can realize much of the potential CPU time saving that parallel processing makes possible. 8 refs., 4 figs., 1 tab

  13. Summary Report for ASC L2 Milestone #4782: Assess Newly Emerging Programming and Memory Models for Advanced Architectures on Integrated Codes

    Energy Technology Data Exchange (ETDEWEB)

    Neely, J. R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hornung, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Black, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Robinson, P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2014-09-29

    This document serves as a detailed companion to the powerpoint slides presented as part of the ASC L2 milestone review for Integrated Codes milestone #4782 titled “Assess Newly Emerging Programming and Memory Models for Advanced Architectures on Integrated Codes”, due on 9/30/2014, and presented for formal program review on 9/12/2014. The program review committee is represented by Mike Zika (A Program Project Lead for Kull), Brian Pudliner (B Program Project Lead for Ares), Scott Futral (DEG Group Lead in LC), and Mike Glass (Sierra Project Lead at Sandia). This document, along with the presentation materials, and a letter of completion signed by the review committee will act as proof of completion for this milestone.

  14. Development and assessment of the COBRA/RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jae Jun; Ha, Kwi Seok; Sim, Seok Ku

    1997-04-01

    The COBRA/RELAP5 code, a merged version of the COBRA-TF and RELAP5/MOD3.2 codes, has been developed to combine the realistic three-dimensional reactor vessel model of COBRA-TF with RELAP5/MOD3, thus to produce an advanced system analysis code with a multidimensional thermal-hydraulic module. This report provides the integration scheme of the two codes and the results of developmental assessments. These includes single channel tests, manometric flow oscillation problem, THTF Test 105, and LOFT L2-3 large-break loss-of-coolant experiment. From the single channel tests the integration scheme and its implementation were proven to be valid. Other simulation results showed good agreement with the experiments. The computational speed was also satisfactory. So it is confirmed that COBRA/RELAP5 can be a promising tool for analysis of complicated, multidimensional two-phase flow transients. The area of further improvements in the code integration are also identified. This report also serves as a user`s manual for the COBRA/RELAP5 code. (author). 6 tabs., 20 figs., 20 refs.

  15. VIM Monte Carlo versus CASMO comparisons for BWR advanced fuel designs

    International Nuclear Information System (INIS)

    Pallotta, A.S.; Blomquist, R.N.

    1994-01-01

    Eigenvalues and two-dimensional fission rate distributions computed with the CASMO-3G lattice physics code and the VIM Monte Carlo Code are compared. The cases assessed are two advanced commercial BWR pin bundle designs. Generally, the two codes show good agreement in K inf , fission rate distributions, and control rod worths

  16. Modeling systems-level dynamics: Understanding without mechanistic explanation in integrative systems biology.

    Science.gov (United States)

    MacLeod, Miles; Nersessian, Nancy J

    2015-02-01

    In this paper we draw upon rich ethnographic data of two systems biology labs to explore the roles of explanation and understanding in large-scale systems modeling. We illustrate practices that depart from the goal of dynamic mechanistic explanation for the sake of more limited modeling goals. These processes use abstract mathematical formulations of bio-molecular interactions and data fitting techniques which we call top-down abstraction to trade away accurate mechanistic accounts of large-scale systems for specific information about aspects of those systems. We characterize these practices as pragmatic responses to the constraints many modelers of large-scale systems face, which in turn generate more limited pragmatic non-mechanistic forms of understanding of systems. These forms aim at knowledge of how to predict system responses in order to manipulate and control some aspects of them. We propose that this analysis of understanding provides a way to interpret what many systems biologists are aiming for in practice when they talk about the objective of a "systems-level understanding." Copyright © 2014 Elsevier Ltd. All rights reserved.

  17. MARS CODE MANUAL VOLUME III - Programmer's Manual

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Hwang, Moon Kyu; Jeong, Jae Jun; Kim, Kyung Doo; Bae, Sung Won; Lee, Young Jin; Lee, Won Jae

    2010-02-01

    Korea Advanced Energy Research Institute (KAERI) conceived and started the development of MARS code with the main objective of producing a state-of-the-art realistic thermal hydraulic systems analysis code with multi-dimensional analysis capability. MARS achieves this objective by very tightly integrating the one dimensional RELAP5/MOD3 with the multi-dimensional COBRA-TF codes. The method of integration of the two codes is based on the dynamic link library techniques, and the system pressure equation matrices of both codes are implicitly integrated and solved simultaneously. In addition, the Equation-Of-State (EOS) for the light water was unified by replacing the EOS of COBRA-TF by that of the RELAP5. This programmer's manual provides a complete list of overall information of code structure and input/output function of MARS. In addition, brief descriptions for each subroutine and major variables used in MARS are also included in this report, so that this report would be very useful for the code maintenance. The overall structure of the manual is modeled on the structure of the RELAP5 and as such the layout of the manual is very similar to that of the RELAP. This similitude to RELAP5 input is intentional as this input scheme will allow minimum modification between the inputs of RELAP5 and MARS3.1. MARS3.1 development team would like to express its appreciation to the RELAP5 Development Team and the USNRC for making this manual possible

  18. Benchmarking of FA2D/PARCS Code Package

    International Nuclear Information System (INIS)

    Grgic, D.; Jecmenica, R.; Pevec, D.

    2006-01-01

    FA2D/PARCS code package is used at Faculty of Electrical Engineering and Computing (FER), University of Zagreb, for static and dynamic reactor core analyses. It consists of two codes: FA2D and PARCS. FA2D is a multigroup two dimensional transport theory code for burn-up calculations based on collision probability method, developed at FER. It generates homogenised cross sections both of single pins and entire fuel assemblies. PARCS is an advanced nodal code developed at Purdue University for US NRC and it is based on neutron diffusion theory for three dimensional whole core static and dynamic calculations. It is modified at FER to enable internal 3D depletion calculation and usage of neutron cross section data in a format produced by FA2D and interface codes. The FA2D/PARCS code system has been validated on NPP Krsko operational data (Cycles 1 and 21). As we intend to use this code package for development of IRIS reactor loading patterns the first logical step was to validate the FA2D/PARCS code package on a set of IRIS benchmarks, starting from simple unit fuel cell, via fuel assembly, to full core benchmark. The IRIS 17x17 fuel with erbium burnable absorber was used in last full core benchmark. The results of modelling the IRIS full core benchmark using FA2D/PARCS code package have been compared with reference data showing the adequacy of FA2D/PARCS code package model for IRIS reactor core design.(author)

  19. Coupled CFD - system-code simulation of a gas cooled reactor

    International Nuclear Information System (INIS)

    Yan, Yizhou; Rizwan-uddin

    2011-01-01

    A generic coupled CFD - system-code thermal hydraulic simulation approach was developed based on FLUENT and RELAP-3D, and applied to LWRs. The flexibility of the coupling methodology enables its application to advanced nuclear energy systems. Gas Turbine - Modular Helium Reactor (GT-MHR) is a Gen IV reactor design which can benefit from this innovative coupled simulation approach. Mixing in the lower plenum of the GT-MHR is investigated here using the CFD - system-code coupled simulation tool. Results of coupled simulations are presented and discussed. The potential of the coupled CFD - system-code approach for next generation of nuclear power plants is demonstrated. (author)

  20. Electrochemical processes and mechanistic aspects of field-effect sensors for biomolecules

    Science.gov (United States)

    Huang, Weiguo; Diallo, Abdou Karim; Dailey, Jennifer L.; Besar, Kalpana

    2017-01-01

    Electronic biosensing is a leading technology for determining concentrations of biomolecules. In some cases, the presence of an analyte molecule induces a measured change in current flow, while in other cases, a new potential difference is established. In the particular case of a field effect biosensor, the potential difference is monitored as a change in conductance elsewhere in the device, such as across a film of an underlying semiconductor. Often, the mechanisms that lead to these responses are not specifically determined. Because improved understanding of these mechanisms will lead to improved performance, it is important to highlight those studies where various mechanistic possibilities are investigated. This review explores a range of possible mechanistic contributions to field-effect biosensor signals. First, we define the field-effect biosensor and the chemical interactions that lead to the field effect, followed by a section on theoretical and mechanistic background. We then discuss materials used in field-effect biosensors and approaches to improving signals from field-effect biosensors. We specifically cover the biomolecule interactions that produce local electric fields, structures and processes at interfaces between bioanalyte solutions and electronic materials, semiconductors used in biochemical sensors, dielectric layers used in top-gated sensors, and mechanisms for converting the surface voltage change to higher signal/noise outputs in circuits. PMID:29238595

  1. A Physics-Inspired Mechanistic Model of Migratory Movement Patterns in Birds.

    Science.gov (United States)

    Revell, Christopher; Somveille, Marius

    2017-08-29

    In this paper, we introduce a mechanistic model of migratory movement patterns in birds, inspired by ideas and methods from physics. Previous studies have shed light on the factors influencing bird migration but have mainly relied on statistical correlative analysis of tracking data. Our novel method offers a bottom up explanation of population-level migratory movement patterns. It differs from previous mechanistic models of animal migration and enables predictions of pathways and destinations from a given starting location. We define an environmental potential landscape from environmental data and simulate bird movement within this landscape based on simple decision rules drawn from statistical mechanics. We explore the capacity of the model by qualitatively comparing simulation results to the non-breeding migration patterns of a seabird species, the Black-browed Albatross (Thalassarche melanophris). This minimal, two-parameter model was able to capture remarkably well the previously documented migration patterns of the Black-browed Albatross, with the best combination of parameter values conserved across multiple geographically separate populations. Our physics-inspired mechanistic model could be applied to other bird and highly-mobile species, improving our understanding of the relative importance of various factors driving migration and making predictions that could be useful for conservation.

  2. Pre-Service Teachers' Perception of Quick Response (QR) Code Integration in Classroom Activities

    Science.gov (United States)

    Ali, Nagla; Santos, Ieda M.; Areepattamannil, Shaljan

    2017-01-01

    Quick Response (QR) codes have been discussed in the literature as adding value to teaching and learning. Despite their potential in education, more research is needed to inform practice and advance knowledge in this field. This paper investigated the integration of the QR code in classroom activities and the perceptions of the integration by…

  3. Structural integrity analyses: can we manage the advances?

    International Nuclear Information System (INIS)

    Sauve, R.

    2006-01-01

    the recognition that traditional methods of analysis can still be utilized to provide improved understanding and also as an independent check on the validity of results. Unfortunately, tight schedule and budgetary constraints, lack of proper understanding of the problem, codes and standards based on older methods of analysis and inadequate training in the use of the advances (e.g. finite element method) can lead to difficulties. In the case of engineering software, the temptation to change the problem to suit the computer code due to its limitations must be avoided. For highly complex structures, computer modelling coupled with testing provides a robust method that can avoid costly and sometimes fatal errors in design. In this presentation, a brief overview of the evolution of state-of-the-art in computer modelling pertaining to structural integrity is provided. Managing the process through understanding the problem, knowing software limitations and establishing appropriate quality assurance for advanced analyses is covered. The impact of the advances in structural integrity evaluations on codes and standards such as the ASME boiler and pressure vessel code is briefly discussed. (author)

  4. Severe accident tests and development of domestic severe accident system codes

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    According to lessons learned from Fukushima-Daiichi NPS accidents, the safety evaluation will be started based on the NRA's New Safety Standards. In parallel with this movement, reinforcement of Severe Accident (SA) Measures and Accident Managements (AMs) has been undertaken and establishments of relevant regulations and standards are recognized as urgent subjects. Strengthening responses against nuclear plant hazards, as well as realistic protection measures and their standardization is also recognized as urgent subjects. Furthermore, decommissioning of Fukushima-Daiichi Unit1 through Unit4 is promoted diligently. Taking into account JNES's mission with regard to these SA Measures, AMs and decommissioning, movement of improving SA evaluation methodologies inside and outside Japan, and prioritization of subjects based on analyses of sequences of Fukushima-Daiichi NPS accidents, three viewpoints was extracted. These viewpoints were substantiated as the following three groups of R and D subjects: (1) Obtaining near term experimental subjects: Containment venting, Seawater injection, Iodine behaviors. (2) Obtaining mid and long experimental subjects: Fuel damage behavior at early phase of core degradation, Core melting and debris formation. (3) Development of a macroscopic level SA code for plant system behaviors and a mechanistic level code for core melting and debris formation. (author)

  5. Severe accident tests and development of domestic severe accident system codes

    International Nuclear Information System (INIS)

    2013-01-01

    According to lessons learned from Fukushima-Daiichi NPS accidents, the safety evaluation will be started based on the NRA's New Safety Standards. In parallel with this movement, reinforcement of Severe Accident (SA) Measures and Accident Managements (AMs) has been undertaken and establishments of relevant regulations and standards are recognized as urgent subjects. Strengthening responses against nuclear plant hazards, as well as realistic protection measures and their standardization is also recognized as urgent subjects. Furthermore, decommissioning of Fukushima-Daiichi Unit1 through Unit4 is promoted diligently. Taking into account JNES's mission with regard to these SA Measures, AMs and decommissioning, movement of improving SA evaluation methodologies inside and outside Japan, and prioritization of subjects based on analyses of sequences of Fukushima-Daiichi NPS accidents, three viewpoints was extracted. These viewpoints were substantiated as the following three groups of R and D subjects: (1) Obtaining near term experimental subjects: Containment venting, Seawater injection, Iodine behaviors. (2) Obtaining mid and long experimental subjects: Fuel damage behavior at early phase of core degradation, Core melting and debris formation. (3) Development of a macroscopic level SA code for plant system behaviors and a mechanistic level code for core melting and debris formation. (author)

  6. 1 CFR 21.14 - Deviations from standard organization of the Code of Federal Regulations.

    Science.gov (United States)

    2010-01-01

    ... 1 General Provisions 1 2010-01-01 2010-01-01 false Deviations from standard organization of the... CODIFICATION General Numbering § 21.14 Deviations from standard organization of the Code of Federal Regulations. (a) Any deviation from standard Code of Federal Regulations designations must be approved in advance...

  7. Coding for effective denial management.

    Science.gov (United States)

    Miller, Jackie; Lineberry, Joe

    2004-01-01

    Nearly everyone will agree that accurate and consistent coding of diagnoses and procedures is the cornerstone for operating a compliant practice. The CPT or HCPCS procedure code tells the payor what service was performed and also (in most cases) determines the amount of payment. The ICD-9-CM diagnosis code, on the other hand, tells the payor why the service was performed. If the diagnosis code does not meet the payor's criteria for medical necessity, all payment for the service will be denied. Implementation of an effective denial management program can help "stop the bleeding." Denial management is a comprehensive process that works in two ways. First, it evaluates the cause of denials and takes steps to prevent them. Second, denial management creates specific procedures for refiling or appealing claims that are initially denied. Accurate, consistent and compliant coding is key to both of these functions. The process of proactively managing claim denials also reveals a practice's administrative strengths and weaknesses, enabling radiology business managers to streamline processes, eliminate duplicated efforts and shift a larger proportion of the staff's focus from paperwork to servicing patients--all of which are sure to enhance operations and improve practice management and office morale. Accurate coding requires a program of ongoing training and education in both CPT and ICD-9-CM coding. Radiology business managers must make education a top priority for their coding staff. Front office staff, technologists and radiologists should also be familiar with the types of information needed for accurate coding. A good staff training program will also cover the proper use of Advance Beneficiary Notices (ABNs). Registration and coding staff should understand how to determine whether the patient's clinical history meets criteria for Medicare coverage, and how to administer an ABN if the exam is likely to be denied. Staff should also understand the restrictions on use of

  8. Overview of the South African mechanistic pavement design analysis method

    CSIR Research Space (South Africa)

    Theyse, HL

    1996-01-01

    Full Text Available A historical overview of the South African mechanistic pavement design method, from its development in the early 1970s to the present, is presented. Material characterization, structural analysis, and pavement life prediction are discussed...

  9. Light water reactor fuel analysis code. FEMAXI-6 (Ver.1). Detailed structure and user's manual

    International Nuclear Information System (INIS)

    Suzuki, Motoe; Saitou, Hiroaki

    2006-02-01

    A light water reactor fuel analysis code FEMAXI-6 is an advanced version which has been produced by integrating the former version FEMAXI-V with numerous functional improvements and extensions. In particular, the FEMAXI-6 code has attained a complete coupled solution of thermal analysis and mechanical analysis, enabling an accurate prediction of pellet-clad gap size and PCMI in high burnup fuel rods. Also, such new models have been implemented as pellet-clad bonding and fission gas bubble swelling, and linkage function with detailed burning analysis code has been enhanced. Furthermore, a number of new materials properties and parameters have been introduced. With these advancements, the FEMAXI-6 code has been upgraded to a versatile analytical tool for high burnup fuel behavior not only in the normal operation but also in anticipated transient conditions. This report describes in detail the design, basic theory and structure, models and numerical method, improvements and extensions, and method of model modification. In order to facilitate effective and wide-ranging application of the code, formats and methods of input/output of the code are also described, and a sample output in an actual form is included. (author)

  10. Ionizing radiation induced cataracts: Recent biological and mechanistic developments and perspectives for future research.

    Science.gov (United States)

    Ainsbury, Elizabeth A; Barnard, Stephen; Bright, Scott; Dalke, Claudia; Jarrin, Miguel; Kunze, Sarah; Tanner, Rick; Dynlacht, Joseph R; Quinlan, Roy A; Graw, Jochen; Kadhim, Munira; Hamada, Nobuyuki

    The lens of the eye has long been considered as a radiosensitive tissue, but recent research has suggested that the radiosensitivity is even greater than previously thought. The 2012 recommendation of the International Commission on Radiological Protection (ICRP) to substantially reduce the annual occupational equivalent dose limit for the ocular lens has now been adopted in the European Union and is under consideration around the rest of the world. However, ICRP clearly states that the recommendations are chiefly based on epidemiological evidence because there are a very small number of studies that provide explicit biological, mechanistic evidence at doses <2Gy. This paper aims to present a review of recently published information on the biological and mechanistic aspects of cataracts induced by exposure to ionizing radiation (IR). The data were compiled by assessing the pertinent literature in several distinct areas which contribute to the understanding of IR induced cataracts, information regarding lens biology and general processes of cataractogenesis. Results from cellular and tissue level studies and animal models, and relevant human studies, were examined. The main focus was the biological effects of low linear energy transfer IR, but dosimetry issues and a number of other confounding factors were also considered. The results of this review clearly highlight a number of gaps in current knowledge. Overall, while there have been a number of recent advances in understanding, it remains unknown exactly how IR exposure contributes to opacification. A fuller understanding of how exposure to relatively low doses of IR promotes induction and/or progression of IR-induced cataracts will have important implications for prevention and treatment of this disease, as well as for the field of radiation protection. Crown Copyright © 2016. Published by Elsevier B.V. All rights reserved.

  11. Mechanistic applicability domain classification of a local lymph node assay dataset for skin sensitization.

    Science.gov (United States)

    Roberts, David W; Patlewicz, Grace; Kern, Petra S; Gerberick, Frank; Kimber, Ian; Dearman, Rebecca J; Ryan, Cindy A; Basketter, David A; Aptula, Aynur O

    2007-07-01

    The goal of eliminating animal testing in the predictive identification of chemicals with the intrinsic ability to cause skin sensitization is an important target, the attainment of which has recently been brought into even sharper relief by the EU Cosmetics Directive and the requirements of the REACH legislation. Development of alternative methods requires that the chemicals used to evaluate and validate novel approaches comprise not only confirmed skin sensitizers and non-sensitizers but also substances that span the full chemical mechanistic spectrum associated with skin sensitization. To this end, a recently published database of more than 200 chemicals tested in the mouse local lymph node assay (LLNA) has been examined in relation to various chemical reaction mechanistic domains known to be associated with sensitization. It is demonstrated here that the dataset does cover the main reaction mechanistic domains. In addition, it is shown that assignment to a reaction mechanistic domain is a critical first step in a strategic approach to understanding, ultimately on a quantitative basis, how chemical properties influence the potency of skin sensitizing chemicals. This understanding is necessary if reliable non-animal approaches, including (quantitative) structure-activity relationships (Q)SARs, read-across, and experimental chemistry based models, are to be developed.

  12. Three-dimensional reactor dynamics code for VVER type nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kyrki-Rajamaeki, R. [VTT Energy, Espoo (Finland)

    1995-10-01

    A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.).

  13. Three-dimensional reactor dynamics code for VVER type nuclear reactors

    International Nuclear Information System (INIS)

    Kyrki-Rajamaeki, R.

    1995-10-01

    A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.)

  14. Self-shielding models of MICROX-2 code: Review and updates

    International Nuclear Information System (INIS)

    Hou, J.; Choi, H.; Ivanov, K.N.

    2014-01-01

    Highlights: • The MICROX-2 code has been improved to expand its application to advanced reactors. • New fine-group cross section libraries based on ENDF/B-VII have been generated. • Resonance self-shielding and spatial self-shielding models have been improved. • The improvements were assessed by a series of benchmark calculations against MCNPX. - Abstract: The MICROX-2 is a transport theory code that solves for the neutron slowing-down and thermalization equations of a two-region lattice cell. The MICROX-2 code has been updated to expand its application to advanced reactor concepts and fuel cycle simulations, including generation of new fine-group cross section libraries based on ENDF/B-VII. In continuation of previous work, the MICROX-2 methods are reviewed and updated in this study, focusing on its resonance self-shielding and spatial self-shielding models for neutron spectrum calculations. The improvement of self-shielding method was assessed by a series of benchmark calculations against the Monte Carlo code, using homogeneous and heterogeneous pin cell models. The results have shown that the implementation of the updated self-shielding models is correct and the accuracy of physics calculation is improved. Compared to the existing models, the updates reduced the prediction error of the infinite multiplication factor by ∼0.1% and ∼0.2% for the homogeneous and heterogeneous pin cell models, respectively, considered in this study

  15. Organophotocatalysis: Insights into the Mechanistic Aspects of Thiourea-Mediated Intermolecular [2+2] Photocycloadditions.

    Science.gov (United States)

    Vallavoju, Nandini; Selvakumar, Sermadurai; Pemberton, Barry C; Jockusch, Steffen; Sibi, Mukund P; Sivaguru, Jayaraman

    2016-04-25

    Mechanistic investigations of the intermolecular [2+2] photocycloaddition of coumarin with tetramethylethylene mediated by thiourea catalysts reveal that the reaction is enabled by a combination of minimized aggregation, enhanced intersystem crossing, and altered excited-state lifetime(s). These results clarify how the excited-state reactivity can be manipulated through catalyst-substrate interactions and reveal a third mechanistic pathway for thiourea-mediated organo-photocatalysis. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  16. From concatenated codes to graph codes

    DEFF Research Database (Denmark)

    Justesen, Jørn; Høholdt, Tom

    2004-01-01

    We consider codes based on simple bipartite expander graphs. These codes may be seen as the first step leading from product type concatenated codes to more complex graph codes. We emphasize constructions of specific codes of realistic lengths, and study the details of decoding by message passing...

  17. Computer code development plant for SMART design

    International Nuclear Information System (INIS)

    Bae, Kyoo Hwan; Choi, S.; Cho, B.H.; Kim, K.K.; Lee, J.C.; Kim, J.P.; Kim, J.H.; Chung, M.; Kang, D.J.; Chang, M.H.

    1999-03-01

    In accordance with the localization plan for the nuclear reactor design driven since the middle of 1980s, various computer codes have been transferred into the korea nuclear industry through the technical transfer program from the worldwide major pressurized water reactor supplier or through the international code development program. These computer codes have been successfully utilized in reactor and reload core design works. As the results, design- related technologies have been satisfactorily accumulated. However, the activities for the native code development activities to substitute the some important computer codes of which usages are limited by the original technique owners have been carried out rather poorly. Thus, it is most preferentially required to secure the native techniques on the computer code package and analysis methodology in order to establish the capability required for the independent design of our own model of reactor. Moreover, differently from the large capacity loop-type commercial reactors, SMART (SYSTEM-integrated Modular Advanced ReacTor) design adopts a single reactor pressure vessel containing the major primary components and has peculiar design characteristics such as self-controlled gas pressurizer, helical steam generator, passive residual heat removal system, etc. Considering those peculiar design characteristics for SMART, part of design can be performed with the computer codes used for the loop-type commercial reactor design. However, most of those computer codes are not directly applicable to the design of an integral reactor such as SMART. Thus, they should be modified to deal with the peculiar design characteristics of SMART. In addition to the modification efforts, various codes should be developed in several design area. Furthermore, modified or newly developed codes should be verified their reliability through the benchmarking or the test for the object design. Thus, it is necessary to proceed the design according to the

  18. Computer code development plant for SMART design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Kyoo Hwan; Choi, S.; Cho, B.H.; Kim, K.K.; Lee, J.C.; Kim, J.P.; Kim, J.H.; Chung, M.; Kang, D.J.; Chang, M.H

    1999-03-01

    In accordance with the localization plan for the nuclear reactor design driven since the middle of 1980s, various computer codes have been transferred into the korea nuclear industry through the technical transfer program from the worldwide major pressurized water reactor supplier or through the international code development program. These computer codes have been successfully utilized in reactor and reload core design works. As the results, design- related technologies have been satisfactorily accumulated. However, the activities for the native code development activities to substitute the some important computer codes of which usages are limited by the original technique owners have been carried out rather poorly. Thus, it is most preferentially required to secure the native techniques on the computer code package and analysis methodology in order to establish the capability required for the independent design of our own model of reactor. Moreover, differently from the large capacity loop-type commercial reactors, SMART (SYSTEM-integrated Modular Advanced ReacTor) design adopts a single reactor pressure vessel containing the major primary components and has peculiar design characteristics such as self-controlled gas pressurizer, helical steam generator, passive residual heat removal system, etc. Considering those peculiar design characteristics for SMART, part of design can be performed with the computer codes used for the loop-type commercial reactor design. However, most of those computer codes are not directly applicable to the design of an integral reactor such as SMART. Thus, they should be modified to deal with the peculiar design characteristics of SMART. In addition to the modification efforts, various codes should be developed in several design area. Furthermore, modified or newly developed codes should be verified their reliability through the benchmarking or the test for the object design. Thus, it is necessary to proceed the design according to the

  19. PBDOWN - a computer code for simulating core material discharge and thermal to mechanical energy conversion in LMFBR hypothetical accidents

    International Nuclear Information System (INIS)

    Royl, P.

    1981-01-01

    PBDOWN is a computer code that simulates the blowdown of confined boiling materials ('pools') into a colder upper coolant plenum as time dependent ejection and expansion with consideration of a few selected exchange processes. Its application is restricted to situations resulting from hypothetical loss of flow (LOF) accidents in LMFBR's, where enough voiding has occured, that in core sodium vapor pressures become negligible. PBDOWN considers one working fluid for the discharge process (either fuel or steel) and a maximum of two working fluids (either fuel and sodium or steel and sodium) for the expansion process in the upper coolant plenum. Entrainment of sodium at the accelerated bubble liquid interfaces is mechanistically calculated by a Taylor instability entrainment model. Simulation of a hemispherical expansion form together with this mechanistic entrainment model gives a new integrated calculation of the time dependent sodium mass in the bubble. The paper summarizes the basic equations and assumptions of this computer model. Sample results compare different heat transfer and Na entrainment models during steel and fuel driven discharge processes. Mechanistic sodium entrainment simulation for SNR-type reactors coupled with a realistic heat transfer model is shown to reduce the integral mechanical work potential by a factor of 1.3 to 2.0 over the isentropic energy of the discharge working fluids. (orig.)

  20. MARS CODE MANAUAL VOLUME IV - Developmental Assessment Report

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Jeong, Jae Jun; Hwang, Moon Kyu; Lee, Won Jae; Lee, Young Jin; Lee, Seung Wook; Kim, Kyung Doo; Bae, Sung Won

    2010-02-01

    Korea Advanced Energy Research Institute (KAERI) conceived and started the development of MARS code with the main objective of producing a state-of-the-art realistic thermal hydraulic systems analysis code with multi-dimensional analysis capability. MARS achieves this objective by very tightly integrating the one dimensional RELAP5/MOD3 with the multi-dimensional COBRA-TF codes. The method of integration of the two codes is based on the dynamic link library techniques, and the system pressure equation matrices of both codes are implicitly integrated and solved simultaneously. In addition, the Equation-Of-State (EOS) for the light water was unified by replacing the EOS of COBRA-TF by that of the RELAP5. This assessment manual provides a complete list of code assessment results of the MARS code for various conceptual problem, separate effect test and integral effect test. From these validation procedures, the soundness and accuracy of the MARS code has been confirmed. The overall structure of the input is modeled on the structure of the RELAP5 and as such the layout of the manual is very similar to that of the RELAP. This similitude to RELAP5 input is intentional as this input scheme will allow minimum modification between the inputs of RELAP5 and MARS3.1. MARS3.1 development team would like to express its appreciation to the RELAP5 Development Team and the USNRC for making this manual possible

  1. Speech coding code- excited linear prediction

    CERN Document Server

    Bäckström, Tom

    2017-01-01

    This book provides scientific understanding of the most central techniques used in speech coding both for advanced students as well as professionals with a background in speech audio and or digital signal processing. It provides a clear connection between the whys hows and whats thus enabling a clear view of the necessity purpose and solutions provided by various tools as well as their strengths and weaknesses in each respect Equivalently this book sheds light on the following perspectives for each technology presented Objective What do we want to achieve and especially why is this goal important Resource Information What information is available and how can it be useful and Resource Platform What kind of platforms are we working with and what are their capabilities restrictions This includes computational memory and acoustic properties and the transmission capacity of devices used. The book goes on to address Solutions Which solutions have been proposed and how can they be used to reach the stated goals and ...

  2. The challenge of making ozone risk assessment for forest trees more mechanistic

    International Nuclear Information System (INIS)

    Matyssek, R.; Sandermann, H.; Wieser, G.; Booker, F.; Cieslik, S.; Musselman, R.; Ernst, D.

    2008-01-01

    Upcoming decades will experience increasing atmospheric CO 2 and likely enhanced O 3 exposure which represents a risk for the carbon sink strength of forests, so that the need for cause-effect related O 3 risk assessment increases. Although assessment will gain in reliability on an O 3 uptake basis, risk is co-determined by the effective dose, i.e. the plant's sensitivity per O 3 uptake. Recent progress in research on the molecular and metabolic control of the effective O 3 dose is reported along with advances in empirically assessing O 3 uptake at the whole-tree and stand level. Knowledge on both O 3 uptake and effective dose (measures of stress avoidance and tolerance, respectively) needs to be understood mechanistically and linked as a pre-requisite before practical use of process-based O 3 risk assessment can be implemented. To this end, perspectives are derived for validating and promoting new O 3 flux-based modelling tools. - Clarifying and linking mechanisms of O 3 uptake and effective dose are research challenges highlighted in view of recent progress and perspectives towards cause-effect based risk assessment

  3. Investigation of alpha experiment by severe accident analysis code SAMPSON

    International Nuclear Information System (INIS)

    Baglietto, Emilio; Ninokata, Hisashi; Naitoh, Masanori

    2006-01-01

    The severe accident analysis code SAMPSON is adopted in this work to evaluate its capability of reproducing the complex gap cooling phenomenon. The ALPHA experiment is adopted for validation, where molten aluminum oxide (Al 2 O 3 ) produced by a thermite reaction is poured into a water filled hemispherical vessel at the ambient pressure of approximately 1.3 MPa. The spreading and cooling of the debris that has relocated into the pressure vessel lower plenum are simulated, including the analysis of the RPV failure. The model included in the core to mimic the water penetration inside the gap is evaluated and improvements are proposed. The importance of the introduction of some mechanistic approach to describe the gap formation and evolution is underlined, where the results show its necessity in order to correctly reproduce the experimental trends. (author)

  4. Quality Assurance for Thermal Hydraulic Analysis Code, TASS/SMR-S

    International Nuclear Information System (INIS)

    Kim, Hee Kyung; Kim, Soo Hyoung; Chung, Young Jong; Kim, Hyeon Soo

    2012-01-01

    Safety analysis for a System-integrated Modular Advanced Reactor (SMART), a computer code called TASS/SMR-S has been developed by Korea Atomic Energy Research Institute (KAERI). To guarantee the quality of the software, a series of software Quality Assurance (QA) procedures has been developed for the TASS/SMR-S code. These procedures are described herein, from the requirement phase to the Verification and Validation (V and V) phase, and representative results of the TASS/SMR-S QA are presented

  5. A Tough Call : Mitigating Advanced Code-Reuse Attacks at the Binary Level

    NARCIS (Netherlands)

    Veen, Victor Van Der; Goktas, Enes; Contag, Moritz; Pawoloski, Andre; Chen, Xi; Rawat, Sanjay; Bos, Herbert; Holz, Thorsten; Athanasopoulos, Ilias; Giuffrida, Cristiano

    2016-01-01

    Current binary-level Control-Flow Integrity (CFI) techniques are weak in determining the set of valid targets for indirect control flow transfers on the forward edge. In particular, the lack of source code forces existing techniques to resort to a conservative address-taken policy that

  6. Does Mechanistic Thinking Improve Student Success in Organic Chemistry?

    Science.gov (United States)

    Grove, Nathaniel P.; Cooper, Melanie M.; Cox, Elizabeth L.

    2012-01-01

    The use of the curved-arrow notation to depict electron flow during mechanistic processes is one of the most important representational conventions in the organic chemistry curriculum. Our previous research documented a disturbing trend: when asked to predict the products of a series of reactions, many students do not spontaneously engage in…

  7. Simulating Coupling Complexity in Space Plasmas: First Results from a new code

    Science.gov (United States)

    Kryukov, I.; Zank, G. P.; Pogorelov, N. V.; Raeder, J.; Ciardo, G.; Florinski, V. A.; Heerikhuisen, J.; Li, G.; Petrini, F.; Shematovich, V. I.; Winske, D.; Shaikh, D.; Webb, G. M.; Yee, H. M.

    2005-12-01

    The development of codes that embrace 'coupling complexity' via the self-consistent incorporation of multiple physical scales and multiple physical processes in models has been identified by the NRC Decadal Survey in Solar and Space Physics as a crucial necessary development in simulation/modeling technology for the coming decade. The National Science Foundation, through its Information Technology Research (ITR) Program, is supporting our efforts to develop a new class of computational code for plasmas and neutral gases that integrates multiple scales and multiple physical processes and descriptions. We are developing a highly modular, parallelized, scalable code that incorporates multiple scales by synthesizing 3 simulation technologies: 1) Computational fluid dynamics (hydrodynamics or magneto-hydrodynamics-MHD) for the large-scale plasma; 2) direct Monte Carlo simulation of atoms/neutral gas, and 3) transport code solvers to model highly energetic particle distributions. We are constructing the code so that a fourth simulation technology, hybrid simulations for microscale structures and particle distributions, can be incorporated in future work, but for the present, this aspect will be addressed at a test-particle level. This synthesis we will provide a computational tool that will advance our understanding of the physics of neutral and charged gases enormously. Besides making major advances in basic plasma physics and neutral gas problems, this project will address 3 Grand Challenge space physics problems that reflect our research interests: 1) To develop a temporal global heliospheric model which includes the interaction of solar and interstellar plasma with neutral populations (hydrogen, helium, etc., and dust), test-particle kinetic pickup ion acceleration at the termination shock, anomalous cosmic ray production, interaction with galactic cosmic rays, while incorporating the time variability of the solar wind and the solar cycle. 2) To develop a coronal

  8. Advanced glycation end-products: a biological consequence of lifestyle contributing to cancer disparity

    OpenAIRE

    Turner, David P.

    2015-01-01

    Low income, poor diet, obesity and a lack of exercise are inter-related lifestyle factors that can profoundly alter our biological make-up to increase cancer risk, growth and development. We recently reported a potential mechanistic link between carbohydrate derived metabolites and cancer which may provide a biological consequence of lifestyle that can directly impact tumor biology. Advanced glycation end-products (AGEs) are reactive metabolites produced as a by-product of sugar metabolism. F...

  9. Description of premixing with the MC3D code including molten jet behavior modeling. Comparison with FARO experimental results

    Energy Technology Data Exchange (ETDEWEB)

    Berthoud, G.; Crecy, F. de; Meignen, R.; Valette, M. [CEA-G, DRN/DTP/SMTH, 17 rue des Martyrs, 38054 Grenoble Cedex 9 (France)

    1998-01-01

    The premixing phase of a molten fuel-coolant interaction is studied by the way of mechanistic multidimensional calculation. Beside water and steam, corium droplet flow and continuous corium jet flow are calculated independent. The 4-field MC3D code and a detailed hot jet fragmentation model are presented. MC3D calculations are compared to the FARO L14 experiment results and are found to give satisfactory results; heat transfer and jet fragmentation models are still to be improved to predict better final debris size values. (author)

  10. Structure of fuel performance audit code for SFR metal fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yong Sik; Kim, Hyo Chan [KAERI, Daejeon (Korea, Republic of); Jeong, Hye Dong; Shin, An Dong; Suh, Nam Duk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2012-10-15

    A Sodium Cooled Fast Reactor (SFR) is a promising option to solve the spent fuel problems, but, there are still much technical issues to commercialize a SFR. One of issues is a development of advanced fuel which can solve the safety and the economic issues at the same time. Since a nuclear fuel is the first barrier to protect radioactive isotope release, the fuel's integrity must be secured. In Korea Institute of Nuclear Safety (KINS), the new project has been started to develop the regulatory technology for SFR system including a fuel area. To evaluate the fuel integrity and safety during an irradiation, the fuel performance code must be used for audit calculation. To develop the new code system, the code structure design and its requirements need to be studied. Various performance models and code systems are reviewed and their characteristics are analyzed in this paper. Based on this study, the fundamental performance models are deduced and basic code requirements and structure are established.

  11. Fuel Performance Experiments and Modeling: Fission Gas Bubble Nucleation and Growth in Alloy Nuclear Fuels

    International Nuclear Information System (INIS)

    McDeavitt, Sean; Shao, Lin; Tsvetkov, Pavel; Wirth, Brian; Kennedy, Rory

    2014-01-01

    Advanced fast reactor systems being developed under the DOE's Advanced Fuel Cycle Initiative are designed to destroy TRU isotopes generated in existing and future nuclear energy systems. Over the past 40 years, multiple experiments and demonstrations have been completed using U-Zr, U-Pu-Zr, U-Mo and other metal alloys. As a result, multiple empirical and semi-empirical relationships have been established to develop empirical performance modeling codes. Many mechanistic questions about fission as mobility, bubble coalescience, and gas release have been answered through industrial experience, research, and empirical understanding. The advent of modern computational materials science, however, opens new doors of development such that physics-based multi-scale models may be developed to enable a new generation of predictive fuel performance codes that are not limited by empiricism.

  12. Fuel Performance Experiments and Modeling: Fission Gas Bubble Nucleation and Growth in Alloy Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    McDeavitt, Sean [Texas A & M Univ., College Station, TX (United States); Shao, Lin [Texas A & M Univ., College Station, TX (United States); Tsvetkov, Pavel [Texas A & M Univ., College Station, TX (United States); Wirth, Brian [Univ. of Tennessee, Knoxville, TN (United States); Kennedy, Rory [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-04-07

    Advanced fast reactor systems being developed under the DOE's Advanced Fuel Cycle Initiative are designed to destroy TRU isotopes generated in existing and future nuclear energy systems. Over the past 40 years, multiple experiments and demonstrations have been completed using U-Zr, U-Pu-Zr, U-Mo and other metal alloys. As a result, multiple empirical and semi-empirical relationships have been established to develop empirical performance modeling codes. Many mechanistic questions about fission as mobility, bubble coalescience, and gas release have been answered through industrial experience, research, and empirical understanding. The advent of modern computational materials science, however, opens new doors of development such that physics-based multi-scale models may be developed to enable a new generation of predictive fuel performance codes that are not limited by empiricism.

  13. Environmental construction of nano-material design codes. The example of simulation codes used in the CMD workshop

    Energy Technology Data Exchange (ETDEWEB)

    Miyazaki, Mikiya [Japan Atomic Energy Research Inst., Center for Promotion of Computational Science and Engineering, Kizu, Kyoto (Japan)

    2003-05-01

    Generally it is well known that the R and D works on new materials or devices will play a central role on the evolution of future society. But, the old ways based on the empirical and experimental approach have already reached the limit, especially for dealing with a strange substance and material. The structure of a substance and material is needed to be dealt with in detail by quantum mechanics, because the limit on accuracy has come in sight in the calculation using a classical theory. The research on the latest electronic state calculation technique founded on quantum mechanics made a great advance as the technique of solving these problems as well as the technique of a computational materials design. It enables the prediction of material properties because it is based on First Principles. Therefore, in the future it is expected to have a very high possibility of becoming a breakthrough in such a situation. In this article, the example calculation results by PC cluster on the codes (MACHIKANEYAMA-2000, OSAKA-2000) used in the CMD (Computational Materials Design) workshop, held on Sep. 19-21, at ITBL-Building and International Institute for Advanced Studies under the auspices of the University of Osaka, are described. Furthermore, the graphical user interfaces on the codes are examined. (author)

  14. Steam condensation modelling in aerosol codes

    International Nuclear Information System (INIS)

    Dunbar, I.H.

    1986-01-01

    The principal subject of this study is the modelling of the condensation of steam into and evaporation of water from aerosol particles. These processes introduce a new type of term into the equation for the development of the aerosol particle size distribution. This new term faces the code developer with three major problems: the physical modelling of the condensation/evaporation process, the discretisation of the new term and the separate accounting for the masses of the water and of the other components. This study has considered four codes which model the condensation of steam into and its evaporation from aerosol particles: AEROSYM-M (UK), AEROSOLS/B1 (France), NAUA (Federal Republic of Germany) and CONTAIN (USA). The modelling in the codes has been addressed under three headings. These are the physical modelling of condensation, the mathematics of the discretisation of the equations, and the methods for modelling the separate behaviour of different chemical components of the aerosol. The codes are least advanced in area of solute effect modelling. At present only AEROSOLS/B1 includes the effect. The effect is greater for more concentrated solutions. Codes without the effect will be more in error (underestimating the total airborne mass) the less condensation they predict. Data are needed on the water vapour pressure above concentrated solutions of the substances of interest (especially CsOH and CsI) if the extent to which aerosols retain water under superheated conditions is to be modelled. 15 refs

  15. TITAN: an advanced three-dimensional neutronics/thermal-hydraulics code for light water reactor safety analysis

    International Nuclear Information System (INIS)

    Griggs, D.P.; Kazimi, M.S.; Henry, A.F.

    1982-01-01

    The initial development of TITAN, a three-dimensional coupled neutronics/thermal-hydraulics code for LWR safety analysis, has been completed. The transient neutronics code QUANDRY has been joined to the two-fluid thermal-hydraulics code THERMIT with the appropriate feedback mechanisms modeled. A detailed steady-state and transient coupling scheme based on the tandem technique was implemented in accordance with the important structural and operational characteristics of QUANDRY and THERMIT. A two channel sample problem formed the basis for steady-state and transient analyses performed with TITAN. TITAN steady-state results were compared with those obtained with MEKIN and showed good agreement. Null transients, simulated turbine trip transients, and a rod withdrawal transient were analyzed with TITAN and reasonable results were obtained

  16. Recent Improvements to the IMPACT-T Parallel Particle Tracking Code

    International Nuclear Information System (INIS)

    Qiang, J.; Pogorelov, I.V.; Ryne, R.

    2006-01-01

    The IMPACT-T code is a parallel three-dimensional quasi-static beam dynamics code for modeling high brightness beams in photoinjectors and RF linacs. Developed under the US DOE Scientific Discovery through Advanced Computing (SciDAC) program, it includes several key features including a self-consistent calculation of 3D space-charge forces using a shifted and integrated Green function method, multiple energy bins for beams with large energy spread, and models for treating RF standing wave and traveling wave structures. In this paper, we report on recent improvements to the IMPACT-T code including modeling traveling wave structures, short-range transverse and longitudinal wakefields, and longitudinal coherent synchrotron radiation through bending magnets

  17. Toric Varieties and Codes, Error-correcting Codes, Quantum Codes, Secret Sharing and Decoding

    DEFF Research Database (Denmark)

    Hansen, Johan Peder

    We present toric varieties and associated toric codes and their decoding. Toric codes are applied to construct Linear Secret Sharing Schemes (LSSS) with strong multiplication by the Massey construction. Asymmetric Quantum Codes are obtained from toric codes by the A.R. Calderbank P.W. Shor and A.......M. Steane construction of stabilizer codes (CSS) from linear codes containing their dual codes....

  18. The MICHELLE 2D/3D ES PIC Code Advances and Applications

    CERN Document Server

    Petillo, John; De Ford, John F; Dionne, Norman J; Eppley, Kenneth; Held, Ben; Levush, Baruch; Nelson, Eric M; Panagos, Dimitrios; Zhai, Xiaoling

    2005-01-01

    MICHELLE is a new 2D/3D steady-state and time-domain particle-in-cell (PIC) code* that employs electrostatic and now magnetostatic finite-element field solvers. The code has been used to design and analyze a wide variety of devices that includes multistage depressed collectors, gridded guns, multibeam guns, annular-beam guns, sheet-beam guns, beam-transport sections, and ion thrusters. Latest additions to the MICHELLE/Voyager tool are as follows: 1) a prototype 3D self magnetic field solver using the curl-curl finite-element formulation for the magnetic vector potential, employing edge basis functions and accumulating current with MICHELLE's new unstructured grid particle tracker, 2) the electrostatic field solver now accommodates dielectric media, 3) periodic boundary conditions are now functional on all grids, not just structured grids, 4) the addition of a global optimization module to the user interface where both electrical parameters (such as electrode voltages)can be optimized, and 5) adaptive mesh ref...

  19. Implantation of TRAC-PF1 computer code of VAX-11/750

    International Nuclear Information System (INIS)

    Madeira, A.A.; Souza Gouvea, L. de; Galetti, M.R.S.

    1988-01-01

    The implantation of TRAC-PF1 code, IBM version, in VAX-11/750 is described. This work provides Reator Department with an advanced best-estimate tool to perform loss-of-coolant accident analysis. (Author) [pt

  20. CSAU (Code Scaling, Applicability and Uncertainty)

    International Nuclear Information System (INIS)

    Wilson, G.E.; Boyack, B.E.

    1989-01-01

    Best Estimate computer codes have been accepted by the U.S. Nuclear Regulatory Commission as an optional tool for performing safety analysis related to the licensing and regulation of current nuclear reactors producing commercial electrical power, providing their uncertainty is quantified. In support of this policy change, the NRC and its contractors and consultants have developed and demonstrated an uncertainty quantification methodology called CSAU. The primary use of the CSAU methodology is to quantify safety margins for existing designs; however, the methodology can also serve an equally important role in advanced reactor research for plants not yet built. This paper describes the CSAU methodology, at the generic process level, and provides the general principles whereby it may be applied to evaluations of advanced reactor designs

  1. Methodology, status and plans for development and assessment of TUF and CATHENA codes

    Energy Technology Data Exchange (ETDEWEB)

    Luxat, J.C.; Liu, W.S.; Leung, R.K. [Ontario Hydro, Toronto (Canada)] [and others

    1997-07-01

    An overview is presented of the Canadian two-fluid computer codes TUF and CATHENA with specific focus on the constraints imposed during development of these codes and the areas of application for which they are intended. Additionally a process for systematic assessment of these codes is described which is part of a broader, industry based initiative for validation of computer codes used in all major disciplines of safety analysis. This is intended to provide both the licensee and the regulator in Canada with an objective basis for assessing the adequacy of codes for use in specific applications. Although focused specifically on CANDU reactors, Canadian experience in developing advanced two-fluid codes to meet wide-ranging application needs while maintaining past investment in plant modelling provides a useful contribution to international efforts in this area.

  2. Methodology, status and plans for development and assessment of TUF and CATHENA codes

    International Nuclear Information System (INIS)

    Luxat, J.C.; Liu, W.S.; Leung, R.K.

    1997-01-01

    An overview is presented of the Canadian two-fluid computer codes TUF and CATHENA with specific focus on the constraints imposed during development of these codes and the areas of application for which they are intended. Additionally a process for systematic assessment of these codes is described which is part of a broader, industry based initiative for validation of computer codes used in all major disciplines of safety analysis. This is intended to provide both the licensee and the regulator in Canada with an objective basis for assessing the adequacy of codes for use in specific applications. Although focused specifically on CANDU reactors, Canadian experience in developing advanced two-fluid codes to meet wide-ranging application needs while maintaining past investment in plant modelling provides a useful contribution to international efforts in this area

  3. A computer code for analysis of severe accidents in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    The ICARE2 computer code, developed and validated since 1988 at IPSN (nuclear safety and protection institute), calculates in a mechanistic way the physical and chemical phenomena involved in the core degradation process during possible severe accidents in LWR's. The coupling between ICARE2 and the best-estimate thermal-hydraulics code CATHARE2 was completed at IPSN and led to the release of a first ICARE/CATHARE V1 version in 1999, followed by 2 successive revisions in 2000 and 2001. This documents gathers all the contributions presented at the first international ICARE/CATHARE users'club seminar that took place in November 2001. This seminar was characterized by a high quality and variety of the presentations, showing an increase of reactor applications and user needs in this area (2D/3D aspects, reflooding, corium slumping into the lower head,...). 2 sessions were organized. The first one was dedicated to the applications of ICARE2 V3mod1 against small-scale experiments such as PHEBUS FPT2 and FPT3 tests, PHEBUS AIC, QUENCH experiments, NRU-FLHT-5 test, ACRR-MP1 and DC1 experiments, CORA-PWR tests, and PBF-SFD1.4 test. The second session involved ICARE/CATHARE V1mod1 reactor applications and users'guidelines. Among reactor applications we found: code applicability to high burn-up fuel rods, simulation of the TMI-2 transient, simulation of a PWR-900 high pressure severe accident sequence, and the simulation of a VVER-1000 large break LOCA scenario. (A.C.)

  4. A computer code for analysis of severe accidents in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    The ICARE2 computer code, developed and validated since 1988 at IPSN (nuclear safety and protection institute), calculates in a mechanistic way the physical and chemical phenomena involved in the core degradation process during possible severe accidents in LWR's. The coupling between ICARE2 and the best-estimate thermal-hydraulics code CATHARE2 was completed at IPSN and led to the release of a first ICARE/CATHARE V1 version in 1999, followed by 2 successive revisions in 2000 and 2001. This documents gathers all the contributions presented at the first international ICARE/CATHARE users'club seminar that took place in November 2001. This seminar was characterized by a high quality and variety of the presentations, showing an increase of reactor applications and user needs in this area (2D/3D aspects, reflooding, corium slumping into the lower head,...). 2 sessions were organized. The first one was dedicated to the applications of ICARE2 V3mod1 against small-scale experiments such as PHEBUS FPT2 and FPT3 tests, PHEBUS AIC, QUENCH experiments, NRU-FLHT-5 test, ACRR-MP1 and DC1 experiments, CORA-PWR tests, and PBF-SFD1.4 test. The second session involved ICARE/CATHARE V1mod1 reactor applications and users'guidelines. Among reactor applications we found: code applicability to high burn-up fuel rods, simulation of the TMI-2 transient, simulation of a PWR-900 high pressure severe accident sequence, and the simulation of a VVER-1000 large break LOCA scenario. (A.C.)

  5. A computer code for analysis of severe accidents in LWRs

    International Nuclear Information System (INIS)

    2001-01-01

    The ICARE2 computer code, developed and validated since 1988 at IPSN (nuclear safety and protection institute), calculates in a mechanistic way the physical and chemical phenomena involved in the core degradation process during possible severe accidents in LWR's. The coupling between ICARE2 and the best-estimate thermal-hydraulics code CATHARE2 was completed at IPSN and led to the release of a first ICARE/CATHARE V1 version in 1999, followed by 2 successive revisions in 2000 and 2001. This documents gathers all the contributions presented at the first international ICARE/CATHARE users'club seminar that took place in November 2001. This seminar was characterized by a high quality and variety of the presentations, showing an increase of reactor applications and user needs in this area (2D/3D aspects, reflooding, corium slumping into the lower head,...). 2 sessions were organized. The first one was dedicated to the applications of ICARE2 V3mod1 against small-scale experiments such as PHEBUS FPT2 and FPT3 tests, PHEBUS AIC, QUENCH experiments, NRU-FLHT-5 test, ACRR-MP1 and DC1 experiments, CORA-PWR tests, and PBF-SFD1.4 test. The second session involved ICARE/CATHARE V1mod1 reactor applications and users'guidelines. Among reactor applications we found: code applicability to high burn-up fuel rods, simulation of the TMI-2 transient, simulation of a PWR-900 high pressure severe accident sequence, and the simulation of a VVER-1000 large break LOCA scenario. (A.C.)

  6. Existing pavement input information for the mechanistic-empirical pavement design guide.

    Science.gov (United States)

    2009-02-01

    The objective of this study is to systematically evaluate the Iowa Department of Transportations (DOTs) existing Pavement Management Information System (PMIS) with respect to the input information required for Mechanistic-Empirical Pavement Des...

  7. Advances in Reactor physics, mathematics and computation. Volume 3

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, volume 3, are divided into sessions bearing on: - poster sessions on benchmark and codes: 35 conferences - review of status of assembly spectrum codes: 9 conferences - Numerical methods in fluid mechanics and thermal hydraulics: 16 conferences - stochastic transport and methods: 7 conferences.

  8. Automatic coding method of the ACR Code

    International Nuclear Information System (INIS)

    Park, Kwi Ae; Ihm, Jong Sool; Ahn, Woo Hyun; Baik, Seung Kook; Choi, Han Yong; Kim, Bong Gi

    1993-01-01

    The authors developed a computer program for automatic coding of ACR(American College of Radiology) code. The automatic coding of the ACR code is essential for computerization of the data in the department of radiology. This program was written in foxbase language and has been used for automatic coding of diagnosis in the Department of Radiology, Wallace Memorial Baptist since May 1992. The ACR dictionary files consisted of 11 files, one for the organ code and the others for the pathology code. The organ code was obtained by typing organ name or code number itself among the upper and lower level codes of the selected one that were simultaneous displayed on the screen. According to the first number of the selected organ code, the corresponding pathology code file was chosen automatically. By the similar fashion of organ code selection, the proper pathologic dode was obtained. An example of obtained ACR code is '131.3661'. This procedure was reproducible regardless of the number of fields of data. Because this program was written in 'User's Defined Function' from, decoding of the stored ACR code was achieved by this same program and incorporation of this program into program in to another data processing was possible. This program had merits of simple operation, accurate and detail coding, and easy adjustment for another program. Therefore, this program can be used for automation of routine work in the department of radiology

  9. Reactive flow simulation in complex 3D geometries using the COM3D code

    International Nuclear Information System (INIS)

    Breitung, W.; Kotchourko, A.; Veser, A.; Scholtyssek, W.

    1999-01-01

    The COM3D code, under development at the Forschungszentrum Karlsruhe (FZK), is a 3-d CFD code to describe turbulent combustion phenomena in complex geometries. It is intended to be part of the advanced integral code system for containment analysis (INCA) which includes in addition GASFLOW for distribution calculations, V3D for slow combustion and DET3D for detonation analysis. COM3D uses a TVD-solver and optional models for turbulence, chemistry and thermodynamics. The hydrodynamic model considers mass, momentum and energy conservation. Advanced procedures were provided to facilitate grid-development for complex 3-d structures. COM3D was validated on experiments performed on different scales with generally good agreement for important physical quantities. The code was applied to combustion analysis of a large PWR. The initial conditions were obtained from a GASFLOW distribution analysis for a LOOP scenario. Results are presented concerning flame propagation and pressure evolution in the containment which clearly demonstrate the effects of internal structures, their influence on turbulence formation and consequences for local loads. (author)

  10. Development of generalized boiling transition model applicable for wide variety of fuel bundle geometries. Basic strategy and numerical approaches

    International Nuclear Information System (INIS)

    Ninokata, Hisashi; Sadatomi, Michio; Okawa, Tomio

    2003-01-01

    In order to establish a key technology to realize advanced BWR fuel designs, a three-year project of the advanced subchannel analysis code development had been started since 2002. The five dominant factors involved in the boiling transitional process in the fuel bundles were focused. They are, (1) inter-subchannel exchanges, (2) influences of obstacles (3) dryout of liquid film, (4) transition of two-phase flow regimes and (5) deposition of droplets. It has been recognized that present physical models or constitutive equations in subchannel formulations need to be improved so that they include geometrical effects in the fuel bundle design more mechanistically and universally. Through reviewing literatures and existent experimental results, underlying elementary processes and geometrical factors that are indispensable for improving subchannel codes were identified. The basic strategy that combines numerical and experimental approaches was proposed aiming at establishment of mechanistic models for the five dominant factors. In this paper, the present status of methodologies for detailed two-phase flow studies has been summarized. According to spatial scales of focused elementary processes, proper numerical approaches were selected. For some promising numerical approaches, preliminary calcitonins were performed for assessing their applicability to investigation of elementary processes involved in the boiling transition. (author)

  11. Mechanistic modeling of reactive soil nitrogen emissions across agricultural management practices

    Science.gov (United States)

    Rasool, Q. Z.; Miller, D. J.; Bash, J. O.; Venterea, R. T.; Cooter, E. J.; Hastings, M. G.; Cohan, D. S.

    2017-12-01

    The global reactive nitrogen (N) budget has increased by a factor of 2-3 from pre-industrial levels. This increase is especially pronounced in highly N fertilized agricultural regions in summer. The reactive N emissions from soil to atmosphere can be in reduced (NH3) or oxidized (NO, HONO, N2O) forms, depending on complex biogeochemical transformations of soil N reservoirs. Air quality models like CMAQ typically neglect soil emissions of HONO and N2O. Previously, soil NO emissions estimated by models like CMAQ remained parametric and inconsistent with soil NH3 emissions. Thus, there is a need to more mechanistically and consistently represent the soil N processes that lead to reactive N emissions to the atmosphere. Our updated approach estimates soil NO, HONO and N2O emissions by incorporating detailed agricultural fertilizer inputs from EPIC, and CMAQ-modeled N deposition, into the soil N pool. EPIC addresses the nitrification, denitrification and volatilization rates along with soil N pools for agricultural soils. Suitable updates to account for factors like nitrite (NO2-) accumulation not addressed in EPIC, will also be made. The NO and N2O emissions from nitrification and denitrification are computed mechanistically using the N sub-model of DAYCENT. These mechanistic definitions use soil water content, temperature, NH4+ and NO3- concentrations, gas diffusivity and labile C availability as dependent parameters at various soil layers. Soil HONO emissions found to be most probable under high NO2- availability will be based on observed ratios of HONO to NO emissions under different soil moistures, pH and soil types. The updated scheme will utilize field-specific soil properties and N inputs across differing manure management practices such as tillage. Comparison of the modeled soil NO emission rates from the new mechanistic and existing schemes against field measurements will be discussed. Our updated framework will help to predict the diurnal and daily variability

  12. SSC-K code user's manual

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Y M; Lee, Y B; Chang, W P; Hahn, D

    2000-07-01

    The Supper System Code of KAERI (SSC-K) is a best-estimate system code for analyzing a variety of off-normal or accidents in the heat transport system of a pool type LMR design. It is being developed at Korea Atomic Energy Research Inititution (KAERI) on the basis of SSC-L, originally developed at BNL to analyze loop-type LMR transients. SSC-K can handle both designs of loop and pool type LMRs. SSC-K contains detailed mechanistic models of transient thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, coolant, fuel elements, and structures to accident conditions. This report provides an overview of recent model developmentsvfor the SSC-K computer code, focusing on phenomenological model descriptions for new thermal, hydraulic, neutronic, and mechnaical modules. A comprehensive description of the models for pool-type reactor is given in Chapters 2 and 3; the steady-state plant characterization, prior to the initiation of transient is described in Chapter 2 and their transient counterparts are discussed in Chapter 3. In Chapter 4, a discussion on the intermediate heat exchanger (IHX) is presented. The IHX model of SSC-K is similar to that used in the SSC-L, except for some changes required for the pool-type configuration of reactor vessel. In Chapter 5, an electromagnetic (EM) pump is modeled as a component. There are two pump choices available in SSC-K; a centrifugal pump which was originally imbedded into the SSC-L, and an EM pump which was introduced for the KALIMER design. In Chapter 6, a model of passive safety decay heat removal system(PSDRS) is discussed, which removes decay heat through the reactor and containment vessel walls to the ambient air heat sink. In Chapter 7, models for various reactivity feedback effects are discussed. Reactivity effects of importance in fast reactor include the Doppler effect, effects of sodium density changes, effects of dimensional changes in core geometry. Finally in Chapter 8

  13. Validation of OPERA3D PCMI Analysis Code

    Energy Technology Data Exchange (ETDEWEB)

    Jeun, Ji Hoon; Choi, Jae Myung; Yoo, Jong Sung [KEPCO Nuclear Fuel Co., Daejeon (Korea, Republic of); Cheng, G.; Sim, K. S.; Chassie, Girma [Candu Energy INC.,Ontario (Canada)

    2013-10-15

    This report will describe introduction of validation of OPERA3D code, and validation results that are directly related with PCMI phenomena. OPERA3D was developed for the PCMI analysis and validated using the in-pile measurement data. Fuel centerline temperature and clad strain calculation results shows close expectations with measurement data. Moreover, 3D FEM fuel model of OPERA3D shows slight hour glassing behavior of fuel pellet in contact case. Further optimization will be conducted for future application of OPERA3D code. Nuclear power plant consists of many complicated systems, and one of the important objects of all the systems is maintaining nuclear fuel integrity. However, it is inevitable to experience PCMI (Pellet Cladding Mechanical Interaction) phenomena at current operating reactors and next generation reactors for advanced safety and economics as well. To evaluate PCMI behavior, many studies are on-going to develop 3-dimensional fuel performance evaluation codes. Moreover, these codes are essential to set the safety limits for the best estimated PCMI phenomena aimed for high burnup fuel.

  14. Profiling the biological activity of oxide nanomaterials with mechanistic models

    NARCIS (Netherlands)

    Burello, E.

    2013-01-01

    In this study we present three mechanistic models for profiling the potential biological and toxicological effects of oxide nanomaterials. The models attempt to describe the reactivity, protein adsorption and membrane adhesion processes of a large range of oxide materials and are based on properties

  15. Generative Mechanistic Explanation Building in Undergraduate Molecular and Cellular Biology

    Science.gov (United States)

    Southard, Katelyn M.; Espindola, Melissa R.; Zaepfel, Samantha D.; Bolger, Molly S.

    2017-01-01

    When conducting scientific research, experts in molecular and cellular biology (MCB) use specific reasoning strategies to construct mechanistic explanations for the underlying causal features of molecular phenomena. We explored how undergraduate students applied this scientific practice in MCB. Drawing from studies of explanation building among…

  16. The APS SASE FEL: modeling and code comparison

    International Nuclear Information System (INIS)

    Biedron, S. G.

    1999-01-01

    A self-amplified spontaneous emission (SASE) free-electron laser (FEL) is under construction at the Advanced Photon Source (APS). Five FEL simulation codes were used in the design phase: GENESIS, GINGER, MEDUSA, RON, and TDA3D. Initial comparisons between each of these independent formulations show good agreement for the parameters of the APS SASE FEL

  17. Melanie Klein's metapsychology: phenomenological and mechanistic perspective.

    Science.gov (United States)

    Mackay, N

    1981-01-01

    Freud's metapsychology is the subject of an important debate. This is over whether psychoanalysis is best construed as a science of the natural science type or as a special human science. The same debate applies to Melanie Klein's work. In Klein's metapsychology are two different and incompatible models of explanation. One is taken over from Freud's structural theory and appears to be similarly mechanistic. The other is clinically based and phenomenological. These two are discussed with special reference to the concepts of "phantasy" and "internal object".

  18. Improvement and test calculation on basic code or sodium-water reaction jet

    Energy Technology Data Exchange (ETDEWEB)

    Saito, Yoshinori; Itooka, Satoshi [Advanced Reactor Engineering Center, Hitachi Works, Hitachi Ltd., Hitachi, Ibaraki (Japan); Okabe, Ayao; Fujimata, Kazuhiro; Sakurai, Tomoo [Consulting Engineering Dept., Hitachi Engineering Co., Ltd., Hitachi, Ibaraki (Japan)

    1999-03-01

    In selecting the reasonable DBL (design basis water leak rate) on steam generator (SG), it is necessary to improve analytical method for estimating the sodium temperature on failure propagation due to overheating. Improvement on the basic code for sodium-water reaction (SWR) jet was performed for an actual scale SG. The improvement points of the code are as follows; (1) introduction of advanced model such as heat transfer between the jet and structure (tube array), cooling effect of the structure, heat transfer between analytic cells, and (2) model improvement for heat transfer between two-phase flow and porous-media. The test calculation using the improved code (LEAP-JET ver.1.30) were carried out with conditions of the SWAT-3{center_dot}Run-19 test and an actual scale SG. It is confirmed that the SWR jet behavior on the results is reasonable and Influence to analysis result of a model. Code integration with the blow down analytic code (LEAP-BLOW) was also studied. It is suitable that LEAP-JET was improved as one of the LEAP-BLOW's models, and it was integrated into this. In addition to above, the improvement for setting of boundary condition and the development of the interface program to transfer the analytical results of LEAP-BLOW have been performed in order to consider the cooling effect of coolant in the tube simply. However, verification of the code by new SWAT-1 and SWAT-3 test data planned in future is necessary because LEAP-JET is under development. And furthermore advancement needs to be planned. (author)

  19. Improvement and test calculation on basic code or sodium-water reaction jet

    International Nuclear Information System (INIS)

    Saito, Yoshinori; Itooka, Satoshi; Okabe, Ayao; Fujimata, Kazuhiro; Sakurai, Tomoo

    1999-03-01

    In selecting the reasonable DBL (design basis water leak rate) on steam generator (SG), it is necessary to improve analytical method for estimating the sodium temperature on failure propagation due to overheating. Improvement on the basic code for sodium-water reaction (SWR) jet was performed for an actual scale SG. The improvement points of the code are as follows; (1) introduction of advanced model such as heat transfer between the jet and structure (tube array), cooling effect of the structure, heat transfer between analytic cells, and (2) model improvement for heat transfer between two-phase flow and porous-media. The test calculation using the improved code (LEAP-JET ver.1.30) were carried out with conditions of the SWAT-3·Run-19 test and an actual scale SG. It is confirmed that the SWR jet behavior on the results is reasonable and Influence to analysis result of a model. Code integration with the blow down analytic code (LEAP-BLOW) was also studied. It is suitable that LEAP-JET was improved as one of the LEAP-BLOW's models, and it was integrated into this. In addition to above, the improvement for setting of boundary condition and the development of the interface program to transfer the analytical results of LEAP-BLOW have been performed in order to consider the cooling effect of coolant in the tube simply. However, verification of the code by new SWAT-1 and SWAT-3 test data planned in future is necessary because LEAP-JET is under development. And furthermore advancement needs to be planned. (author)

  20. Quantitative assessment of biological impact using transcriptomic data and mechanistic network models

    International Nuclear Information System (INIS)

    Thomson, Ty M.; Sewer, Alain; Martin, Florian; Belcastro, Vincenzo; Frushour, Brian P.; Gebel, Stephan; Park, Jennifer; Schlage, Walter K.; Talikka, Marja; Vasilyev, Dmitry M.; Westra, Jurjen W.; Hoeng, Julia; Peitsch, Manuel C.

    2013-01-01

    Exposure to biologically active substances such as therapeutic drugs or environmental toxicants can impact biological systems at various levels, affecting individual molecules, signaling pathways, and overall cellular processes. The ability to derive mechanistic insights from the resulting system responses requires the integration of experimental measures with a priori knowledge about the system and the interacting molecules therein. We developed a novel systems biology-based methodology that leverages mechanistic network models and transcriptomic data to quantitatively assess the biological impact of exposures to active substances. Hierarchically organized network models were first constructed to provide a coherent framework for investigating the impact of exposures at the molecular, pathway and process levels. We then validated our methodology using novel and previously published experiments. For both in vitro systems with simple exposure and in vivo systems with complex exposures, our methodology was able to recapitulate known biological responses matching expected or measured phenotypes. In addition, the quantitative results were in agreement with experimental endpoint data for many of the mechanistic effects that were assessed, providing further objective confirmation of the approach. We conclude that our methodology evaluates the biological impact of exposures in an objective, systematic, and quantifiable manner, enabling the computation of a systems-wide and pan-mechanistic biological impact measure for a given active substance or mixture. Our results suggest that various fields of human disease research, from drug development to consumer product testing and environmental impact analysis, could benefit from using this methodology. - Highlights: • The impact of biologically active substances is quantified at multiple levels. • The systems-level impact integrates the perturbations of individual networks. • The networks capture the relationships between

  1. Methods for studying fuel management in advanced gas cooled reactors

    International Nuclear Information System (INIS)

    Buckler, A.N.; Griggs, C.F.; Tyror, J.G.

    1971-07-01

    The methods used for studying fuel and absorber management problems in AGRs are described. The basis of the method is the use of ARGOSY lattice data in reactor calculations performed at successive time steps. These reactor calculations may be quite crude but for advanced design calculations a detailed channel-by-channel representation of the whole core is required. The main emphasis of the paper is in describing such an advanced approach - the ODYSSEUS-6 code. This code evaluates reactor power distributions as a function of time and uses the information to select refuelling moves and determine controller positions. (author)

  2. Time parallelization of advanced operation scenario simulations of ITER plasma

    International Nuclear Information System (INIS)

    Samaddar, D; Casper, T A; Kim, S H; Houlberg, W A; Berry, L A; Elwasif, W R; Batchelor, D

    2013-01-01

    This work demonstrates that simulations of advanced burning plasma operation scenarios can be successfully parallelized in time using the parareal algorithm. CORSICA -an advanced operation scenario code for tokamak plasmas is used as a test case. This is a unique application since the parareal algorithm has so far been applied to relatively much simpler systems except for the case of turbulence. In the present application, a computational gain of an order of magnitude has been achieved which is extremely promising. A successful implementation of the Parareal algorithm to codes like CORSICA ushers in the possibility of time efficient simulations of ITER plasmas.

  3. Coding in pigeons: Multiple-coding versus single-code/default strategies.

    Science.gov (United States)

    Pinto, Carlos; Machado, Armando

    2015-05-01

    To investigate the coding strategies that pigeons may use in a temporal discrimination tasks, pigeons were trained on a matching-to-sample procedure with three sample durations (2s, 6s and 18s) and two comparisons (red and green hues). One comparison was correct following 2-s samples and the other was correct following both 6-s and 18-s samples. Tests were then run to contrast the predictions of two hypotheses concerning the pigeons' coding strategies, the multiple-coding and the single-code/default. According to the multiple-coding hypothesis, three response rules are acquired, one for each sample. According to the single-code/default hypothesis, only two response rules are acquired, one for the 2-s sample and a "default" rule for any other duration. In retention interval tests, pigeons preferred the "default" key, a result predicted by the single-code/default hypothesis. In no-sample tests, pigeons preferred the key associated with the 2-s sample, a result predicted by multiple-coding. Finally, in generalization tests, when the sample duration equaled 3.5s, the geometric mean of 2s and 6s, pigeons preferred the key associated with the 6-s and 18-s samples, a result predicted by the single-code/default hypothesis. The pattern of results suggests the need for models that take into account multiple sources of stimulus control. © Society for the Experimental Analysis of Behavior.

  4. Perspectives on the development of next generation reactor systems safety analysis codes

    International Nuclear Information System (INIS)

    Zhang, H.

    2015-01-01

    'Full text:' Existing reactor system analysis codes, such as RELAP5-3D and TRAC, have gained worldwide success in supporting reactor safety analyses, as well as design and licensing of new reactors. These codes are important assets to the nuclear engineering research community, as well as to the nuclear industry. However, most of these codes were originally developed during the 1970s', and it becomes necessary to develop next-generation reactor system analysis codes for several reasons. Firstly, as new reactor designs emerge, there are new challenges emerging in numerical simulations of reactor systems such as long lasting transients and multi-physics phenomena. These new requirements are beyond the range of applicability of the existing system analysis codes. Advanced modeling and numerical methods must be taken into consideration to improve the existing capabilities. Secondly, by developing next-generation reactor system analysis codes, the knowledge (know how) in two phase flow modeling and the highly complex constitutive models will be transferred to the young generation of nuclear engineers. And thirdly, all computer codes have limited shelf life. It becomes less and less cost-effective to maintain a legacy code, due to the fast change of computer hardware and software environment. There are several critical perspectives in terms of developing next-generation reactor system analysis codes: 1) The success of the next-generation codes must be built upon the success of the existing codes. The knowledge of the existing codes, not just simply the manuals and codes, but knowing why and how, must be transferred to the next-generation codes. The next-generation codes should encompass the capability of the existing codes. The shortcomings of existing codes should be identified, understood, and properly categorized, for example into model deficiencies or numerical method deficiencies. 2) State-of-the-art models and numerical methods must be considered to

  5. Perspectives on the development of next generation reactor systems safety analysis codes

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, H., E-mail: Hongbin.Zhang@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States)

    2015-07-01

    'Full text:' Existing reactor system analysis codes, such as RELAP5-3D and TRAC, have gained worldwide success in supporting reactor safety analyses, as well as design and licensing of new reactors. These codes are important assets to the nuclear engineering research community, as well as to the nuclear industry. However, most of these codes were originally developed during the 1970s', and it becomes necessary to develop next-generation reactor system analysis codes for several reasons. Firstly, as new reactor designs emerge, there are new challenges emerging in numerical simulations of reactor systems such as long lasting transients and multi-physics phenomena. These new requirements are beyond the range of applicability of the existing system analysis codes. Advanced modeling and numerical methods must be taken into consideration to improve the existing capabilities. Secondly, by developing next-generation reactor system analysis codes, the knowledge (know how) in two phase flow modeling and the highly complex constitutive models will be transferred to the young generation of nuclear engineers. And thirdly, all computer codes have limited shelf life. It becomes less and less cost-effective to maintain a legacy code, due to the fast change of computer hardware and software environment. There are several critical perspectives in terms of developing next-generation reactor system analysis codes: 1) The success of the next-generation codes must be built upon the success of the existing codes. The knowledge of the existing codes, not just simply the manuals and codes, but knowing why and how, must be transferred to the next-generation codes. The next-generation codes should encompass the capability of the existing codes. The shortcomings of existing codes should be identified, understood, and properly categorized, for example into model deficiencies or numerical method deficiencies. 2) State-of-the-art models and numerical methods must be considered to

  6. Analysis of ATLAS Cold Leg SBLOCA Using SPACE Code

    International Nuclear Information System (INIS)

    Kang, Doo Hyuk; Suh, Jae Seung; Kim, Se Yun

    2012-01-01

    SPACE Code has been developed to predict the thermal-hydraulic responses of nuclear steam supply system to the anticipated transients and postulated accidents and adopted advanced physical modeling of two-phase flows, mainly two-fluid, three-field models that comprise gas, continuous liquid, and droplet fields and has the capability to simulate 3D effects by the use of structured and/or non-structured meshes. In this paper, a cold-leg SBLOCA which is the experiment, SB-CL-09, of the ATLAS integral effect test facility during the second domestic stand problem (DSP-02) was analyzed. The results were compared with those of MARS-KS code simulations. The SPACE code with a 1.0 version was released by KHNP in 2012. The analysis has been performed in a desktop PC with Windows 7 environment

  7. Lattices applied to coding for reliable and secure communications

    CERN Document Server

    Costa, Sueli I R; Campello, Antonio; Belfiore, Jean-Claude; Viterbo, Emanuele

    2017-01-01

    This book provides a first course on lattices – mathematical objects pertaining to the realm of discrete geometry, which are of interest to mathematicians for their structure and, at the same time, are used by electrical and computer engineers working on coding theory and cryptography. The book presents both fundamental concepts and a wealth of applications, including coding and transmission over Gaussian channels, techniques for obtaining lattices from finite prime fields and quadratic fields, constructions of spherical codes, and hard lattice problems used in cryptography. The topics selected are covered in a level of detail not usually found in reference books. As the range of applications of lattices continues to grow, this work will appeal to mathematicians, electrical and computer engineers, and graduate or advanced undergraduate in these fields.

  8. ASTEC V2 severe accident integral code: Fission product modelling and validation

    International Nuclear Information System (INIS)

    Cantrel, L.; Cousin, F.; Bosland, L.; Chevalier-Jabet, K.; Marchetto, C.

    2014-01-01

    One main goal of the severe accident integral code ASTEC V2, jointly developed since almost more than 15 years by IRSN and GRS, is to simulate the overall behaviour of fission products (FP) in a damaged nuclear facility. ASTEC applications are source term determinations, level 2 Probabilistic Safety Assessment (PSA2) studies including the determination of uncertainties, accident management studies and physical analyses of FP experiments to improve the understanding of the phenomenology. ASTEC is a modular code and models of a part of the phenomenology are implemented in each module: the release of FPs and structural materials from degraded fuel in the ELSA module; the transport through the reactor coolant system approximated as a sequence of control volumes in the SOPHAEROS module; and the radiochemistry inside the containment nuclear building in the IODE module. Three other modules, CPA, ISODOP and DOSE, allow respectively computing the deposition rate of aerosols inside the containment, the activities of the isotopes as a function of time, and the gaseous dose rate which is needed to model radiochemistry in the gaseous phase. In ELSA, release models are semi-mechanistic and have been validated for a wide range of experimental data, and noticeably for VERCORS experiments. For SOPHAEROS, the models can be divided into two parts: vapour phase phenomena and aerosol phase phenomena. For IODE, iodine and ruthenium chemistry are modelled based on a semi-mechanistic approach, these FPs can form some volatile species and are particularly important in terms of potential radiological consequences. The models in these 3 modules are based on a wide experimental database, resulting for a large part from international programmes, and they are considered at the state of the art of the R and D knowledge. This paper illustrates some FPs modelling capabilities of ASTEC and computed values are compared to some experimental results, which are parts of the validation matrix

  9. Advancing Risk Assessment through the Application of Systems Toxicology

    Science.gov (United States)

    Sauer, John Michael; Kleensang, André; Peitsch, Manuel C.; Hayes, A. Wallace

    2016-01-01

    Risk assessment is the process of quantifying the probability of a harmful effect to individuals or populations from human activities. Mechanistic approaches to risk assessment have been generally referred to as systems toxicology. Systems toxicology makes use of advanced analytical and computational tools to integrate classical toxicology and quantitative analysis of large networks of molecular and functional changes occurring across multiple levels of biological organization. Three presentations including two case studies involving both in vitro and in vivo approaches described the current state of systems toxicology and the potential for its future application in chemical risk assessment. PMID:26977253

  10. Development of CAP code for nuclear power plant containment: Lumped model

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Soon Joon, E-mail: sjhong90@fnctech.com [FNC Tech. Co. Ltd., Heungdeok 1 ro 13, Giheung-gu, Yongin-si, Gyeonggi-do 446-908 (Korea, Republic of); Choo, Yeon Joon; Hwang, Su Hyun; Lee, Byung Chul [FNC Tech. Co. Ltd., Heungdeok 1 ro 13, Giheung-gu, Yongin-si, Gyeonggi-do 446-908 (Korea, Republic of); Ha, Sang Jun [Central Research Institute, Korea Hydro & Nuclear Power Company, Ltd., 70, 1312-gil, Yuseong-daero, Yuseong-gu, Daejeon 305-343 (Korea, Republic of)

    2015-09-15

    Highlights: • State-of-art containment analysis code, CAP, has been developed. • CAP uses 3-field equations, water level oriented upwind scheme, local head model. • CAP has a function of linked calculation with reactor coolant system code. • CAP code assessments showed appropriate prediction capabilities. - Abstract: CAP (nuclear Containment Analysis Package) code has been developed in Korean nuclear society for the analysis of nuclear containment thermal hydraulic behaviors including pressure and temperature trends and hydrogen concentration. Lumped model of CAP code uses 2-phase, 3-field equations for fluid behaviors, and has appropriate constitutive equations, 1-dimensional heat conductor model, component models, trip and control models, and special process models. CAP can run in a standalone mode or a linked mode with a reactor coolant system analysis code. The linked mode enables the more realistic calculation of a containment response and is expected to be applicable to a more complicated advanced plant design calculation. CAP code assessments were carried out by gradual approaches: conceptual problems, fundamental phenomena, component and principal phenomena, experimental validation, and finally comparison with other code calculations on the base of important phenomena identifications. The assessments showed appropriate prediction capabilities of CAP.

  11. Development of CAP code for nuclear power plant containment: Lumped model

    International Nuclear Information System (INIS)

    Hong, Soon Joon; Choo, Yeon Joon; Hwang, Su Hyun; Lee, Byung Chul; Ha, Sang Jun

    2015-01-01

    Highlights: • State-of-art containment analysis code, CAP, has been developed. • CAP uses 3-field equations, water level oriented upwind scheme, local head model. • CAP has a function of linked calculation with reactor coolant system code. • CAP code assessments showed appropriate prediction capabilities. - Abstract: CAP (nuclear Containment Analysis Package) code has been developed in Korean nuclear society for the analysis of nuclear containment thermal hydraulic behaviors including pressure and temperature trends and hydrogen concentration. Lumped model of CAP code uses 2-phase, 3-field equations for fluid behaviors, and has appropriate constitutive equations, 1-dimensional heat conductor model, component models, trip and control models, and special process models. CAP can run in a standalone mode or a linked mode with a reactor coolant system analysis code. The linked mode enables the more realistic calculation of a containment response and is expected to be applicable to a more complicated advanced plant design calculation. CAP code assessments were carried out by gradual approaches: conceptual problems, fundamental phenomena, component and principal phenomena, experimental validation, and finally comparison with other code calculations on the base of important phenomena identifications. The assessments showed appropriate prediction capabilities of CAP

  12. Code Cactus; Code Cactus

    Energy Technology Data Exchange (ETDEWEB)

    Fajeau, M; Nguyen, L T; Saunier, J [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

    1966-09-01

    This code handles the following problems: -1) Analysis of thermal experiments on a water loop at high or low pressure; steady state or transient behavior; -2) Analysis of thermal and hydrodynamic behavior of water-cooled and moderated reactors, at either high or low pressure, with boiling permitted; fuel elements are assumed to be flat plates: - Flowrate in parallel channels coupled or not by conduction across plates, with conditions of pressure drops or flowrate, variable or not with respect to time is given; the power can be coupled to reactor kinetics calculation or supplied by the code user. The code, containing a schematic representation of safety rod behavior, is a one dimensional, multi-channel code, and has as its complement (FLID), a one-channel, two-dimensional code. (authors) [French] Ce code permet de traiter les problemes ci-dessous: 1. Depouillement d'essais thermiques sur boucle a eau, haute ou basse pression, en regime permanent ou transitoire; 2. Etudes thermiques et hydrauliques de reacteurs a eau, a plaques, a haute ou basse pression, ebullition permise: - repartition entre canaux paralleles, couples on non par conduction a travers plaques, pour des conditions de debit ou de pertes de charge imposees, variables ou non dans le temps; - la puissance peut etre couplee a la neutronique et une representation schematique des actions de securite est prevue. Ce code (Cactus) a une dimension d'espace et plusieurs canaux, a pour complement Flid qui traite l'etude d'un seul canal a deux dimensions. (auteurs)

  13. Comparison of Two Commercial FE-Codes for Sheet Metal Forming

    International Nuclear Information System (INIS)

    Revuelta, A.; Larkiola, J.; Kanervo, K.; Korhonen, A. S.; Myllykoski, P.

    2007-01-01

    There is urgent need to develop new advanced fast and cost-effective mass-production methods for small sheet metal components. Traditionally progressive dies have been designed by using various CAD techniques. Recent results in mass production of small sheet metal parts using progressive dies and a transfer press showed that the tool design time may be cut in up to a half by using 3D finite element simulation of forming. In numerical simulation of sheet metal forming better constitutive models are required to obtain more accurate results, reduce the time for tool design and cut the production costs further. Accurate models are needed to describe the initial yielding, subsequent work hardening and to predict the formability. In this work two commercially available finite element simulation codes, PAM-STAMP and LS-DYNA, were compared in forming of small austenitic stainless steel sheet part for electronic industry. Several constitutive models were used in both codes and the results were compared. Comparisons were made between the same models in each of the codes and also between different models in the same code. Material models ranged from very simple to advanced ones, which took into account anisotropy and both isotropic and kinematic hardening behavior. In order to make a valid comparison we employed similar finite element meshes. The effects of the material models parameters were studied and the results were compared with experiments. The effects of the computational time were also studied

  14. International Code Assessment and Applications Program: Annual report

    International Nuclear Information System (INIS)

    Ting, P.; Hanson, R.; Jenks, R.

    1987-03-01

    This is the first annual report of the International Code Assessment and Applications Program (ICAP). The ICAP was organized by the Office of Nuclear Regulatory Research, United States Nuclear Regulatory Commission (USNRC) in 1985. The ICAP is an international cooperative reactor safety research program planned to continue over a period of approximately five years. To date, eleven European and Asian countries/organizations have joined the program through bilateral agreements with the USNRC. Seven proposed agreements are currently under negotiation. The primary mission of the ICAP is to provide independent assessment of the three major advanced computer codes (RELAP5, TRAC-PWR, and TRAC-BWR) developed by the USNRC. However, program activities can be expected to enhance the assessment process throughout member countries. The codes were developed to calculate the reactor plant response to transients and loss-of-coolant accidents. Accurate prediction of normal and abnormal plant response using the codes enhances procedures and regulations used for the safe operation of the plant and also provides technical basis for assessing the safety margin of future reactor plant designs. The ICAP is providing required assessment data that will contribute to quantification of the code uncertainty for each code. The first annual report is devoted to coverage of program activities and accomplishments during the period between April 1985 and March 1987

  15. MARS CODE MANUAL VOLUME V: Models and Correlations

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Bae, Sung Won; Lee, Seung Wook; Yoon, Churl; Hwang, Moon Kyu; Kim, Kyung Doo; Jeong, Jae Jun

    2010-02-01

    Korea Advanced Energy Research Institute (KAERI) conceived and started the development of MARS code with the main objective of producing a state-of-the-art realistic thermal hydraulic systems analysis code with multi-dimensional analysis capability. MARS achieves this objective by very tightly integrating the one dimensional RELAP5/MOD3 with the multi-dimensional COBRA-TF codes. The method of integration of the two codes is based on the dynamic link library techniques, and the system pressure equation matrices of both codes are implicitly integrated and solved simultaneously. In addition, the Equation-Of-State (EOS) for the light water was unified by replacing the EOS of COBRA-TF by that of the RELAP5. This models and correlations manual provides a complete list of detailed information of the thermal-hydraulic models used in MARS, so that this report would be very useful for the code users. The overall structure of the manual is modeled on the structure of the RELAP5 and as such the layout of the manual is very similar to that of the RELAP. This similitude to RELAP5 input is intentional as this input scheme will allow minimum modification between the inputs of RELAP5 and MARS3.1. MARS3.1 development team would like to express its appreciation to the RELAP5 Development Team and the USNRC for making this manual possible

  16. The impact of medical tourism and the code of medical ethics on advertisement in Nigeria.

    Science.gov (United States)

    Makinde, Olusesan Ayodeji; Brown, Brandon; Olaleye, Olalekan

    2014-01-01

    Advances in management of clinical conditions are being made in several resource poor countries including Nigeria. Yet, the code of medical ethics which bars physician and health practices from advertising the kind of services they render deters these practices. This is worsened by the incursion of medical tourism facilitators (MTF) who continue to market healthcare services across countries over the internet and social media thereby raising ethical questions. A significant review of the advertisement ban in the code of ethics is long overdue. Limited knowledge about advances in medical practice among physicians and the populace, the growing medical tourism industry and its attendant effects, and the possibility of driving brain gain provide evidence to repeal the code. Ethical issues, resistance to change and elitist ideas are mitigating factors working in the opposite direction. The repeal of the code of medical ethics against advertising will undoubtedly favor health facilities in the country that currently cannot advertise the kind of services they render. A repeal or review of this code of medical ethics is necessary with properly laid down guidelines on how advertisements can be and cannot be done.

  17. A mechanistic model on methane oxidation in the rice rhizosphere

    NARCIS (Netherlands)

    Bodegom, van P.M.; Leffelaar, P.A.; Goudriaan, J.

    2001-01-01

    A mechanistic model is presented on the processes leading to methane oxidation in rice rhizosphere. The model is driven by oxygen release from a rice root into anaerobic rice soil. Oxygen is consumed by heterotrophic and methanotrophic respiration, described by double Monod kinetics, and by iron

  18. Wyner-Ziv Coding of Depth Maps Exploiting Color Motion Information

    DEFF Research Database (Denmark)

    Salmistraro, Matteo; Zamarin, Marco; Forchhammer, Søren

    2013-01-01

    Distributed Video Coding of multi-view data and depth maps is an interesting and challenging research field, whose interest is growing thanks to the recent advances in depth estimation and the development of affordable devices able to acquire depth information. In applications like video surveill...

  19. Empirical Analysis of Using Erasure Coding in Outsourcing Data Storage With Provable Security

    Science.gov (United States)

    2016-06-01

    computing and communication technologies become powerful and advanced , people are exchanging a huge amount of data, and they are de- manding more storage...NAVAL POSTGRADUATE SCHOOL MONTEREY, CALIFORNIA THESIS EMPIRICAL ANALYSIS OF USING ERASURE CODING IN OUTSOURCING DATA STORAGEWITH PROVABLE SECURITY by...2015 to 06-17-2016 4. TITLE AND SUBTITLE EMPIRICAL ANALYSIS OF USING ERASURE CODING IN OUTSOURCING DATA STORAGE WITH PROVABLE SECURITY 5. FUNDING

  20. The Altered Hepatic Tubulin Code in Alcoholic Liver Disease.

    Science.gov (United States)

    Groebner, Jennifer L; Tuma, Pamela L

    2015-09-18

    The molecular mechanisms that lead to the progression of alcoholic liver disease have been actively examined for decades. Because the hepatic microtubule cytoskeleton supports innumerable cellular processes, it has been the focus of many such mechanistic studies. It has long been appreciated that α-tubulin is a major target for modification by highly reactive ethanol metabolites and reactive oxygen species. It is also now apparent that alcohol exposure induces post-translational modifications that are part of the natural repertoire, mainly acetylation. In this review, the modifications of the "tubulin code" are described as well as those adducts by ethanol metabolites. The potential cellular consequences of microtubule modification are described with a focus on alcohol-induced defects in protein trafficking and enhanced steatosis. Possible mechanisms that can explain hepatic dysfunction are described and how this relates to the onset of liver injury is discussed. Finally, we propose that agents that alter the cellular acetylation state may represent a novel therapeutic strategy for treating liver disease.

  1. Recent Progress on the Marylie/Impact Beam Dynamics Code

    International Nuclear Information System (INIS)

    Ryne, R.D.; Qiang, J.; Bethel, E.W.; Pogorelov, I.; Shalf, J.; Siegerist, C.; Venturini, M.; Dragt, A.J.; Adelmann, A.; Abell, D.; Amundson, J.; Spentzouris, P.; Neri, F.; Walstrom, P.; Mottershead, C.T.; Samulyak, R.

    2006-01-01

    MARYLIE/IMPACT (ML/I) is a hybrid code that combines the beam optics capabilities of MARYLIE with the parallel Particle-In-Cell capabilities of IMPACT. In addition to combining the capabilities of these codes, ML/I has a number of powerful features, including a choice of Poisson solvers, a fifth-order rf cavity model, multiple reference particles for rf cavities, a library of soft-edge magnet models, representation of magnet systems in terms of coil stacks with possibly overlapping fields, and wakefield effects. The code allows for map production, map analysis, particle tracking, and 3D envelope tracking, all within a single, coherent user environment. ML/I has a front end that can read both MARYLIE input and MAD lattice descriptions. The code can model beams with or without acceleration, and with or without space charge. Developed under a US DOE Scientific Discovery through Advanced Computing (SciDAC) project, ML/I is well suited to large-scale modeling, simulations having been performed with up to 100M macroparticles. The code inherits the powerful fitting and optimizing capabilities of MARYLIE augmented for the new features of ML/I. The combination of soft-edge magnet models, high-order capability, space charge effects, and fitting/optimization capabilities, make ML/I a powerful code for a wide range of beam optics design problems. This paper provides a description of the code and its unique capabilities

  2. Fast neutron fluence evaluation of the smart reactor pressure vessel by using the GEOSHIELD code

    International Nuclear Information System (INIS)

    Kim, K.Y.; Kim, K.S.; Kim, H.Y.; Lee, C.C.; Zee, S.Q.

    2007-01-01

    In Korea, the design of an advanced integral reactor system called SMART has been developed by KAERI to supply energy for seawater desalination as well as an electricity generation. A fast neutron fluence distribution at the SMART reactor pressure vessel was evaluated to confirm the integrity of the vessel by using the GEOSHIELD code. The GEOSHIELD code was developed by KAERI in order to prepare an input list including a geometry modeling of the DORT code and to process results from the DORT code output list. Results by a GEOSHIELD code processing and by a manual processing of the DORT show a good agreement. (author)

  3. The UK core performance code package

    International Nuclear Information System (INIS)

    Hutt, P.K.; Gaines, N.; McEllin, M.; White, R.J.; Halsall, M.J.

    1991-01-01

    Over the last few years work has been co-ordinated by Nuclear Electric, originally part of the Central Electricity Generating Board, with contributions from the United Kingdom Atomic Energy Authority and British Nuclear Fuels Limited, to produce a generic, easy-to-use and integrated package of core performance codes able to perform a comprehensive range of calculations for fuel cycle design, safety analysis and on-line operational support for Light Water Reactor and Advanced Gas Cooled Reactor plant. The package consists of modern rationalized generic codes for lattice physics (WIMS), whole reactor calculations (PANTHER), thermal hydraulics (VIPRE) and fuel performance (ENIGMA). These codes, written in FORTRAN77, are highly portable and new developments have followed modern quality assurance standards. These codes can all be run ''stand-alone'' but they are also being integrated within a new UNIX-based interactive system called the Reactor Physics Workbench (RPW). The RPW provides an interactive user interface and a sophisticated data management system. It offers quality assurance features to the user and has facilities for defining complex calculational sequences. The Paper reviews the current capabilities of these components, their integration within the package and outlines future developments underway. Finally, the Paper describes the development of an on-line version of this package which is now being commissioned on UK AGR stations. (author)

  4. Mathematical Description and Mechanistic Reasoning: A Pathway toward STEM Integration

    Science.gov (United States)

    Weinberg, Paul J.

    2017-01-01

    Because reasoning about mechanism is critical to disciplined inquiry in science, technology, engineering, and mathematics (STEM) domains, this study focuses on ways to support the development of this form of reasoning. This study attends to how mechanistic reasoning is constituted through mathematical description. This study draws upon Smith's…

  5. Integrated analysis of core debris interactions and their effects on containment integrity using the CONTAIN computer code

    International Nuclear Information System (INIS)

    Carroll, D.E.; Bergeron, K.D.; Williams, D.C.; Tills, J.L.; Valdez, G.D.

    1987-01-01

    The CONTAIN computer code includes a versatile system of phenomenological models for analyzing the physical, chemical and radiological conditions inside the containment building during severe reactor accidents. Important contributors to these conditions are the interactions which may occur between released corium and cavity concrete. The phenomena associated with interactions between ejected corium debris and the containment atmosphere (Direct Containment Heating or DCH) also pose a potential threat to containment integrity. In this paper, we describe recent enhancements of the CONTAIN code which allow an integrated analysis of these effects in the presence of other mitigating or aggravating physical processes. In particular, the recent inclusion of the CORCON and VANESA models is described and a calculation example presented. With this capability CONTAIN can model core-concrete interactions occurring simultaneously in multiple compartments and can couple the aerosols thereby generated to the mechanistic description of all atmospheric aerosol components. Also discussed are some recent results of modeling the phenomena involved in Direct Containment Heating. (orig.)

  6. INCORPORATION OF MECHANISTIC INFORMATION IN THE ARSENIC PBPK MODEL DEVELOPMENT PROCESS

    Science.gov (United States)

    INCORPORATING MECHANISTIC INSIGHTS IN A PBPK MODEL FOR ARSENICElaina M. Kenyon, Michael F. Hughes, Marina V. Evans, David J. Thomas, U.S. EPA; Miroslav Styblo, University of North Carolina; Michael Easterling, Analytical Sciences, Inc.A physiologically based phar...

  7. Evaluation Codes from an Affine Veriety Code Perspective

    DEFF Research Database (Denmark)

    Geil, Hans Olav

    2008-01-01

    Evaluation codes (also called order domain codes) are traditionally introduced as generalized one-point geometric Goppa codes. In the present paper we will give a new point of view on evaluation codes by introducing them instead as particular nice examples of affine variety codes. Our study...... includes a reformulation of the usual methods to estimate the minimum distances of evaluation codes into the setting of affine variety codes. Finally we describe the connection to the theory of one-pointgeometric Goppa codes. Contents 4.1 Introduction...... . . . . . . . . . . . . . . . . . . . . . . . 171 4.9 Codes form order domains . . . . . . . . . . . . . . . . . . . . . . . . . . . . 173 4.10 One-point geometric Goppa codes . . . . . . . . . . . . . . . . . . . . . . . . 176 4.11 Bibliographical Notes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 178 References...

  8. Evaluation of the analysis models in the ASTRA nuclear design code system

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Nam Jin; Park, Chang Jea; Kim, Do Sam; Lee, Kyeong Taek; Kim, Jong Woon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2000-11-15

    In the field of nuclear reactor design, main practice was the application of the improved design code systems. During the process, a lot of basis and knowledge were accumulated in processing input data, nuclear fuel reload design, production and analysis of design data, et al. However less efforts were done in the analysis of the methodology and in the development or improvement of those code systems. Recently, KEPO Nuclear Fuel Company (KNFC) developed the ASTRA (Advanced Static and Transient Reactor Analyzer) code system for the purpose of nuclear reactor design and analysis. In the code system, two group constants were generated from the CASMO-3 code system. The objective of this research is to analyze the analysis models used in the ASTRA/CASMO-3 code system. This evaluation requires indepth comprehension of the models, which is important so much as the development of the code system itself. Currently, most of the code systems used in domestic Nuclear Power Plant were imported, so it is very difficult to maintain and treat the change of the situation in the system. Therefore, the evaluation of analysis models in the ASTRA nuclear reactor design code system in very important.

  9. Assessment of the SPACE Code Using the ATLAS SLB-GB-01 Test

    International Nuclear Information System (INIS)

    Kim, Yo Han; Yang, Chang Keun; Kim, Seyun

    2013-01-01

    The Korea Nuclear Hydro and Nuclear Power Co. (KHNP) has developed a safety analysis code, called the Safety and Performance Analysis Code for Nuclear Power Plants (SPACE) by collaborative works with other Korean nuclear industries. The SPACE is a general-purpose best-estimated two-phase three-field thermal-hydraulic analysis code to analyze the safety and performance of pressurized water reactors (PWRs). The SPACE code has sufficient functions and capabilities to replace outdated vendor supplied codes and to be used for the safety analysis of operating PWRs and the design of advanced reactors. As a result of the second phase of the SPACE code development project, the 2.14 version of the code was released through the successive various V and V works using integral loop test data or plant operating data. In this study, the ATLAS main steam-line break (MSLB) test, SLB-GB-01, was simulated as a V and V work. The results were compared with the measured data. The ATALS MSLB test, SLB-GB-01, was simulated using the SPACE code. The results were compared with experimental data. Through the simulation, it was concluded that the SPACE code can effectively simulate MSLB accidents

  10. Mechanistic insight into oxide-promoted palladium catalysts for the electro-oxidation of ethanol.

    Science.gov (United States)

    Martinez, Ulises; Serov, Alexey; Padilla, Monica; Atanassov, Plamen

    2014-08-01

    Recent advancements in the development of alternatives to proton exchange membrane fuel cells utilizing less-expensive catalysts and renewable liquid fuels, such as alcohols, has been observed for alkaline fuel cell systems. Alcohol fuels present the advantage of not facing the challenge of storage and transportation encountered with hydrogen fuel. Oxidation of alcohols has been improved by the promotion of alloyed or secondary phases. Nevertheless, currently, there is no experimental understanding of the difference between an intrinsic and a synergistic promotion effect in high-pH environments. This report shows evidence of different types of promotion effects on palladium electrocatalysts obtained from the presence of an oxide phase for the oxidation of ethanol. The correlation of mechanistic in situ IR spectroscopic studies with electrochemical voltammetry studies on two similar electrocatalytic systems allow the role of either an alloyed or a secondary phase on the mechanism of oxidation of ethanol to be elucidated. Evidence is presented for the difference between an intrinsic effect obtained from an alloyed system and a synergistic effect produced by the presence of an oxide phase. © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  11. Conceptual models for waste tank mechanistic analysis. Status report, January 1991

    Energy Technology Data Exchange (ETDEWEB)

    Allemann, R. T.; Antoniak, Z. I.; Eyler, L. L.; Liljegren, L. M.; Roberts, J. S.

    1992-02-01

    Pacific Northwest Laboratory (PNL) is conducting a study for Westinghouse Hanford Company (Westinghouse Hanford), a contractor for the US Department of Energy (DOE). The purpose of the work is to study possible mechanisms and fluid dynamics contributing to the periodic release of gases from double-shell waste storage tanks at the Hanford Site in Richland, Washington. This interim report emphasizing the modeling work follows two other interim reports, Mechanistic Analysis of Double-Shell Tank Gas Release Progress Report -- November 1990 and Collection and Analysis of Existing Data for Waste Tank Mechanistic Analysis Progress Report -- December 1990, that emphasized data correlation and mechanisms. The approach in this study has been to assemble and compile data that are pertinent to the mechanisms, analyze the data, evaluate physical properties and parameters, evaluate hypothetical mechanisms, and develop mathematical models of mechanisms.

  12. Visual and intelligent transients and accidents analyzer based on thermal-hydraulic system code

    International Nuclear Information System (INIS)

    Meng Lin; Rui Hu; Yun Su; Ronghua Zhang; Yanhua Yang

    2005-01-01

    Full text of publication follows: Many thermal-hydraulic system codes were developed in the past twenty years, such as RELAP5, RETRAN, ATHLET, etc. Because of their general and advanced features in thermal-hydraulic computation, they are widely used in the world to analyze transients and accidents. But there are following disadvantages for most of these original thermal-hydraulic system codes. Firstly, because models are built through input decks, so the input files are complex and non-figurative, and the style of input decks is various for different users and models. Secondly, results are shown in off-line data file form. It is not convenient for analysts who may pay more attention to dynamic parameters trend and changing. Thirdly, there are few interfaces with other program in these original thermal-hydraulic system codes. This restricts the codes expanding. The subject of this paper is to develop a powerful analyzer based on these thermal-hydraulic system codes to analyze transients and accidents more simply, accurately and fleetly. Firstly, modeling is visual and intelligent. Users build the thermalhydraulic system model using component objects according to their needs, and it is not necessary for them to face bald input decks. The style of input decks created automatically by the analyzer is unified and can be accepted easily by other people. Secondly, parameters concerned by analyst can be dynamically communicated to show or even change. Thirdly, the analyzer provide interface with other programs for the thermal-hydraulic system code. Thus parallel computation between thermal-hydraulic system code and other programs become possible. In conclusion, through visual and intelligent method, the analyzer based on general and advanced thermal-hydraulic system codes can be used to analysis transients and accidents more effectively. The main purpose of this paper is to present developmental activities, assessment and application results of the visual and intelligent

  13. Development of the containment transient analysis code for the passive reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Young Dong; Kim, Young In; Bae, Yoon Young; Chang, Moon Hi [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-05-01

    This study was performed to develop the analysis tools for the passively cooled steel containment and to construct the integrated code system which can analyze a thermal hydraulic behavior of the containment and reactor system during a loss of coolant accident. The computer code CONTEMPT4/MOD5/PCCS was developed by incorporating the passive containment cooling models to the containment pressure and temperature transient analysis computer code CONTEMPT4/MOD5. The integrated reactor thermal hydraulic analysis code system for passive reactor was constructed by coupling the best estimate thermal hydraulic system analysis code RELAP5/MOD3 and CONTEMPT4/MOD5/PCCS through the process control method. In addition, to evaluate the applicability of the code the CONTEMPT4/MOD5/PCCS was applied to the SMART(System-Integrated Modular Advanced Reactor). The pressure and temperature transient following the small break LOCA of SMART was analysed by modeling the safeguard vessel using both the newly added passive containment cooling model and existing pool model. (author). 16 refs., 22 figs., 7 tabs.

  14. Validation of Advanced Computer Codes for VVER Technology: LB-LOCA Transient in PSB-VVER Facility

    Directory of Open Access Journals (Sweden)

    A. Del Nevo

    2012-01-01

    Full Text Available The OECD/NEA PSB-VVER project provided unique and useful experimental data for code validation from PSB-VVER test facility. This facility represents the scaled-down layout of the Russian-designed pressurized water reactor, namely, VVER-1000. Five experiments were executed, dealing with loss of coolant scenarios (small, intermediate, and large break loss of coolant accidents, a primary-to-secondary leak, and a parametric study (natural circulation test aimed at characterizing the VVER system at reduced mass inventory conditions. The comparative analysis, presented in the paper, regards the large break loss of coolant accident experiment. Four participants from three different institutions were involved in the benchmark and applied their own models and set up for four different thermal-hydraulic system codes. The benchmark demonstrated the performances of such codes in predicting phenomena relevant for safety on the basis of fixed criteria.

  15. Advance Noise Control Fan II: Test Rig Fan Risk Management Study

    Science.gov (United States)

    Lucero, John

    2013-01-01

    Since 1995 the Advanced Noise Control Fan (ANCF) has significantly contributed to the advancement of the understanding of the physics of fan tonal noise generation. The 9'x15' WT has successfully tested multiple high speed fan designs over the last several decades. This advanced several tone noise reduction concepts to higher TRL and the validation of fan tone noise prediction codes.

  16. Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) verification and validation plan. version 1.

    Energy Technology Data Exchange (ETDEWEB)

    Bartlett, Roscoe Ainsworth; Arguello, Jose Guadalupe, Jr.; Urbina, Angel; Bouchard, Julie F.; Edwards, Harold Carter; Freeze, Geoffrey A.; Knupp, Patrick Michael; Wang, Yifeng; Schultz, Peter Andrew; Howard, Robert (Oak Ridge National Laboratory, Oak Ridge, TN); McCornack, Marjorie Turner

    2011-01-01

    The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. To meet this objective, NEAMS Waste IPSC M&S capabilities will be applied to challenging spatial domains, temporal domains, multiphysics couplings, and multiscale couplings. A strategic verification and validation (V&V) goal is to establish evidence-based metrics for the level of confidence in M&S codes and capabilities. Because it is economically impractical to apply the maximum V&V rigor to each and every M&S capability, M&S capabilities will be ranked for their impact on the performance assessments of various components of the repository systems. Those M&S capabilities with greater impact will require a greater level of confidence and a correspondingly greater investment in V&V. This report includes five major components: (1) a background summary of the NEAMS Waste IPSC to emphasize M&S challenges; (2) the conceptual foundation for verification, validation, and confidence assessment of NEAMS Waste IPSC M&S capabilities; (3) specifications for the planned verification, validation, and confidence-assessment practices; (4) specifications for the planned evidence information management system; and (5) a path forward for the incremental implementation of this V&V plan.

  17. Current Status of the LIFE Fast Reactors Fuel Performance Codes

    International Nuclear Information System (INIS)

    Yacout, A.M.; Billone, M.C.

    2013-01-01

    The LIFE-4 (Rev. 1) code was calibrated and validated using data from (U,Pu)O2 mixed-oxide fuel pins and UO2 blanket rods which were irradiation tested under steady-state and transient conditions. – It integrates a broad material and fuel-pin irradiation database into a consistent framework for use and extrapolation of the database to reactor design applications. – The code is available and running on different computer platforms (UNIX & PC) – Detailed documentations of the code’s models, routines, calibration and validation data sets are available. LIFE-METAL code is based on LIFE4 with modifications to include key phenomena applicable to metallic fuel, and metallic fuel properties – Calibrated with large database from irradiations in EBR-II – Further effort for calibration and detailed documentation. Recent activities with the codes are related to reactor design studies and support of licensing efforts for 4S and KAERI SFR designs. Future activities are related to re-assessment of the codes calibration and validation and inclusion of models for advanced fuels (transmutation fuels)

  18. An Optimal Linear Coding for Index Coding Problem

    OpenAIRE

    Pezeshkpour, Pouya

    2015-01-01

    An optimal linear coding solution for index coding problem is established. Instead of network coding approach by focus on graph theoric and algebraic methods a linear coding program for solving both unicast and groupcast index coding problem is presented. The coding is proved to be the optimal solution from the linear perspective and can be easily utilize for any number of messages. The importance of this work is lying mostly on the usage of the presented coding in the groupcast index coding ...

  19. Validation and uncertainty analysis of the Athlet thermal-hydraulic computer code

    International Nuclear Information System (INIS)

    Glaeser, H.

    1995-01-01

    The computer code ATHLET is being developed by GRS as an advanced best-estimate code for the simulation of breaks and transients in Pressurized Water Reactor (PWRs) and Boiling Water Reactor (BWRs) including beyond design basis accidents. A systematic validation of ATHLET is based on a well balanced set of integral and separate effects tests emphasizing the German combined Emergency Core Cooling (ECC) injection system. When using best estimate codes for predictions of reactor plant states during assumed accidents, qualification of the uncertainty in these calculations is highly desirable. A method for uncertainty and sensitivity evaluation has been developed by GRS where the computational effort is independent of the number of uncertain parameters. (author)

  20. ATHENA [Advanced Thermal Hydraulic Energy Network Analyzer] solutions to developmental assessment problems

    International Nuclear Information System (INIS)

    Carlson, K.E.; Ransom, V.H.; Roth, P.A.

    1987-03-01

    The ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) code has been developed to perform transient simulation of the thermal hydraulic systems that may be found in fusion reactors, space reactors, and other advanced systems. As an assessment of current capability the code was applied to a number of physical problems, both conceptual and actual experiments. Results indicate that the numerical solution to the basic conservation equations is technically sound, and that generally good agreement can be obtained when modeling relevant hydrodynamic experiments. The assessment also demonstrates basic fusion system modeling capability and verifies compatibility of the code with both CDC and CRAY mainframes. Areas where improvements could be made include constitutive modeling, which describes the interfacial exchange term. 13 refs., 84 figs

  1. Mechanistic kinetic models of enzymatic cellulose hydrolysis-A review.

    Science.gov (United States)

    Jeoh, Tina; Cardona, Maria J; Karuna, Nardrapee; Mudinoor, Akshata R; Nill, Jennifer

    2017-07-01

    Bioconversion of lignocellulose forms the basis for renewable, advanced biofuels, and bioproducts. Mechanisms of hydrolysis of cellulose by cellulases have been actively studied for nearly 70 years with significant gains in understanding of the cellulolytic enzymes. Yet, a full mechanistic understanding of the hydrolysis reaction has been elusive. We present a review to highlight new insights gained since the most recent comprehensive review of cellulose hydrolysis kinetic models by Bansal et al. (2009) Biotechnol Adv 27:833-848. Recent models have taken a two-pronged approach to tackle the challenge of modeling the complex heterogeneous reaction-an enzyme-centric modeling approach centered on the molecularity of the cellulase-cellulose interactions to examine rate limiting elementary steps and a substrate-centric modeling approach aimed at capturing the limiting property of the insoluble cellulose substrate. Collectively, modeling results suggest that at the molecular-scale, how rapidly cellulases can bind productively (complexation) and release from cellulose (decomplexation) is limiting, while the overall hydrolysis rate is largely insensitive to the catalytic rate constant. The surface area of the insoluble substrate and the degrees of polymerization of the cellulose molecules in the reaction both limit initial hydrolysis rates only. Neither enzyme-centric models nor substrate-centric models can consistently capture hydrolysis time course at extended reaction times. Thus, questions of the true reaction limiting factors at extended reaction times and the role of complexation and decomplexation in rate limitation remain unresolved. Biotechnol. Bioeng. 2017;114: 1369-1385. © 2017 Wiley Periodicals, Inc. © 2017 Wiley Periodicals, Inc.

  2. Ruthenium-Catalyzed Transformations of Alcohols: Mechanistic Investigations and Methodology Development

    DEFF Research Database (Denmark)

    Makarov, Ilya; Madsen, Robert; Fristrup, Peter

    with dimethoxyisopropylidene and pyridilidene ligands could be more active than RuCl2(IiPr)(p-cymene) used in the mechanistic investigation. Two analogs of the calculated complexes were synthesized but were not isolated in a pure form. The amidation reaction catalyzed by a mixture containing the N-ethyl pyridilidene...

  3. Mechanistic investigation on the oxidation of kinetin by Ag(III)

    Indian Academy of Sciences (India)

    Home; Journals; Journal of Chemical Sciences; Volume 122; Issue 6. Mechanistic investigation on the oxidation of kinetin by Ag(III) periodate complex in aqueous alkaline media: A kinetic approach. S D Lamani A M Tatagar S T Nandibewoor. Full Papers Volume 122 Issue 6 November 2010 pp 891-900 ...

  4. Finite mixture models for sensitivity analysis of thermal hydraulic codes for passive safety systems analysis

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, Francesco, E-mail: francesco.dimaio@polimi.it [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Nicola, Giancarlo [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Zio, Enrico [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Chair on System Science and Energetic Challenge Fondation EDF, Ecole Centrale Paris and Supelec, Paris (France); Yu, Yu [School of Nuclear Science and Engineering, North China Electric Power University, 102206 Beijing (China)

    2015-08-15

    Highlights: • Uncertainties of TH codes affect the system failure probability quantification. • We present Finite Mixture Models (FMMs) for sensitivity analysis of TH codes. • FMMs approximate the pdf of the output of a TH code with a limited number of simulations. • The approach is tested on a Passive Containment Cooling System of an AP1000 reactor. • The novel approach overcomes the results of a standard variance decomposition method. - Abstract: For safety analysis of Nuclear Power Plants (NPPs), Best Estimate (BE) Thermal Hydraulic (TH) codes are used to predict system response in normal and accidental conditions. The assessment of the uncertainties of TH codes is a critical issue for system failure probability quantification. In this paper, we consider passive safety systems of advanced NPPs and present a novel approach of Sensitivity Analysis (SA). The approach is based on Finite Mixture Models (FMMs) to approximate the probability density function (i.e., the uncertainty) of the output of the passive safety system TH code with a limited number of simulations. We propose a novel Sensitivity Analysis (SA) method for keeping the computational cost low: an Expectation Maximization (EM) algorithm is used to calculate the saliency of the TH code input variables for identifying those that most affect the system functional failure. The novel approach is compared with a standard variance decomposition method on a case study considering a Passive Containment Cooling System (PCCS) of an Advanced Pressurized reactor AP1000.

  5. RELAP5 kinetics model development for the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Judd, J.L.; Terry, W.K.

    1990-01-01

    A point-kinetics model of the Advanced Test Reactor has been developed for the RELAP5 code. Reactivity feedback parameters were calculated by a three-dimensional analysis with the PDQ neutron diffusion code. Analyses of several hypothetical reactivity insertion events by the new model and two earlier models are discussed. 3 refs., 10 figs., 6 tabs

  6. Advanced subgrid modeling for Multiphase CFD in CASL VERA tools

    International Nuclear Information System (INIS)

    Baglietto, Emilio; Gilman, Lindsey; Sugrue, Rosie

    2014-01-01

    This work introduces advanced modeling capabilities that are being developed to improve the accuracy and extend the applicability of Multiphase CFD. Specifics of the advanced and hardened boiling closure model are described in this work. The development has been driven by new physical understanding, derived from the innovative experimental techniques available at MIT. A new experimental-based mechanistic approach to heat partitioning is proposed. The model introduces a new description of the bubble evaporation, sliding and interaction on the heated surface to accurately capture the evaporation occurring at the heated surface, while also tracking the local surface conditions. The model is being assembled to cover an extended application area, up to Critical Heat Flux (CHF). The accurate description of the bubble interaction, effective microlayer and dry surface area are considered to be the enabling quantities towards innovated CHF capturing methodologies. Further, improved mechanistic force-balance models for bubble departure predictions and lift-off diameter predictions are implemented in the model. Studies demonstrate the influence of the newly implemented partitioning components. Finally, the development work towards a more consistent and integrated hydrodynamic closure is presented. The main objective here is to develop a set of robust momentum closure relations which focuses on the specific application to PWR conditions, but will facilitate the application to other geometries, void fractions, and flow regimes. The innovative approach considers local flow conditions on a cell-by-cell basis to ensure robustness. Closure relations of interest initially include drag, lift, and turbulence dispersion, with near wall corrections applied for both drag and lift. (author)

  7. Recent activities on nuclear codes and standards

    International Nuclear Information System (INIS)

    Minematsu, Akiyoshi; Ishimoto, Shozaburo; Honjin, Masao

    2000-01-01

    The technical codes and standards relating to the nuclear power stations in Japan are prepared by shapes of laws (ministerial ordinances and bulletins) issued by the government and obliged to comply with by 'the Law concerning the Regulations of Nuclear Material Substances, Nuclear Fuel Substances and Nuclear Reactors' and 'the Electricity Business Act' and of guides defined by the Nuclear Safety Commission, and further some private standards have been issued at a shape of complement of these laws and guides by receiving national recommendation. On the other hand, in the fields of electricity and heat facilities except atomic energy, simplification and feature stipulation of the national technical codes and standards was recently carried out, by which a system usable for the private standards in and out of Japan were prepared through approval of the private Japan Electrotechnical Standards and Codes Committee (JESC). As the nuclear field was now excepted from simultaneous transfer to the private standard and the standard application system, it is expected in future to realize similar transfer if possible and preparation of the private standards is now being advanced. Here were introduced on present state on technical codes and standards relating to the nuclear power generation facilities and recent trends on their private standardization. (G.K.)

  8. Validation of an advanced analytical procedure applied to the measurement of environmental radioactivity.

    Science.gov (United States)

    Thanh, Tran Thien; Vuong, Le Quang; Ho, Phan Long; Chuong, Huynh Dinh; Nguyen, Vo Hoang; Tao, Chau Van

    2018-04-01

    In this work, an advanced analytical procedure was applied to calculate radioactivity in spiked water samples in a close geometry gamma spectroscopy. It included MCNP-CP code in order to calculate the coincidence summing correction factor (CSF). The CSF results were validated by a deterministic method using ETNA code for both p-type HPGe detectors. It showed that a good agreement for both codes. Finally, the validity of the developed procedure was confirmed by a proficiency test to calculate the activities of various radionuclides. The results of the radioactivity measurement with both detectors using the advanced analytical procedure were received the ''Accepted'' statuses following the proficiency test. Copyright © 2018 Elsevier Ltd. All rights reserved.

  9. Probable mode prediction for H.264 advanced video coding P slices using removable SKIP mode distortion estimation

    Science.gov (United States)

    You, Jongmin; Jeong, Jechang

    2010-02-01

    The H.264/AVC (advanced video coding) is used in a wide variety of applications including digital broadcasting and mobile applications, because of its high compression efficiency. The variable block mode scheme in H.264/AVC contributes much to its high compression efficiency but causes a selection problem. In general, rate-distortion optimization (RDO) is the optimal mode selection strategy, but it is computationally intensive. For this reason, the H.264/AVC encoder requires a fast mode selection algorithm for use in applications that require low-power and real-time processing. A probable mode prediction algorithm for the H.264/AVC encoder is proposed. To reduce the computational complexity of RDO, the proposed method selects probable modes among all allowed block modes using removable SKIP mode distortion estimation. Removable SKIP mode distortion is used to estimate whether or not a further divided block mode is appropriate for a macroblock. It is calculated using a no-motion reference block with a few computations. Then the proposed method reduces complexity by performing the RDO process only for probable modes. Experimental results show that the proposed algorithm can reduce encoding time by an average of 55.22% without significant visual quality degradation and increased bit rate.

  10. Adaption, validation and application of advanced codes with 3-dimensional neutron kinetics for accident analysis calculations - STC with Bulgaria

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.; Mittag, S.; Rohde, U.; Seidel, A.; Panayotov, D.; Ilieva, B.

    2001-08-01

    In the frame of a project on scientific-technical co-operation funded by BMBF/BMWi, the program code DYN3D and the coupled code ATHLET-DYN3D have been transferred to the Institute for Nuclear Research and Nuclear Energy (INRNE) Sofia. The coupled code represents an implementation of the 3D core model DYN3D developed by FZR into the GRS thermal-hydraulics code system ATHLET. For the purpose of validation of these codes, a measurement data base about a start-up experiment obtained at the unit 6 of Kozloduy NPP (VVER-1000/V-320) has been generated. The results of performed validation calculations were compared with measurement values from the data base. A simplified model for estimation of cross flow mixing between fuel assemblies has been implemented into the program code DYN3D by Bulgarian experts. Using this cross flow model, transient processes with asymmetrical boundary conditions can be analysed more realistic. The validation of the implemented model were performed with help of comparison calculations between modified DYD3D code and thermal-hydraulics code COBRA-4I, and also on the base of the collected measurement data from Kozloduy NPP. (orig.) [de

  11. Optimizing zonal advection of the Advanced Research WRF (ARW) dynamics for Intel MIC

    Science.gov (United States)

    Mielikainen, Jarno; Huang, Bormin; Huang, Allen H.

    2014-10-01

    The Weather Research and Forecast (WRF) model is the most widely used community weather forecast and research model in the world. There are two distinct varieties of WRF. The Advanced Research WRF (ARW) is an experimental, advanced research version featuring very high resolution. The WRF Nonhydrostatic Mesoscale Model (WRF-NMM) has been designed for forecasting operations. WRF consists of dynamics code and several physics modules. The WRF-ARW core is based on an Eulerian solver for the fully compressible nonhydrostatic equations. In the paper, we will use Intel Intel Many Integrated Core (MIC) architecture to substantially increase the performance of a zonal advection subroutine for optimization. It is of the most time consuming routines in the ARW dynamics core. Advection advances the explicit perturbation horizontal momentum equations by adding in the large-timestep tendency along with the small timestep pressure gradient tendency. We will describe the challenges we met during the development of a high-speed dynamics code subroutine for MIC architecture. Furthermore, lessons learned from the code optimization process will be discussed. The results show that the optimizations improved performance of the original code on Xeon Phi 5110P by a factor of 2.4x.

  12. ASPECT: An advanced specified-profile evaluation code for tokamaks

    International Nuclear Information System (INIS)

    Stotler, D.P.; Reiersen, W.T.; Bateman, G.

    1993-03-01

    A specified-profile, global analysis code has been developed to evaluate the performance of fusion reactor designs. Both steady-state and time-dependent calculations are carried out; the results of the former can be used in defining the parameters of the latter, if desired. In the steady-state analysis, the performance is computed at a density and temperature chosen to be consistent with input limits (e.g., density and beta) of several varieties. The calculation can be made at either the intersection of the two limits or at the point of optimum performance as the density and temperature are varied along the limiting boundaries. Two measures of performance are available for this purpose: the ignition margin or the confinement level required to achieve a prescribed ignition margin. The time-dependent calculation can be configured to yield either the evolution of plasma energy as a function of time or, via an iteration scheme, the amount of auxiliary power required to achieve a desired final plasma energy

  13. Comparative analysis of nine structural codes used in the second WIPP benchmark problem

    International Nuclear Information System (INIS)

    Morgan, H.S.; Krieg, R.D.; Matalucci, R.V.

    1981-11-01

    In the Waste Isolation Pilot Plant (WIPP) Benchmark II study, various computer codes were compared on the basis of their capabilities for calculating the response of hypothetical drift configurations for nuclear waste experiments and storage demonstration. The codes used by participants in the study were ANSALT, DAPROK, JAC, REM, SANCHO, SPECTROM, STEALTH, and two different implementations of MARC. Errors were found in the preliminary results, and several calculations were revised. Revised solutions were in reasonable agreement except for the REM solution. The Benchmark II study allowed significant advances in understanding the relative behavior of computer codes available for WIPP calculations. The study also pointed out the possible need for performing critical design calculations with more than one code. Lastly, it indicated the magnitude of the code-to-code spread in results which is to be expected even when a model has been explicitly defined

  14. Modular ORIGEN-S for multi-physics code systems

    Energy Technology Data Exchange (ETDEWEB)

    Yesilyurt, Gokhan; Clarno, Kevin T.; Gauld, Ian C., E-mail: yesilyurtg@ornl.gov, E-mail: clarnokt@ornl.gov, E-mail: gauldi@ornl.gov [Oak Ridge National Laboratory, TN (United States); Galloway, Jack, E-mail: jack@galloways.net [Los Alamos National Laboratory, Los Alamos, NM (United States)

    2011-07-01

    the AMP Nuclear Fuel Performance code and the NESTLE advanced nodal code. (author)

  15. Modular ORIGEN-S for multi-physics code systems

    International Nuclear Information System (INIS)

    Yesilyurt, Gokhan; Clarno, Kevin T.; Gauld, Ian C.; Galloway, Jack

    2011-01-01

    the AMP Nuclear Fuel Performance code and the NESTLE advanced nodal code. (author)

  16. Mode-dependent templates and scan order for H.264/AVC-based intra lossless coding.

    Science.gov (United States)

    Gu, Zhouye; Lin, Weisi; Lee, Bu-Sung; Lau, Chiew Tong; Sun, Ming-Ting

    2012-09-01

    In H.264/advanced video coding (AVC), lossless coding and lossy coding share the same entropy coding module. However, the entropy coders in the H.264/AVC standard were original designed for lossy video coding and do not yield adequate performance for lossless video coding. In this paper, we analyze the problem with the current lossless coding scheme and propose a mode-dependent template (MD-template) based method for intra lossless coding. By exploring the statistical redundancy of the prediction residual in the H.264/AVC intra prediction modes, more zero coefficients are generated. By designing a new scan order for each MD-template, the scanned coefficients sequence fits the H.264/AVC entropy coders better. A fast implementation algorithm is also designed. With little computation increase, experimental results confirm that the proposed fast algorithm achieves about 7.2% bit saving compared with the current H.264/AVC fidelity range extensions high profile.

  17. Refuelling design and core calculations at NPP Paks: codes and methods

    International Nuclear Information System (INIS)

    Pos, I.; Nemes, I.; Javor, E.; Korpas, L.; Szecsenyi, Z.; Patai-Szabo, S.

    2001-01-01

    This article gives a brief review of the computer codes used in the fuel management practice at NPP Paks. The code package consist of the HELIOS neutron and gamma transport code for preparation of few-group cross section library, the CERBER code to determine the optimal core loading patterns and the C-PORCA code for detailed reactor physical analysis of different reactor states. The last two programs have been developed at the NPP Paks. HELIOS gives sturdy basis for our neutron physical calculation, CERBER and C-PORCA programs have been enhanced in great extent for last years. Methods and models have become more detailed and accurate as regards the calculated parameters and space resolution. Introduction of a more advanced data handling algorithm arbitrary move of fuel assemblies can be followed either in the reactor core or storage pool. The new interactive WINDOWS applications allow easier and more reliable use of codes. All these computer code developments made possible to handle and calculate new kind of fuels as profiled Russian and BNFL fuel with burnable poison or to support the reliable reuse of fuel assemblies stored in the storage pool. To extend thermo-hydraulic capability, with KFKI contribution the COBRA code will also be coupled to the system (Authors)

  18. RADTRAN 5: A computer code for transportation risk analysis

    International Nuclear Information System (INIS)

    Neuhauser, K.S.; Kanipe, F.L.

    1991-01-01

    RADTRAN 5 is a computer code developed at Sandia National Laboratories (SNL) in Albuquerque, NM, to estimate radiological and nonradiological risks of radioactive materials transportation. RADTRAN 5 is written in ANSI Standard FORTRAN 77 and contains significant advances in the methodology for route-specific analysis first developed by SNL for RADTRAN 4 (Neuhauser and Kanipe, 1992). Like the previous RADTRAN codes, RADTRAN 5 contains two major modules for incident-free and accident risk amlysis, respectively. All commercially important transportation modes may be analyzed with RADTRAN 5: highway by combination truck; highway by light-duty vehicle; rail; barge; ocean-going ship; cargo air; and passenger air

  19. Turbine Internal and Film Cooling Modeling For 3D Navier-Stokes Codes

    Science.gov (United States)

    DeWitt, Kenneth; Garg Vijay; Ameri, Ali

    2005-01-01

    The aim of this research project is to make use of NASA Glenn on-site computational facilities in order to develop, validate and apply aerodynamic, heat transfer, and turbine cooling models for use in advanced 3D Navier-Stokes Computational Fluid Dynamics (CFD) codes such as the Glenn-" code. Specific areas of effort include: Application of the Glenn-HT code to specific configurations made available under Turbine Based Combined Cycle (TBCC), and Ultra Efficient Engine Technology (UEET) projects. Validating the use of a multi-block code for the time accurate computation of the detailed flow and heat transfer of cooled turbine airfoils. The goal of the current research is to improve the predictive ability of the Glenn-HT code. This will enable one to design more efficient turbine components for both aviation and power generation. The models will be tested against specific configurations provided by NASA Glenn.

  20. Advanced compiler design and implementation

    CERN Document Server

    Muchnick, Steven S

    1997-01-01

    From the Foreword by Susan L. Graham: This book takes on the challenges of contemporary languages and architectures, and prepares the reader for the new compiling problems that will inevitably arise in the future. The definitive book on advanced compiler design This comprehensive, up-to-date work examines advanced issues in the design and implementation of compilers for modern processors. Written for professionals and graduate students, the book guides readers in designing and implementing efficient structures for highly optimizing compilers for real-world languages. Covering advanced issues in fundamental areas of compiler design, this book discusses a wide array of possible code optimizations, determining the relative importance of optimizations, and selecting the most effective methods of implementation. * Lays the foundation for understanding the major issues of advanced compiler design * Treats optimization in-depth * Uses four case studies of commercial compiling suites to illustrate different approache...