Collaborative Multi-Layer Network Coding in Hybrid Cellular Cognitive Radio Networks

KAUST Repository

Moubayed, Abdallah J.; Sorour, Sameh; Al-Naffouri, Tareq Y.; Alouini, Mohamed-Slim

2015-01-01

In this paper, as an extension to [1], we propose a prioritized multi-layer network coding scheme for collaborative packet recovery in hybrid (interweave and underlay) cellular cognitive radio networks. This scheme allows the uncoordinated

Collaborative Multi-Layer Network Coding For Hybrid Cellular Cognitive Radio Networks

KAUST Repository

Moubayed, Abdallah J.

2014-01-01

In this thesis, as an extension to [1], we propose a prioritized multi-layer network coding scheme for collaborative packet recovery in hybrid (interweave and underlay) cellular cognitive radio networks. This scheme allows the uncoordinated

Advanced burnup calculation code system in a subcritical state with continuous-energy Monte Carlo code for fusion-fission hybrid reactor

International Nuclear Information System (INIS)

Matsunaka, Masayuki; Ohta, Masayuki; Miyamaru, Hiroyuki; Murata, Isao

2009-01-01

The fusion-fission (FF) hybrid reactor is a promising energy source that is thought to act as a bridge between the existing fission reactor and the genuine fusion reactor in the future. The burnup calculation system that aims at precise burnup calculations of a subcritical system was developed for the detailed design of the FF hybrid reactor, and the system consists of MCNP, ORIGEN, and postprocess codes. In the present study, the calculation system was substantially modified to improve the calculation accuracy and at the same time the calculation speed as well. The reaction rate estimation can be carried out accurately with the present system that uses track-length (TL) data in the continuous-energy treatment. As for the speed-up of the reaction rate calculation, a new TL data bunching scheme was developed so that only necessary TL data are used as long as the accuracy of the point-wise nuclear data is conserved. With the present system, an example analysis result for our proposed FF hybrid reactor is described, showing that the computation time could really be saved with the same accuracy as before. (author)

Low complexity source and channel coding for mm-wave hybrid fiber-wireless links

DEFF Research Database (Denmark)

Lebedev, Alexander; Vegas Olmos, Juan José; Pang, Xiaodan

2014-01-01

We report on the performance of channel and source coding applied for an experimentally realized hybrid fiber-wireless W-band link. Error control coding performance is presented for a wireless propagation distance of 3 m and 20 km fiber transmission. We report on peak signal-to-noise ratio perfor...

Noise suppression system of OCDMA with spectral/spatial 2D hybrid code

Science.gov (United States)

Matem, Rima; Aljunid, S. A.; Junita, M. N.; Rashidi, C. B. M.; Shihab Aqrab, Israa

2017-11-01

In this paper, we propose a novel 2D spectral/spatial hybrid code based on 1D ZCC and 1D MD where the both present a zero cross correlation property analyzed and the influence of the noise of optical as Phase Induced Intensity Noise (PIIN), shot and thermal noise. This new code is shown effectively to mitigate the PIIN and suppresses MAI. Using 2D ZCC/MD code the performance of the system can be improved in term of as well as to support more simultaneous users compared of the 2D FCC/MDW and 2D DPDC codes.

Noise suppression system of OCDMA with spectral/spatial 2D hybrid code

Directory of Open Access Journals (Sweden)

Matem Rima

2017-01-01

Full Text Available In this paper, we propose a novel 2D spectral/spatial hybrid code based on 1D ZCC and 1D MD where the both present a zero cross correlation property analyzed and the influence of the noise of optical as Phase Induced Intensity Noise (PIIN, shot and thermal noise. This new code is shown effectively to mitigate the PIIN and suppresses MAI. Using 2D ZCC/MD code the performance of the system can be improved in term of as well as to support more simultaneous users compared of the 2D FCC/MDW and 2D DPDC codes.

Studies of Planet Formation using a Hybrid N-body + Planetesimal Code

Science.gov (United States)

Kenyon, Scott J.; Bromley, Benjamin C.; Salamon, Michael (Technical Monitor)

2005-01-01

The goal of our proposal was to use a hybrid multi-annulus planetesimal/n-body code to examine the planetesimal theory, one of the two main theories of planet formation. We developed this code to follow the evolution of numerous 1 m to 1 km planetesimals as they collide, merge, and grow into full-fledged planets. Our goal was to apply the code to several well-posed, topical problems in planet formation and to derive observational consequences of the models. We planned to construct detailed models to address two fundamental issues: 1) icy planets - models for icy planet formation will demonstrate how the physical properties of debris disks, including the Kuiper Belt in our solar system, depend on initial conditions and input physics; and 2) terrestrial planets - calculations following the evolution of 1-10 km planetesimals into Earth-mass planets and rings of dust will provide a better understanding of how terrestrial planets form and interact with their environment. During the past year, we made progress on each issue. Papers published in 2004 are summarized. Summaries of work to be completed during the first half of 2005 and work planned for the second half of 2005 are included.

Benchmark of coupling codes (ALOHA, TOPLHA and GRILL3D) with ITER-relevant Lower Hybrid antenna

International Nuclear Information System (INIS)

Milanesio, D.; Hillairet, J.; Panaccione, L.; Maggiora, R.; Artaud, J.F.; Bae, Y.S.; Barbera, A.M.A.; Belo, J.; Berger-By, G.; Bernard, J.M.; Cara, Ph.; Cardinali, A.; Castaldo, C.; Ceccuzzi, S.; Cesario, R.; Decker, J.; Delpech, L.; Ekedahl, A.; Garcia, J.; Garibaldi, P.

2011-01-01

In order to assist the design of the future ITER Lower Hybrid launcher, coupling codes ALOHA, from CEA/IRFM, TOPLHA, from Politecnico di Torino, and GRILL3D, developed by Dr. Mikhail Irzak (A.F. Ioffe Physico-Technical Institute, St. Petersburg, Russia) and operated by ENEA Frascati, have been compared with the updated (six modules with four active waveguides per module) Passive-Active Multi-junction (PAM) Lower Hybrid antennas. Both ALOHA and GRILL3D formulate the problem in terms of rectangular waveguides modes, while TOPLHA is based on boundary-value problem with the adoption of a triangular cell-mesh to represent the relevant waveguides surfaces. Several plasma profiles, with varying edge density and density increase, have been adopted to provide a complete description of the simulated launcher in terms of reflection coefficient, computed at the beginning of each LH module, and of power spectra. Good agreement can be observed among codes for all the simulated profiles.

Hybrid MPI-OpenMP Parallelism in the ONETEP Linear-Scaling Electronic Structure Code: Application to the Delamination of Cellulose Nanofibrils.

Science.gov (United States)

Wilkinson, Karl A; Hine, Nicholas D M; Skylaris, Chris-Kriton

2014-11-11

We present a hybrid MPI-OpenMP implementation of Linear-Scaling Density Functional Theory within the ONETEP code. We illustrate its performance on a range of high performance computing (HPC) platforms comprising shared-memory nodes with fast interconnect. Our work has focused on applying OpenMP parallelism to the routines which dominate the computational load, attempting where possible to parallelize different loops from those already parallelized within MPI. This includes 3D FFT box operations, sparse matrix algebra operations, calculation of integrals, and Ewald summation. While the underlying numerical methods are unchanged, these developments represent significant changes to the algorithms used within ONETEP to distribute the workload across CPU cores. The new hybrid code exhibits much-improved strong scaling relative to the MPI-only code and permits calculations with a much higher ratio of cores to atoms. These developments result in a significantly shorter time to solution than was possible using MPI alone and facilitate the application of the ONETEP code to systems larger than previously feasible. We illustrate this with benchmark calculations from an amyloid fibril trimer containing 41,907 atoms. We use the code to study the mechanism of delamination of cellulose nanofibrils when undergoing sonification, a process which is controlled by a large number of interactions that collectively determine the structural properties of the fibrils. Many energy evaluations were needed for these simulations, and as these systems comprise up to 21,276 atoms this would not have been feasible without the developments described here.

Development of 3-D Flow Analysis Code for Fuel Assembly using Unstructured Grid System

Energy Technology Data Exchange (ETDEWEB)

Myong, Hyon Kook; Kim, Jong Eun; Ahn, Jong Ki; Yang, Seung Yong [Kookmin Univ., Seoul (Korea, Republic of)

2007-03-15

The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Final objective of this study is to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system. In order to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system, the following researches are made: - Development of numerical algorithm for CFD code's solver - Grid and geometric connectivity data - Development of software(PowerCFD code) for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system - Modulation of software(PowerCFD code) - Development of turbulence model - Development of analysis module of RANS/LES hybrid models - Analysis of turbulent flow and heat transfer - Basic study on LES analysis - Development of main frame on pre/post processors based on GUI - Algorithm for fully-developed flow.

Development of 3-D Flow Analysis Code for Fuel Assembly using Unstructured Grid System

International Nuclear Information System (INIS)

Myong, Hyon Kook; Kim, Jong Eun; Ahn, Jong Ki; Yang, Seung Yong

2007-03-01

The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Final objective of this study is to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system. In order to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system, the following researches are made: - Development of numerical algorithm for CFD code's solver - Grid and geometric connectivity data - Development of software(PowerCFD code) for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system - Modulation of software(PowerCFD code) - Development of turbulence model - Development of analysis module of RANS/LES hybrid models - Analysis of turbulent flow and heat transfer - Basic study on LES analysis - Development of main frame on pre/post processors based on GUI - Algorithm for fully-developed flow

Parallel Computing Characteristics of CUPID code under MPI and Hybrid environment

Energy Technology Data Exchange (ETDEWEB)

Lee, Jae Ryong; Yoon, Han Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Jeon, Byoung Jin; Choi, Hyoung Gwon [Seoul National Univ. of Science and Technology, Seoul (Korea, Republic of)

2014-05-15

In this paper, a characteristic of parallel algorithm is presented for solving an elliptic type equation of CUPID via domain decomposition method using the MPI and the parallel performance is estimated in terms of a scalability which shows the speedup ratio. In addition, the time-consuming pattern of major subroutines is studied. Two different grid systems are taken into account: 40,000 meshes for coarse system and 320,000 meshes for fine system. Since the matrix of the CUPID code differs according to whether the flow is single-phase or two-phase, the effect of matrix shape is evaluated. Finally, the effect of the preconditioner for matrix solver is also investigated. Finally, the hybrid (OpenMP+MPI) parallel algorithm is introduced and discussed in detail for solving pressure solver. Component-scale thermal-hydraulics code, CUPID has been developed for two-phase flow analysis, which adopts a three-dimensional, transient, three-field model, and parallelized to fulfill a recent demand for long-transient and highly resolved multi-phase flow behavior. In this study, the parallel performance of the CUPID code was investigated in terms of scalability. The CUPID code was parallelized with domain decomposition method. The MPI library was adopted to communicate the information at the neighboring domain. For managing the sparse matrix effectively, the CSR storage format is used. To take into account the characteristics of the pressure matrix which turns to be asymmetric for two-phase flow, both single-phase and two-phase calculations were run. In addition, the effect of the matrix size and preconditioning was also investigated. The fine mesh calculation shows better scalability than the coarse mesh because the number of coarse mesh does not need to decompose the computational domain excessively. The fine mesh can be present good scalability when dividing geometry with considering the ratio between computation and communication time. For a given mesh, single-phase flow

Software development infrastructure for the HYBRID modeling and simulation project

International Nuclear Information System (INIS)

Epiney, Aaron S.; Kinoshita, Robert A.; Kim, Jong Suk; Rabiti, Cristian; Greenwood, M. Scott

2016-01-01

One of the goals of the HYBRID modeling and simulation project is to assess the economic viability of hybrid systems in a market that contains renewable energy sources like wind. The idea is that it is possible for the nuclear plant to sell non-electric energy cushions, which absorb (at least partially) the volatility introduced by the renewable energy sources. This system is currently modeled in the Modelica programming language. To assess the economics of the system, an optimization procedure is trying to find the minimal cost of electricity production. The RAVEN code is used as a driver for the whole problem. It is assumed that at this stage, the HYBRID modeling and simulation framework can be classified as non-safety “research and development” software. The associated quality level is Quality Level 3 software. This imposes low requirements on quality control, testing and documentation. The quality level could change as the application development continues.Despite the low quality requirement level, a workflow for the HYBRID developers has been defined that include a coding standard and some documentation and testing requirements. The repository performs automated unit testing of contributed models. The automated testing is achieved via an open-source python script called BuildingsP from Lawrence Berkeley National Lab. BuildingsPy runs Modelica simulation tests using Dymola in an automated manner and generates and runs unit tests from Modelica scripts written by developers. In order to assure effective communication between the different national laboratories a biweekly videoconference has been set-up, where developers can report their progress and issues. In addition, periodic face-face meetings are organized intended to discuss high-level strategy decisions with management. A second means of communication is the developer email list. This is a list to which everybody can send emails that will be received by the collective of the developers and managers

Software development infrastructure for the HYBRID modeling and simulation project

Energy Technology Data Exchange (ETDEWEB)

Epiney, Aaron S. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinoshita, Robert A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kim, Jong Suk [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Greenwood, M. Scott [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-09-01

One of the goals of the HYBRID modeling and simulation project is to assess the economic viability of hybrid systems in a market that contains renewable energy sources like wind. The idea is that it is possible for the nuclear plant to sell non-electric energy cushions, which absorb (at least partially) the volatility introduced by the renewable energy sources. This system is currently modeled in the Modelica programming language. To assess the economics of the system, an optimization procedure is trying to find the minimal cost of electricity production. The RAVEN code is used as a driver for the whole problem. It is assumed that at this stage, the HYBRID modeling and simulation framework can be classified as non-safety “research and development” software. The associated quality level is Quality Level 3 software. This imposes low requirements on quality control, testing and documentation. The quality level could change as the application development continues.Despite the low quality requirement level, a workflow for the HYBRID developers has been defined that include a coding standard and some documentation and testing requirements. The repository performs automated unit testing of contributed models. The automated testing is achieved via an open-source python script called BuildingsP from Lawrence Berkeley National Lab. BuildingsPy runs Modelica simulation tests using Dymola in an automated manner and generates and runs unit tests from Modelica scripts written by developers. In order to assure effective communication between the different national laboratories a biweekly videoconference has been set-up, where developers can report their progress and issues. In addition, periodic face-face meetings are organized intended to discuss high-level strategy decisions with management. A second means of communication is the developer email list. This is a list to which everybody can send emails that will be received by the collective of the developers and managers

Sensitivity analysis of the titan hybrid deterministic transport code for SPECT simulation

International Nuclear Information System (INIS)

Royston, Katherine K.; Haghighat, Alireza

2011-01-01

Single photon emission computed tomography (SPECT) has been traditionally simulated using Monte Carlo methods. The TITAN code is a hybrid deterministic transport code that has recently been applied to the simulation of a SPECT myocardial perfusion study. For modeling SPECT, the TITAN code uses a discrete ordinates method in the phantom region and a combined simplified ray-tracing algorithm with a fictitious angular quadrature technique to simulate the collimator and generate projection images. In this paper, we compare the results of an experiment with a physical phantom with predictions from the MCNP5 and TITAN codes. While the results of the two codes are in good agreement, they differ from the experimental data by ∼ 21%. In order to understand these large differences, we conduct a sensitivity study by examining the effect of different parameters including heart size, collimator position, collimator simulation parameter, and number of energy groups. (author)

Hybrid PAPR reduction scheme with Huffman coding and DFT-spread technique for direct-detection optical OFDM systems

Science.gov (United States)

Peng, Miao; Chen, Ming; Zhou, Hui; Wan, Qiuzhen; Jiang, LeYong; Yang, Lin; Zheng, Zhiwei; Chen, Lin

2018-01-01

High peak-to-average power ratio (PAPR) of the transmit signal is a major drawback in optical orthogonal frequency division multiplexing (OOFDM) system. In this paper, we propose and experimentally demonstrate a novel hybrid scheme, combined the Huffman coding and Discrete Fourier Transmission-Spread (DFT-spread), in order to reduce high PAPR in a 16-QAM short-reach intensity-modulated and direct-detection OOFDM (IMDD-OOFDM) system. The experimental results demonstrated that the hybrid scheme can reduce the PAPR by about 1.5, 2, 3 and 6 dB, and achieve 1.5, 1, 2.5 and 3 dB receiver sensitivity improvement compared to clipping, DFT-spread and Huffman coding and original OFDM signals, respectively, at an error vector magnitude (EVM) of -10 dB after transmission over 20 km standard single-mode fiber (SSMF). Furthermore, the throughput gain can be of the order of 30% by using the hybrid scheme compared with the cases of without applying the Huffman coding.

Lower hybrid current drive in shaped tokamaks

International Nuclear Information System (INIS)

Kesner, J.

1993-01-01

A time dependent lower hybrid current drive tokamak simulation code has been developed. This code combines the BALDUR tokamak simulation code and the Bonoli/Englade lower hybrid current drive code and permits the study of the interaction of lower hybrid current drive with neutral beam heating in shaped cross-section plasmas. The code is time dependent and includes the beam driven and bootstrap currents in addition to the current driven by the lower hybrid system. Examples of simulations are shown for the PBX-M experiment which include the effect of cross section shaping on current drive, ballooning mode stabilization by current profile control and sawtooth stabilization. A critical question in current drive calculations is the radial transport of the energetic electrons. The authors have developed a response function technique to calculate radial transport in the presence of an electric field. The consequences of the combined influences of radial diffusion and electric field acceleration are discussed

Parametric systems analysis for ICF hybrid reactors

International Nuclear Information System (INIS)

Berwald, D.H.; Maniscalco, J.A.; Chapin, D.L.

1981-01-01

Parametric design and systems analysis for inertial confinement fusion-fission hybrids are presented. These results were generated as part of the Electric Power Research Institute (EPRI) sponsored Feasibility Assessment of Fusion-Fission Hybrids, using an Inertial Confinement Fusion (ICF) hybrid power plant design code developed in conjunction with the feasibility assessment. The SYMECON systems analysis code, developed by Westinghouse, was used to generate economic results for symbiotic electricity generation systems consisting of the hybrid and its client Light Water Reactors (LWRs). These results explore the entire fusion parameter space for uranium fast fission blanket hybrids, thorium fast fission blanket hybrids, and thorium suppressed fission blanket types are discussed, and system sensitivities to design uncertainties are explored

1-D hybrid code for FRM dynamics

International Nuclear Information System (INIS)

Stark, R.A.; Miley, G.H.

1985-01-01

A 1-D radial hybrid code has been written to study the start-up of the FRM via neutral-beam injection. This code, named FROST (Field Reversed One-dimensional STart-up), models the plasma as azimuthal symmetric with no axial dependence. A multi-group method in energy and canonical angular momentum describes the large-orbit ions from the beam. This method is designed to be more efficient than those employing particle tracking, since the characteristic timescale of the simulation is the ion slowing down time, rather than the much shorter cyclotron period. A time-differentiated Grad-Shafranov equation couples the ion current to massless fluid equations describing electrons and low energy ions. Flux coordinates are used in this fluid model, in preference to an Eulerian framework, so that coupling of plasma at the two different radii of a closed flux surface may be treated with ease. Since a fluid treatment for electrons is invalid near a field null, a separate model for the electron current has been included for this region, a unique feature. Results of simulation of injection into a 2XIIB-like plasma are discussed. Electron currents are found to retard, but not prevent reversal of the magnetic field at the plasma center

An integrated PCR colony hybridization approach to screen cDNA libraries for full-length coding sequences.

Science.gov (United States)

Pollier, Jacob; González-Guzmán, Miguel; Ardiles-Diaz, Wilson; Geelen, Danny; Goossens, Alain

2011-01-01

cDNA-Amplified Fragment Length Polymorphism (cDNA-AFLP) is a commonly used technique for genome-wide expression analysis that does not require prior sequence knowledge. Typically, quantitative expression data and sequence information are obtained for a large number of differentially expressed gene tags. However, most of the gene tags do not correspond to full-length (FL) coding sequences, which is a prerequisite for subsequent functional analysis. A medium-throughput screening strategy, based on integration of polymerase chain reaction (PCR) and colony hybridization, was developed that allows in parallel screening of a cDNA library for FL clones corresponding to incomplete cDNAs. The method was applied to screen for the FL open reading frames of a selection of 163 cDNA-AFLP tags from three different medicinal plants, leading to the identification of 109 (67%) FL clones. Furthermore, the protocol allows for the use of multiple probes in a single hybridization event, thus significantly increasing the throughput when screening for rare transcripts. The presented strategy offers an efficient method for the conversion of incomplete expressed sequence tags (ESTs), such as cDNA-AFLP tags, to FL-coding sequences.

Nuclear Characteristics of SPNDs and Preliminary Calculation of Hybrid Fixed Incore Detector with Monte Carlo Code

International Nuclear Information System (INIS)

Koo, Bon Seung; Lee, Kyung Hoon; Song, Jae Seung; Park, Sang Yoon

2013-01-01

In this paper, the basic nuclear characteristics of major emitter materials were surveyed. In addition, preliminary calculations of Cobalt-Vanadium fixed incore detector were performed using the Monte Carlo code. Calculational results were cross-checked by KARMA. KARMA is a two-dimensional multigroup transport theory code developed by the KAERI and approved by Korean regularity agency to be employed as a nuclear design tool for a Korean commercial pressurizer water reactor. The nuclear characteristics of the major emitter materials were surveyed, and preliminary calculations of the hybrid fixed incore detector were performed with the MCNP code. The eigenvalue and pin-by-pin fission power distributions were calculated and showed good agreement with the KARMA calculation results. As future work, gamma power distributions as well as several types of XS of the emitter, insulator, and collector regions for a Co-V ICI assembly will be evaluated and compared

Hybrid2 - The hybrid power system simulation model

Energy Technology Data Exchange (ETDEWEB)

Baring-Gould, E.I.; Green, H.J.; Dijk, V.A.P. van [National Renewable Energy Lab., Golden, CO (United States); Manwell, J.F. [Univ. of Massachusetts, Amherst, MA (United States)

1996-12-31

There is a large-scale need and desire for energy in remote communities, especially in the developing world; however the lack of a user friendly, flexible performance prediction model for hybrid power systems incorporating renewables hindered the analysis of hybrids as options to conventional solutions. A user friendly model was needed with the versatility to simulate the many system locations, widely varying hardware configurations, and differing control options for potential hybrid power systems. To meet these ends, researchers from the National Renewable Energy Laboratory (NREL) and the University of Massachusetts (UMass) developed the Hybrid2 software. This paper provides an overview of the capabilities, features, and functionality of the Hybrid2 code, discusses its validation and future plans. Model availability and technical support provided to Hybrid2 users are also discussed. 12 refs., 3 figs., 4 tabs.

Hybrid code simulation on mode conversion in the second harmonic ICRF heating

International Nuclear Information System (INIS)

Sakai, K.; Takeuchi, S.; Matsumoto, M.; Sugihara, R.

1985-01-01

ICRF second harmonic heating of a single-species plasma is studied by using a 1-1/2 dimensional quasi-neutral hybrid code. Mode conversion, transmission and reflection of the magnetosonic waves are confirmed, both for the high- and low-field-side excitations. The ion heating by waves propagating perpendicularly to the static magnetic field is also observed

HyDEn: A Hybrid Steganocryptographic Approach for Data Encryption Using Randomized Error-Correcting DNA Codes

Directory of Open Access Journals (Sweden)

Dan Tulpan

2013-01-01

Full Text Available This paper presents a novel hybrid DNA encryption (HyDEn approach that uses randomized assignments of unique error-correcting DNA Hamming code words for single characters in the extended ASCII set. HyDEn relies on custom-built quaternary codes and a private key used in the randomized assignment of code words and the cyclic permutations applied on the encoded message. Along with its ability to detect and correct errors, HyDEn equals or outperforms existing cryptographic methods and represents a promising in silico DNA steganographic approach.

Burnup calculation for a tokamak commercial hybrid reactor

International Nuclear Information System (INIS)

Feng Kaiming; Xie Zhongyou

1990-08-01

A computer code ISOGEN-III and its associated data library BULIB have been developed for fusion-fission hybrid reactor burnup calculations. These are used to calcuate burnup of a tokamak commercial hybrid reactor. The code and library are introduced briefly, and burnup calculation results are given

A HYDROCHEMICAL HYBRID CODE FOR ASTROPHYSICAL PROBLEMS. I. CODE VERIFICATION AND BENCHMARKS FOR A PHOTON-DOMINATED REGION (PDR)

International Nuclear Information System (INIS)

Motoyama, Kazutaka; Morata, Oscar; Hasegawa, Tatsuhiko; Shang, Hsien; Krasnopolsky, Ruben

2015-01-01

A two-dimensional hydrochemical hybrid code, KM2, is constructed to deal with astrophysical problems that would require coupled hydrodynamical and chemical evolution. The code assumes axisymmetry in a cylindrical coordinate system and consists of two modules: a hydrodynamics module and a chemistry module. The hydrodynamics module solves hydrodynamics using a Godunov-type finite volume scheme and treats included chemical species as passively advected scalars. The chemistry module implicitly solves nonequilibrium chemistry and change of energy due to thermal processes with transfer of external ultraviolet radiation. Self-shielding effects on photodissociation of CO and H 2 are included. In this introductory paper, the adopted numerical method is presented, along with code verifications using the hydrodynamics module and a benchmark on the chemistry module with reactions specific to a photon-dominated region (PDR). Finally, as an example of the expected capability, the hydrochemical evolution of a PDR is presented based on the PDR benchmark

Review of Hybrid (Deterministic/Monte Carlo) Radiation Transport Methods, Codes, and Applications at Oak Ridge National Laboratory

International Nuclear Information System (INIS)

Wagner, John C.; Peplow, Douglas E.; Mosher, Scott W.; Evans, Thomas M.

2010-01-01

This paper provides a review of the hybrid (Monte Carlo/deterministic) radiation transport methods and codes used at the Oak Ridge National Laboratory and examples of their application for increasing the efficiency of real-world, fixed-source Monte Carlo analyses. The two principal hybrid methods are (1) Consistent Adjoint Driven Importance Sampling (CADIS) for optimization of a localized detector (tally) region (e.g., flux, dose, or reaction rate at a particular location) and (2) Forward Weighted CADIS (FW-CADIS) for optimizing distributions (e.g., mesh tallies over all or part of the problem space) or multiple localized detector regions (e.g., simultaneous optimization of two or more localized tally regions). The two methods have been implemented and automated in both the MAVRIC sequence of SCALE 6 and ADVANTG, a code that works with the MCNP code. As implemented, the methods utilize the results of approximate, fast-running 3-D discrete ordinates transport calculations (with the Denovo code) to generate consistent space- and energy-dependent source and transport (weight windows) biasing parameters. These methods and codes have been applied to many relevant and challenging problems, including calculations of PWR ex-core thermal detector response, dose rates throughout an entire PWR facility, site boundary dose from arrays of commercial spent fuel storage casks, radiation fields for criticality accident alarm system placement, and detector response for special nuclear material detection scenarios and nuclear well-logging tools. Substantial computational speed-ups, generally O(10 2-4 ), have been realized for all applications to date. This paper provides a brief review of the methods, their implementation, results of their application, and current development activities, as well as a considerable list of references for readers seeking more information about the methods and/or their applications.

Review of Hybrid (Deterministic/Monte Carlo) Radiation Transport Methods, Codes, and Applications at Oak Ridge National Laboratory

International Nuclear Information System (INIS)

Wagner, John C.; Peplow, Douglas E.; Mosher, Scott W.; Evans, Thomas M.

2010-01-01

This paper provides a review of the hybrid (Monte Carlo/deterministic) radiation transport methods and codes used at the Oak Ridge National Laboratory and examples of their application for increasing the efficiency of real-world, fixed-source Monte Carlo analyses. The two principal hybrid methods are (1) Consistent Adjoint Driven Importance Sampling (CADIS) for optimization of a localized detector (tally) region (e.g., flux, dose, or reaction rate at a particular location) and (2) Forward Weighted CADIS (FW-CADIS) for optimizing distributions (e.g., mesh tallies over all or part of the problem space) or multiple localized detector regions (e.g., simultaneous optimization of two or more localized tally regions). The two methods have been implemented and automated in both the MAVRIC sequence of SCALE 6 and ADVANTG, a code that works with the MCNP code. As implemented, the methods utilize the results of approximate, fast-running 3-D discrete ordinates transport calculations (with the Denovo code) to generate consistent space- and energy-dependent source and transport (weight windows) biasing parameters. These methods and codes have been applied to many relevant and challenging problems, including calculations of PWR ex-core thermal detector response, dose rates throughout an entire PWR facility, site boundary dose from arrays of commercial spent fuel storage casks, radiation fields for criticality accident alarm system placement, and detector response for special nuclear material detection scenarios and nuclear well-logging tools. Substantial computational speed-ups, generally O(102-4), have been realized for all applications to date. This paper provides a brief review of the methods, their implementation, results of their application, and current development activities, as well as a considerable list of references for readers seeking more information about the methods and/or their applications.

Review of hybrid (deterministic/Monte Carlo) radiation transport methods, codes, and applications at Oak Ridge National Laboratory

International Nuclear Information System (INIS)

Wagner, J.C.; Peplow, D.E.; Mosher, S.W.; Evans, T.M.

2010-01-01

This paper provides a review of the hybrid (Monte Carlo/deterministic) radiation transport methods and codes used at the Oak Ridge National Laboratory and examples of their application for increasing the efficiency of real-world, fixed-source Monte Carlo analyses. The two principal hybrid methods are (1) Consistent Adjoint Driven Importance Sampling (CADIS) for optimization of a localized detector (tally) region (e.g., flux, dose, or reaction rate at a particular location) and (2) Forward Weighted CADIS (FW-CADIS) for optimizing distributions (e.g., mesh tallies over all or part of the problem space) or multiple localized detector regions (e.g., simultaneous optimization of two or more localized tally regions). The two methods have been implemented and automated in both the MAVRIC sequence of SCALE 6 and ADVANTG, a code that works with the MCNP code. As implemented, the methods utilize the results of approximate, fast-running 3-D discrete ordinates transport calculations (with the Denovo code) to generate consistent space- and energy-dependent source and transport (weight windows) biasing parameters. These methods and codes have been applied to many relevant and challenging problems, including calculations of PWR ex-core thermal detector response, dose rates throughout an entire PWR facility, site boundary dose from arrays of commercial spent fuel storage casks, radiation fields for criticality accident alarm system placement, and detector response for special nuclear material detection scenarios and nuclear well-logging tools. Substantial computational speed-ups, generally O(10 2-4 ), have been realized for all applications to date. This paper provides a brief review of the methods, their implementation, results of their application, and current development activities, as well as a considerable list of references for readers seeking more information about the methods and/or their applications. (author)

The FLUKA code for space applications Recent developments

CERN Document Server

Andersen, V; Battistoni, G; Campanella, M; Carboni, M; Cerutti, F; Empl, A; Fassò, A; Ferrari, A; Gadioli, E; Garzelli, M V; Lee, K; Ottolenghi, A; Pelliccioni, M; Pinsky, L S; Ranft, J; Roesler, S; Sala, P R; Wilson, T L

2004-01-01

The FLUKA Monte Carlo transport code is widely used for fundamental research, radioprotection and dosimetry, hybrid nuclear energy system and cosmic ray calculations. The validity of its physical models has been benchmarked against a variety of experimental data over a wide range of energies, ranging from accelerator data to cosmic ray showers in the earth atmosphere. The code is presently undergoing several developments in order to better fit the needs of space applications. The generation of particle spectra according to up-to- date cosmic ray data as well as the effect of the solar and geomagnetic modulation have been implemented and already successfully applied to a variety of problems. The implementation of suitable models for heavy ion nuclear interactions has reached an operational stage. At medium/high energy FLUKA is using the DPMJET model. The major task of incorporating heavy ion interactions from a few GeV/n down to the threshold for inelastic collisions is also progressing and promising results h...

Mobile, hybrid Compton/coded aperture imaging for detection, identification and localization of gamma-ray sources at stand-off distances

Science.gov (United States)

Tornga, Shawn R.

The Stand-off Radiation Detection System (SORDS) program is an Advanced Technology Demonstration (ATD) project through the Department of Homeland Security's Domestic Nuclear Detection Office (DNDO) with the goal of detection, identification and localization of weak radiological sources in the presence of large dynamic backgrounds. The Raytheon-SORDS Tri-Modal Imager (TMI) is a mobile truck-based, hybrid gamma-ray imaging system able to quickly detect, identify and localize, radiation sources at standoff distances through improved sensitivity while minimizing the false alarm rate. Reconstruction of gamma-ray sources is performed using a combination of two imaging modalities; coded aperture and Compton scatter imaging. The TMI consists of 35 sodium iodide (NaI) crystals 5x5x2 in3 each, arranged in a random coded aperture mask array (CA), followed by 30 position sensitive NaI bars each 24x2.5x3 in3 called the detection array (DA). The CA array acts as both a coded aperture mask and scattering detector for Compton events. The large-area DA array acts as a collection detector for both Compton scattered events and coded aperture events. In this thesis, developed coded aperture, Compton and hybrid imaging algorithms will be described along with their performance. It will be shown that multiple imaging modalities can be fused to improve detection sensitivity over a broader energy range than either alone. Since the TMI is a moving system, peripheral data, such as a Global Positioning System (GPS) and Inertial Navigation System (INS) must also be incorporated. A method of adapting static imaging algorithms to a moving platform has been developed. Also, algorithms were developed in parallel with detector hardware, through the use of extensive simulations performed with the Geometry and Tracking Toolkit v4 (GEANT4). Simulations have been well validated against measured data. Results of image reconstruction algorithms at various speeds and distances will be presented as well as

Development of new two-dimensional spectral/spatial code based on dynamic cyclic shift code for OCDMA system

Science.gov (United States)

Jellali, Nabiha; Najjar, Monia; Ferchichi, Moez; Rezig, Houria

2017-07-01

In this paper, a new two-dimensional spectral/spatial codes family, named two dimensional dynamic cyclic shift codes (2D-DCS) is introduced. The 2D-DCS codes are derived from the dynamic cyclic shift code for the spectral and spatial coding. The proposed system can fully eliminate the multiple access interference (MAI) by using the MAI cancellation property. The effect of shot noise, phase-induced intensity noise and thermal noise are used to analyze the code performance. In comparison with existing two dimensional (2D) codes, such as 2D perfect difference (2D-PD), 2D Extended Enhanced Double Weight (2D-Extended-EDW) and 2D hybrid (2D-FCC/MDW) codes, the numerical results show that our proposed codes have the best performance. By keeping the same code length and increasing the spatial code, the performance of our 2D-DCS system is enhanced: it provides higher data rates while using lower transmitted power and a smaller spectral width.

A hybrid WDM/OCDMA ring with a dynamic add/drop function based on Fourier code for local area networks.

Science.gov (United States)

Choi, Yong-Kyu; Hosoya, Kenta; Lee, Chung Ghiu; Hanawa, Masanori; Park, Chang-Soo

2011-03-28

We propose and experimentally demonstrate a hybrid WDM/OCDMA ring with a dynamic add/drop function based on Fourier code for local area networks. Dynamic function is implemented by mechanically tuning the Fourier encoder/decoder for optical code division multiple access (OCDMA) encoding/decoding. Wavelength division multiplexing (WDM) is utilized for node assignment and 4-chip Fourier code recovers the matched signal from the codes. For an optical source well adapted to WDM channels and its short optical pulse generation, reflective semiconductor optical amplifiers (RSOAs) are used with a fiber Bragg grating (FBG) and gain-switched. To demonstrate we experimentally investigated a two-node hybrid WDM/OCDMA ring with a 4-chip Fourier encoder/decoder fabricated by cascading four FBGs with the bit error rate (BER) of <10(-9) for the node span of 10.64 km at 1.25 Gb/s.

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

Hybrid digital-analog coding with bandwidth expansion for correlated Gaussian sources under Rayleigh fading

Science.gov (United States)

Yahampath, Pradeepa

2017-12-01

Consider communicating a correlated Gaussian source over a Rayleigh fading channel with no knowledge of the channel signal-to-noise ratio (CSNR) at the transmitter. In this case, a digital system cannot be optimal for a range of CSNRs. Analog transmission however is optimal at all CSNRs, if the source and channel are memoryless and bandwidth matched. This paper presents new hybrid digital-analog (HDA) systems for sources with memory and channels with bandwidth expansion, which outperform both digital-only and analog-only systems over a wide range of CSNRs. The digital part is either a predictive quantizer or a transform code, used to achieve a coding gain. Analog part uses linear encoding to transmit the quantization error which improves the performance under CSNR variations. The hybrid encoder is optimized to achieve the minimum AMMSE (average minimum mean square error) over the CSNR distribution. To this end, analytical expressions are derived for the AMMSE of asymptotically optimal systems. It is shown that the outage CSNR of the channel code and the analog-digital power allocation must be jointly optimized to achieve the minimum AMMSE. In the case of HDA predictive quantization, a simple algorithm is presented to solve the optimization problem. Experimental results are presented for both Gauss-Markov sources and speech signals.

Hybrid threshold adaptable quantum secret sharing scheme with reverse Huffman-Fibonacci-tree coding.

Science.gov (United States)

Lai, Hong; Zhang, Jun; Luo, Ming-Xing; Pan, Lei; Pieprzyk, Josef; Xiao, Fuyuan; Orgun, Mehmet A

2016-08-12

With prevalent attacks in communication, sharing a secret between communicating parties is an ongoing challenge. Moreover, it is important to integrate quantum solutions with classical secret sharing schemes with low computational cost for the real world use. This paper proposes a novel hybrid threshold adaptable quantum secret sharing scheme, using an m-bonacci orbital angular momentum (OAM) pump, Lagrange interpolation polynomials, and reverse Huffman-Fibonacci-tree coding. To be exact, we employ entangled states prepared by m-bonacci sequences to detect eavesdropping. Meanwhile, we encode m-bonacci sequences in Lagrange interpolation polynomials to generate the shares of a secret with reverse Huffman-Fibonacci-tree coding. The advantages of the proposed scheme is that it can detect eavesdropping without joint quantum operations, and permits secret sharing for an arbitrary but no less than threshold-value number of classical participants with much lower bandwidth. Also, in comparison with existing quantum secret sharing schemes, it still works when there are dynamic changes, such as the unavailability of some quantum channel, the arrival of new participants and the departure of participants. Finally, we provide security analysis of the new hybrid quantum secret sharing scheme and discuss its useful features for modern applications.

LiveCode mobile development

CERN Document Server

Lavieri, Edward D

2013-01-01

A practical guide written in a tutorial-style, ""LiveCode Mobile Development Hotshot"" walks you step-by-step through 10 individual projects. Every project is divided into sub tasks to make learning more organized and easy to follow along with explanations, diagrams, screenshots, and downloadable material.This book is great for anyone who wants to develop mobile applications using LiveCode. You should be familiar with LiveCode and have access to a smartphone. You are not expected to know how to create graphics or audio clips.

Color-coded Live Imaging of Heterokaryon Formation and Nuclear Fusion of Hybridizing Cancer Cells.

Science.gov (United States)

Suetsugu, Atsushi; Matsumoto, Takuro; Hasegawa, Kosuke; Nakamura, Miki; Kunisada, Takahiro; Shimizu, Masahito; Saji, Shigetoyo; Moriwaki, Hisataka; Bouvet, Michael; Hoffman, Robert M

2016-08-01

Fusion of cancer cells has been studied for over half a century. However, the steps involved after initial fusion between cells, such as heterokaryon formation and nuclear fusion, have been difficult to observe in real time. In order to be able to visualize these steps, we have established cancer-cell sublines from the human HT-1080 fibrosarcoma, one expressing green fluorescent protein (GFP) linked to histone H2B in the nucleus and a red fluorescent protein (RFP) in the cytoplasm and the other subline expressing RFP in the nucleus (mCherry) linked to histone H2B and GFP in the cytoplasm. The two reciprocal color-coded sublines of HT-1080 cells were fused using the Sendai virus. The fused cells were cultured on plastic and observed using an Olympus FV1000 confocal microscope. Multi-nucleate (heterokaryotic) cancer cells, in addition to hybrid cancer cells with single-or multiple-fused nuclei, including fused mitotic nuclei, were observed among the fused cells. Heterokaryons with red, green, orange and yellow nuclei were observed by confocal imaging, even in single hybrid cells. The orange and yellow nuclei indicate nuclear fusion. Red and green nuclei remained unfused. Cell fusion with heterokaryon formation and subsequent nuclear fusion resulting in hybridization may be an important natural phenomenon between cancer cells that may make them more malignant. The ability to image the complex processes following cell fusion using reciprocal color-coded cancer cells will allow greater understanding of the genetic basis of malignancy. Copyright© 2016 International Institute of Anticancer Research (Dr. John G. Delinassios), All rights reserved.

Analysis on applicable error-correcting code strength of storage class memory and NAND flash in hybrid storage

Science.gov (United States)

Matsui, Chihiro; Kinoshita, Reika; Takeuchi, Ken

2018-04-01

A hybrid of storage class memory (SCM) and NAND flash is a promising technology for high performance storage. Error correction is inevitable on SCM and NAND flash because their bit error rate (BER) increases with write/erase (W/E) cycles, data retention, and program/read disturb. In addition, scaling and multi-level cell technologies increase BER. However, error-correcting code (ECC) degrades storage performance because of extra memory reading and encoding/decoding time. Therefore, applicable ECC strength of SCM and NAND flash is evaluated independently by fixing ECC strength of one memory in the hybrid storage. As a result, weak BCH ECC with small correctable bit is recommended for the hybrid storage with large SCM capacity because SCM is accessed frequently. In contrast, strong and long-latency LDPC ECC can be applied to NAND flash in the hybrid storage with large SCM capacity because large-capacity SCM improves the storage performance.

Comparison of TITAN hybrid deterministic transport code and MCNP5 for simulation of SPECT

International Nuclear Information System (INIS)

Royston, K.; Haghighat, A.; Yi, C.

2010-01-01

Traditionally, Single Photon Emission Computed Tomography (SPECT) simulations use Monte Carlo methods. The hybrid deterministic transport code TITAN has recently been applied to the simulation of a SPECT myocardial perfusion study. The TITAN SPECT simulation uses the discrete ordinates formulation in the phantom region and a simplified ray-tracing formulation outside of the phantom. A SPECT model has been created in the Monte Carlo Neutral particle (MCNP)5 Monte Carlo code for comparison. In MCNP5 the collimator is directly modeled, but TITAN instead simulates the effect of collimator blur using a circular ordinate splitting technique. Projection images created using the TITAN code are compared to results using MCNP5 for three collimator acceptance angles. Normalized projection images for 2.97 deg, 1.42 deg and 0.98 deg collimator acceptance angles had maximum relative differences of 21.3%, 11.9% and 8.3%, respectively. Visually the images are in good agreement. Profiles through the projection images were plotted to find that the TITAN results followed the shape of the MCNP5 results with some differences in magnitude. A timing comparison on 16 processors found that the TITAN code completed the calculation 382 to 2787 times faster than MCNP5. Both codes exhibit good parallel performance. (author)

Development status of TUF code

International Nuclear Information System (INIS)

Liu, W.S.; Tahir, A.; Zaltsgendler

1996-01-01

An overview of the important development of the TUF code in 1995 is presented. The development in the following areas is presented: control of round-off error propagation, gas resolution and release models, and condensation induced water hammer. This development is mainly generated from station requests for operational support and code improvement. (author)

Development of three-dimensional neoclassical transport simulation code with high performance Fortran on a vector-parallel computer

International Nuclear Information System (INIS)

Satake, Shinsuke; Okamoto, Masao; Nakajima, Noriyoshi; Takamaru, Hisanori

2005-11-01

A neoclassical transport simulation code (FORTEC-3D) applicable to three-dimensional configurations has been developed using High Performance Fortran (HPF). Adoption of computing techniques for parallelization and a hybrid simulation model to the δf Monte-Carlo method transport simulation, including non-local transport effects in three-dimensional configurations, makes it possible to simulate the dynamism of global, non-local transport phenomena with a self-consistent radial electric field within a reasonable computation time. In this paper, development of the transport code using HPF is reported. Optimization techniques in order to achieve both high vectorization and parallelization efficiency, adoption of a parallel random number generator, and also benchmark results, are shown. (author)

Recent activities in accelerator code development

International Nuclear Information System (INIS)

Copper, R.K.; Ryne, R.D.

1992-01-01

In this paper we will review recent activities in the area of code development as it affects the accelerator community. We will first discuss the changing computing environment. We will review how the computing environment has changed in the last 10 years, with emphasis on computing power, operating systems, computer languages, graphics standards, and massively parallel processing. Then we will discuss recent code development activities in the areas of electromagnetics codes and beam dynamics codes

Hybrid parallelization of the XTOR-2F code for the simulation of two-fluid MHD instabilities in tokamaks

Science.gov (United States)

Marx, Alain; Lütjens, Hinrich

2017-03-01

A hybrid MPI/OpenMP parallel version of the XTOR-2F code [Lütjens and Luciani, J. Comput. Phys. 229 (2010) 8130] solving the two-fluid MHD equations in full tokamak geometry by means of an iterative Newton-Krylov matrix-free method has been developed. The present work shows that the code has been parallelized significantly despite the numerical profile of the problem solved by XTOR-2F, i.e. a discretization with pseudo-spectral representations in all angular directions, the stiffness of the two-fluid stability problem in tokamaks, and the use of a direct LU decomposition to invert the physical pre-conditioner at every Krylov iteration of the solver. The execution time of the parallelized version is an order of magnitude smaller than the sequential one for low resolution cases, with an increasing speedup when the discretization mesh is refined. Moreover, it allows to perform simulations with higher resolutions, previously forbidden because of memory limitations.

Using Coding Apps to Support Literacy Instruction and Develop Coding Literacy

Science.gov (United States)

Hutchison, Amy; Nadolny, Larysa; Estapa, Anne

2016-01-01

In this article the authors present the concept of Coding Literacy and describe the ways in which coding apps can support the development of Coding Literacy and disciplinary and digital literacy skills. Through detailed examples, we describe how coding apps can be integrated into literacy instruction to support learning of the Common Core English…

A NEW HYBRID N-BODY-COAGULATION CODE FOR THE FORMATION OF GAS GIANT PLANETS

International Nuclear Information System (INIS)

Bromley, Benjamin C.; Kenyon, Scott J.

2011-01-01

We describe an updated version of our hybrid N-body-coagulation code for planet formation. In addition to the features of our 2006-2008 code, our treatment now includes algorithms for the one-dimensional evolution of the viscous disk, the accretion of small particles in planetary atmospheres, gas accretion onto massive cores, and the response of N-bodies to the gravitational potential of the gaseous disk and the swarm of planetesimals. To validate the N-body portion of the algorithm, we use a battery of tests in planetary dynamics. As a first application of the complete code, we consider the evolution of Pluto-mass planetesimals in a swarm of 0.1-1 cm pebbles. In a typical evolution time of 1-3 Myr, our calculations transform 0.01-0.1 M sun disks of gas and dust into planetary systems containing super-Earths, Saturns, and Jupiters. Low-mass planets form more often than massive planets; disks with smaller α form more massive planets than disks with larger α. For Jupiter-mass planets, masses of solid cores are 10-100 M + .

Development of time dependent safety analysis code for plasma anomaly events in fusion reactors

International Nuclear Information System (INIS)

Honda, Takuro; Okazaki, Takashi; Bartels, H.W.; Uckan, N.A.; Seki, Yasushi.

1997-01-01

A safety analysis code SAFALY has been developed to analyze plasma anomaly events in fusion reactors, e.g., a loss of plasma control. The code is a hybrid code comprising a zero-dimensional plasma dynamics and a one-dimensional thermal analysis of in-vessel components. The code evaluates the time evolution of plasma parameters and temperature distributions of in-vessel components. As the plasma-safety interface model, we proposed a robust plasma physics model taking into account updated data for safety assessment. For example, physics safety guidelines for beta limit, density limit and H-L mode confinement transition threshold power, etc. are provided in the model. The model of the in-vessel components are divided into twenty temperature regions in the poloidal direction taking account of radiative heat transfer between each surface of each region. This code can also describe the coolant behavior under hydraulic accidents with the results by hydraulics code and treat vaporization (sublimation) from plasma facing components (PFCs). Furthermore, the code includes the model of impurity transport form PFCs by using a transport probability and a time delay. Quantitative analysis based on the model is possible for a scenario of plasma passive shutdown. We examined the possibility of the code as a safety analysis code for plasma anomaly events in fusion reactors and had a prospect that it would contribute to the safety analysis of the International Thermonuclear Experimental Reactor (ITER). (author)

Theoretical Atomic Physics code development II: ACE: Another collisional excitation code

International Nuclear Information System (INIS)

Clark, R.E.H.; Abdallah, J. Jr.; Csanak, G.; Mann, J.B.; Cowan, R.D.

1988-12-01

A new computer code for calculating collisional excitation data (collision strengths or cross sections) using a variety of models is described. The code uses data generated by the Cowan Atomic Structure code or CATS for the atomic structure. Collisional data are placed on a random access file and can be displayed in a variety of formats using the Theoretical Atomic Physics Code or TAPS. All of these codes are part of the Theoretical Atomic Physics code development effort at Los Alamos. 15 refs., 10 figs., 1 tab

Development and validation of sodium fire codes

International Nuclear Information System (INIS)

Morii, Tadashi; Himeno Yoshiaki; Miyake, Osamu

1989-01-01

Development, verification, and validation of the spray fire code, SPRAY-3M, the pool fire codes, SOFIRE-M2 and SPM, the aerosol behavior code, ABC-INTG, and the simultaneous spray and pool fires code, ASSCOPS, are presented. In addition, the state-of-the-art of development of the multi-dimensional natural convection code, SOLFAS, for the analysis of heat-mass transfer during a fire, is presented. (author)

Reactor safety computer code development at INEL

International Nuclear Information System (INIS)

Johnsen, G.W.

1985-01-01

This report provides a brief overview of the computer code development programs being conducted at EG and G Idaho, Inc. on behalf of US Nuclear Regulatory Commission and the Department of Energy, Idaho Operations Office. Included are descriptions of the codes being developed, their development status as of the date of this report, and resident code development expertise

Development of a coupled code system based on system transient code, RETRAN, and 3-D neutronics code, MASTER

International Nuclear Information System (INIS)

Kim, K. D.; Jung, J. J.; Lee, S. W.; Cho, B. O.; Ji, S. K.; Kim, Y. H.; Seong, C. K.

2002-01-01

A coupled code system of RETRAN/MASTER has been developed for best-estimate simulations of interactions between reactor core neutron kinetics and plant thermal-hydraulics by incorporation of a 3-D reactor core kinetics analysis code, MASTER into system transient code, RETRAN. The soundness of the consolidated code system is confirmed by simulating the MSLB benchmark problem developed to verify the performance of a coupled kinetics and system transient codes by OECD/NEA

Fluid-structure-interaction analyses of reactor vessel using improved hybrid Lagrangian Eulerian code ALICE-II

Energy Technology Data Exchange (ETDEWEB)

Wang, C.Y.

1993-06-01

This paper describes fluid-structure-interaction and structure response analyses of a reactor vessel subjected to loadings associated with postulated accidents, using the hybrid Lagrangian-Eulerian code ALICE-II. This code has been improved recently to accommodate many features associated with innovative designs of reactor vessels. Calculational capabilities have been developed to treat water in the reactor cavity outside the vessel, internal shield structures and internal thin shells. The objective of the present analyses is to study the cover response and potential for missile generation in response to a fuel-coolant interaction in the core region. Three calculations were performed using the cover weight as a parameter. To study the effect of the cavity water, vessel response calculations for both wet- and dry-cavity designs are compared. Results indicate that for all cases studied and for the design parameters assumed, the calculated cover displacements are all smaller than the bolts` ultimate displacement and no missile generation of the closure head is predicted. Also, solutions reveal that the cavity water of the wet-cavity design plays an important role of restraining the downward displacement of the bottom head. Based on these studies, the analyses predict that the structure integrity is maintained throughout the postulated accident for the wet-cavity design.

Fluid-structure-interaction analyses of reactor vessel using improved hybrid Lagrangian Eulerian code ALICE-II

Energy Technology Data Exchange (ETDEWEB)

Wang, C.Y.

1993-01-01

This paper describes fluid-structure-interaction and structure response analyses of a reactor vessel subjected to loadings associated with postulated accidents, using the hybrid Lagrangian-Eulerian code ALICE-II. This code has been improved recently to accommodate many features associated with innovative designs of reactor vessels. Calculational capabilities have been developed to treat water in the reactor cavity outside the vessel, internal shield structures and internal thin shells. The objective of the present analyses is to study the cover response and potential for missile generation in response to a fuel-coolant interaction in the core region. Three calculations were performed using the cover weight as a parameter. To study the effect of the cavity water, vessel response calculations for both wet- and dry-cavity designs are compared. Results indicate that for all cases studied and for the design parameters assumed, the calculated cover displacements are all smaller than the bolts' ultimate displacement and no missile generation of the closure head is predicted. Also, solutions reveal that the cavity water of the wet-cavity design plays an important role of restraining the downward displacement of the bottom head. Based on these studies, the analyses predict that the structure integrity is maintained throughout the postulated accident for the wet-cavity design.

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

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

Status of SPACE Safety Analysis Code Development

International Nuclear Information System (INIS)

Lee, Dong Hyuk; Yang, Chang Keun; Kim, Se Yun; Ha, Sang Jun

2009-01-01

In 2006, the Korean the Korean nuclear industry started developing a thermal-hydraulic analysis code for safety analysis of PWR(Pressurized Water Reactor). The new code is named as SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). The SPACE code can solve two-fluid, three-field governing equations in one dimensional or three dimensional geometry. The SPACE code has many component models required for modeling a PWR, such as reactor coolant pump, safety injection tank, etc. The programming language used in the new code is C++, for new generation of engineers who are more comfortable with C/C++ than old FORTRAN language. This paper describes general characteristics of SPACE code and current status of SPACE code development

Towards Product Lining Model-Driven Development Code Generators

OpenAIRE

Roth, Alexander; Rumpe, Bernhard

2015-01-01

A code generator systematically transforms compact models to detailed code. Today, code generation is regarded as an integral part of model-driven development (MDD). Despite its relevance, the development of code generators is an inherently complex task and common methodologies and architectures are lacking. Additionally, reuse and extension of existing code generators only exist on individual parts. A systematic development and reuse based on a code generator product line is still in its inf...

Description and applicability of the BEFEM-CODE

Energy Technology Data Exchange (ETDEWEB)

Groth, T.

1980-05-15

The BEFEM-CODE, developed for rock mechanics problems in hard rock with joints, is a simple FEM code constructed using triangular and quadrilateral elements. As an option, a joint element of the Goodman type may be used. The Cook-Pian type quadrilateral stress hybrid element was introduced into the version of the code used for the Naesliden project, to replace the constant stress quadrilateral elements. This hybrid element, derived with assumed stress distributions, simplifies the excavation process for use in non-linear models. The shear behavior of the Goodman 1976 joint element has been replaced by Goodman's 1968 formulation. This element makes it possible to take dilation into account, but it was not considered necessary to use dilation to simulate proper joint behavior in the Naesliden project. The code uses Barton's shear strength criteria. Excessive nodal forces due to failure and non-linearities in the joint elements are redistributed with stress transfer iterations. Convergence can be speeded up by dividing each excavation sequence into several loadsteps in which the stiffness matrix is recalculated.

Development of 2-d cfd code

International Nuclear Information System (INIS)

Mirza, S.A.

1999-01-01

In the present study, a two-dimensional computer code has been developed in FORTRAN using CFD technique, which is basically a numerical scheme. This computer code solves the Navier Stokes equations and continuity equation to find out the velocity and pressure fields within a given domain. This analysis has been done for the developed within a square cavity driven by the upper wall which has become a bench mark for testing and comparing the newly developed numerical schemes. Before to handle this task, different one-dimensional cases have been studied by CFD technique and their FORTRAN programs written. The cases studied are Couette flow, Poiseuille flow with and without using symmetric boundary condition. Finally a comparison between CFD results and analytical results has also been made. For the cavity flow the results from the developed code have been obtained for different Reynolds numbers which are finally presented in the form of velocity vectors. The comparison of the developed code results have been made with the results obtained from the share ware version of a commercially available code for Reynolds number of 10.0. The disagreement in the results quantitatively and qualitatively at some grid points of the calculation domain have been discussed and future recommendations in this regard have also been made. (author)

Simulation of lower hybrid current drive in enhanced reversed shear plasmas in the tokamak fusion test reactor using the lower hybrid simulation code

International Nuclear Information System (INIS)

Kaita, R.; Bernabei, S.; Budny, R.

1996-01-01

The Enhanced Reversed Shear (ERS) mode has already shown great potential for improving the performance of the Tokamak Fusion Test Reactor (TFTR) and other devices. Sustaining the ERS, however, remains an outstanding problem. Lower hybrid (LH) current drive is a possible method for modifying the current profile and controlling its time evolution. To predict its effectiveness in TFTR, the Lower Hybrid Simulation Code (LSC) model is used in the TRANSP code and the Tokamak Simulation Code (TSC). Among the results from the simulations are the following. (1) Single-pass absorption is expected in TFTR ERS plasmas. The simulations show that the LH current follows isotherms of the electron temperature. The ability to control the location of the minimum in the q profile (q min ) has been demonstrated by varying the phase velocity of the launched LH waves and observing the change in the damping location. (2) LH current drive can been used to sustain the q min location. The tendency of qmin to drift inward, as the inductive current diffuses during the formation phase of the reversed shear discharge, is prevented by the LH current driven at a fixed radial location. If this results in an expanded plasma volume with improved confinement as high power neutral beam injection is applied, the high bootstrap currents induced during this phase can then maintain the larger qmin radius. (3) There should be no LH wave damping on energetic beam particles. The values of perpendicular index of refraction in the calculations never exceed about 20, while ions at TFR injection energies are resonant with waves having values closer to 100. Other issues being addressed in the study include the LH current drive efficiency in the presence of high bootstrap currents, and the effect of fast electron diffusion on LH current localization

Computer-assisted Particle-in-Cell code development

International Nuclear Information System (INIS)

Kawata, S.; Boonmee, C.; Teramoto, T.; Drska, L.; Limpouch, J.; Liska, R.; Sinor, M.

1997-12-01

This report presents a new approach for an electromagnetic Particle-in-Cell (PIC) code development by a computer: in general PIC codes have a common structure, and consist of a particle pusher, a field solver, charge and current density collections, and a field interpolation. Because of the common feature, the main part of the PIC code can be mechanically developed on a computer. In this report we use the packages FIDE and GENTRAN of the REDUCE computer algebra system for discretizations of field equations and a particle equation, and for an automatic generation of Fortran codes. The approach proposed is successfully applied to the development of 1.5-dimensional PIC code. By using the generated PIC code the Weibel instability in a plasma is simulated. The obtained growth rate agrees well with the theoretical value. (author)

Hardware in the loop platform development for hybrid vehicles

Energy Technology Data Exchange (ETDEWEB)

Wilhelm, E. [ETH Zurich, Zurich (Switzerland); Fowler, E.; Stevens, M.B. [Waterloo Univ., ON (Canada). Dept. of Chemical Engineering; Fraser, M.W. [Waterloo Univ., ON (Canada). Dept. of Mechanical Engineering

2007-07-01

This paper described a hardware-in-the-loop (HIL) validation simulation system designed to evaluate hybrid control strategies. The system was designed to reduce development costs and improve the safety of hybrid vehicle control systems. Model-based design processes for power trains typically include a series of processes to assess the real time and physical limitations of control systems prior to in-vehicle testing. The study used a 70 kW nickel metal hydride battery; a 67 kW 3-phase induction traction motor; and, a high voltage DC-DC converter within a fuel cell Chevrolet Equinox. Two physical vehicle controllers were used to interface with the virtual vehicle simulation in real time. System performance was monitored with a supervisory computer. A software in the loop (SIL) process was conducted to assess torque control and regenerative braking algorithm validation. An analysis of the controller code showed that a Simulink-native integrator block was updating too slowly. A custom integration term calculation was written. The charge control was then validated and tuned. It was concluded that use of the HIL system mitigated the risk of component damage through the identification and correction of unstable control logic. 10 refs., 2 tabs., 10 figs.

Hybrid and Parallel Domain-Decomposition Methods Development to Enable Monte Carlo for Reactor Analyses

International Nuclear Information System (INIS)

Wagner, John C.; Mosher, Scott W.; Evans, Thomas M.; Peplow, Douglas E.; Turner, John A.

2010-01-01

This paper describes code and methods development at the Oak Ridge National Laboratory focused on enabling high-fidelity, large-scale reactor analyses with Monte Carlo (MC). Current state-of-the-art tools and methods used to perform real commercial reactor analyses have several undesirable features, the most significant of which is the non-rigorous spatial decomposition scheme. Monte Carlo methods, which allow detailed and accurate modeling of the full geometry and are considered the gold standard for radiation transport solutions, are playing an ever-increasing role in correcting and/or verifying the deterministic, multi-level spatial decomposition methodology in current practice. However, the prohibitive computational requirements associated with obtaining fully converged, system-wide solutions restrict the role of MC to benchmarking deterministic results at a limited number of state-points for a limited number of relevant quantities. The goal of this research is to change this paradigm by enabling direct use of MC for full-core reactor analyses. The most significant of the many technical challenges that must be overcome are the slow, non-uniform convergence of system-wide MC estimates and the memory requirements associated with detailed solutions throughout a reactor (problems involving hundreds of millions of different material and tally regions due to fuel irradiation, temperature distributions, and the needs associated with multi-physics code coupling). To address these challenges, our research has focused on the development and implementation of (1) a novel hybrid deterministic/MC method for determining high-precision fluxes throughout the problem space in k-eigenvalue problems and (2) an efficient MC domain-decomposition (DD) algorithm that partitions the problem phase space onto multiple processors for massively parallel systems, with statistical uncertainty estimation. The hybrid method development is based on an extension of the FW-CADIS method, which

Hybrid and parallel domain-decomposition methods development to enable Monte Carlo for reactor analyses

International Nuclear Information System (INIS)

Wagner, J.C.; Mosher, S.W.; Evans, T.M.; Peplow, D.E.; Turner, J.A.

2010-01-01

This paper describes code and methods development at the Oak Ridge National Laboratory focused on enabling high-fidelity, large-scale reactor analyses with Monte Carlo (MC). Current state-of-the-art tools and methods used to perform 'real' commercial reactor analyses have several undesirable features, the most significant of which is the non-rigorous spatial decomposition scheme. Monte Carlo methods, which allow detailed and accurate modeling of the full geometry and are considered the 'gold standard' for radiation transport solutions, are playing an ever-increasing role in correcting and/or verifying the deterministic, multi-level spatial decomposition methodology in current practice. However, the prohibitive computational requirements associated with obtaining fully converged, system-wide solutions restrict the role of MC to benchmarking deterministic results at a limited number of state-points for a limited number of relevant quantities. The goal of this research is to change this paradigm by enabling direct use of MC for full-core reactor analyses. The most significant of the many technical challenges that must be overcome are the slow, non-uniform convergence of system-wide MC estimates and the memory requirements associated with detailed solutions throughout a reactor (problems involving hundreds of millions of different material and tally regions due to fuel irradiation, temperature distributions, and the needs associated with multi-physics code coupling). To address these challenges, our research has focused on the development and implementation of (1) a novel hybrid deterministic/MC method for determining high-precision fluxes throughout the problem space in k-eigenvalue problems and (2) an efficient MC domain-decomposition (DD) algorithm that partitions the problem phase space onto multiple processors for massively parallel systems, with statistical uncertainty estimation. The hybrid method development is based on an extension of the FW-CADIS method

Development of Coolant Radioactivity Interpretation Code

International Nuclear Information System (INIS)

Kim, Kiyoung; Jung, Youngsuk; Kim, Kyounghyun; Kim, Jangwook

2013-01-01

In Korea, the coolant radioactivity analysis has been performed by using the computer codes of foreign companies such as CADE (Westinghouse), IODYNE and CESIUM (ABB-CE). However, these computer codes are too conservative and have involved considerable errors. Furthermore, since these codes are DOS-based program, their easy operability is not satisfactory. Therefore it is required development of an enhanced analysis algorithm applying an analytical method reflecting the change of operational environments of domestic nuclear power plants and a fuel failure evaluation software considering user' conveniences. We have developed a nuclear fuel failure evaluation code able to estimate the number of failed fuel rods and the burn-up of failed fuels during nuclear power plant operation cycle. A Coolant Radio-activity Interpretation Code (CRIC) for LWR has been developed as the output of the project 'Development of Fuel Reliability Enhanced Technique' organized by Korea Institute of Energy Technology Evaluation and Planning (KETEP). The CRIC is Windows based-software able to evaluate the number of failed fuel rods and the burn-up of failed fuel region by analyzing coolant radioactivity of LWR in operation. The CRIC is based on the model of fission products release commonly known as 'three region model' (pellet region, gap region, and coolant region), and we are verifying the CRIC results based on the cases of domestic fuel failures. CRIC users are able to estimate the number of failed fuel rods, burn-up and regions of failed fuel considered enrichment and power distribution of fuel region by using operational cycle data, coolant activity data, fuel loading pattern, Cs-134/Cs-137 ratio according to burn-up and U-235 enrichment provided in the code. Due to development of the CRIC, it is secured own unique fuel failure evaluation code. And, it is expected to have the following significant meaning. This is that the code reflecting a proprietary technique for quantitatively

Development on hybrid evaluated nuclear data library HENDL1.0/MG/MC

International Nuclear Information System (INIS)

Xu Dezheng; Gao Chunjing; Zheng Shanliang; Liu Haibo; Zhu Xiaoxiang; Li Jingjing; Wu Yican

2004-01-01

A Hybrid Evaluated Nuclear Data Library (HENDL) named as HENDL1.0 has been developed by Fusion Design Study (FDS) team of Institute of Plasma Physics, Academia Sinica (ASIPP) to take into account the requirements in design and research relevant to fusion, fission and fusion-fission sub-critical hybrid reactor. HENDLI1.0 contains one basic evaluated sub-library naming HENDL1.0/E and to processed working sub-libraries naming HENDL1.0/MG and HENDL1.0/MC, respectively. Through carefully comparing, distinguishing and choosing, HENDL1.0/E integrated basic evaluated neutron data files of 213 nuclides from the several main data libraries for evaluated neutron reaction cross sections including ENDF/B-VI (USA), JEF-2.2 (OECD/NEA, Europe), JENDL-3.2 (Japan), CENDL-2 (China), BROND-2 (Russia) and FENDL-2 (IAEA/NDS, ITER program). Based on this, 175-group neutron and 42-group photon neutron-photon coupled multi-group working library HENDL1.0/MG used for discrete ordinate Sn method transport calculation (such as ANISN code) and a compact ENDF form (ACE), continuous energy structure (pointwise) neutron cross section library HENDL1.0/MC for Monte Carlo method transport simulation (as MCMP code) can be attainable with the current group constants processing system NJOY and transport cross section preparation code TRANSX referring to the Vitamin-J energy group structure. In addition, two special bases i.e. transmutation (burnup) library BURNUP. DAT and response function library RESPONSE.DAT, have been also made for fuel cycle calculation and reactivity analyses of nuclear reactor. The relevant sample testing, benchmark checking and primary confirmation are also carried out to assess the validity of multi-purpose data library HENDL1.0. (authors)

Variable weight Khazani-Syed code using hybrid fixed-dynamic technique for optical code division multiple access system

Science.gov (United States)

Anas, Siti Barirah Ahmad; Seyedzadeh, Saleh; Mokhtar, Makhfudzah; Sahbudin, Ratna Kalos Zakiah

2016-10-01

Future Internet consists of a wide spectrum of applications with different bit rates and quality of service (QoS) requirements. Prioritizing the services is essential to ensure that the delivery of information is at its best. Existing technologies have demonstrated how service differentiation techniques can be implemented in optical networks using data link and network layer operations. However, a physical layer approach can further improve system performance at a prescribed received signal quality by applying control at the bit level. This paper proposes a coding algorithm to support optical domain service differentiation using spectral amplitude coding techniques within an optical code division multiple access (OCDMA) scenario. A particular user or service has a varying weight applied to obtain the desired signal quality. The properties of the new code are compared with other OCDMA codes proposed for service differentiation. In addition, a mathematical model is developed for performance evaluation of the proposed code using two different detection techniques, namely direct decoding and complementary subtraction.

Graphical user interface development for the MARS code

International Nuclear Information System (INIS)

Jeong, J.-J.; Hwang, M.; Lee, Y.J.; Kim, K.D.; Chung, B.D.

2003-01-01

KAERI has developed the best-estimate thermal-hydraulic system code MARS using the RELAP5/MOD3 and COBRA-TF codes. To exploit the excellent features of the two codes, we consolidated the two codes. Then, to improve the readability, maintainability, and portability of the consolidated code, all the subroutines were completely restructured by employing a modular data structure. At present, a major part of the MARS code development program is underway to improve the existing capabilities. The code couplings with three-dimensional neutron kinetics, containment analysis, and transient critical heat flux calculations have also been carried out. At the same time, graphical user interface (GUI) tools have been developed for user friendliness. This paper presents the main features of the MARS GUI. The primary objective of the GUI development was to provide a valuable aid for all levels of MARS users in their output interpretation and interactive controls. Especially, an interactive control function was designed to allow operator actions during simulation so that users can utilize the MARS code like conventional nuclear plant analyzers (NPAs). (author)

Thermal safety analysis for pebble bed blanket fusion-fission hybrid reactor

International Nuclear Information System (INIS)

Wei Renjie

1998-01-01

Pebble bed blanket hybrid reactor may have more advantages than slab element blanket hybrid reactor in nuclear fuel production and nuclear safety. The thermo-hydraulic calculations of the blanket in the Tokamak helium cooling pebble bed blanket fusion-fission hybrid reactor developed in China are carried out using the Code THERMIX and auxiliary code. In the calculations different fuel pebble material and steady state, depressurization and total loss of flow accident conditions are included. The results demonstrate that the conceptual design of the Tokamak helium cooling pebble bed blanket fusion-fission hybrid reactor with dump tank is feasible and safe enough only if the suitable fuel pebble material is selected and the suitable control system and protection system are established. Some recommendations for due conceptual design are also presented

CATHARE code development and assessment methodologies

International Nuclear Information System (INIS)

Micaelli, J.C.; Barre, F.; Bestion, D.

1995-01-01

The CATHARE thermal-hydraulic code has been developed jointly by Commissariat a l'Energie Atomique (CEA), Electricite de France (EdF), and Framatorne for safety analysis. Since the beginning of the project (September 1979), development and assessment activities have followed a methodology supported by two series of experimental tests: separate effects tests and integral effects tests. The purpose of this paper is to describe this methodology, the code assessment status, and the evolution to take into account two new components of this program: the modeling of three-dimensional phenomena and the requirements of code uncertainty evaluation

MIDAS/PK code development using point kinetics model

International Nuclear Information System (INIS)

Song, Y. M.; Park, S. H.

1999-01-01

In this study, a MIDAS/PK code has been developed for analyzing the ATWS (Anticipated Transients Without Scram) which can be one of severe accident initiating events. The MIDAS is an integrated computer code based on the MELCOR code to develop a severe accident risk reduction strategy by Korea Atomic Energy Research Institute. In the mean time, the Chexal-Layman correlation in the current MELCOR, which was developed under a BWR condition, is appeared to be inappropriate for a PWR. So as to provide ATWS analysis capability to the MIDAS code, a point kinetics module, PKINETIC, has first been developed as a stand-alone code whose reference model was selected from the current accident analysis codes. In the next step, the MIDAS/PK code has been developed via coupling PKINETIC with the MIDAS code by inter-connecting several thermal hydraulic parameters between the two codes. Since the major concern in the ATWS analysis is the primary peak pressure during the early few minutes into the accident, the peak pressure from the PKINETIC module and the MIDAS/PK are compared with the RETRAN calculations showing a good agreement between them. The MIDAS/PK code is considered to be valuable for analyzing the plant response during ATWS deterministically, especially for the early domestic Westinghouse plants which rely on the operator procedure instead of an AMSAC (ATWS Mitigating System Actuation Circuitry) against ATWS. This capability of ATWS analysis is also important from the view point of accident management and mitigation

Development of bio-hybrid material based on Salmonella ...

African Journals Online (AJOL)

The immobilization of a whole microbial cell is an important process used in nanotechnology of biosensors and other related fields, especially the development of bio-hybrid materials based on live organisms and inorganic compounds. Here, we described an essay to develop a bio-hybrid material based on Salmonella ...

Program Hybrid/GDH. Revision

International Nuclear Information System (INIS)

Blann, M.; Bisplinghoff, J.

1975-10-01

This code is the most recent in a series of codes for doing a-priori pre-equilibrium decay calculations. It has been written to permit the user to exercise many options at time of execution. It will, for example, permit calculation with either Hybrid model or the geometry dependent Hybrid model (GDH). Intranuclear transition rates can be calculated using either a nucleon-nucleon scattering approach (improved over earlier results) or based on the imaginary optical potential. Transition rates based on exciton lifetimes can be selected (as suggested in the Hybrid model formulation) or an average lifetime for each n-exciton configuration may be selected

Accident consequence assessment code development

International Nuclear Information System (INIS)

Homma, T.; Togawa, O.

1991-01-01

This paper describes the new computer code system, OSCAAR developed for off-site consequence assessment of a potential nuclear accident. OSCAAR consists of several modules which have modeling capabilities in atmospheric transport, foodchain transport, dosimetry, emergency response and radiological health effects. The major modules of the consequence assessment code are described, highlighting the validation and verification of the models. (author)

Development of MCNP interface code in HFETR

International Nuclear Information System (INIS)

Qiu Liqing; Fu Rong; Deng Caiyu

2007-01-01

In order to describe the HFETR core with MCNP method, the interface code MCNPIP for HFETR and MCNP code is developed. This paper introduces the core DXSY and flowchart of MCNPIP code, and the handling of compositions of fuel elements and requirements on hardware and software. Finally, MCNPIP code is validated against the practical application. (authors)

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.

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

Development of the New Light-Duty Hybrid Truck

OpenAIRE

Yamaguchi, Koichi

2008-01-01

Hino Motors, Ltd., developed the new light-duty hybrid truck whose traction motor, inverter, and traction battery were completely redesigned for maximizing output and efficiency. It also succeeds in balancing low fuel economy and low exhaust emissions by utilizing a combination of a new hybrid system control with a specially developed diesel engine.

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

Development of Premacy Hydrogen RE Hybrid

Energy Technology Data Exchange (ETDEWEB)

Wakayama, N. [Mazda Motor Corporation, Hiroshima (Japan)

2010-07-01

Hydrogen powered ICE (internal combustion engine) vehicles can play an important role as an automotive power source in the future, because of its higher reliability and cost performance than those of fuel cell vehicles. Combined with hydrogen, Mazda's unique rotary engine (RE) has merits such as a prevention of hydrogen pre-ignition. Mazda has been developing hydrogen vehicles with the hydrogen RE from the early 1990s. Premacy (Mazda5) Hydrogen RE Hybrid was developed and launched in 2009, following RX-8 Hydrogen RE delivered in 2006. A series hybrid system was adopted in Premacy Hydrogen RE Hybrid. A traction motor switches its windings while the vehicle is moving. This switching technology allows the motor to be small and high-efficient. The lithium-ion high voltage battery, which has excellent input-output characteristics, was installed. These features extend the hydrogen fuel driving range to 200 km and obtain excellent acceleration performance. The hydrogen RE can be also operated by gasoline (Dual Fuel System). The additional gasoline operation makes hydrogen vehicles possible to drive in non-hydrogen station area. With approval from the Japanese Ministry of Land Infrastructure and Transport, Mazda Premacy Hydrogen RE Hybrid was delivered successfully to the Japanese market in the form of leasing. (orig.)

Reactor systems modeling for ICF hybrids

International Nuclear Information System (INIS)

Berwald, D.H.; Meier, W.R.

1980-10-01

The computational models of ICF reactor subsystems developed by LLNL and TRW are described and a computer program was incorporated for use in the EPRI-sponsored Feasibility Assessment of Fusion-Fission Hybrids. Representative parametric variations have been examined. Many of the ICF subsystem models are very preliminary and more quantitative models need to be developed and included in the code

COSINE software development based on code generation technology

International Nuclear Information System (INIS)

Ren Hao; Mo Wentao; Liu Shuo; Zhao Guang

2013-01-01

The code generation technology can significantly improve the quality and productivity of software development and reduce software development risk. At present, the code generator is usually based on UML model-driven technology, which can not satisfy the development demand of nuclear power calculation software. The feature of scientific computing program was analyzed and the FORTRAN code generator (FCG) based on C# was developed in this paper. FCG can generate module variable definition FORTRAN code automatically according to input metadata. FCG also can generate memory allocation interface for dynamic variables as well as data access interface. FCG was applied to the core and system integrated engine for design and analysis (COSINE) software development. The result shows that FCG can greatly improve the development efficiency of nuclear power calculation software, and reduce the defect rate of software development. (authors)

Development and application of the waste code

International Nuclear Information System (INIS)

Morison, I.W.

1984-01-01

This paper discusses the objectives and general approach underlying the Australian Code of Practice on the Management of Radioactive Wastes arising from the Mining and Milling of Radioactive Ores 1982. Background to the development of the Code is provided and the guidelines which supplement the Code are considered

Health Code Number (HCN) Development Procedure

Energy Technology Data Exchange (ETDEWEB)

Petrocchi, Rocky; Craig, Douglas K.; Bond, Jayne-Anne; Trott, Donna M.; Yu, Xiao-Ying

2013-09-01

This report provides the detailed description of health code numbers (HCNs) and the procedure of how each HCN is assigned. It contains many guidelines and rationales of HCNs. HCNs are used in the chemical mixture methodology (CMM), a method recommended by the department of energy (DOE) for assessing health effects as a result of exposures to airborne aerosols in an emergency. The procedure is a useful tool for proficient HCN code developers. Intense training and quality assurance with qualified HCN developers are required before an individual comprehends the procedure to develop HCNs for DOE.

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

Development of the DTNTES code

International Nuclear Information System (INIS)

Ortega Prieto, P.; Morales Dorado, M.D.; Alonso Santos, A.

1987-01-01

The DTNTES code has been developed in the Department of Nuclear Technology of the Polytechnical University in Madrid as a part of the Research Program on Quantitative Risk Analysis. DTNTES code calculates several time-dependent probabilistic characteristics of basic events, minimal cut sets and the top event of a fault tree. The code assumes that basic events are statistically independent, and they have failure and repair distributions. It computes the minimal cut upper bound approach for the top event unavailability, and the time-dependent unreliability of the top event by means of different methods, selected by the user. These methods are: expected number of system failures, failure rate, Barlow-Proschan bound, steady-state upper bound, and T* method. (author)

Breaking and Fixing Origin-Based Access Control in Hybrid Web/Mobile Application Frameworks.

Science.gov (United States)

Georgiev, Martin; Jana, Suman; Shmatikov, Vitaly

2014-02-01

Hybrid mobile applications (apps) combine the features of Web applications and "native" mobile apps. Like Web applications, they are implemented in portable, platform-independent languages such as HTML and JavaScript. Like native apps, they have direct access to local device resources-file system, location, camera, contacts, etc. Hybrid apps are typically developed using hybrid application frameworks such as PhoneGap. The purpose of the framework is twofold. First, it provides an embedded Web browser (for example, WebView on Android) that executes the app's Web code. Second, it supplies "bridges" that allow Web code to escape the browser and access local resources on the device. We analyze the software stack created by hybrid frameworks and demonstrate that it does not properly compose the access-control policies governing Web code and local code, respectively. Web code is governed by the same origin policy, whereas local code is governed by the access-control policy of the operating system (for example, user-granted permissions in Android). The bridges added by the framework to the browser have the same local access rights as the entire application, but are not correctly protected by the same origin policy. This opens the door to fracking attacks, which allow foreign-origin Web content included into a hybrid app (e.g., ads confined in iframes) to drill through the layers and directly access device resources. Fracking vulnerabilities are generic: they affect all hybrid frameworks, all embedded Web browsers, all bridge mechanisms, and all platforms on which these frameworks are deployed. We study the prevalence of fracking vulnerabilities in free Android apps based on the PhoneGap framework. Each vulnerability exposes sensitive local resources-the ability to read and write contacts list, local files, etc.-to dozens of potentially malicious Web domains. We also analyze the defenses deployed by hybrid frameworks to prevent resource access by foreign-origin Web content

Breaking and Fixing Origin-Based Access Control in Hybrid Web/Mobile Application Frameworks

Science.gov (United States)

Georgiev, Martin; Jana, Suman; Shmatikov, Vitaly

2014-01-01

Hybrid mobile applications (apps) combine the features of Web applications and “native” mobile apps. Like Web applications, they are implemented in portable, platform-independent languages such as HTML and JavaScript. Like native apps, they have direct access to local device resources—file system, location, camera, contacts, etc. Hybrid apps are typically developed using hybrid application frameworks such as PhoneGap. The purpose of the framework is twofold. First, it provides an embedded Web browser (for example, WebView on Android) that executes the app's Web code. Second, it supplies “bridges” that allow Web code to escape the browser and access local resources on the device. We analyze the software stack created by hybrid frameworks and demonstrate that it does not properly compose the access-control policies governing Web code and local code, respectively. Web code is governed by the same origin policy, whereas local code is governed by the access-control policy of the operating system (for example, user-granted permissions in Android). The bridges added by the framework to the browser have the same local access rights as the entire application, but are not correctly protected by the same origin policy. This opens the door to fracking attacks, which allow foreign-origin Web content included into a hybrid app (e.g., ads confined in iframes) to drill through the layers and directly access device resources. Fracking vulnerabilities are generic: they affect all hybrid frameworks, all embedded Web browsers, all bridge mechanisms, and all platforms on which these frameworks are deployed. We study the prevalence of fracking vulnerabilities in free Android apps based on the PhoneGap framework. Each vulnerability exposes sensitive local resources—the ability to read and write contacts list, local files, etc.—to dozens of potentially malicious Web domains. We also analyze the defenses deployed by hybrid frameworks to prevent resource access by foreign

Development of an Auto-Validation Program for MARS Code Assessments

International Nuclear Information System (INIS)

Lee, Young Jin; Chung, Bub Dong

2006-01-01

MARS (Multi-dimensional Analysis of Reactor Safety) code is a best-estimate thermal hydraulic system analysis code developed at KAERI. It is important for a thermal hydraulic computer code to be assessed against theoretical and experimental data to verify and validate the performance and the integrity of the structure, models and correlations of the code. The code assessment efforts for complex thermal hydraulics code such as MARS code can be tedious, time-consuming and require large amount of human intervention in data transfer to see the results in graphic forms. Code developers produce many versions of a code during development and each version need to be verified for integrity. Thus, for MARS code developers, it is desirable to have an automatic way of carrying out the code assessment calculations. In the present work, an Auto-Validation program that carries out the code assessment efforts has been developed. The program uses the user supplied configuration file (with '.vv' extension) which contain commands to read input file, to execute the user selected MARS program, and to generate result graphs. The program can be useful if a same set of code assessments is repeated with different versions of the code. The program is written with the Delphi program language. The program runs under the Microsoft Windows environment

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

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

Subchannel analysis code development for CANDU fuel channel

International Nuclear Information System (INIS)

Park, J. H.; Suk, H. C.; Jun, J. S.; Oh, D. J.; Hwang, D. H.; Yoo, Y. J.

1998-07-01

Since there are several subchannel codes such as COBRA and TORC codes for a PWR fuel channel but not for a CANDU fuel channel in our country, the subchannel analysis code for a CANDU fuel channel was developed for the prediction of flow conditions on the subchannels, for the accurate assessment of the thermal margin, the effect of appendages, and radial/axial power profile of fuel bundles on flow conditions and CHF and so on. In order to develop the subchannel analysis code for a CANDU fuel channel, subchannel analysis methodology and its applicability/pertinence for a fuel channel were reviewed from the CANDU fuel channel point of view. Several thermalhydraulic and numerical models for the subchannel analysis on a CANDU fuel channel were developed. The experimental data of the CANDU fuel channel were collected, analyzed and used for validation of a subchannel analysis code developed in this work. (author). 11 refs., 3 tabs., 50 figs

Developing a Blended Type Course of Introduction to Hybrid Vehicles

OpenAIRE

Na Zhu

2016-01-01

An innovative course of introduction to hybrid vehicles is developed for both associate and bachelor degree programs for engineering technology with automotive/mechanical concentration. The hybrid vehicle course content includes several topics, such as the rational of pure electric vehicle and hybrid vehicle, hybrid vehicle propulsion systems, fundamentals of motor/generator systems, fundamentals of battery and energy management system, and introduction to various configurations of hybrid veh...

Developments of fuel performance analysis codes in KEPCO NF

International Nuclear Information System (INIS)

Han, H. T.; Choi, J. M.; Jung, C. D.; Yoo, J. S.

2012-01-01

The KEPCO NF has developed fuel performance analysis and design code named as ROPER, and utility codes of XGCOL and XDNB in order to perform fuel rod design evaluation for Korean nuclear power plants. The ROPER code intends to cover full range of fuel performance evaluation. The XGCOL code is for the clad flattening evaluation and the XDNB code is for the extensive DNB propagation evaluation. In addition to these, the KEPCO NF is now in the developing stage for 3-dimensional fuel performance analysis code, named as OPER3D, using 3-dimensional FEM for the nest generation within the joint project CANDU ENERGY in order to analyze PCMI behavior and fuel performance under load following operation. Of these, the ROPER code is now in the stage of licensing activities by Korean regulatory body and the other two are almost in the final developing stage. After finishing the developing, licensing activities are to be performed. These activities are intending to acquire competitiveness, originality, vendor-free ownership of fuel performance codes in the KEPCO NF

High performance 3D neutron transport on peta scale and hybrid architectures within APOLLO3 code

International Nuclear Information System (INIS)

Jamelot, E.; Dubois, J.; Lautard, J-J.; Calvin, C.; Baudron, A-M.

2011-01-01

APOLLO3 code is a common project of CEA, AREVA and EDF for the development of a new generation system for core physics analysis. We present here the parallelization of two deterministic transport solvers of APOLLO3: MINOS, a simplified 3D transport solver on structured Cartesian and hexagonal grids, and MINARET, a transport solver based on triangular meshes on 2D and prismatic ones in 3D. We used two different techniques to accelerate MINOS: a domain decomposition method, combined with an accelerated algorithm using GPU. The domain decomposition is based on the Schwarz iterative algorithm, with Robin boundary conditions to exchange information. The Robin parameters influence the convergence and we detail how we optimized the choice of these parameters. MINARET parallelization is based on angular directions calculation using explicit message passing. Fine grain parallelization is also available for each angular direction using shared memory multithreaded acceleration. Many performance results are presented on massively parallel architectures using more than 103 cores and on hybrid architectures using some tens of GPUs. This work contributes to the HPC development in reactor physics at the CEA Nuclear Energy Division. (author)

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)

NASA space radiation transport code development consortium

International Nuclear Information System (INIS)

Townsend, L. W.

2005-01-01

Recently, NASA established a consortium involving the Univ. of Tennessee (lead institution), the Univ. of Houston, Roanoke College and various government and national laboratories, to accelerate the development of a standard set of radiation transport computer codes for NASA human exploration applications. This effort involves further improvements of the Monte Carlo codes HETC and FLUKA and the deterministic code HZETRN, including developing nuclear reaction databases necessary to extend the Monte Carlo codes to carry out heavy ion transport, and extending HZETRN to three dimensions. The improved codes will be validated by comparing predictions with measured laboratory transport data, provided by an experimental measurements consortium, and measurements in the upper atmosphere on the balloon-borne Deep Space Test Bed (DSTB). In this paper, we present an overview of the consortium members and the current status and future plans of consortium efforts to meet the research goals and objectives of this extensive undertaking. (authors)

Development of Scientific Simulation 3D Full Wave ICRF Code for Stellarators and Heating/CD Scenarios Development

Energy Technology Data Exchange (ETDEWEB)

Vdovin V.L.

2005-08-15

In this report we describe theory and 3D full wave code description for the wave excitation, propagation and absorption in 3-dimensional (3D) stellarator equilibrium high beta plasma in ion cyclotron frequency range (ICRF). This theory forms a basis for a 3D code creation, urgently needed for the ICRF heating scenarios development for the operated LHD, constructed W7-X, NCSX and projected CSX3 stellarators, as well for re evaluation of ICRF scenarios in operated tokamaks and in the ITER . The theory solves the 3D Maxwell-Vlasov antenna-plasma-conducting shell boundary value problem in the non-orthogonal flux coordinates ({Psi}, {theta}, {var_phi}), {Psi} being magnetic flux function, {theta} and {var_phi} being the poloidal and toroidal angles, respectively. All basic physics, like wave refraction, reflection and diffraction are self consistently included, along with the fundamental ion and ion minority cyclotron resonances, two ion hybrid resonance, electron Landau and TTMP absorption. Antenna reactive impedance and loading resistance are also calculated and urgently needed for an antenna -generator matching. This is accomplished in a real confining magnetic field being varying in a plasma major radius direction, in toroidal and poloidal directions, through making use of the hot dense plasma wave induced currents with account to the finite Larmor radius effects. We expand the solution in Fourier series over the toroidal ({var_phi}) and poloidal ({theta}) angles and solve resulting ordinary differential equations in a radial like {Psi}-coordinate by finite difference method. The constructed discretization scheme is divergent-free one, thus retaining the basic properties of original equations. The Fourier expansion over the angle coordinates has given to us the possibility to correctly construct the ''parallel'' wave number k{sub //}, and thereby to correctly describe the ICRF waves absorption by a hot plasma. The toroidal harmonics are tightly

Development of thermal hydraulic evaluation code for CANDU reactors

International Nuclear Information System (INIS)

Kim, Man Woong; Yu, Seon Oh; Choi, Yong Seog; Shin, Chull; Hwang, Soo Hyun

2004-02-01

To enhance the safety of operating CANDU reactors, the establishment of the safety analysis codes system for CANDU reactors is in progress. As for the development of thermal-hydraulic analysis code for CANDU system, the studies for improvement of evaluation model inside RELAP/CANDU code and the development of safety assessment methodology for GSI (Generic Safety Issues) are in progress as a part of establishment of CANDU safety assessment system. To develop the 3-D thermal-hydraulic analysis code for moderator system, the CFD models for analyzing the CANDU-6 moderator circulation are developed. One model uses a structured grid system with the porous media approach for the 380 Calandria tubes in the core region. The other uses a unstructured grid system on the real geometry of 380 Calandria tubes, so that the detailed fluid flow between the Calandria tubes can be observed. As to the development of thermal-hydraulic analysis code for containment, the study on the applicability of CONTAIN 2.0 code to a CANDU containment was conducted and a simulation of the thermal-hydraulic phenomena during the accident was performed. Besides, the model comparison of ESFs (Engineered Safety Features) inside CONTAIN 2.0 code and PRESCON code has also conducted

ICARE/CATHARE and ASTEC code development trends

International Nuclear Information System (INIS)

Chatelard, P.; Dorsselaere, J.-P. van

2000-01-01

Regarding the computer code development for simulation of LWR severe accidents, IPSN developed a two-tier approach based on detailed codes such as ICARE/CATHARE and simplified models to be assembled in the ASTEC integral code. The ICARE/CATHARE code results from the coupling between the ICARE2 code modelling the core degradation phenomena and the thermalhydraulics code CATHARE2. It allows to calculate PWR and VVER severe accident sequences in the whole RCS. The modelling of the early degradation phase can be considered as rather complete in the ICARE/CATHARE V1 mod1 version (to be released by mid-2000) whereas some models are still missing for the late phase. The main future developments (ICARE/CATHARE V2) will concern the multi-dimensional thermalhydraulics, the quenching of partially damaged cores (mechanical and chemical effects), the debris bed two-phase thermalhydraulics (including reflooding) and the corium behaviour in the lower head. The main other physical improvements should concern the behaviour of boron carbide control rods, the processes governing the core loss of geometry (transition phase) and the oxidation of relocated melts. The ASTEC (Accident Source Term Evaluation Code) integral code, commonly developed by IPSN and GRS, aims to predict an entire LWR (PWR, VVER and BWR) severe accident sequence from the initiating event through to FP release out of the containment, for source term, PSA level 2, or accident management studies. The version ASTEC VO.3 to be released by mid-2000 can be considered now as robust and fast-running enough (between 2 and 12 hours for a one day accident) and allows to perform, with a containment multi-compartment configuration, any scenario accident study accounting for the main safety systems and operator procedures (spray, recombiner, etc.). The next version ASTEC V1, to be released beginning of 2002, will include the frontend simulation and improve modelling of in-vessel core degradation. A large validation activity will

Fatigue of hybrid glass/carbon composites: 3D computational studies

DEFF Research Database (Denmark)

Dai, Gaoming; Mishnaevsky, Leon

2014-01-01

3D computational simulations of fatigue of hybrid carbon/glass fiber reinforced composites is carried out using X-FEM and multifiber unit cell models. A new software code for the automatic generation of unit cell multifiber models of composites with randomly misaligned fibers of various properties...... and geometrical parameters is developed. With the use of this program code and the X-FEM method, systematic investigations of the effect of microstructure of hybrid composites (fraction of carbon versus glass fibers, misalignment, and interface strength) and the loading conditions (tensile versus compression...... cyclic loading effects) on fatigue behavior of the materials are carried out. It was demonstrated that the higher fraction of carbon fibers in hybrid composites is beneficial for the fatigue lifetime of the composites under tension-tension cyclic loading, but might have negative effect on the lifetime...

Village power hybrid systems development in the United States

Energy Technology Data Exchange (ETDEWEB)

Flowers, L.; Green, J. [National Renewable Energy Lab., Golden, CO (United States); Bergey, M. [Bergey Windpower Co., Norman, OK (United States); Lilley, A. [Westinghouse Electric Corp., Pittsburgh, PA (United States); Mott, L. [Northern Power Systems, Moretown, VT (United States)

1994-11-01

The energy demand in developing countries is growing at a rate seven times that of the OECD countries, even though there are still 2 billion people living in developing countries without electricity. Many developing countries have social and economic development programs aimed at stemming the massive migration from the rural communities to the overcrowded, environmentally problematic, unemployment-bound urban centers. To address the issue of providing social, educational, health, and economic benefits to the rural communities of the developing world, a number of government and nongovernment agencies are sponsoring pilot programs to install and evaluate renewable energy systems as alternatives to line extension, diesels, kerosene, and batteries. The use of renewables in remote villages has yielded mixed results over the last 20 years. However, recently, photovoltaics, small wind turbines, and microhydro system shave gained increasing recognition as reliable, cost-effective alternatives to grid extension and diesel gensets for village-electricity applications. At the same time, hybrid systems based on combinations of PV/wind/batteries/diesel gensets have proven reliable and economic for remote international telecommunications markets. With the growing emphasis on environmentally and economically sustainable development of international rural communities, the US hybrid industry is responding with the development and demonstration of hybrid systems and architectures that will directly compete with conventional alternatives for village electrification. Assisting the US industry in this development, the National Renewable Energy Laboratory (NREL) has embarked on a program of collaborative technology development and technical assistance in the area of hybrid systems for village power. Following a brief review of village-power hybrid systems application and design issues, this paper presents the present industry development activities of three US suppliers and the NREL.

Aeroelastic code development activities in the United States

Energy Technology Data Exchange (ETDEWEB)

Wright, A.D. [National Renewable Energy Lab., Golden, Colorado (United States)

1996-09-01

Designing wind turbines to be fatigue resistant and to have long lifetimes at minimal cost is a major goal of the federal wind program and the wind industry in the United States. To achieve this goal, we must be able to predict critical loads for a wide variety of different wind turbines operating under extreme conditions. The codes used for wind turbine dynamic analysis must be able to analyze a wide range of different wind turbine configurations as well as rapidly predict the loads due to turbulent wind inflow with a minimal set of degrees of freedom. Code development activities in the US have taken a two-pronged approach in order to satisfy both of these criteria: (1) development of a multi-purpose code which can be used to analyze a wide variety of wind turbine configurations without having to develop new equations of motion with each configuration change, and (2) development of specialized codes with minimal sets of specific degrees of freedom for analysis of two- and three-bladed horizontal axis wind turbines and calculation of machine loads due to turbulent inflow. In the first method we have adapted a commercial multi-body dynamics simulation package for wind turbine analysis. In the second approach we are developing specialized codes with limited degrees of freedom, usually specified in the modal domain. This paper will summarize progress to date in the development, validation, and application of these codes. (au) 13 refs.

Monte Carlo code development in Los Alamos

International Nuclear Information System (INIS)

Carter, L.L.; Cashwell, E.D.; Everett, C.J.; Forest, C.A.; Schrandt, R.G.; Taylor, W.M.; Thompson, W.L.; Turner, G.D.

1974-01-01

The present status of Monte Carlo code development at Los Alamos Scientific Laboratory is discussed. A brief summary is given of several of the most important neutron, photon, and electron transport codes. 17 references. (U.S.)

Neutron transport-burnup code MCORGS and its application in fusion fission hybrid blanket conceptual research

Science.gov (United States)

Shi, Xue-Ming; Peng, Xian-Jue

2016-09-01

Fusion science and technology has made progress in the last decades. However, commercialization of fusion reactors still faces challenges relating to higher fusion energy gain, irradiation-resistant material, and tritium self-sufficiency. Fusion Fission Hybrid Reactors (FFHR) can be introduced to accelerate the early application of fusion energy. Traditionally, FFHRs have been classified as either breeders or transmuters. Both need partition of plutonium from spent fuel, which will pose nuclear proliferation risks. A conceptual design of a Fusion Fission Hybrid Reactor for Energy (FFHR-E), which can make full use of natural uranium with lower nuclear proliferation risk, is presented. The fusion core parameters are similar to those of the International Thermonuclear Experimental Reactor. An alloy of natural uranium and zirconium is adopted in the fission blanket, which is cooled by light water. In order to model blanket burnup problems, a linkage code MCORGS, which couples MCNP4B and ORIGEN-S, is developed and validated through several typical benchmarks. The average blanket energy Multiplication and Tritium Breeding Ratio can be maintained at 10 and 1.15 respectively over tens of years of continuous irradiation. If simple reprocessing without separation of plutonium from uranium is adopted every few years, FFHR-E can achieve better neutronic performance. MCORGS has also been used to analyze the ultra-deep burnup model of Laser Inertial Confinement Fusion Fission Energy (LIFE) from LLNL, and a new blanket design that uses Pb instead of Be as the neutron multiplier is proposed. In addition, MCORGS has been used to simulate the fluid transmuter model of the In-Zinerater from Sandia. A brief comparison of LIFE, In-Zinerater, and FFHR-E will be given.

Developing an Australian code of construction ethics

Directory of Open Access Journals (Sweden)

Sean Francis McCarthy

2012-05-01

Full Text Available This article looks at the increasing need to consider the role of ethics in construction. The industry, historically, has been challenged by allegations of a serious shortfall in ethical standards. Only limited attempts to date in Australia have been made to address that concern. Any ethical analysis should consider the definition of ethics and its historical development. This paper considers major historical developments in ethical thinking as well as contemporary thinking on ethics for professional sub-sets. A code could be developed specific to construction. Current methods of addressing ethics in construction and in other industries are also reviewed. This paper argues that developing a code of ethics, supported by other measures is the way forward. The author’s aim is to promote further discussion and promote the drafting of a code. This paper includes a summary of other ethical codes that may provide a starting point. The time for reform is upon us, and there is an urgent need for an independent body to take the lead, for fear of floundering and having only found ‘another debating topic’ (Uff 2006.

Design and evaluation of hybrid wind/PV/diesel power systems for Brazilian applications

Energy Technology Data Exchange (ETDEWEB)

McGowan, J.G.; Manwell, J.F.; Avelar, C. [Univ. of Massachusetts, Amherst, MA (United States); Warner, C. [National Renewable Energy Lab., Golden, CO (United States)

1996-12-31

This paper presents a summary of a study centered on the design and evaluation of hybrid wind/PV/diesel systems for remote locations in Brazil. The objective of this work was to evaluate high reliability hybrid power systems that have been designed for the lowest life cycle costs. The technical and economic analysis of the hybrid wind/PV/diesel systems was carried out using HYBRID2, a computational code developed at the University of Massachusetts in conjunction with the National Renewable Energy Laboratory (NREL). After a summary of a generalized design procedure for such systems based on the use of this code, a systematic parametric evaluation of a representative design case for a village power system in Brazil is presented. As summarized in the paper, the performance and economic effects of key design parameters are illustrated. 8 refs., 10 figs.

One dimensional neutron kinetics in the TRAC-BF1 code

International Nuclear Information System (INIS)

Weaver, W.L. III; Wagner, K.C.

1987-01-01

The TRAC-BWR code development program at the Idaho National Engineering Laboratory is developing a version of the TRAC code for the U.S. Nuclear Regulatory Commission (USNRC) to provide a best-estimate analysis capability for the simulation of postulated accidents in boiling water reactor (BWR) power systems and related experimental facilities. Recent development efforts in the TRAC-BWR program have focused on improving the computational efficiency through the incorporation of a hybrid Courant- limit-violating numerical solution scheme in the one-dimensional component models and on improving code accuracy through the development of a one-dimensional neutron kinetics model. Many other improvements have been incorporated into TRAC-BWR to improve code portability, accuracy, efficiency, and maintainability. This paper will describe the one- dimensional neutron kinetics model, the generation of the required input data for this model, and present results of the first calculations using the model

Developing a Blended Type Course of Introduction to Hybrid Vehicles

Directory of Open Access Journals (Sweden)

Na Zhu

2016-02-01

Full Text Available An innovative course of introduction to hybrid vehicles is developed for both associate and bachelor degree programs for engineering technology with automotive/mechanical concentration. The hybrid vehicle course content includes several topics, such as the rational of pure electric vehicle and hybrid vehicle, hybrid vehicle propulsion systems, fundamentals of motor/generator systems, fundamentals of battery and energy management system, and introduction to various configurations of hybrid vehicle systems available in market and under development. Hybrid vehicle technology is a new area and developed rapidly in the field of automotive and mechanical engineering. Students need not only the fundamentals and concepts from college, but also the ability to keep up with the latest technology after their graduation. Therefore, a blended course type is employed to help students have a better understanding of the fundamentals of hybrid vehicle and developing their self-studying ability. Topics in the course have three steps of learning. Firstly, on-ground lecture is given in class, where the instructor explains basic knowledge, such as principles, equations, and design rules.  In this way, the students will have enough background knowledge and be able to conduct further self-reading and research work. Secondly, students are required to go to university’s desire to learn (D2L online system and finish the online part of the topic. In the D2L system, students will find a quiz and its supporting materials. Thirdly, students come back to the on-ground lecture and discuss the quiz in groups with instructor. After the discussion, the instructor gives students a conclusion of the topic and moves forward to the next topic. A computer simulation class is also given to help student better understand the operation strategies of the hybrid vehicle systems and have a trial of design of hybrid vehicle.

Development of a 1 D hybrid HTC model using CFD simulations for the analysis of direct contact condensation as the driving force for water hammers

Energy Technology Data Exchange (ETDEWEB)

Ceuca, Christian Sabin; Macian-Juan, Rafael [Technische Univ. Muenchen (Germany). Lehrstuhl fuer Nukleartechnik

2013-03-15

A Hybrid Heat Transfer Coefficient module has been developed based on two Surface Renewal Theory models using CFD simulations. The validation of the model has been done on a meso-scale computational grid for CFD simulations and on a macro-scale computational grid for System Code analysis. The CFD simulation was performed for a stratified co-current two phase flow between saturated steam and sub-cooled water while the System Code analysis was performed for a Condensation Induced Water Hammer experiment. (orig.)

Methodology, status and plans for development and assessment of Cathare code

Energy Technology Data Exchange (ETDEWEB)

Bestion, D.; Barre, F.; Faydide, B. [CEA - Grenoble (France)

1997-07-01

This paper presents the methodology, status and plans for the development, assessment and uncertainty evaluation of the Cathare code. Cathare is a thermalhydraulic code developed by CEA (DRN), IPSN, EDF and FRAMATOME for PWR safety analysis. First, the status of the code development and assessment is presented. The general strategy used for the development and the assessment of the code is presented. Analytical experiments with separate effect tests, and component tests are used for the development and the validation of closure laws. Successive Revisions of constitutive laws are implemented in successive Versions of the code and assessed. System tests or integral tests are used to validate the general consistency of the Revision. Each delivery of a code Version + Revision is fully assessed and documented. A methodology is being developed to determine the uncertainty on all constitutive laws of the code using calculations of many analytical tests and applying the Discrete Adjoint Sensitivity Method (DASM). At last, the plans for the future developments of the code are presented. They concern the optimization of the code performance through parallel computing - the code will be used for real time full scope plant simulators - the coupling with many other codes (neutronic codes, severe accident codes), the application of the code for containment thermalhydraulics. Also, physical improvements are required in the field of low pressure transients and in the modeling for the 3-D model.

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.)

Development of ADINA-J-integral code

International Nuclear Information System (INIS)

Kurihara, Ryoichi

1988-07-01

A general purpose finite element program ADINA (Automatic Dynamic Incremental Nonlinear Analysis), which was developed by Bathe et al., was revised to be able to calculate the J- and J-integral. This report introduced the numerical method to add this capability to the code, and the evaluation of the revised ADINA-J code by using a few of examples of the J estimation model, i.e. a compact tension specimen, a center cracked panel subjected to dynamic load, and a thick shell cylinder having inner axial crack subjected to thermal load. The evaluation testified the function of the revised code. (author)

LMBFR and LWR in-core thermal-hydraulic codes: the state-of-the-art and research and development needs

International Nuclear Information System (INIS)

Khan, E.U.; Coomes, E.P.; Rowe, D.S.; Trent, D.S.

1981-04-01

A review of analytical design methods used for predicting reactor core flow and temperature distributions is presented with emphasis on LMFBR's. The paper also briefly describes and contrasts the methods used for LWR's. These methods are global analysis, subchannel analysis, distributed parameter, and hybrid analysis. The evolution of the local and subchannel analysis methods is presented. Data used for code validation are also presented. Current research and development needs are identified and discussed. Areas identified for future research and development include methods and expermental data for analysis of distorted bundles and natural convection. Methods that have been developed for predicting the safety performance of LMFBR's and LWR's are not within the scope of this paper

Development and application of best-estimate LWR safety analysis codes

International Nuclear Information System (INIS)

Reocreux, M.

1997-01-01

This paper is a review of the status and the future orientations of the development and application of best estimate LWR safety analysis codes. The present status of these codes exhibits a large success and almost a complete fulfillment of the objectives which were assigned in the 70s. The applications of Best Estimate codes are numerous and cover a large variety of safety questions. However these applications raised a number of problems. The first ones concern the need to have a better control of the quality of the results. This means requirements on code assessment and on uncertainties evaluation. The second ones concern needs for code development and specifically regarding physical models, numerics, coupling with other codes and programming. The analysis of the orientations for code developments and applications in the next years, shows that some developments should be made without delay in order to solve today questions whereas some others are more long term and should be tested for example in some pilot programmes before being eventually applied in main code development. Each of these development programmes are analyzed in the paper by detailing their main content and their possible interest. (author)

Development of codes for physical calculations of WWER

International Nuclear Information System (INIS)

Novikov, A.N.

2000-01-01

A package of codes for physical calculations of WWER reactors, used at the RRC 'Kurchatov Institute' is discussed including the purpose of these codes, approximations used, degree of data verification, possibilities of automation of calculations and presentation of results, trends of further development of the codes. (Authors)

SCDAP/RELAP5 code development and assessment

International Nuclear Information System (INIS)

Allison, C.M.; Hohorst, J.K.

1996-01-01

The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The current version of the code is SCDAP/RELAP5/MOD3.1e. Although MOD3.1e contains a number of significant improvements since the initial version of MOD3.1 was released, new models to treat the behavior of the fuel and cladding during reflood have had the most dramatic impact on the code's calculations. This paper provides a brief description of the new reflood models, presents highlights of the assessment of the current version of MOD3.1, and discusses future SCDAP/RELAP5/MOD3.2 model development activities

AREVA Developments for an Efficient and Reliable use of Monte Carlo codes for Radiation Transport Applications

Science.gov (United States)

Chapoutier, Nicolas; Mollier, François; Nolin, Guillaume; Culioli, Matthieu; Mace, Jean-Reynald

2017-09-01

In the context of the rising of Monte Carlo transport calculations for any kind of application, AREVA recently improved its suite of engineering tools in order to produce efficient Monte Carlo workflow. Monte Carlo codes, such as MCNP or TRIPOLI, are recognized as reference codes to deal with a large range of radiation transport problems. However the inherent drawbacks of theses codes - laboring input file creation and long computation time - contrast with the maturity of the treatment of the physical phenomena. The goals of the recent AREVA developments were to reach similar efficiency as other mature engineering sciences such as finite elements analyses (e.g. structural or fluid dynamics). Among the main objectives, the creation of a graphical user interface offering CAD tools for geometry creation and other graphical features dedicated to the radiation field (source definition, tally definition) has been reached. The computations times are drastically reduced compared to few years ago thanks to the use of massive parallel runs, and above all, the implementation of hybrid variance reduction technics. From now engineering teams are capable to deliver much more prompt support to any nuclear projects dealing with reactors or fuel cycle facilities from conceptual phase to decommissioning.

Specification of real-time automation systems with HybridUML; Spezifikation von Echtzeit-Automatisierungssystemen mit HybridUML

Energy Technology Data Exchange (ETDEWEB)

Berkenkoetter, K.; Bisanz, S.; Hannemann, U.; Peleska, J. [Univ. Bremen (Germany)

2004-07-01

Complex automation systems require specification formalisms supporting the description of real-time requirements with respect to both discrete and time-continuous observables. For this purpose, the authors have designed the HybridUML specification language. Discrete events, communication, and variable assignments are specified by state machines, timers, and invariant conditions. The time-continuous aspects of system behaviour are described by associating differential equations or time-dependent algebraic conditions with system states. The complexity of large systems is controlled by decomposing the specification into parallel components and hierarchical state machines. Instead of inventing a new language syntax, HybridUML is represented as a profile of the Unified Modeling Language UML 2.0. This allows to re-use the syntactic framework of well-accepted graphical UML constructs and development support provided by various UML case tools. The profile is associated with a precise language semantics linking unambiguous meaning to all HybridUML specifications. As a consequence, HybridUML specifications can be compiled into executable code which is suitable for execution in hard realtime on multi-processor computers. This serves both for the development of automation systems and for specification-based testing in real-time. This paper contains an introduction to HybridUML which is illustrated by an example from the field of automated train control. (orig.)

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.

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)

Process development and tooling design for intrinsic hybrid composites

Science.gov (United States)

Riemer, M.; Müller, R.; Drossel, W. G.; Landgrebe, D.

2017-09-01

Hybrid parts, which combine the advantages of different material classes, are moving into the focus of lightweight applications. This development is amplified by their high potential for usage in the field of crash relevant structures. By the current state of the art, hybrid parts are mainly made in separate, subsequent forming and joining processes. By using the concept of an intrinsic hybrid, the shaping of the part and the joining of the different materials are performed in a single process step for shortening the overall processing time and thereby the manufacturing costs. The investigated hybrid part is made from continuous fibre reinforced plastic (FRP), in which a metallic reinforcement structure is integrated. The connection between these layered components is realized by a combination of adhesive bonding and a geometrical form fit. The form fit elements are intrinsically generated during the forming process. This contribution regards the development of the forming process and the design of the forming tool for the single step production of a hybrid part. To this end a forming tool, which combines the thermo-forming and the metal forming process, is developed. The main challenge by designing the tool is the temperature management of the tool elements for the variothermal forming process. The process parameters are determined in basic tests and finite element (FE) simulation studies. On the basis of these investigations a control concept for the steering of the motion axes and the tool temperature is developed. Forming tests are carried out with the developed tool and the manufactured parts are analysed by computer assisted tomography (CT) scans.

Development of the point-depletion code DEPTH

International Nuclear Information System (INIS)

She, Ding; Wang, Kan; Yu, Ganglin

2013-01-01

Highlights: ► The DEPTH code has been developed for the large-scale depletion system. ► DEPTH uses the data library which is convenient to couple with MC codes. ► TTA and matrix exponential methods are implemented and compared. ► DEPTH is able to calculate integral quantities based on the matrix inverse. ► Code-to-code comparisons prove the accuracy and efficiency of DEPTH. -- Abstract: The burnup analysis is an important aspect in reactor physics, which is generally done by coupling of transport calculations and point-depletion calculations. DEPTH is a newly-developed point-depletion code of handling large burnup depletion systems and detailed depletion chains. For better coupling with Monte Carlo transport codes, DEPTH uses data libraries based on the combination of ORIGEN-2 and ORIGEN-S and allows users to assign problem-dependent libraries for each depletion step. DEPTH implements various algorithms of treating the stiff depletion systems, including the Transmutation trajectory analysis (TTA), the Chebyshev Rational Approximation Method (CRAM), the Quadrature-based Rational Approximation Method (QRAM) and the Laguerre Polynomial Approximation Method (LPAM). Three different modes are supported by DEPTH to execute the decay, constant flux and constant power calculations. In addition to obtaining the instantaneous quantities of the radioactivity, decay heats and reaction rates, DEPTH is able to calculate the integral quantities by a time-integrated solver. Through calculations compared with ORIGEN-2, the validity of DEPTH in point-depletion calculations is proved. The accuracy and efficiency of depletion algorithms are also discussed. In addition, an actual pin-cell burnup case is calculated to illustrate the DEPTH code performance in coupling with the RMC Monte Carlo code

Algorithms and computer codes for atomic and molecular quantum scattering theory

International Nuclear Information System (INIS)

Thomas, L.

1979-01-01

This workshop has succeeded in bringing up 11 different coupled equation codes on the NRCC computer, testing them against a set of 24 different test problems and making them available to the user community. These codes span a wide variety of methodologies, and factors of up to 300 were observed in the spread of computer times on specific problems. A very effective method was devised for examining the performance of the individual codes in the different regions of the integration range. Many of the strengths and weaknesses of the codes have been identified. Based on these observations, a hybrid code has been developed which is significantly superior to any single code tested. Thus, not only have the original goals been fully met, the workshop has resulted directly in an advancement of the field. All of the computer programs except VIVS are available upon request from the NRCC. Since an improved version of VIVS is contained in the hybrid program, VIVAS, it was not made available for distribution. The individual program LOGD is, however, available. In addition, programs which compute the potential energy matrices of the test problems are also available. The software library names for Tests 1, 2 and 4 are HEH2, LICO, and EN2, respectively

Challenges on innovations of newly-developed safety analysis codes

International Nuclear Information System (INIS)

Yang, Yanhua; Zhang, Hao

2016-01-01

With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.

Challenges on innovations of newly-developed safety analysis codes

Energy Technology Data Exchange (ETDEWEB)

Yang, Yanhua [Shanghai Jiao Tong Univ. (China). School of Nuclear Science and Engineering; Zhang, Hao [State Nuclear Power Software Development Center, Beijing (China). Beijing Future Science and Technology City

2016-05-15

With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.

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

HELIAS module development for systems codes

Energy Technology Data Exchange (ETDEWEB)

Warmer, F., E-mail: Felix.Warmer@ipp.mpg.de; Beidler, C.D.; Dinklage, A.; Egorov, K.; Feng, Y.; Geiger, J.; Schauer, F.; Turkin, Y.; Wolf, R.; Xanthopoulos, P.

2015-02-15

In order to study and design next-step fusion devices such as DEMO, comprehensive systems codes are commonly employed. In this work HELIAS-specific models are proposed which are designed to be compatible with systems codes. The subsequently developed models include: a geometry model based on Fourier coefficients which can represent the complex 3-D plasma shape, a basic island divertor model which assumes diffusive cross-field transport and high radiation at the X-point, and a coil model which combines scaling aspects based on the HELIAS 5-B reactor design in combination with analytic inductance and field calculations. In addition, stellarator-specific plasma transport is discussed. A strategy is proposed which employs a predictive confinement time scaling derived from 1-D neoclassical and 3-D turbulence simulations. This paper reports on the progress of the development of the stellarator-specific models while an implementation and verification study within an existing systems code will be presented in a separate work. This approach is investigated to ultimately allow one to conduct stellarator system studies, develop design points of HELIAS burning plasma devices, and to facilitate a direct comparison between tokamak and stellarator DEMO and power plant designs.

Frontier battery development for hybrid vehicles

Directory of Open Access Journals (Sweden)

Lewis Heather

2012-04-01

Full Text Available Abstract Background Interest in hybrid-electric vehicles (HEVs has recently spiked, partly due to an increasingly negative view toward the U.S. foreign oil dependency and environmental concerns. Though HEVs are becoming more common, they have a significant price premium over gasoline-powered vehicles. One of the primary drivers of this “hybrid premium” is the cost of the vehicles’ batteries. This paper focuses on these batteries used in hybrid vehicles, examines the types of batteries used for transportation applications and addresses some of the technological, environmental and political drivers in battery development and the deployment of HEVs. Methods This paper examines the claim, often voiced by HEV proponents, that by taking into account savings on gasoline and vehicle maintenance, hybrid cars are cheaper than traditional gasoline cars. This is done by a quantitative benefit-cost analysis, in addition to qualitative benefit-cost analysis from political, technological and environmental perspectives. Results The quantitative benefit-cost analysis shows that, taking account of all costs for the life of the vehicle, hybrid cars are in fact more expensive than gasoline-powered vehicles; however, after five years, HEVs will break even with gasoline cars. Conclusions Our results show that it is likely that after 5 years, using hybrid vehicles should be cheaper in effect and yield a positive net benefit to society. There are a number of externalities that could significantly impact the total social cost of the car. These externalities can be divided into four categories: environmental, industrial, R&D and political. Despite short-term implications and hurdles, increased HEV usage forecasts a generally favorable long-term net benefit to society. Most notably, increasing HEV usage could decrease greenhouse gas emissions, while also decreasing U.S. dependence on foreign oil.

Frontier battery development for hybrid vehicles.

Science.gov (United States)

Lewis, Heather; Park, Haram; Paolini, Maion

2012-04-23

Interest in hybrid-electric vehicles (HEVs) has recently spiked, partly due to an increasingly negative view toward the U.S. foreign oil dependency and environmental concerns. Though HEVs are becoming more common, they have a significant price premium over gasoline-powered vehicles. One of the primary drivers of this "hybrid premium" is the cost of the vehicles' batteries. This paper focuses on these batteries used in hybrid vehicles, examines the types of batteries used for transportation applications and addresses some of the technological, environmental and political drivers in battery development and the deployment of HEVs. This paper examines the claim, often voiced by HEV proponents, that by taking into account savings on gasoline and vehicle maintenance, hybrid cars are cheaper than traditional gasoline cars. This is done by a quantitative benefit-cost analysis, in addition to qualitative benefit-cost analysis from political, technological and environmental perspectives. The quantitative benefit-cost analysis shows that, taking account of all costs for the life of the vehicle, hybrid cars are in fact more expensive than gasoline-powered vehicles; however, after five years, HEVs will break even with gasoline cars. Our results show that it is likely that after 5 years, using hybrid vehicles should be cheaper in effect and yield a positive net benefit to society. There are a number of externalities that could significantly impact the total social cost of the car. These externalities can be divided into four categories: environmental, industrial, R&D and political. Despite short-term implications and hurdles, increased HEV usage forecasts a generally favorable long-term net benefit to society. Most notably, increasing HEV usage could decrease greenhouse gas emissions, while also decreasing U.S. dependence on foreign oil.

Frontier battery development for hybrid vehicles

Science.gov (United States)

2012-01-01

Background Interest in hybrid-electric vehicles (HEVs) has recently spiked, partly due to an increasingly negative view toward the U.S. foreign oil dependency and environmental concerns. Though HEVs are becoming more common, they have a significant price premium over gasoline-powered vehicles. One of the primary drivers of this “hybrid premium” is the cost of the vehicles’ batteries. This paper focuses on these batteries used in hybrid vehicles, examines the types of batteries used for transportation applications and addresses some of the technological, environmental and political drivers in battery development and the deployment of HEVs. Methods This paper examines the claim, often voiced by HEV proponents, that by taking into account savings on gasoline and vehicle maintenance, hybrid cars are cheaper than traditional gasoline cars. This is done by a quantitative benefit-cost analysis, in addition to qualitative benefit-cost analysis from political, technological and environmental perspectives. Results The quantitative benefit-cost analysis shows that, taking account of all costs for the life of the vehicle, hybrid cars are in fact more expensive than gasoline-powered vehicles; however, after five years, HEVs will break even with gasoline cars. Conclusions Our results show that it is likely that after 5 years, using hybrid vehicles should be cheaper in effect and yield a positive net benefit to society. There are a number of externalities that could significantly impact the total social cost of the car. These externalities can be divided into four categories: environmental, industrial, R&D and political. Despite short-term implications and hurdles, increased HEV usage forecasts a generally favorable long-term net benefit to society. Most notably, increasing HEV usage could decrease greenhouse gas emissions, while also decreasing U.S. dependence on foreign oil. PMID:22540987

Development and application of a deterministic-realistic hybrid methodology for LOCA licensing analysis

International Nuclear Information System (INIS)

Liang, Thomas K.S.; Chou, Ling-Yao; Zhang, Zhongwei; Hsueh, Hsiang-Yu; Lee, Min

2011-01-01

Highlights: → A new LOCA licensing methodology (DRHM, deterministic-realistic hybrid methodology) was developed. → DRHM involves conservative Appendix K physical models and statistical treatment of plant status uncertainties. → DRHM can generate 50-100 K PCT margin as compared to a traditional Appendix K methodology. - Abstract: It is well recognized that a realistic LOCA analysis with uncertainty quantification can generate greater safety margin as compared with classical conservative LOCA analysis using Appendix K evaluation models. The associated margin can be more than 200 K. To quantify uncertainty in BELOCA analysis, generally there are two kinds of uncertainties required to be identified and quantified, which involve model uncertainties and plant status uncertainties. Particularly, it will take huge effort to systematically quantify individual model uncertainty of a best estimate LOCA code, such as RELAP5 and TRAC. Instead of applying a full ranged BELOCA methodology to cover both model and plant status uncertainties, a deterministic-realistic hybrid methodology (DRHM) was developed to support LOCA licensing analysis. Regarding the DRHM methodology, Appendix K deterministic evaluation models are adopted to ensure model conservatism, while CSAU methodology is applied to quantify the effect of plant status uncertainty on PCT calculation. Generally, DRHM methodology can generate about 80-100 K margin on PCT as compared to Appendix K bounding state LOCA analysis.

Development of Traction Drive Motors for the Toyota Hybrid System

Science.gov (United States)

Kamiya, Munehiro

Toyota Motor Corporation developed in 2005 a new hybrid system for a large SUV. This system included the new development of a high-speed traction drive motor achieving a significant increase in power weight ratio. This paper provides an overview of the hybrid system, discusses the characteristics required of a traction drive motor, and presents the technologies employed in the developed motor.

Hybrid Micro-Depletion method in the DYN3D code

Energy Technology Data Exchange (ETDEWEB)

Bilodid, Yurii [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Div. Reactor Safety

2016-07-01

A new method for accounting spectral history effects was developed and implemented in the reactor dynamics code DYN3D. Detailed nuclide content is calculated for each region of the reactor core and used to correct fuel properties. The new method demonstrates excellent results in test cases.

Constructing LDPC Codes from Loop-Free Encoding Modules

Science.gov (United States)

Divsalar, Dariush; Dolinar, Samuel; Jones, Christopher; Thorpe, Jeremy; Andrews, Kenneth

2009-01-01

A method of constructing certain low-density parity-check (LDPC) codes by use of relatively simple loop-free coding modules has been developed. The subclasses of LDPC codes to which the method applies includes accumulate-repeat-accumulate (ARA) codes, accumulate-repeat-check-accumulate codes, and the codes described in Accumulate-Repeat-Accumulate-Accumulate Codes (NPO-41305), NASA Tech Briefs, Vol. 31, No. 9 (September 2007), page 90. All of the affected codes can be characterized as serial/parallel (hybrid) concatenations of such relatively simple modules as accumulators, repetition codes, differentiators, and punctured single-parity check codes. These are error-correcting codes suitable for use in a variety of wireless data-communication systems that include noisy channels. These codes can also be characterized as hybrid turbolike codes that have projected graph or protograph representations (for example see figure); these characteristics make it possible to design high-speed iterative decoders that utilize belief-propagation algorithms. The present method comprises two related submethods for constructing LDPC codes from simple loop-free modules with circulant permutations. The first submethod is an iterative encoding method based on the erasure-decoding algorithm. The computations required by this method are well organized because they involve a parity-check matrix having a block-circulant structure. The second submethod involves the use of block-circulant generator matrices. The encoders of this method are very similar to those of recursive convolutional codes. Some encoders according to this second submethod have been implemented in a small field-programmable gate array that operates at a speed of 100 megasymbols per second. By use of density evolution (a computational- simulation technique for analyzing performances of LDPC codes), it has been shown through some examples that as the block size goes to infinity, low iterative decoding thresholds close to

Modified hybrid subcarrier/amplitude/ phase/polarization LDPC-coded modulation for 400 Gb/s optical transmission and beyond.

Science.gov (United States)

Batshon, Hussam G; Djordjevic, Ivan; Xu, Lei; Wang, Ting

2010-06-21

In this paper, we present a modified coded hybrid subcarrier/ amplitude/phase/polarization (H-SAPP) modulation scheme as a technique capable of achieving beyond 400 Gb/s single-channel transmission over optical channels. The modified H-SAPP scheme profits from the available resources in addition to geometry to increase the bandwidth efficiency of the transmission system, and so increases the aggregate rate of the system. In this report we present the modified H-SAPP scheme and focus on an example that allows 11 bits/Symbol that can achieve 440 Gb/s transmission using components of 50 Giga Symbol/s (GS/s).

Hybrid electric vehicles energy management strategies

CERN Document Server

Onori, Simona; Rizzoni, Giorgio

2016-01-01

This SpringerBrief deals with the control and optimization problem in hybrid electric vehicles. Given that there are two (or more) energy sources (i.e., battery and fuel) in hybrid vehicles, it shows the reader how to implement an energy-management strategy that decides how much of the vehicle’s power is provided by each source instant by instant. Hybrid Electric Vehicles: •introduces methods for modeling energy flow in hybrid electric vehicles; •presents a standard mathematical formulation of the optimal control problem; •discusses different optimization and control strategies for energy management, integrating the most recent research results; and •carries out an overall comparison of the different control strategies presented. Chapter by chapter, a case study is thoroughly developed, providing illustrative numerical examples that show the basic principles applied to real-world situations. In addition to the examples, simulation code is provided via a website, so that readers can work on the actua...

Development of the next generation reactor analysis code system, MARBLE

International Nuclear Information System (INIS)

Yokoyama, Kenji; Hazama, Taira; Nagaya, Yasunobu; Chiba, Go; Kugo, Teruhiko; Ishikawa, Makoto; Tatsumi, Masahiro; Hirai, Yasushi; Hyoudou, Hideaki; Numata, Kazuyuki; Iwai, Takehiko; Jin, Tomoyuki

2011-03-01

A next generation reactor analysis code system, MARBLE, has been developed. MARBLE is a successor of the fast reactor neutronics analysis code systems, JOINT-FR and SAGEP-FR (conventional systems), which were developed for so-called JUPITER standard analysis methods. MARBLE has the equivalent analysis capability to the conventional system because MARBLE can utilize sub-codes included in the conventional system without any change. On the other hand, burnup analysis functionality for power reactors is improved compared with the conventional system by introducing models on fuel exchange treatment and control rod operation and so on. In addition, MARBLE has newly developed solvers and some new features of burnup calculation by the Krylov sub-space method and nuclear design accuracy evaluation by the extended bias factor method. In the development of MARBLE, the object oriented technology was adopted from the view-point of improvement of the software quality such as flexibility, expansibility, facilitation of the verification by the modularization and assistance of co-development. And, software structure called the two-layer system consisting of scripting language and system development language was applied. As a result, MARBLE is not an independent analysis code system which simply receives input and returns output, but an assembly of components for building an analysis code system (i.e. framework). Furthermore, MARBLE provides some pre-built analysis code systems such as the fast reactor neutronics analysis code system. SCHEME, which corresponds to the conventional code and the fast reactor burnup analysis code system, ORPHEUS. (author)

Accelerator-driven transmutation reactor analysis code system (ATRAS)

Energy Technology Data Exchange (ETDEWEB)

Sasa, Toshinobu; Tsujimoto, Kazufumi; Takizuka, Takakazu; Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1999-03-01

JAERI is proceeding a design study of the hybrid type minor actinide transmutation system which mainly consist of an intense proton accelerator and a fast subcritical core. Neutronics and burnup characteristics of the accelerator-driven system is important from a view point of the maintenance of subcriticality and energy balance during the system operation. To determine those characteristics accurately, it is necessary to involve reactions at high-energy region, which are not treated on ordinary reactor analysis codes. The authors developed a code system named ATRAS to analyze the neutronics and burnup characteristics of accelerator-driven subcritical reactor systems. ATRAS has a function of burnup analysis taking account of the effect of spallation neutron source. ATRAS consists of a spallation analysis code, a neutron transport codes and a burnup analysis code. Utility programs for fuel exchange, pre-processing and post-processing are also incorporated. (author)

Hybrid model for simulation of plasma jet injection in tokamak

Science.gov (United States)

Galkin, Sergei A.; Bogatu, I. N.

2016-10-01

Hybrid kinetic model of plasma treats the ions as kinetic particles and the electrons as charge neutralizing massless fluid. The model is essentially applicable when most of the energy is concentrated in the ions rather than in the electrons, i.e. it is well suited for the high-density hyper-velocity C60 plasma jet. The hybrid model separates the slower ion time scale from the faster electron time scale, which becomes disregardable. That is why hybrid codes consistently outperform the traditional PIC codes in computational efficiency, still resolving kinetic ions effects. We discuss 2D hybrid model and code with exact energy conservation numerical algorithm and present some results of its application to simulation of C60 plasma jet penetration through tokamak-like magnetic barrier. We also examine the 3D model/code extension and its possible applications to tokamak and ionospheric plasmas. The work is supported in part by US DOE DE-SC0015776 Grant.

Development and Application of a Code for Internal Exposure (CINEX) based on the CINDY code

International Nuclear Information System (INIS)

Kravchik, T.; Duchan, N.; Sarah, R.; Gabay, Y.; Kol, R.

2004-01-01

Internal exposure to radioactive materials at the NRCN is evaluated using the CINDY (Code for Internal Dosimetry) Package. The code was developed by the Pacific Northwest Laboratory to assist the interpretation of bioassay data, provide bioassay projections and evaluate committed and calendar-year doses from intake or bioassay measurement data. It provides capabilities to calculate organ dose and effective dose equivalents using the International Commission on Radiological Protection (ICRP) 30 approach. The CINDY code operates under DOS operating system and consequently its operation needs a relatively long procedure which also includes a lot of manual typing that can lead to personal human mistakes. A new code has been developed at the NRCN, the CINEX (Code for Internal Exposure), which is an Excel application and leads to a significant reduction in calculation time (in the order of 5-10 times) and in the risk of personal human mistakes. The code uses a database containing tables which were constructed by the CINDY and contain the bioassay values predicted by the ICRP30 model after an intake of an activity unit of each isotope. Using the database, the code than calculates the appropriate intake and consequently the committed effective dose and organ dose. Calculations with the CINEX code were compared to similar calculations with the CINDY code. The discrepancies were less than 5%, which is the rounding error of the CINDY code. Attached is a table which compares parameters calculated with the CINEX and the CINDY codes (for a class Y uranium). The CINEX is now used at the NRCN to calculate occupational intakes and doses to workers with radioactive materials

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

[INVITED] Luminescent QR codes for smart labelling and sensing

Science.gov (United States)

Ramalho, João F. C. B.; António, L. C. F.; Correia, S. F. H.; Fu, L. S.; Pinho, A. S.; Brites, C. D. S.; Carlos, L. D.; André, P. S.; Ferreira, R. A. S.

2018-05-01

QR (Quick Response) codes are two-dimensional barcodes composed of special geometric patterns of black modules in a white square background that can encode different types of information with high density and robustness, correct errors and physical damages, thus keeping the stored information protected. Recently, these codes have gained increased attention as they offer a simple physical tool for quick access to Web sites for advertising and social interaction. Challenges encompass the increase of the storage capacity limit, even though they can store approximately 350 times more information than common barcodes, and encode different types of characters (e.g., numeric, alphanumeric, kanji and kana). In this work, we fabricate luminescent QR codes based on a poly(methyl methacrylate) substrate coated with organic-inorganic hybrid materials doped with trivalent terbium (Tb3+) and europium (Eu3+) ions, demonstrating the increase of storage capacity per unit area by a factor of two by using the colour multiplexing, when compared to conventional QR codes. A novel methodology to decode the multiplexed QR codes is developed based on a colour separation threshold where a decision level is calculated through a maximum-likelihood criteria to minimize the error probability of the demultiplexed modules, maximizing the foreseen total storage capacity. Moreover, the thermal dependence of the emission colour coordinates of the Eu3+/Tb3+-based hybrids enables the simultaneously QR code colour-multiplexing and may be used to sense temperature (reproducibility higher than 93%), opening new fields of applications for QR codes as smart labels for sensing.

Development of a tritium dispersion code

International Nuclear Information System (INIS)

Bell, R.P.; Davis, M.W.; Joseph, S.; Wong, K.Y.

1985-01-01

This paper describes the development and verification of a computer code designed to calculate the radiation dose to man following acute or chronic atmospheric releases of tritium gas and oxide from a point source. The Ontario Hydro Tritium Dispersion Code calculates tritium concentrations in air, soil, and vegetation and doses to man resulting from inhalation/immersion and ingestion of food, milk meat and water. The deposition of HT to soil, conversion of HT to HTO by soil enzymes and resuspension of HTO to air have been incorporated into the terrestrial compartment model and are unique features of the code. Sensitivity analysis has identified the HT deposition velocity and the equivalent water depth of the vegetation compartment as two parameters which have a strong influence on dose calculations. Tritium concentrations in vegetation and soil calculated by the code were in reasonable agreement with experimental results. The radiological significance of including the mechanisms of HT to HTO conversion and resuspension of HTO to air is illustrated

Recent developments in KTF. Code optimization and improved numerics

International Nuclear Information System (INIS)

Jimenez, Javier; Avramova, Maria; Sanchez, Victor Hugo; Ivanov, Kostadin

2012-01-01

The rapid increase of computer power in the last decade facilitated the development of high fidelity simulations in nuclear engineering allowing a more realistic and accurate optimization as well as safety assessment of reactor cores and power plants compared to the legacy codes. Thermal hydraulic subchannel codes together with time dependent neutron transport codes are the options of choice for an accurate prediction of local safety parameters. Moreover, fast running codes with the best physical models are needed for high fidelity coupled thermal hydraulic / neutron kinetic solutions. Hence at KIT, different subchannel codes such as SUBCHANFLOW and KTF are being improved, validated and coupled with different neutron kinetics solutions. KTF is a subchannel code developed for best-estimate analysis of both Pressurized Water Reactor (PWR) and BWR. It is based on the Pennsylvania State University (PSU) version of COBRA-TF (Coolant Boling in Rod Arrays Two Fluids) named CTF. In this paper, the investigations devoted to the enhancement of the code numeric and informatics structure are presented and discussed. By some examples the gain on code speed-up will be demonstrated and finally an outlook of further activities concentrated on the code improvements will be given. (orig.)

Recent developments in KTF. Code optimization and improved numerics

Energy Technology Data Exchange (ETDEWEB)

Jimenez, Javier; Avramova, Maria; Sanchez, Victor Hugo; Ivanov, Kostadin [Karlsruhe Institute of Technology (KIT) (Germany). Inst. for Neutron Physics and Reactor Technology (INR)

2012-11-01

The rapid increase of computer power in the last decade facilitated the development of high fidelity simulations in nuclear engineering allowing a more realistic and accurate optimization as well as safety assessment of reactor cores and power plants compared to the legacy codes. Thermal hydraulic subchannel codes together with time dependent neutron transport codes are the options of choice for an accurate prediction of local safety parameters. Moreover, fast running codes with the best physical models are needed for high fidelity coupled thermal hydraulic / neutron kinetic solutions. Hence at KIT, different subchannel codes such as SUBCHANFLOW and KTF are being improved, validated and coupled with different neutron kinetics solutions. KTF is a subchannel code developed for best-estimate analysis of both Pressurized Water Reactor (PWR) and BWR. It is based on the Pennsylvania State University (PSU) version of COBRA-TF (Coolant Boling in Rod Arrays Two Fluids) named CTF. In this paper, the investigations devoted to the enhancement of the code numeric and informatics structure are presented and discussed. By some examples the gain on code speed-up will be demonstrated and finally an outlook of further activities concentrated on the code improvements will be given. (orig.)

Development of code SFINEL (Spent fuel integrity evaluator)

Energy Technology Data Exchange (ETDEWEB)

Kim, Yong Soo; Min, Chin Young; Ohk, Young Kil; Yang, Yong Sik; Kim, Dong Ju; Kim, Nam Ku [Hanyang University, Seoul (Korea)

1999-01-01

SFINEL code, an integrated computer program for predicting the spent fuel rod integrity based on burn-up history and major degradation mechanisms, has been developed through this project. This code can sufficiently simulate the power history of a fuel rod during the reactor operation and estimate the degree of deterioration of spent fuel cladding using the recently-developed models on the degradation mechanisms. SFINEL code has been thoroughly benchmarked against the collected in-pile data and operating experiences: deformation and rupture, and cladding oxidation, rod internal pressure creep, then comprehensive whole degradation process. (author). 75 refs., 51 figs., 5 tabs.

Development of FBR integrity system code. Basic concept

International Nuclear Information System (INIS)

Asayama, Tai

2001-05-01

For fast breeder reactors to be commercialized, they must be more reliable, safer, and at the same, economically competitive with future light water reactors. Innovation of elevated temperature structural design standard is necessary to achieve this goal. The most powerful way is to enlarge the scope of structural integrity code to cover items other than design evaluation that has been addressed in existing codes. Items that must be newly covered are prerequisites of design, fabrication, examination, operation and maintenance, etc. This allows designers to choose the most economical combination of design variations to achieve specific reliability that is needed for a particular component. Designing components by this concept, a cost-minimum design of a whole plant can be realized. By determining the reliability that must be achieved for a component by risk technologies, further economical improvement can be expected by avoiding excessive quality. Recognizing the necessity for the codes based on the new concept, the development of 'FBR integrity system code' began in 2000. Research and development will last 10 years. For this development, the basic logistics and system as well as technologies that materialize the concept are necessary. Original logistics and system must be developed, because no existing researches are available in and out of Japan. This reports presents the results of the work done in the first year regarding the basic idea, methodology, and structure of the code. (author)

Development of authentication code for multi-access optical code division multiplexing based quantum key distribution

Science.gov (United States)

Taiwo, Ambali; Alnassar, Ghusoon; Bakar, M. H. Abu; Khir, M. F. Abdul; Mahdi, Mohd Adzir; Mokhtar, M.

2018-05-01

One-weight authentication code for multi-user quantum key distribution (QKD) is proposed. The code is developed for Optical Code Division Multiplexing (OCDMA) based QKD network. A unique address assigned to individual user, coupled with degrading probability of predicting the source of the qubit transmitted in the channel offer excellent secure mechanism against any form of channel attack on OCDMA based QKD network. Flexibility in design as well as ease of modifying the number of users are equally exceptional quality presented by the code in contrast to Optical Orthogonal Code (OOC) earlier implemented for the same purpose. The code was successfully applied to eight simultaneous users at effective key rate of 32 bps over 27 km transmission distance.

Development of HTGR plant dynamics simulation code

International Nuclear Information System (INIS)

Ohashi, Kazutaka; Tazawa, Yujiro; Mitake, Susumu; Suzuki, Katsuo.

1987-01-01

Plant dynamics simulation analysis plays an important role in the design work of nuclear power plant especially in the plant safety analysis, control system analysis, and transient condition analysis. The authors have developed the plant dynamics simulation code named VESPER, which is applicable to the design work of High Temperature Engineering Test Reactor, and have been improving the code corresponding to the design changes made in the subsequent design works. This paper describes the outline of VESPER code and shows its sample calculation results selected from the recent design work. (author)

The possibility of developing hybrid PV/T solar system

Science.gov (United States)

Dobrnjac, M.; Zivkovic, P.; Babic, V.

2017-05-01

An alternative and cost-effective solution to developing integrated PV system is to use hybrid photovoltaic/thermal (PV/T) solar system. The temperature of PV modules increases due to the absorbed solar radiation that is not converted into electricity, causing a decrease in their efficiency. In hybrid PV/T solar systems the reduction of PV module temperature can be combined with a useful fluid heating. In this paper we present the possibility of developing a new hybrid PV/T solar system. Hybrid PV/T system can provide electrical and thermal energy, thus achieving a higher energy conversion rate of the absorbed solar radiation. We developed PV/T prototype consisted of commercial PV module and thermal panel with our original solution of aluminium absorber with special geometric shapes. The main advantages of our combined PV/T system are: removing of heat from the PV panel; extending the lifetime of photovoltaic cells; excess of the removing heat from PV part is used to heat the fluid in the thermal part of the panel; the possibility of using on the roof and facade constructions because less weight.

Integrated code development for studying laser driven plasmas

Energy Technology Data Exchange (ETDEWEB)

Takabe, Hideaki; Nagatomo, Hideo; Sunahara, Atsusi; Ohnishi, Naofumi; Naruo, Syuji; Mima, Kunioki [Osaka Univ., Suita (Japan). Inst. of Laser Engineering

1998-03-01

Present status and plan for developing an integrated implosion code are briefly explained by focusing on motivation, numerical scheme and issues to be developed more. Highly nonlinear stage of Rayleigh-Taylor instability of ablation front by laser irradiation has been simulated so as to be compared with model experiments. Improvement in transport and rezoning/remapping algorithms in ILESTA code is described. (author)

Non-coding RNAs and plant male sterility: current knowledge and future prospects.

Science.gov (United States)

Mishra, Ankita; Bohra, Abhishek

2018-02-01

Latest outcomes assign functional role to non-coding (nc) RNA molecules in regulatory networks that confer male sterility to plants. Male sterility in plants offers great opportunity for improving crop performance through application of hybrid technology. In this respect, cytoplasmic male sterility (CMS) and sterility induced by photoperiod (PGMS)/temperature (TGMS) have greatly facilitated development of high-yielding hybrids in crops. Participation of non-coding (nc) RNA molecules in plant reproductive development is increasingly becoming evident. Recent breakthroughs in rice definitively associate ncRNAs with PGMS and TGMS. In case of CMS, the exact mechanism through which the mitochondrial ORFs exert influence on the development of male gametophyte remains obscure in several crops. High-throughput sequencing has enabled genome-wide discovery and validation of these regulatory molecules and their target genes, describing their potential roles performed in relation to CMS. Discovery of ncRNA localized in plant mtDNA with its possible implication in CMS induction is intriguing in this respect. Still, conclusive evidences linking ncRNA with CMS phenotypes are currently unavailable, demanding complementing genetic approaches like transgenics to substantiate the preliminary findings. Here, we review the recent literature on the contribution of ncRNAs in conferring male sterility to plants, with an emphasis on microRNAs. Also, we present a perspective on improved understanding about ncRNA-mediated regulatory pathways that control male sterility in plants. A refined understanding of plant male sterility would strengthen crop hybrid industry to deliver hybrids with improved performance.

Coupled geochemical and solute transport code development

International Nuclear Information System (INIS)

Morrey, J.R.; Hostetler, C.J.

1985-01-01

A number of coupled geochemical hydrologic codes have been reported in the literature. Some of these codes have directly coupled the source-sink term to the solute transport equation. The current consensus seems to be that directly coupling hydrologic transport and chemical models through a series of interdependent differential equations is not feasible for multicomponent problems with complex geochemical processes (e.g., precipitation/dissolution reactions). A two-step process appears to be the required method of coupling codes for problems where a large suite of chemical reactions must be monitored. Two-step structure requires that the source-sink term in the transport equation is supplied by a geochemical code rather than by an analytical expression. We have developed a one-dimensional two-step coupled model designed to calculate relatively complex geochemical equilibria (CTM1D). Our geochemical module implements a Newton-Raphson algorithm to solve heterogeneous geochemical equilibria, involving up to 40 chemical components and 400 aqueous species. The geochemical module was designed to be efficient and compact. A revised version of the MINTEQ Code is used as a parent geochemical code

Development and Meiosis of Three Interspecific Hybrids with Cultivated Barley (Hordeum vulgare L.)

DEFF Research Database (Denmark)

Von Bothmer, R.; Flink, J.; Linde-Laursen, Ib

1986-01-01

The development and meiosis of three interspecific hybrids between cultivated barley (Hordeum vulgare L.) and H. secalinum Schreb., H. tetraploidum Covas, and H. parodii Covas, respectively, were studied. All three hybrid combinations developed very slowly vegetatively. Meiosis of the hybrids...

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

AREVA Developments for an Efficient and Reliable use of Monte Carlo codes for Radiation Transport Applications

Directory of Open Access Journals (Sweden)

Chapoutier Nicolas

2017-01-01

Full Text Available In the context of the rising of Monte Carlo transport calculations for any kind of application, AREVA recently improved its suite of engineering tools in order to produce efficient Monte Carlo workflow. Monte Carlo codes, such as MCNP or TRIPOLI, are recognized as reference codes to deal with a large range of radiation transport problems. However the inherent drawbacks of theses codes - laboring input file creation and long computation time - contrast with the maturity of the treatment of the physical phenomena. The goals of the recent AREVA developments were to reach similar efficiency as other mature engineering sciences such as finite elements analyses (e.g. structural or fluid dynamics. Among the main objectives, the creation of a graphical user interface offering CAD tools for geometry creation and other graphical features dedicated to the radiation field (source definition, tally definition has been reached. The computations times are drastically reduced compared to few years ago thanks to the use of massive parallel runs, and above all, the implementation of hybrid variance reduction technics. From now engineering teams are capable to deliver much more prompt support to any nuclear projects dealing with reactors or fuel cycle facilities from conceptual phase to decommissioning.

The development of code benchmarks

International Nuclear Information System (INIS)

Glass, R.E.

1986-01-01

Sandia National Laboratories has undertaken a code benchmarking effort to define a series of cask-like problems having both numerical solutions and experimental data. The development of the benchmarks includes: (1) model problem definition, (2) code intercomparison, and (3) experimental verification. The first two steps are complete and a series of experiments are planned. The experiments will examine the elastic/plastic behavior of cylinders for both the end and side impacts resulting from a nine meter drop. The cylinders will be made from stainless steel and aluminum to give a range of plastic deformations. This paper presents the results of analyses simulating the model's behavior using materials properties for stainless steel and aluminum

Development of Evaluation Code for MUF Uncertainty

International Nuclear Information System (INIS)

Won, Byung Hee; Han, Bo Young; Shin, Hee Sung; Ahn, Seong-Kyu; Park, Geun-Il; Park, Se Hwan

2015-01-01

Material Unaccounted For (MUF) is the material balance evaluated by measured nuclear material in a Material Balance Area (MBA). Assuming perfect measurements and no diversion from a facility, one can expect a zero MUF. However, non-zero MUF is always occurred because of measurement uncertainty even though the facility is under normal operation condition. Furthermore, there are many measurements using different equipment at various Key Measurement Points (KMPs), and the MUF uncertainty is affected by errors of those measurements. Evaluating MUF uncertainty is essentially required to develop safeguards system including nuclear measurement system in pyroprocessing, which is being developed for reducing radioactive waste from spent fuel in Korea Atomic Energy Research Institute (KAERI). The evaluation code for analyzing MUF uncertainty has been developed and it was verified using sample problem from the IAEA reference. MUF uncertainty can be simply and quickly calculated by using this evaluation code which is made based on graphical user interface for user friendly. It is also expected that the code will make the sensitivity analysis on the MUF uncertainty for the various safeguards systems easy and more systematic. It is suitable for users who want to evaluate the conventional safeguards system as well as to develop a new system for developing facilities

Development of Evaluation Code for MUF Uncertainty

Energy Technology Data Exchange (ETDEWEB)

Won, Byung Hee; Han, Bo Young; Shin, Hee Sung; Ahn, Seong-Kyu; Park, Geun-Il; Park, Se Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

Material Unaccounted For (MUF) is the material balance evaluated by measured nuclear material in a Material Balance Area (MBA). Assuming perfect measurements and no diversion from a facility, one can expect a zero MUF. However, non-zero MUF is always occurred because of measurement uncertainty even though the facility is under normal operation condition. Furthermore, there are many measurements using different equipment at various Key Measurement Points (KMPs), and the MUF uncertainty is affected by errors of those measurements. Evaluating MUF uncertainty is essentially required to develop safeguards system including nuclear measurement system in pyroprocessing, which is being developed for reducing radioactive waste from spent fuel in Korea Atomic Energy Research Institute (KAERI). The evaluation code for analyzing MUF uncertainty has been developed and it was verified using sample problem from the IAEA reference. MUF uncertainty can be simply and quickly calculated by using this evaluation code which is made based on graphical user interface for user friendly. It is also expected that the code will make the sensitivity analysis on the MUF uncertainty for the various safeguards systems easy and more systematic. It is suitable for users who want to evaluate the conventional safeguards system as well as to develop a new system for developing facilities.

Hybrid information privacy system: integration of chaotic neural network and RSA coding

Science.gov (United States)

Hsu, Ming-Kai; Willey, Jeff; Lee, Ting N.; Szu, Harold H.

2005-03-01

Electronic mails are adopted worldwide; most are easily hacked by hackers. In this paper, we purposed a free, fast and convenient hybrid privacy system to protect email communication. The privacy system is implemented by combining private security RSA algorithm with specific chaos neural network encryption process. The receiver can decrypt received email as long as it can reproduce the specified chaos neural network series, so called spatial-temporal keys. The chaotic typing and initial seed value of chaos neural network series, encrypted by the RSA algorithm, can reproduce spatial-temporal keys. The encrypted chaotic typing and initial seed value are hidden in watermark mixed nonlinearly with message media, wrapped with convolution error correction codes for wireless 3rd generation cellular phones. The message media can be an arbitrary image. The pattern noise has to be considered during transmission and it could affect/change the spatial-temporal keys. Since any change/modification on chaotic typing or initial seed value of chaos neural network series is not acceptable, the RSA codec system must be robust and fault-tolerant via wireless channel. The robust and fault-tolerant properties of chaos neural networks (CNN) were proved by a field theory of Associative Memory by Szu in 1997. The 1-D chaos generating nodes from the logistic map having arbitrarily negative slope a = p/q generating the N-shaped sigmoid was given first by Szu in 1992. In this paper, we simulated the robust and fault-tolerance properties of CNN under additive noise and pattern noise. We also implement a private version of RSA coding and chaos encryption process on messages.

The Development of the World Anti-Doping Code.

Science.gov (United States)

Young, Richard

2017-01-01

This chapter addresses both the development and substance of the World Anti-Doping Code, which came into effect in 2003, as well as the subsequent Code amendments, which came into effect in 2009 and 2015. Through an extensive process of stakeholder input and collaboration, the World Anti-Doping Code has transformed the hodgepodge of inconsistent and competing pre-2003 anti-doping rules into a harmonized and effective approach to anti-doping. The Code, as amended, is now widely recognized worldwide as the gold standard in anti-doping. The World Anti-Doping Code originally went into effect on January 1, 2004. The first amendments to the Code went into effect on January 1, 2009, and the second amendments on January 1, 2015. The Code and the related international standards are the product of a long and collaborative process designed to make the fight against doping more effective through the adoption and implementation of worldwide harmonized rules and best practices. © 2017 S. Karger AG, Basel.

Experimental neutronic science and instrumentation: from hybrid reactors to fourth generation reactors

International Nuclear Information System (INIS)

Jammes, Ch.

2010-07-01

After an overview of his academic career and scientific and research activities, the author proposes a rather detailed synthesis and overview of his scientific activities in the fields of cross sections and Doppler effect (development and validation of a code), on the MUSE-4 hybrid reactor (experiments, static and dynamic measurements), on the TRADE hybrid reactor (experimental means, sub-critical reactivity measurement), on the RACE hybrid reactor (experimental results, modelling and interpretation), and on neutron detection (design and modelling of fission chamber, on-line measurement of the fast flow). The next part gives an overview of some research programs (neutron monitoring in sodium-cool fast reactors, research and development on fission chambers, improvement of effective delayed neutron measurements)

Development of probabilistic fracture mechanics code PASCAL and user's manual

Energy Technology Data Exchange (ETDEWEB)

Shibata, Katsuyuki; Onizawa, Kunio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Li, Yinsheng; Kato, Daisuke [Fuji Research Institute Corporation, Tokyo (Japan)

2001-03-01

As a part of the aging and structural integrity research for LWR components, a new PFM (Probabilistic Fracture Mechanics) code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed since FY1996. This code evaluates the failure probability of an aged reactor pressure vessel subjected to transient loading such as PTS (Pressurized Thermal Shock). The development of the code has been aimed to improve the accuracy and reliability of analysis by introducing new analysis methodologies and algorithms considering the recent development in the fracture mechanics methodologies and computer performance. The code has some new functions in optimized sampling and cell dividing procedure in stratified Monte Carlo simulation, elastic-plastic fracture criterion of R6 method, extension analysis models in semi-elliptical crack, evaluation of effect of thermal annealing and etc. In addition, an input data generator of temperature and stress distribution time histories was also prepared in the code. Functions and performance of the code have been confirmed based on the verification analyses and some case studies on the influence parameters. The present phase of the development will be completed in FY2000. Thus this report provides the user's manual and theoretical background of the code. (author)

Status and modeling improvements of hybrid wind/PV/diesel power systems for Brazilian applications

Energy Technology Data Exchange (ETDEWEB)

McGowan, J.G.; Manwell, J.F.; Avelar, C. [Univ. of Massachusetts, Amherst, MA (United States); Taylor, R. [National Renewable Energy Lab., Golden, CO (United States)

1997-12-31

This paper present a summary of the ongoing work on the modeling and system design of hybrid wind/PV/diesel systems for two different sites in the Amazonia region of Brazil. The work incorporates the latest resource data and is based on the use of the Hybrid2 simulation code developed by the University of Massachusetts and NREL. Details of the baseline operating hybrid systems are reviewed, and the results of the latest detailed hybrid system evaluation for each site are summarized. Based on the system modeling results, separate recommendations for system modification and improvements are made.

SWAAM code development, verification and application to steam generator design

International Nuclear Information System (INIS)

Shin, Y.W.; Valentin, R.A.

1990-01-01

This paper describes the family of SWAAM codes developed by Argonne National Laboratory to analyze the effects of sodium/water reactions on LMR steam generators. The SWAAM codes were developed as design tools for analyzing various phenomena related to steam generator leaks and to predict the resulting thermal and hydraulic effects on the steam generator and the intermediate heat transport system (IHTS). The theoretical foundations and numerical treatments on which the codes are based are discussed, followed by a description of code capabilities and limitations, verification of the codes by comparison with experiment, and applications to steam generator and IHTS design. (author). 25 refs, 14 figs

Development of the versatile reactor analysis code system, MARBLE2

International Nuclear Information System (INIS)

Yokoyama, Kenji; Jin, Tomoyuki; Hazama, Taira; Hirai, Yasushi

2015-07-01

The second version of the versatile reactor analysis code system, MARBLE2, has been developed. A lot of new functions have been added in MARBLE2 by using the base technology developed in the first version (MARBLE1). Introducing the remaining functions of the conventional code system (JOINT-FR and SAGEP-FR), MARBLE2 enables one to execute almost all analysis functions of the conventional code system with the unified user interfaces of its subsystem, SCHEME. In particular, the sensitivity analysis functionality is available in MARBLE2. On the other hand, new built-in solvers have been developed, and existing ones have been upgraded. Furthermore, some other analysis codes and libraries developed in JAEA have been consolidated and prepared in SCHEME. In addition, several analysis codes developed in the other institutes have been additionally introduced as plug-in solvers. Consequently, gamma-ray transport calculation and heating evaluation become available. As for another subsystem, ORPHEUS, various functionality updates and speed-up techniques have been applied based on user experience of MARBLE1 to enhance its usability. (author)

Development of TIME2 code

International Nuclear Information System (INIS)

1986-02-01

The paper reviews the progress on the development of a computer model TIME2, for modelling the long term evolution of shallow burial site environments for low- and intermediate-level radioactive waste disposal. The subject is discussed under the five topic headings: 1) background studies, including geomorphology, climate, human-induced effects, and seismicity, 2) development of the TIME2 code, 3) verification and testing, 4) documentation, and, 5) role of TIME2 in radiological risk assessment. (U.K.)

Collaborative Multi-Layer Network Coding in Hybrid Cellular Cognitive Radio Networks

KAUST Repository

Moubayed, Abdallah J.

2015-05-01

In this paper, as an extension to [1], we propose a prioritized multi-layer network coding scheme for collaborative packet recovery in hybrid (interweave and underlay) cellular cognitive radio networks. This scheme allows the uncoordinated collaboration between the collocated primary and cognitive radio base-stations in order to minimize their own as well as each other\\'s packet recovery overheads, thus by improving their throughput. The proposed scheme ensures that each network\\'s performance is not degraded by its help to the other network. Moreover, it guarantees that the primary network\\'s interference threshold is not violated in the same and adjacent cells. Yet, the scheme allows the reduction of the recovery overhead in the collocated primary and cognitive radio networks. The reduction in the cognitive radio network is further amplified due to the perfect detection of spectrum holes which allows the cognitive radio base station to transmit at higher power without fear of violating the interference threshold of the primary network. For the secondary network, simulation results show reductions of 20% and 34% in the packet recovery overhead, compared to the non-collaborative scheme, for low and high probabilities of primary packet arrivals, respectively. For the primary network, this reduction was found to be 12%. © 2015 IEEE.

Theoretical atomic physics code development III TAPS: A display code for atomic physics data

International Nuclear Information System (INIS)

Clark, R.E.H.; Abdallah, J. Jr.; Kramer, S.P.

1988-12-01

A large amount of theoretical atomic physics data is becoming available through use of the computer codes CATS and ACE developed at Los Alamos National Laboratory. A new code, TAPS, has been written to access this data, perform averages over terms and configurations, and display information in graphical or text form. 7 refs., 13 figs., 1 tab

Development of LWR fuel performance code FEMAXI-6

International Nuclear Information System (INIS)

Suzuki, Motoe

2006-01-01

LWR fuel performance code: FEMAXI-6 (Finite Element Method in AXIs-symmetric system) is a representative fuel analysis code in Japan. Development history, background, design idea, features of model, and future are stated. Characteristic performance of LWR fuel and analysis code, what is model, development history of FEMAXI, use of FEMAXI code, fuel model, and a special feature of FEMAXI model is described. As examples of analysis, PCMI (Pellet-Clad Mechanical Interaction), fission gas release, gap bonding, and fission gas bubble swelling are reported. Thermal analysis and dynamic analysis system of FEMAXI-6, function block at one time step of FEMAXI-6, analytical example of PCMI in the output increase test by FEMAXI-III, analysis of fission gas release in Halden reactor by FEMAXI-V, comparison of the center temperature of fuel in Halden reactor, and analysis of change of diameter of fuel rod in high burn up BWR fuel are shown. (S.Y.)

Code development for nuclear reactor simulation

International Nuclear Information System (INIS)

Chauliac, C.; Verwaerde, D.; Pavageau, O.

2006-01-01

Full text of publication follows: Since several years, CEA, EDF and FANP have developed several numerical codes which are currently used for nuclear industry applications and will be remain in use for the coming years. Complementary to this set of codes and in order to better meet the present and future needs, a new system is being developed through a joint venture between CEA, EDF and FANP, with a ten year prospect and strong intermediate milestones. The focus is put on a multi-scale and multi-physics approach enabling to take into account phenomena from microscopic to macroscopic scale, and to describe interactions between various physical fields such as neutronics (DESCARTES), thermal-hydraulics (NEPTUNE) and fuel behaviour (PLEIADES). This approach is based on a more rational design of the softwares and uses a common integration platform providing pre-processing, supervision of computation and post-processing. This paper will describe the overall system under development and present the first results obtained. (authors)

Neutron flux investigation on certain alternative fluids in a hybrid system by using MCNPX Monte Carlo transport code

Energy Technology Data Exchange (ETDEWEB)

Guenay, Mehtap [Inoenue Univ., Malatya (Turkey). Physics Dept.

2014-04-15

In this study, the molten salt-heavy metal mixtures 93-85 % Li{sub 20}Sn{sub 80} + 5 % SFG-PuO{sub 2} and 2-10 % UO{sub 2}, 93-85 % Li{sub 20}Sn{sub 80} + 5 % SFG-PuO{sub 2} and 2-10 % NpO{sub 2}, 93-85 % Li{sub 20}Sn{sub 80} + 5 % SFG-PuO{sub 2} and 2-10 % UCO were used as fluids. The fluids were used in the liquid first wall, blanket and shield zones of the designed hybrid reactor system. Four centimeter thick 9Cr2WVTa ferritic steel was used as the structural material. In this study, the effect of mixture components on the neutron flux was investigated in a designed fusion-fission hybrid reactor system. The neutron flux was investigated according to the mixture components, radial flux distribution and energy spectrum in the designed system. Three-dimensional analyses were performed using the most recent MCNPX-2.7.0 Monte Carlo radiation transport code and the ENDF/B-VII.0 nuclear data library. (orig.)

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.

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.

Hybrid-PIC Computer Simulation of the Plasma and Erosion Processes in Hall Thrusters

Science.gov (United States)

Hofer, Richard R.; Katz, Ira; Mikellides, Ioannis G.; Gamero-Castano, Manuel

2010-01-01

HPHall software simulates and tracks the time-dependent evolution of the plasma and erosion processes in the discharge chamber and near-field plume of Hall thrusters. HPHall is an axisymmetric solver that employs a hybrid fluid/particle-in-cell (Hybrid-PIC) numerical approach. HPHall, originally developed by MIT in 1998, was upgraded to HPHall-2 by the Polytechnic University of Madrid in 2006. The Jet Propulsion Laboratory has continued the development of HPHall-2 through upgrades to the physical models employed in the code, and the addition of entirely new ones. Primary among these are the inclusion of a three-region electron mobility model that more accurately depicts the cross-field electron transport, and the development of an erosion sub-model that allows for the tracking of the erosion of the discharge chamber wall. The code is being developed to provide NASA science missions with a predictive tool of Hall thruster performance and lifetime that can be used to validate Hall thrusters for missions.

FDA Developments: Food Code 2013 and Proposed Trans Fat Determination

NARCIS (Netherlands)

Grossman, M.R.

2014-01-01

268 Reports EFFL 4|2014 USA FDA Developments: Food Code 2013 and Proposed Trans Fat Determination Margaret Rosso Grossman* I. Food Code 2013 and Food Code Reference System Since 1993, the US Food and Drug Administration has published a Food Code, now updated every four years. In November 2013, the

Development of a Fully-Automated Monte Carlo Burnup Code Monteburns

International Nuclear Information System (INIS)

Poston, D.I.; Trellue, H.R.

1999-01-01

Several computer codes have been developed to perform nuclear burnup calculations over the past few decades. In addition, because of advances in computer technology, it recently has become more desirable to use Monte Carlo techniques for such problems. Monte Carlo techniques generally offer two distinct advantages over discrete ordinate methods: (1) the use of continuous energy cross sections and (2) the ability to model detailed, complex, three-dimensional (3-D) geometries. These advantages allow more accurate burnup results to be obtained, provided that the user possesses the required computing power (which is required for discrete ordinate methods as well). Several linkage codes have been written that combine a Monte Carlo N-particle transport code (such as MCNP TM ) with a radioactive decay and burnup code. This paper describes one such code that was written at Los Alamos National Laboratory: monteburns. Monteburns links MCNP with the isotope generation and depletion code ORIGEN2. The basis for the development of monteburns was the need for a fully automated code that could perform accurate burnup (and other) calculations for any 3-D system (accelerator-driven or a full reactor core). Before the initial development of monteburns, a list of desired attributes was made and is given below. o The code should be fully automated (that is, after the input is set up, no further user interaction is required). . The code should allow for the irradiation of several materials concurrently (each material is evaluated collectively in MCNP and burned separately in 0RIGEN2). o The code should allow the transfer of materials (shuffling) between regions in MCNP. . The code should allow any materials to be added or removed before, during, or after each step in an automated fashion. . The code should not require the user to provide input for 0RIGEN2 and should have minimal MCNP input file requirements (other than a working MCNP deck). . The code should be relatively easy to use

Hybrid stage I palliation for hypo-plastic left heart condition without a hybrid suite: Suggestions for developing nations

Directory of Open Access Journals (Sweden)

S. Anuradha

2012-07-01

Full Text Available Cardiac hybrid procedures are performed in modern, spacious, and highly equipped hybrid suites in developed countries. Organizing such expensive suites in countries with an emerging economy is difficult from both a financial and logistics point of view. We share our experience of safely performing a Hybrid stage I palliation procedure for Aortic atresia with ventricular septal defect on a 2-month-old infant weighing 3.35 kg using minimal resources in a conventional catheterization laboratory.

Development and verification of a coupled code system RETRAN-MASTER-TORC

International Nuclear Information System (INIS)

Cho, J.Y.; Song, J.S.; Joo, H.G.; Zee, S.Q.

2004-01-01

Recently, coupled thermal-hydraulics (T-H) and three-dimensional kinetics codes have been widely used for the best-estimate simulations such as the main steam line break (MSLB) and locked rotor problems. This work is to develop and verify one of such codes by coupling the system T-H code RETRAN, the 3-D kinetics code MASTER and sub-channel analysis code TORC. The MASTER code has already been applied to such simulations after coupling with the MARS or RETRAN-3D multi-dimensional system T-H codes. The MASTER code contains a sub-channel analysis code COBRA-III C/P, and the coupled systems MARSMASTER-COBRA and RETRAN-MASTER-COBRA had been already developed and verified. With these previous studies, a new coupled system of RETRAN-MASTER-TORC is to be developed and verified for the standard best-estimate simulation code package in Korea. The TORC code has already been applied to the thermal hydraulics design of the several ABB/CE type plants and Korean Standard Nuclear Power Plants (KSNP). This justifies the choice of TORC rather than COBRA. Because the coupling between RETRAN and MASTER codes are already established and verified, this work is simplified to couple the TORC sub-channel T-H code with the MASTER neutronics code. The TORC code is a standalone code that solves the T-H equations for a given core problem from reading the input file and finally printing the converged solutions. However, in the coupled system, because TORC receives the pin power distributions from the neutronics code MASTER and transfers the T-H results to MASTER iteratively, TORC needs to be controlled by the MASTER code and does not need to solve the given problem completely at each iteration step. By this reason, the coupling of the TORC code with the MASTER code requires several modifications in the I/O treatment, flow iteration and calculation logics. The next section of this paper describes the modifications in the TORC code. The TORC control logic of the MASTER code is then followed. The

Development and Verification of Behavior of Tritium Analytic Code (BOTANIC)

International Nuclear Information System (INIS)

Park, Min Young; Kim, Eung Soo

2014-01-01

VHTR, one of the Generation IV reactor concepts, has a relatively high operation temperature and is usually suggested as a heat source for many industrial processes, including hydrogen production process. Thus, it is vital to trace tritium behavior in the VHTR system and the potential permeation rate to the industrial process. In other words, tritium is a crucial issue in terms of safety in the fission reactor system. Therefore, it is necessary to understand the behavior of tritium and the development of the tool to enable this is vital.. In this study, a Behavior of Tritium Analytic Code (BOTANIC) an analytic tool which is capable of analyzing tritium behavior is developed using a chemical process code called gPROMS. BOTANIC was then further verified using the analytic solutions and benchmark codes such as Tritium Permeation Analysis Code (TPAC) and COMSOL. In this study, the Behavior of Tritium Analytic Code, BOTANIC, has been developed using a chemical process code called gPROMS. The code has several distinctive features including non-diluted assumption, flexible applications and adoption of distributed permeation model. Due to these features, BOTANIC has the capability to analyze a wide range of tritium level systems and has a higher accuracy as it has the capacity to solve distributed models. BOTANIC was successfully developed and verified using analytical solution and the benchmark code calculation result. The results showed very good agreement with the analytical solutions and the calculation results of TPAC and COMSOL. Future work will be focused on the total system verification

Reactor Systems Technology Division code development and configuration/quality control procedures

International Nuclear Information System (INIS)

Johnson, E.C.

1985-06-01

Procedures are prescribed for executing a code development task and implementing the resulting coding in an official version of a computer code. The responsibilities of the project manager, development staff members, and the Code Configuration/Quality Control Group are defined. Examples of forms, logs, computer job control language, and suggested outlines for reports associated with software production and implementation are included in Appendix A. 1 raf., 2 figs

Study on fission blanket fuel cycling of a fusion-fission hybrid energy generation system

International Nuclear Information System (INIS)

Zhou, Z.; Yang, Y.; Xu, H.

2011-01-01

This paper presents a preliminary study on neutron physics characteristics of a light water cooled fission blanket for a new type subcritical fusion-fission hybrid reactor aiming at electric power generation with low technical limits of fission fuel. The major objective is to study the fission fuel cycling performance in the blanket, which may possess significant impacts on the feasibility of the new concept of fusion-fission hybrid reactor with a high energy gain (M) and tritium breeding ratio (TBR). The COUPLE2 code developed by the Institute of Nuclear and New Energy Technology of Tsinghua University is employed to simulate the neutronic behaviour in the blanket. COUPLE2 combines the particle transport code MCNPX with the fuel depletion code ORIGEN2. The code calculation results show that soft neutron spectrum can yield M > 20 while maintaining TBR >1.15 and the conversion ratio of fissile materials CR > 1 in a reasonably long refuelling cycle (>five years). The preliminary results also indicate that it is rather promising to design a high-performance light water cooled fission blanket of fusion-fission hybrid reactor for electric power generation by directly loading natural or depleted uranium if an ITER-scale tokamak fusion neutron source is achievable.

Practical requirements for software tools to assist in the validation and verification of hybrid expert systems

International Nuclear Information System (INIS)

Singh, G.P.; Cadena, D.; Burgess, J.

1992-01-01

Any practical software development effort must remain focused on verification and validation of user requirements. Knowledge-based system development is no different in this regard. In industry today, most expert systems being produced are, in reality, hybrid software systems which, in addition to those components that provide the knowledge base and expert reasoning over the problem domain using various rule-based and object-oriented paradigms, incorporate significant bodies of code based on more traditional software techniques such as database management, graphical user interfaces, hypermedia, spreadsheets, as well as specially developed sequential code. Validation and verification of such hybrid systems must perforce integrate suitable methodologies from all such fields. This paper attempts to provide a broad overview of the practical requirements for methodologies and the concomitant groupware tools which would assist in such an enterprise. These methodologies and groupware tools would facilitate the teamwork efforts necessary to validate and verify all components of such hybrid systems by emphasizing cooperative recording of requirements and negotiated resolutions of any conflicts grounded in a solid understanding of the semantics of such a system

Theoretical Atomic Physics code development IV: LINES, A code for computing atomic line spectra

International Nuclear Information System (INIS)

Abdallah, J. Jr.; Clark, R.E.H.

1988-12-01

A new computer program, LINES, has been developed for simulating atomic line emission and absorption spectra using the accurate fine structure energy levels and transition strengths calculated by the (CATS) Cowan Atomic Structure code. Population distributions for the ion stages are obtained in LINES by using the Local Thermodynamic Equilibrium (LTE) model. LINES is also useful for displaying the pertinent atomic data generated by CATS. This report describes the use of LINES. Both CATS and LINES are part of the Theoretical Atomic PhysicS (TAPS) code development effort at Los Alamos. 11 refs., 9 figs., 1 tab

Development of hybrid frequency couplers for non-inductive current drive in a tokamak; Developpement de coupleurs a la frequence hybride pour la generation non inductive du courant dans un tokamak

Energy Technology Data Exchange (ETDEWEB)

Berio, St.

1996-11-04

Used at its first time as an heating method in order to reach the temperature requisite for the fusion of a thermonuclear plasma, the hybrid waves has shown that they were the more efficient method for non-inductive current drive in a tokamak. The size and the objectives of a next machine such as ITER lead of the design of new antennae (in process of realisation on Tore Supra) made of oversized waveguides. This new concept of antenna will be more simple, more robust and will be able to transmit the same if not much power than the present antennae. This thesis contribute to the development of a new code called ALOHA (for `Advanced LOwer Hybrid Antenna`) which, at the end, will be able to give the characteristics and the behaviours of this new oversized antennae in front of a tokamak plasma. This thesis is also a first step in the interpretation of some experimental data concerning the measurement of coupling, absorption and current drive of the actual hybrid wave launched by a grill with rectangular waveguides. Moreover, this thesis lay some foundations of the study of these new antennae in front of a non-parallel confinement magnetic field and/or in front of poloidal inhomogeneities of plasma. (author). 53 refs.

Krylov solvers preconditioned with the low-order red-black algorithm for the PN hybrid FEM for the instant code

Energy Technology Data Exchange (ETDEWEB)

Wang, Yaqi; Rabiti, Cristian; Palmiotti, Giuseppe, E-mail: yaqi.wang@inl.gov, E-mail: cristian.rabiti@inl.gov, E-mail: giuseppe.palmiotti@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States)

2011-07-01

This paper proposes a new set of Krylov solvers, CG and GMRes, as an alternative of the Red-Black (RB) algorithm on on solving the steady-state one-speed neutron transport equation discretized with PN in angle and hybrid FEM (Finite Element Method) in space. A pre conditioner with the low-order RB iteration is designed to improve their convergence. These Krylov solvers can reduce the cost of pre-assembling the response matrices greatly. Numerical results with the INSTANT code are presented in order to show that they can be a good supplement on solving the PN-HFEM system. (author)

Krylov solvers preconditioned with the low-order red-black algorithm for the PN hybrid FEM for the instant code

International Nuclear Information System (INIS)

Wang, Yaqi; Rabiti, Cristian; Palmiotti, Giuseppe

2011-01-01

This paper proposes a new set of Krylov solvers, CG and GMRes, as an alternative of the Red-Black (RB) algorithm on on solving the steady-state one-speed neutron transport equation discretized with PN in angle and hybrid FEM (Finite Element Method) in space. A pre conditioner with the low-order RB iteration is designed to improve their convergence. These Krylov solvers can reduce the cost of pre-assembling the response matrices greatly. Numerical results with the INSTANT code are presented in order to show that they can be a good supplement on solving the PN-HFEM system. (author)

On transform coding tools under development for VP10

Science.gov (United States)

Parker, Sarah; Chen, Yue; Han, Jingning; Liu, Zoe; Mukherjee, Debargha; Su, Hui; Wang, Yongzhe; Bankoski, Jim; Li, Shunyao

2016-09-01

Google started the WebM Project in 2010 to develop open source, royaltyfree video codecs designed specifically for media on the Web. The second generation codec released by the WebM project, VP9, is currently served by YouTube, and enjoys billions of views per day. Realizing the need for even greater compression efficiency to cope with the growing demand for video on the web, the WebM team embarked on an ambitious project to develop a next edition codec, VP10, that achieves at least a generational improvement in coding efficiency over VP9. Starting from VP9, a set of new experimental coding tools have already been added to VP10 to achieve decent coding gains. Subsequently, Google joined a consortium of major tech companies called the Alliance for Open Media to jointly develop a new codec AV1. As a result, the VP10 effort is largely expected to merge with AV1. In this paper, we focus primarily on new tools in VP10 that improve coding of the prediction residue using transform coding techniques. Specifically, we describe tools that increase the flexibility of available transforms, allowing the codec to handle a more diverse range or residue structures. Results are presented on a standard test set.

SWAAM-code development and verification and application to steam generator designs

International Nuclear Information System (INIS)

Shin, Y.W.; Valentin, R.A.

1990-01-01

This paper describes the family of SWAAM codes which were developed by Argonne National Laboratory to analyze the effects of sodium-water reactions on LMR steam generators. The SWAAM codes were developed as design tools for analyzing various phenomena related to steam generator leaks and the resulting thermal and hydraulic effects on the steam generator and the intermediate heat transport system (IHTS). The paper discusses the theoretical foundations and numerical treatments on which the codes are based, followed by a description of code capabilities and limitations, verification of the codes and applications to steam generator and IHTS designs. 25 refs., 14 figs

Qualifying codes under software quality assurance: Two examples as guidelines for codes that are existing or under development

Energy Technology Data Exchange (ETDEWEB)

Mangold, D.

1993-05-01

Software quality assurance is an area of concem for DOE, EPA, and other agencies due to the poor quality of software and its documentation they have received in the past. This report briefly summarizes the software development concepts and terminology increasingly employed by these agencies and provides a workable approach to scientific programming under the new requirements. Following this is a practical description of how to qualify a simulation code, based on a software QA plan that has been reviewed and officially accepted by DOE/OCRWM. Two codes have recently been baselined and qualified, so that they can be officially used for QA Level 1 work under the DOE/OCRWM QA requirements. One of them was baselined and qualified within one week. The first of the codes was the multi-phase multi-component flow code TOUGH version 1, an already existing code, and the other was a geochemistry transport code STATEQ that was under development The way to accomplish qualification for both types of codes is summarized in an easy-to-follow step-by step fashion to illustrate how to baseline and qualify such codes through a relatively painless procedure.

Qualifying codes under software quality assurance: Two examples as guidelines for codes that are existing or under development

International Nuclear Information System (INIS)

Mangold, D.

1993-05-01

Software quality assurance is an area of concern for DOE, EPA, and other agencies due to the poor quality of software and its documentation they have received in the past. This report briefly summarizes the software development concepts and terminology increasingly employed by these agencies and provides a workable approach to scientific programming under the new requirements. Following this is a practical description of how to qualify a simulation code, based on a software QA plan that has been reviewed and officially accepted by DOE/OCRWM. Two codes have recently been baselined and qualified, so that they can be officially used for QA Level 1 work under the DOE/OCRWM QA requirements. One of them was baselined and qualified within one week. The first of the codes was the multi-phase multi-component flow code TOUGH version 1, an already existing code, and the other was a geochemistry transport code STATEQ that was under development The way to accomplish qualification for both types of codes is summarized in an easy-to-follow step-by step fashion to illustrate how to baseline and qualify such codes through a relatively painless procedure

Real-time hybrid simulation in a shaking table configuration for parametric studies of high-voltage equipment and IEEE693 development

Energy Technology Data Exchange (ETDEWEB)

Günay, Selim [nees@berkeley, UC Berkeley, Richmond, CA (United States); Mosalam, Khalid [Department of Civil and Environmental Engineering, UC Berkeley, Berkeley, CA (United States); Takhirov, Shakhzod, E-mail: takhirov@berkeley.edu [nees@berkeley, UC Berkeley, Richmond, CA (United States)

2015-12-15

Highlights: • A real-time hybrid simulation (RTHS) system for high-voltage (HV) equipment is developed. • The system is a cost effective and timely efficient approach for seismic testing and evaluation. • The coupled system of equipment and modeled support structure is tested/analyzed in real time. • The system is validated by comparing the RTHS test results with the shaking table results. • The effect of support structure on the equipment response is analyzed in a parametric study. - Abstract: This paper presents extensive discussion on seismic qualification of substation equipment in conventional shake table tests and its comparison to real-time hybrid simulation (RTHS). The hybrid simulation technique is based on a sub-structuring idea where a portion of a test specimen with well-predicted performance can be replaced by its finite element model. The rest of the test specimen is experimentally studied as part of the coupled system, where the test object and the mathematical model are interacting with each other in real time. The real-time hybrid simulation technique has a strong potential of complementing and in some cases replacing seismic qualification testing. In addition to that, it has a strong potential as a comprehensive and reliable tool for IEEE693 development, where code provisions can be developed from parametric hybrid simulation studies of actual pieces of substation equipment which are otherwise difficult to model. As a typical example of successful application of hybrid simulation, a comprehensive study related to RTHS of electrical disconnect switches is discussed in the paper. First, the RTHS system developed for this purpose is described and the results of a RTHS test are compared with a benchmark conventional shaking table test as a validation of the system. Second, effect of the support structures of the disconnect switches on the global and local responses of different insulator types is evaluated using the results of a series of

Real-time hybrid simulation in a shaking table configuration for parametric studies of high-voltage equipment and IEEE693 development

International Nuclear Information System (INIS)

Günay, Selim; Mosalam, Khalid; Takhirov, Shakhzod

2015-01-01

Highlights: • A real-time hybrid simulation (RTHS) system for high-voltage (HV) equipment is developed. • The system is a cost effective and timely efficient approach for seismic testing and evaluation. • The coupled system of equipment and modeled support structure is tested/analyzed in real time. • The system is validated by comparing the RTHS test results with the shaking table results. • The effect of support structure on the equipment response is analyzed in a parametric study. - Abstract: This paper presents extensive discussion on seismic qualification of substation equipment in conventional shake table tests and its comparison to real-time hybrid simulation (RTHS). The hybrid simulation technique is based on a sub-structuring idea where a portion of a test specimen with well-predicted performance can be replaced by its finite element model. The rest of the test specimen is experimentally studied as part of the coupled system, where the test object and the mathematical model are interacting with each other in real time. The real-time hybrid simulation technique has a strong potential of complementing and in some cases replacing seismic qualification testing. In addition to that, it has a strong potential as a comprehensive and reliable tool for IEEE693 development, where code provisions can be developed from parametric hybrid simulation studies of actual pieces of substation equipment which are otherwise difficult to model. As a typical example of successful application of hybrid simulation, a comprehensive study related to RTHS of electrical disconnect switches is discussed in the paper. First, the RTHS system developed for this purpose is described and the results of a RTHS test are compared with a benchmark conventional shaking table test as a validation of the system. Second, effect of the support structures of the disconnect switches on the global and local responses of different insulator types is evaluated using the results of a series of

Development of fast and accurate Monte Carlo code MVP

International Nuclear Information System (INIS)

Mori, Takamasa

2001-01-01

The development work of fast and accurate Monte Carlo code MVP has started at JAERI in late 80s. From the beginning, the code was designed to utilize vector supercomputers and achieved higher computation speed by a factor of 10 or more compared with conventional codes. In 1994, the first version of MVP was released together with cross section libraries based on JENDL-3.1 and JENDL-3.2. In 1996, minor revision was made by adding several functions such as treatments of ENDF-B6 file 6 data, time dependent problem, and so on. Since 1996, several works have been carried out for the next version of MVP. The main works are (1) the development of continuous energy Monte Carlo burn-up calculation code MVP-BURN, (2) the development of a system to generate cross section libraries at arbitrary temperature, and (3) the study on error estimations and their biases in Monte Carlo eigenvalue calculations. This paper summarizes the main features of MVP, results of recent studies and future plans for MVP. (author)

Development of throughflow calculation code for axial flow compressors

International Nuclear Information System (INIS)

Kim, Ji Hwan; Kim, Hyeun Min; No, Hee Cheon

2005-01-01

The power conversion systems of the current HTGRs are based on closed Brayton cycle and major concern is thermodynamic performance of the axial flow helium gas turbines. Particularly, the helium compressor has some unique design challenges compared to the air-breathing compressor such as high hub-to-tip ratios throughout the machine and a large number of stages due to the physical property of the helium and thermodynamic cycle. Therefore, it is necessary to develop a design and analysis code for helium compressor that can estimate the design point and off-design performance accurately. KAIST nuclear system laboratory has developed a compressor design and analysis code by means of throughflow calculation and several loss models. This paper presents the outline of the development of a throughflow calculation code and its verification results

Developing and modifying behavioral coding schemes in pediatric psychology: a practical guide.

Science.gov (United States)

Chorney, Jill MacLaren; McMurtry, C Meghan; Chambers, Christine T; Bakeman, Roger

2015-01-01

To provide a concise and practical guide to the development, modification, and use of behavioral coding schemes for observational data in pediatric psychology. This article provides a review of relevant literature and experience in developing and refining behavioral coding schemes. A step-by-step guide to developing and/or modifying behavioral coding schemes is provided. Major steps include refining a research question, developing or refining the coding manual, piloting and refining the coding manual, and implementing the coding scheme. Major tasks within each step are discussed, and pediatric psychology examples are provided throughout. Behavioral coding can be a complex and time-intensive process, but the approach is invaluable in allowing researchers to address clinically relevant research questions in ways that would not otherwise be possible. © The Author 2014. Published by Oxford University Press on behalf of the Society of Pediatric Psychology. All rights reserved. For permissions, please e-mail: journals.permissions@oup.com.

Developments in the Generation and Interpretation of Wire Codes (invited paper)

International Nuclear Information System (INIS)

Ebi, K.L.

1999-01-01

Three new developments in the generation and interpretation of wire codes are discussed. First, a method was developed to computer generate wire codes using data gathered from a utility database of the local distribution system and from tax assessor records. This method was used to wire code more than 250,000 residences in the greater Denver metropolitan area. There was an approximate 75% agreement with field wire coding. Other research in Denver suggests that wire codes predict some characteristics of a residence and its neighbourhood, including age, assessed value, street layout and traffic density. A third new development is the case-specular method to study the association between wire codes and childhood cancers. Recent results from applying the method to the Savitz et al and London et al studies suggest that the associations between childhood cancer and VHCC residences were strongest for residences with a backyard rather than street service drop, and for VHCC residences with LCC speculars. (author)

Development of LMR basic design technology - Development of 3-D multi-group nodal kinetics code for liquid metal reactors

Energy Technology Data Exchange (ETDEWEB)

Kim, Myung Hyun [Kyunghee University, Seoul (Korea, Republic of)

1996-07-01

A development project of 3-dimensional kinetics code for ALMR has three level of works. In the first level, a multi-group, nodal kinetics code for the HEX-Z geometry has been developed. A code showed very good results for the static analysis as well as for the kinetics problems. At the second level, a core thermal-hydraulic analysis code was developed for the temperature feedback calculation in ALMR transients analysis. This code is coupled with kinetics code. A sodium property table was programmed and tested to the KAERI data and thermal feedback model was developed and coupled in code. Benchmarking of T/H calculation has been performed and showed fairly good results. At the third level of research work, reactivity feedback model for structure thermal expansion is developed and added to the code. At present, basic model was studied. However, code development in now on going. Benchmarking of this model developed can not be done because of lack of data. 31 refs., 17 tabs., 38 figs. (author)

Application of software to development of reactor-safety codes

International Nuclear Information System (INIS)

Wilburn, N.P.; Niccoli, L.G.

1980-09-01

Over the past two-and-a-half decades, the application of new techniques has reduced hardware cost for digital computer systems and increased computational speed by several orders of magnitude. A corresponding cost reduction in business and scientific software development has not occurred. The same situation is seen for software developed to model the thermohydraulic behavior of nuclear systems under hypothetical accident situations. For all cases this is particularly noted when costs over the total software life cycle are considered. A solution to this dilemma for reactor safety code systems has been demonstrated by applying the software engineering techniques which have been developed over the course of the last few years in the aerospace and business communities. These techniques have been applied recently with a great deal of success in four major projects at the Hanford Engineering Development Laboratory (HEDL): 1) a rewrite of a major safety code (MELT); 2) development of a new code system (CONACS) for description of the response of LMFBR containment to hypothetical accidents, and 3) development of two new modules for reactor safety analysis

Development of 1D Liner Compression Code for IDL

Science.gov (United States)

Shimazu, Akihisa; Slough, John; Pancotti, Anthony

2015-11-01

A 1D liner compression code is developed to model liner implosion dynamics in the Inductively Driven Liner Experiment (IDL) where FRC plasmoid is compressed via inductively-driven metal liners. The driver circuit, magnetic field, joule heating, and liner dynamics calculations are performed at each time step in sequence to couple these effects in the code. To obtain more realistic magnetic field results for a given drive coil geometry, 2D and 3D effects are incorporated into the 1D field calculation through use of correction factor table lookup approach. Commercial low-frequency electromagnetic fields solver, ANSYS Maxwell 3D, is used to solve the magnetic field profile for static liner condition at various liner radius in order to derive correction factors for the 1D field calculation in the code. The liner dynamics results from the code is verified to be in good agreement with the results from commercial explicit dynamics solver, ANSYS Explicit Dynamics, and previous liner experiment. The developed code is used to optimize the capacitor bank and driver coil design for better energy transfer and coupling. FRC gain calculations are also performed using the liner compression data from the code for the conceptual design of the reactor sized system for fusion energy gains.

Development and application of methods to characterize code uncertainty

International Nuclear Information System (INIS)

Wilson, G.E.; Burtt, J.D.; Case, G.S.; Einerson, J.J.; Hanson, R.G.

1985-01-01

The United States Nuclear Regulatory Commission sponsors both international and domestic studies to assess its safety analysis codes. The Commission staff intends to use the results of these studies to quantify the uncertainty of the codes with a statistically based analysis method. Development of the methodology is underway. The Idaho National Engineering Laboratory contributions to the early development effort, and testing of two candidate methods are the subjects of this paper

The development of the code package PERMAK--3D//SC--1

International Nuclear Information System (INIS)

Bolobov, P. A.; Oleksuk, D. A.

2011-01-01

Code package PERMAK-3D//SC-1 was developed for performing pin-by-pin coupled neutronic and thermal hydraulic calculation of the core fragment of seven fuel assemblies and was designed on the basis of 3D multigroup pin-by-pin code PERMAK-3D and 3D (subchannel) thermal hydraulic code SC-1 The code package predicts axial and radial pin-by-pin power distribution and coolant parameters in stimulated region (enthalpies,, velocities,, void fractions,, boiling and DNBR margins).. The report describes some new steps in code package development. Some PERMAK-3D//SC-1 outcomes of WWER calculations are presented in the report. (Authors)

What Are They Talking About? Analyzing Code Reviews in Pull-Based Development Model

Institute of Scientific and Technical Information of China (English)

Zhi-Xing Li; Yue Yu; Gang Yin; Tao Wang; Huai-Min Wang

2017-01-01

Code reviews in pull-based model are open to community users on GitHub. Various participants are taking part in the review discussions and the review topics are not only about the improvement of code contributions but also about project evolution and social interaction. A comprehensive understanding of the review topics in pull-based model would be useful to better organize the code review process and optimize review tasks such as reviewer recommendation and pull-request prioritization. In this paper, we first conduct a qualitative study on three popular open-source software projects hosted on GitHub and construct a fine-grained two-level taxonomy covering four level-1 categories (code correctness, pull-request decision-making, project management, and social interaction) and 11 level-2 subcategories (e.g., defect detecting, reviewer assigning, contribution encouraging). Second, we conduct preliminary quantitative analysis on a large set of review comments that were labeled by TSHC (a two-stage hybrid classification algorithm), which is able to automatically classify review comments by combining rule-based and machine-learning techniques. Through the quantitative study, we explore the typical review patterns. We find that the three projects present similar comments distribution on each subcategory. Pull-requests submitted by inexperienced contributors tend to contain potential issues even though they have passed the tests. Furthermore, external contributors are more likely to break project conventions in their early contributions.

Cooperation of experts' opinion, experiment and computer code development

International Nuclear Information System (INIS)

Wolfert, K.; Hicken, E.

The connection between code development, code assessment and confidence in the analysis of transients will be discussed. In this manner, the major sources of errors in the codes and errors in applications of the codes will be shown. Standard problem results emphasize that, in order to have confidence in licensing statements, the codes must be physically realistic and the code user must be qualified and experienced. We will discuss why there is disagreement between the licensing authority and vendor concerning assessment of the fullfillment of safety goal requirements. The answer to the question lies in the different confidence levels of the assessment of transient analysis. It is expected that a decrease in the disagreement will result from an increased confidence level. Strong efforts will be made to increase this confidence level through improvements in the codes, experiments and related organizational strcutures. Because of the low probability for loss-of-coolant-accidents in the nuclear industry, assessment must rely on analytical techniques and experimental investigations. (orig./HP) [de

Development of parallel Fokker-Planck code ALLAp

International Nuclear Information System (INIS)

Batishcheva, A.A.; Sigmar, D.J.; Koniges, A.E.

1996-01-01

We report on our ongoing development of the 3D Fokker-Planck code ALLA for a highly collisional scrape-off-layer (SOL) plasma. A SOL with strong gradients of density and temperature in the spatial dimension is modeled. Our method is based on a 3-D adaptive grid (in space, magnitude of the velocity, and cosine of the pitch angle) and a second order conservative scheme. Note that the grid size is typically 100 x 257 x 65 nodes. It was shown in our previous work that only these capabilities make it possible to benchmark a 3D code against a spatially-dependent self-similar solution of a kinetic equation with the Landau collision term. In the present work we show results of a more precise benchmarking against the exact solutions of the kinetic equation using a new parallel code ALLAp with an improved method of parallelization and a modified boundary condition at the plasma edge. We also report first results from the code parallelization using Message Passing Interface for a Massively Parallel CRI T3D platform. We evaluate the ALLAp code performance versus the number of T3D processors used and compare its efficiency against a Work/Data Sharing parallelization scheme and a workstation version

Development of chemical equilibrium analysis code 'CHEEQ'

International Nuclear Information System (INIS)

Nagai, Shuichiro

2006-08-01

'CHEEQ' code which calculates the partial pressure and the mass of the system consisting of ideal gas and pure condensed phase compounds, was developed. Characteristics of 'CHEEQ' code are as follows. All the chemical equilibrium equations were described by the formation reactions from the mono-atomic gases in order to simplify the code structure and input preparation. Chemical equilibrium conditions, Σν i μ i =0 for the gaseous compounds and precipitated condensed phase compounds and Σν i μ i > 0 for the non-precipitated condensed phase compounds, were applied. Where, ν i and μ i are stoichiometric coefficient and chemical potential of component i. Virtual solid model was introduced to perform the calculation of constant partial pressure condition. 'CHEEQ' was consisted of following 3 parts, (1) analysis code, zc132. f. (2) thermodynamic data base, zmdb01 and (3) input data file, zindb. 'CHEEQ' code can calculate the system which consisted of elements (max.20), condensed phase compounds (max.100) and gaseous compounds. (max.200). Thermodynamic data base, zmdb01 contains about 1000 elements and compounds, and 200 of them were Actinide elements and their compounds. This report describes the basic equations, the outline of the solution procedure and instructions to prepare the input data and to evaluate the calculation results. (author)

Development of code PRETOR for stellarator simulation

International Nuclear Information System (INIS)

Dies, J.; Fontanet, J.; Fontdecaba, J.M.; Castejon, F.; Alejandre, C.

1998-01-01

The Department de Fisica i Enginyeria Nuclear (DFEN) of the UPC has some experience in the development of the transport code PRETOR. This code has been validated with shots of DIII-D, JET and TFTR, it has also been used in the simulation of operational scenarios of ITER fast burnt termination. Recently, the association EURATOM-CIEMAT has started the operation of the TJ-II stellarator. Due to the need of validating the results given by others transport codes applied to stellarators and because all of them made some approximations, as a averaging magnitudes in each magnetic surface, it was thought suitable to adapt the PRETOR code to devices without axial symmetry, like stellarators, which is very suitable for the specific needs of the study of TJ-II. Several modifications are required in PRETOR; the main concerns to the models of: magnetic equilibrium, geometry and transport of energy and particles. In order to solve the complex magnetic equilibrium geometry the powerful numerical code VMEC has been used. This code gives the magnetic surface shape as a Fourier series in terms of the harmonics (m,n). Most of the geometric magnitudes are also obtained from the VMEC results file. The energy and particle transport models will be replaced by other phenomenological models that are better adapted to stellarator simulation. Using the proposed models, it is pretended to reproduce experimental data available from present stellarators, given especial attention to the TJ-II of the association EURATOM-CIEMAT. (Author)

Multiple application coded switch development report

International Nuclear Information System (INIS)

Bernal, E.L.; Kestly, J.D.

1979-03-01

The development of the Multiple Application Coded Switch (MACS) and its related controller are documented; the functional and electrical characteristics are described; the interface requirements defined, and a troubleshooting guide provided. The system was designed for the Safe Secure Trailer System used for secure transportation of nuclear material

Development of steam explosion simulation code JASMINE

Energy Technology Data Exchange (ETDEWEB)

Moriyama, Kiyofumi; Yamano, Norihiro; Maruyama, Yu; Kudo, Tamotsu; Sugimoto, Jun [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nagano, Katsuhiro; Araki, Kazuhiro

1995-11-01

A steam explosion is considered as a phenomenon which possibly threatens the integrity of the containment vessel of a nuclear power plant in a severe accident condition. A numerical calculation code JASMINE (JAeri Simulator for Multiphase INteraction and Explosion) purposed to simulate the whole process of steam explosions has been developed. The premixing model is based on a multiphase flow simulation code MISTRAL by Fuji Research Institute Co. In JASMINE code, the constitutive equations and the flow regime map are modified for the simulation of premixing related phenomena. The numerical solution method of the original code is succeeded, i.e. the basic equations are discretized semi-implicitly, BCGSTAB method is used for the matrix solver to improve the stability and convergence, also TVD scheme is applied to capture a steep phase distribution accurately. Test calculations have been performed for the conditions correspond to the experiments by Gilbertson et al. and Angelini et al. in which mixing of solid particles and water were observed in iso-thermal condition and with boiling, respectively. (author).

Development of steam explosion simulation code JASMINE

International Nuclear Information System (INIS)

Moriyama, Kiyofumi; Yamano, Norihiro; Maruyama, Yu; Kudo, Tamotsu; Sugimoto, Jun; Nagano, Katsuhiro; Araki, Kazuhiro.

1995-11-01

A steam explosion is considered as a phenomenon which possibly threatens the integrity of the containment vessel of a nuclear power plant in a severe accident condition. A numerical calculation code JASMINE (JAeri Simulator for Multiphase INteraction and Explosion) purposed to simulate the whole process of steam explosions has been developed. The premixing model is based on a multiphase flow simulation code MISTRAL by Fuji Research Institute Co. In JASMINE code, the constitutive equations and the flow regime map are modified for the simulation of premixing related phenomena. The numerical solution method of the original code is succeeded, i.e. the basic equations are discretized semi-implicitly, BCGSTAB method is used for the matrix solver to improve the stability and convergence, also TVD scheme is applied to capture a steep phase distribution accurately. Test calculations have been performed for the conditions correspond to the experiments by Gilbertson et al. and Angelini et al. in which mixing of solid particles and water were observed in iso-thermal condition and with boiling, respectively. (author)

Present status of transport code development based on Monte Carlo method

International Nuclear Information System (INIS)

Nakagawa, Masayuki

1985-01-01

The present status of development in Monte Carlo code is briefly reviewed. The main items are the followings; Application fields, Methods used in Monte Carlo code (geometry spectification, nuclear data, estimator and variance reduction technique) and unfinished works, Typical Monte Carlo codes and Merits of continuous energy Monte Carlo code. (author)

Development of a power train for the hybrid automobile - the Civic IMA

Energy Technology Data Exchange (ETDEWEB)

Matsuki, Masato; Sato, Toshiyuki; Wakashiro, Teruo; Kaku, Toshiaki; Kamiyama, Toshihiro; Kanda, Masahiro [Tochigi R and D Center (Japan); Brachmann, T. [Tochigi Offenbach R und D Center (Germany)

2003-07-01

The Civic Hybrid was developed as a compact passenger hybrid car that achieves both low fuel consumption and cleaner operation from the viewpoints of preserving the global environment and conserving resources. The engine has been improved for Hybrid applications, which were added to the base i-DSI, 4-cylinder, 1.3-liter SOHC, 2-ignition plugs/cylinder engine mounted in the Honda 'Jazz'. In addition, the cylinder idling system has been adopted to increase the regenerated energy during deceleration. The hybrid system is based on the Honda IMA system, and the maximum regenerative torque has been increased by approximately 30% by improving the magnetic circuits of an ultra-thin DC brushless motor and adopting a new rotor manufacturing method. Fuel economy is improved by a new hybrid power train, thus achieving low fuel consumption of 4.9 1/100 km in the European UDC+EUCD combined mode by at the same time meeting EURO IV standards. The power control unit, which is the IMA system control unit, was downsized and located behind the rear seat, thus ensuring comparable trunk capacity to the base vehicle of the Civic 4-Door. Hybrid vehicles have a lot to offer. This paper introduces evolutionary developments of Hybrid vehicles within the Honda Motor Company. (orig.)

Python-Assisted MODFLOW Application and Code Development

Science.gov (United States)

Langevin, C.

2013-12-01

The U.S. Geological Survey (USGS) has a long history of developing and maintaining free, open-source software for hydrological investigations. The MODFLOW program is one of the most popular hydrologic simulation programs released by the USGS, and it is considered to be the most widely used groundwater flow simulation code. MODFLOW was written using a modular design and a procedural FORTRAN style, which resulted in code that could be understood, modified, and enhanced by many hydrologists. The code is fast, and because it uses standard FORTRAN it can be run on most operating systems. Most MODFLOW users rely on proprietary graphical user interfaces for constructing models and viewing model results. Some recent efforts, however, have focused on construction of MODFLOW models using open-source Python scripts. Customizable Python packages, such as FloPy (https://code.google.com/p/flopy), can be used to generate input files, read simulation results, and visualize results in two and three dimensions. Automating this sequence of steps leads to models that can be reproduced directly from original data and rediscretized in space and time. Python is also being used in the development and testing of new MODFLOW functionality. New packages and numerical formulations can be quickly prototyped and tested first with Python programs before implementation in MODFLOW. This is made possible by the flexible object-oriented design capabilities available in Python, the ability to call FORTRAN code from Python, and the ease with which linear systems of equations can be solved using SciPy, for example. Once new features are added to MODFLOW, Python can then be used to automate comprehensive regression testing and ensure reliability and accuracy of new versions prior to release.

Development of thermal hydraulic models for the reliable regulatory auditing code

Energy Technology Data Exchange (ETDEWEB)

Chung, B. D.; Song, C. H.; Lee, Y. J.; Kwon, T. S.; Lee, S. W. [Korea Automic Energy Research Institute, Taejon (Korea, Republic of)

2004-02-15

The objective of this project is to develop thermal hydraulic models for use in improving the reliability of the regulatory auditing codes. The current year fall under the second step of the 3 year project, and the main researches were focused on the development of downcorner boiling model. During the current year, the bubble stream model of downcorner has been developed and installed in he auditing code. The model sensitivity analysis has been performed for APR1400 LBLOCA scenario using the modified code. The preliminary calculation has been performed for the experimental test facility using FLUENT and MARS code. The facility for air bubble experiment has been installed. The thermal hydraulic phenomena for VHTR and super critical reactor have been identified for the future application and model development.

A Hybrid Task Graph Scheduler for High Performance Image Processing Workflows.

Science.gov (United States)

Blattner, Timothy; Keyrouz, Walid; Bhattacharyya, Shuvra S; Halem, Milton; Brady, Mary

2017-12-01

Designing applications for scalability is key to improving their performance in hybrid and cluster computing. Scheduling code to utilize parallelism is difficult, particularly when dealing with data dependencies, memory management, data motion, and processor occupancy. The Hybrid Task Graph Scheduler (HTGS) improves programmer productivity when implementing hybrid workflows for multi-core and multi-GPU systems. The Hybrid Task Graph Scheduler (HTGS) is an abstract execution model, framework, and API that increases programmer productivity when implementing hybrid workflows for such systems. HTGS manages dependencies between tasks, represents CPU and GPU memories independently, overlaps computations with disk I/O and memory transfers, keeps multiple GPUs occupied, and uses all available compute resources. Through these abstractions, data motion and memory are explicit; this makes data locality decisions more accessible. To demonstrate the HTGS application program interface (API), we present implementations of two example algorithms: (1) a matrix multiplication that shows how easily task graphs can be used; and (2) a hybrid implementation of microscopy image stitching that reduces code size by ≈ 43% compared to a manually coded hybrid workflow implementation and showcases the minimal overhead of task graphs in HTGS. Both of the HTGS-based implementations show good performance. In image stitching the HTGS implementation achieves similar performance to the hybrid workflow implementation. Matrix multiplication with HTGS achieves 1.3× and 1.8× speedup over the multi-threaded OpenBLAS library for 16k × 16k and 32k × 32k size matrices, respectively.

Hybrid Obfuscation Using Signals and Encryption

Directory of Open Access Journals (Sweden)

Bahare Hashemzade

2018-01-01

Full Text Available Obfuscation of software and data is one of the subcategories of software security. Hence, the outlines of the obfuscation problem and its various methods have been studied in this article. This paper proposes a hybrid of two signals and encryption obfuscation to hide the behaviour program and prevent reconstruction of the normal code by hackers. The usual signal method is strong enough for obfuscation, but its problem is the high complexity because of a lot of call and return instructions. In this study, a new dispatcher was added to the source code to reconstruct the original control flow graph from the hidden one to solve the problem of the signal method. This dispatcher code is encrypted to preclude access by the hacker. In this paper, the potency that makes the obfuscation strong has been increased and the resilience that makes the obfuscation poor has been decreased. The results of a comparison of the similarity among the ambiguous data with its original code and with available efficient methods present a performance advantage of the proposed hybrid obfuscation algorithm.

Safety, codes and standards for hydrogen installations. Metrics development and benchmarking

Energy Technology Data Exchange (ETDEWEB)

Harris, Aaron P. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Dedrick, Daniel E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); LaFleur, Angela Christine [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); San Marchi, Christopher W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2014-04-01

Automakers and fuel providers have made public commitments to commercialize light duty fuel cell electric vehicles and fueling infrastructure in select US regions beginning in 2014. The development, implementation, and advancement of meaningful codes and standards is critical to enable the effective deployment of clean and efficient fuel cell and hydrogen solutions in the energy technology marketplace. Metrics pertaining to the development and implementation of safety knowledge, codes, and standards are important to communicate progress and inform future R&D investments. This document describes the development and benchmarking of metrics specific to the development of hydrogen specific codes relevant for hydrogen refueling stations. These metrics will be most useful as the hydrogen fuel market transitions from pre-commercial to early-commercial phases. The target regions in California will serve as benchmarking case studies to quantify the success of past investments in research and development supporting safety codes and standards R&D.

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)

Development of the PRO-LOCA Probabilistic Fracture Mechanics Code, MERIT Final Report

International Nuclear Information System (INIS)

Scott, Paul; Kurth, Robert; Cox, Andrew; Olson, Rick; Rudland, Dave

2010-12-01

The MERIT project has been an internationally financed program with the main purpose of developing probabilistic models for piping failure of nuclear components and to include these models in a probabilistic code named PRO-LOCA. The principal objective of the project has been to develop probabilistic models for piping failure of nuclear components and to include these models in a probabilistic code. The MERIT program has produced a code named PRO-LOCA with the following features: - Crack initiation models for fatigue or stress corrosion cracking for previously unflawed material. - Subcritical crack growth models for fatigue and stress corrosion cracking for both initiated and pre-existing circumferential defects. - Models for flaw detection by inspections and leak detection. - Crack stability. The PRO-LOCA code can thus predict the leak or break frequency for the whole sequence of initiation, subcritical crack growth until wall penetration and leakage, instability of the through-wall crack (pipe rupture). The outcome of the PRO-LOCA code are a sequence of failure frequencies which represents the probability of surface crack developing, a through-wall crack developing and six different sizes of crack opening areas corresponding to different leak flow rates or LOCA categories. Note that the level of quality assurance of the PRO-LOCA code is such that the code in its current state of development is considered to be more of a research code than a regulatory tool.

Development of the PRO-LOCA Probabilistic Fracture Mechanics Code, MERIT Final Report

Energy Technology Data Exchange (ETDEWEB)

Scott, Paul; Kurth, Robert; Cox, Andrew; Olson, Rick (Battelle Columbus (United States)); Rudland, Dave (Nuclear Regulatory Commission (United States))

2010-12-15

The MERIT project has been an internationally financed program with the main purpose of developing probabilistic models for piping failure of nuclear components and to include these models in a probabilistic code named PRO-LOCA. The principal objective of the project has been to develop probabilistic models for piping failure of nuclear components and to include these models in a probabilistic code. The MERIT program has produced a code named PRO-LOCA with the following features: - Crack initiation models for fatigue or stress corrosion cracking for previously unflawed material. - Subcritical crack growth models for fatigue and stress corrosion cracking for both initiated and pre-existing circumferential defects. - Models for flaw detection by inspections and leak detection. - Crack stability. The PRO-LOCA code can thus predict the leak or break frequency for the whole sequence of initiation, subcritical crack growth until wall penetration and leakage, instability of the through-wall crack (pipe rupture). The outcome of the PRO-LOCA code are a sequence of failure frequencies which represents the probability of surface crack developing, a through-wall crack developing and six different sizes of crack opening areas corresponding to different leak flow rates or LOCA categories. Note that the level of quality assurance of the PRO-LOCA code is such that the code in its current state of development is considered to be more of a research code than a regulatory tool.

Theoretical atomic physics code development I: CATS: Cowan Atomic Structure Code

International Nuclear Information System (INIS)

Abdallah, J. Jr.; Clark, R.E.H.; Cowan, R.D.

1988-12-01

An adaptation of R.D. Cowan's Atomic Structure program, CATS, has been developed as part of the Theoretical Atomic Physics (TAPS) code development effort at Los Alamos. CATS has been designed to be easy to run and to produce data files that can interface with other programs easily. The CATS produced data files currently include wave functions, energy levels, oscillator strengths, plane-wave-Born electron-ion collision strengths, photoionization cross sections, and a variety of other quantities. This paper describes the use of CATS. 10 refs

Recent developments in the Los Alamos radiation transport code system

International Nuclear Information System (INIS)

Forster, R.A.; Parsons, K.

1997-01-01

A brief progress report on updates to the Los Alamos Radiation Transport Code System (LARTCS) for solving criticality and fixed-source problems is provided. LARTCS integrates the Diffusion Accelerated Neutral Transport (DANT) discrete ordinates codes with the Monte Carlo N-Particle (MCNP) code. The LARCTS code is being developed with a graphical user interface for problem setup and analysis. Progress in the DANT system for criticality applications include a two-dimensional module which can be linked to a mesh-generation code and a faster iteration scheme. Updates to MCNP Version 4A allow statistical checks of calculated Monte Carlo results

Improved Hybrid Opponent System for Professional Military Training

Directory of Open Access Journals (Sweden)

Michael Pelosi

2017-10-01

Full Text Available Described herein is a general-purpose software engineering architecture for autonomous, computer controlled opponent implementation in modern maneuver warfare simulation and training. The implementation has been developed, refined, and tested in the user crucible for several years. The approach represents a hybrid application of various well-known AI techniques, including domain modeling, agent modeling, and object-oriented programming. Inspired by computer chess approaches, the methodology combines this theoretical foundation with a hybrid and scalable portfolio of additional techniques. The result remains simple enough to be maintainable, comprehensible for the code writers as well as the end-users, and robust enough to handle a wide spectrum of possible mission scenarios and circumstances without modification.

Development of Nuclear Energy Security Code

International Nuclear Information System (INIS)

Shimamura, Takehisa; Suzuki, Atsuyuki; Okubo, Hiroo; Kikuchi, Masahiro.

1990-01-01

In establishing of the nuclear fuel cycle in Japan that have a vulnerability in own energy structure, an effectiveness of energy security should be taken into account as well as an economy based on the balance of supply and demand of nuclear fuels. NMCC develops the 'Nuclear Energy Security Code' which was able to evaluate the effectiveness of energy security. Evaluation method adopted in this code is 'Import Premium' which was proposed in 'World Oil', EMF Report 6. The viewpoints of evaluation are as follows: 1. How much uranium fuel quantity can be reduced by using plutonium fuel? 2. How much a sudden rise of fuel cost can be absorbed by establishing the plutonium cycle beforehand the energy crisis? (author)

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

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

Fatigue Life Analysis of Tapered Hybrid Composite Flexbeams

Science.gov (United States)

Murri, Gretchen B.; Schaff, Jeffery R.; Dobyns, Alan L.

2002-01-01

Nonlinear-tapered flexbeam laminates from a full-size composite helicopter rotor hub flexbeam were tested under combined constant axial tension and cyclic bending loads. The two different graphite/glass hybrid configurations tested under cyclic loading failed by delamination in the tapered region. A 2-D finite element model was developed which closely approximated the flexbeam geometry, boundary conditions, and loading. The analysis results from two geometrically nonlinear finite element codes, ANSYS and ABAQUS, are presented and compared. Strain energy release rates (G) obtained from the above codes using the virtual crack closure technique (VCCT) at a resin crack location in the flexbeams are presented for both hybrid material types. These results compare well with each other and suggest that the initial delamination growth from the resin crack toward the thick region of the flexbeam is strongly mode II. The peak calculated G values were used with material characterization data to calculate fatigue life curves and compared with test data. A curve relating maximum surface strain to number of loading cycles at delamination onset compared reasonably well with the test results.

Open-Source Development of the Petascale Reactive Flow and Transport Code PFLOTRAN

Science.gov (United States)

Hammond, G. E.; Andre, B.; Bisht, G.; Johnson, T.; Karra, S.; Lichtner, P. C.; Mills, R. T.

2013-12-01

Open-source software development has become increasingly popular in recent years. Open-source encourages collaborative and transparent software development and promotes unlimited free redistribution of source code to the public. Open-source development is good for science as it reveals implementation details that are critical to scientific reproducibility, but generally excluded from journal publications. In addition, research funds that would have been spent on licensing fees can be redirected to code development that benefits more scientists. In 2006, the developers of PFLOTRAN open-sourced their code under the U.S. Department of Energy SciDAC-II program. Since that time, the code has gained popularity among code developers and users from around the world seeking to employ PFLOTRAN to simulate thermal, hydraulic, mechanical and biogeochemical processes in the Earth's surface/subsurface environment. PFLOTRAN is a massively-parallel subsurface reactive multiphase flow and transport simulator designed from the ground up to run efficiently on computing platforms ranging from the laptop to leadership-class supercomputers, all from a single code base. The code employs domain decomposition for parallelism and is founded upon the well-established and open-source parallel PETSc and HDF5 frameworks. PFLOTRAN leverages modern Fortran (i.e. Fortran 2003-2008) in its extensible object-oriented design. The use of this progressive, yet domain-friendly programming language has greatly facilitated collaboration in the code's software development. Over the past year, PFLOTRAN's top-level data structures were refactored as Fortran classes (i.e. extendible derived types) to improve the flexibility of the code, ease the addition of new process models, and enable coupling to external simulators. For instance, PFLOTRAN has been coupled to the parallel electrical resistivity tomography code E4D to enable hydrogeophysical inversion while the same code base can be used as a third

Coronal mass ejection hits mercury: A.I.K.E.F. hybrid-code results compared to MESSENGER data

Science.gov (United States)

Exner, W.; Heyner, D.; Liuzzo, L.; Motschmann, U.; Shiota, D.; Kusano, K.; Shibayama, T.

2018-04-01

Mercury is the closest orbiting planet around the sun and is therefore embedded in an intensive and highly varying solar wind. In-situ data from the MESSENGER spacecraft of the plasma environment near Mercury indicates that a coronal mass ejection (CME) passed the planet on 23 November 2011 over the span of the 12 h MESSENGER orbit. Slavin et al. (2014) derived the upstream parameters of the solar wind at the time of that orbit, and were able to explain the observed MESSENGER data in the cusp and magnetopause segments of MESSENGER's trajectory. These upstream parameters will be used for our first simulation run. We use the hybrid code A.I.K.E.F. which treats ions as individual particles and electrons as a mass-less fluid, to conduct hybrid simulations of Mercury's magnetospheric response to the impact of the CME on ion gyro time scales. Results from the simulation are in agreement with magnetic field measurements from the inner day-side magnetosphere and the bow-shock region. However, at the planet's nightside, Mercury's plasma environment seemed to be governed by different solar wind conditions, in conclusion, Mercury's interaction with the CME is not sufficiently describable by only one set of upstream parameters. Therefore, to simulate the magnetospheric response while MESSENGER was located in the tail region, we use parameters obtained from the MHD solar wind simulation code SUSANOO (Shiota et al. (2014)) for our second simulation run. The parameters of the SUSANOO model achieve a good agreement of the data concerning the plasma tail crossing and the night-side approach to Mercury. However, the polar and closest approach are hardly described by both upstream parameters, namely, neither upstream dataset is able to reproduce the MESSENGER crossing of Mercury's magnetospheric cusp. We conclude that the respective CME was too variable on the timescale of the MESSENGER orbit to be described by only two sets of upstream conditions. Our results suggest locally strong

Development of DEMO-FNS tokamak for fusion and hybrid technologies

Science.gov (United States)

Kuteev, B. V.; Azizov, E. A.; Alexeev, P. N.; Ignatiev, V. V.; Subbotin, S. A.; Tsibulskiy, V. F.

2015-07-01

The history of fusion-fission hybrid systems based on a tokamak device as an extremely efficient DT-fusion neutron source has passed through several periods of ample research activity in the world since the very beginning of fusion research in the 1950s. Recently, a new roadmap of the hybrid program has been proposed with the goal to build a pilot hybrid plant (PHP) in Russia by 2030. Development of the DEMO-FNS tokamak for fusion and hybrid technologies, which is planned to be built by 2023, is the key milestone on the path to the PHP. This facility is in the phase of conceptual design aimed at providing feasibility studies for a full set of steady state tokamak technologies at a fusion energy gain factor Q ˜ 1, fusion power of ˜40 MW and opportunities for testing a wide range of hybrid technologies with the emphasis on continuous nuclide processing in molten salts. This paper describes the project motivations, its current status and the key issues of the design.

Modelling of data uncertainties on hybrid computers

Energy Technology Data Exchange (ETDEWEB)

Schneider, Anke (ed.)

2016-06-15

The codes d{sup 3}f and r{sup 3}t are well established for modelling density-driven flow and nuclide transport in the far field of repositories for hazardous material in deep geological formations. They are applicable in porous media as well as in fractured rock or mudstone, for modelling salt- and heat transport as well as a free groundwater surface. Development of the basic framework of d{sup 3}f and r{sup 3}t had begun more than 20 years ago. Since that time significant advancements took place in the requirements for safety assessment as well as for computer hardware development. The period of safety assessment for a repository of high-level radioactive waste was extended to 1 million years, and the complexity of the models is steadily growing. Concurrently, the demands on accuracy increase. Additionally, model and parameter uncertainties become more and more important for an increased understanding of prediction reliability. All this leads to a growing demand for computational power that requires a considerable software speed-up. An effective way to achieve this is the use of modern, hybrid computer architectures which requires basically the set-up of new data structures and a corresponding code revision but offers a potential speed-up by several orders of magnitude. The original codes d{sup 3}f and r{sup 3}t were applications of the software platform UG /BAS 94/ whose development had begun in the early nineteennineties. However, UG had recently been advanced to the C++ based, substantially revised version UG4 /VOG 13/. To benefit also in the future from state-of-the-art numerical algorithms and to use hybrid computer architectures, the codes d{sup 3}f and r{sup 3}t were transferred to this new code platform. Making use of the fact that coupling between different sets of equations is natively supported in UG4, d{sup 3}f and r{sup 3}t were combined to one conjoint code d{sup 3}f++. A direct estimation of uncertainties for complex groundwater flow models with the

Development and validation of sodium fire analysis code ASSCOPS

International Nuclear Information System (INIS)

Ohno, Shuji

2001-01-01

A version 2.1 of the ASSCOPS sodium fire analysis code was developed to evaluate the thermal consequences of a sodium leak and consequent fire in LMFBRs. This report describes the computational models and the validation studies using the code. The ASSCOPS calculates sodium droplet and pool fire, and consequential heat/mass transfer behavior. Analyses of sodium pool or spray fire experiments confirmed that this code and parameters used in the validation studies gave valid results on the thermal consequences of sodium leaks and fires. (author)

Collaborative Multi-Layer Network Coding For Hybrid Cellular Cognitive Radio Networks

KAUST Repository

Moubayed, Abdallah J.

2014-05-01

In this thesis, as an extension to [1], we propose a prioritized multi-layer network coding scheme for collaborative packet recovery in hybrid (interweave and underlay) cellular cognitive radio networks. This scheme allows the uncoordinated collaboration between the collocated primary and cognitive radio base-stations in order to minimize their own as well as each other’s packet recovery overheads, thus by improving their throughput. The proposed scheme ensures that each network’s performance is not degraded by its help to the other network. Moreover, it guarantees that the primary network’s interference threshold is not violated in the same and adjacent cells. Yet, the scheme allows the reduction of the recovery overhead in the collocated primary and cognitive radio networks. The reduction in the cognitive radio network is further amplified due to the perfect detection of spectrum holes which allows the cognitive radio base station to transmit at higher power without fear of violating the interference threshold of the primary network. For the secondary network, simulation results show reductions of 20% and 34% in the packet recovery overhead, compared to the non-collaborative scheme, for low and high probabilities of primary packet arrivals, respectively. For the primary network, this reduction was found to be 12%. Furthermore, with the use of fractional cooperation, the average recovery overhead is further reduced by around 5% for the primary network and around 10% for the secondary network when a high fractional cooperation probability is used.

Investigation of Fluctuation-Induced Electron Transport in Hall Thrusters with a 2D Hybrid Code in the Azimuthal and Axial Coordinates

Science.gov (United States)

Fernandez, Eduardo; Borelli, Noah; Cappelli, Mark; Gascon, Nicolas

2003-10-01

Most current Hall thruster simulation efforts employ either 1D (axial), or 2D (axial and radial) codes. These descriptions crucially depend on the use of an ad-hoc perpendicular electron mobility. Several models for the mobility are typically invoked: classical, Bohm, empirically based, wall-induced, as well as combinations of the above. Experimentally, it is observed that fluctuations and electron transport depend on axial distance and operating parameters. Theoretically, linear stability analyses have predicted a number of unstable modes; yet the nonlinear character of the fluctuations and/or their contribution to electron transport remains poorly understood. Motivated by these observations, a 2D code in the azimuthal and axial coordinates has been written. In particular, the simulation self-consistently calculates the azimuthal disturbances resulting in fluctuating drifts, which in turn (if properly correlated with plasma density disturbances) result in fluctuation-driven electron transport. The characterization of the turbulence at various operating parameters and across the channel length is also the object of this study. A description of the hybrid code used in the simulation as well as the initial results will be presented.

Development of eggplant hybrid cultivar ‘BATEM FILIZI’ and determination of yield performance

Directory of Open Access Journals (Sweden)

Hatice Filiz BOYACI

2014-12-01

Full Text Available Eggplant (Solanum melongena L. cultivation is an important part of vegetable production in Turkey. It is highly affected by abiotic stress factors such as low temperatures during growth and development. The present study was aimed to improve new hybrid varieties with high quality and fruit setting under undesirable environmental conditions in eggplant. The 39 inbred lines were developed from eggplant genepool by using pedigree breeding method at the Bati Akdeniz Agricultural Research Institute. They were characterized morphologically and evaluated for phylogenetic relationship among them. Selected seven lines were crossed each other to obtain F1 hybrids. Thirteen hybrids from these crosses were examined for their rate of heterosis and heterobeltiosis. These hybrids were not suitable for market demands although F1 hybrids including P350 as parental line had high heterosis rate. Oval fruit shaped P350 inbred line as female parent showed good performance and crossed with long fruit shaped P599 inbred line. The resultant F1 hybrid cultivar higher yielding, suitable for production under greenhouse and highly adaptable was registered with the name of BATEM FILIZI. It is recommended that this hybrid can be used in greenhouse production especially in single crop season.

Development and benchmark analysis of the hybrid evaluated nuclear data library HENDL1.0

International Nuclear Information System (INIS)

Xu Dezheng; Wu Yican; Gao Chunjing; Zheng Shanliang; Li Jingjing; Zhu Xiaoxiang; Liu Haibo

2004-01-01

To meet the requirements of fusion-fission sub-critical hybrid reactor design and the other related studies, the evaluate nuclear data library named HENDL1.0/E has been constituted based on the several main national evaluated data libraries. The relevant working libraries including transport sub-libraries HENDL1.0/MG in groupwise form, HENDL1.0/MC in pointwise form, and the burnup sub-library HENDL1.0/BU and response function sub-library HENDL1.0/RF are generated using the nuclear data processing codes NJOY97 and TRANSX2. The simulating calculation and comparative analysis are carried out against a series of existing benchmark test experiments with popular neutron transport codes, in order to validate the correctness and availability of the HENDL1.0. (authors)

Development of statistical analysis code for meteorological data (W-View)

International Nuclear Information System (INIS)

Tachibana, Haruo; Sekita, Tsutomu; Yamaguchi, Takenori

2003-03-01

A computer code (W-View: Weather View) was developed to analyze the meteorological data statistically based on 'the guideline of meteorological statistics for the safety analysis of nuclear power reactor' (Nuclear Safety Commission on January 28, 1982; revised on March 29, 2001). The code gives statistical meteorological data to assess the public dose in case of normal operation and severe accident to get the license of nuclear reactor operation. This code was revised from the original code used in a large office computer code to enable a personal computer user to analyze the meteorological data simply and conveniently and to make the statistical data tables and figures of meteorology. (author)

Development of REFLA/TRAC code for engineering work station

International Nuclear Information System (INIS)

Ohnuki, Akira; Akimoto, Hajime; Murao, Yoshio

1994-03-01

The REFLA/TRAC code is a best-estimate code which is expected to check reactor safety analysis codes for light water reactors (LWRs) and to perform accident analyses for LWRs and also for an advanced LWR. Therefore, a high predictive capability is required and the assessment of each physical model becomes important because the models govern the predictive capability. In the case of the assessment of three-dimensional models in REFLA/TRAC code, a conventional large computer is being used and it is difficult to perform the assessment efficiently because the turnaround time for the calculation and the analysis is long. Then, a REFLA/TRAC code which can run on an engineering work station (EWS) was developed. Calculational speed of the current EWS is the same order as that of large computers and the EWS has an excellent function for multidimensional graphical drawings. Besides, the plotting processors for X-Y drawing and for two-dimensional graphical drawing were developed in order to perform efficient analyses for three-dimensional calculations. In future, we can expect that the assessment of three-dimensional models becomes more efficient by introducing an EWS with higher calculational speed and with improved graphical drawings. In this report, each outline for the following three programs is described: (1) EWS version of REFLA/TRAC code, (2) Plot processor for X-Y drawing and (3) Plot processor for two-dimensional graphical drawing. (author)

Human Resource Development in Hybrid Libraries

OpenAIRE

Prakasan, E. R.; Swarna, T.; Vijai Kumar, *

2000-01-01

This paper explores the human resources and development implications in hybrid libraries. Due to technological changes in libraries, which is a result of the proliferation of electronic resources, there has been a shift in workloads and workflow, requiring staff with different skills and educational backgrounds. Training of staff at all levels in information technology is the key to manage change, alleviate anxiety in the workplace and assure quality service in the libraries. Staff developmen...

Development of particle and heavy ion transport code system

International Nuclear Information System (INIS)

Niita, Koji

2004-01-01

Particle and heavy ion transport code system (PHITS) is 3 dimension general purpose Monte Carlo simulation codes for description of transport and reaction of particle and heavy ion in materials. It is developed on the basis of NMTC/JAM for design and safety of J-PARC. What is PHITS, it's physical process, physical models and development process of PHITC code are described. For examples of application, evaluation of neutron optics, cancer treatment by heavy particle ray and cosmic radiation are stated. JAM and JQMD model are used as the physical model. Neutron motion in six polar magnetic field and gravitational field, PHITC simulation of trace of C 12 beam and secondary neutron track of small model of cancer treatment device in HIMAC and neutron flux in Space Shuttle are explained. (S.Y.)

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.

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

Development of AGNES, a kinetics code for fissile solutions, 1

International Nuclear Information System (INIS)

Nakajima, Ken; Ohnishi, Nobuaki

1986-01-01

A kinetics code for fissile solutions, AGNES (Accidentally Generated Nuclear Excursion Simulation code), has been developed. This code calculates the radiolytic gas void effect as a reactivity feedback. Physical and calculative models of the radiolytic gas void are summarized and the usage of AGNES is described. In addition, some benchmark calculations were performed and results of calculations show good agreement with those of experiments. (author)

Development of System Based Code: Case Study of Life-Cycle Margin Evaluation

International Nuclear Information System (INIS)

Tai Asayama; Masaki Morishita; Masanori Tashimo

2006-01-01

For a leap of progress in structural deign of nuclear plant components, The late Professor Emeritus Yasuhide Asada proposed the System Based Code. The key concepts of the System Based Code are; (1) life-cycle margin optimization, (2) expansion of technical options as well as combinations of technical options beyond the current codes and standards, and (3) designing to clearly defined target reliabilities. Those concepts are very new to most of the nuclear power plant designers who are naturally obliged to design to current codes and standards; the application of the concepts of the System Based Code to design will lead to entire change of practices that designers have long been accustomed to. On the other hand, experienced designers are supposed to have expertise that can support and accelerate the development of the System Based Code. Therefore, interfacing with experienced designers is of crucial importance for the development of the System Based Code. The authors conducted a survey on the acceptability of the System Based Code concept. The results were analyzed from the possibility of improving structural design both in terms of reliability and cost effectiveness by the introduction of the System Based Code concept. It was concluded that the System Based Code is beneficial for those purposes. Also described is the expertise elicited from the results of the survey that can be reflected to the development of the System Based Code. (authors)

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)

Development of safety analysis codes for light water reactor

International Nuclear Information System (INIS)

Akimoto, Masayuki

1985-01-01

An overview is presented of currently used major codes for the prediction of thermohydraulic transients in nuclear power plants. The overview centers on the two-phase fluid dynamics of the coolant system and the assessment of the codes. Some of two-phase phenomena such as phase separation are not still predicted with engineering accuracy. MINCS-PIPE are briefly introduced. The MINCS-PIPE code is to assess constitutive relations and to aid development of various experimental correlations for 1V1T model to 2V2T model. (author)

Frontier battery development for hybrid vehicles

OpenAIRE

Lewis, Heather; Park, Haram; Paolini, Maion

2012-01-01

Abstract Background Interest in hybrid-electric vehicles (HEVs) has recently spiked, partly due to an increasingly negative view toward the U.S. foreign oil dependency and environmental concerns. Though HEVs are becoming more common, they have a significant price premium over gasoline-powered vehicles. One of the primary drivers of this “hybrid premium” is the cost of the vehicles’ batteries. This paper focuses on these batteries used in hybrid vehicles, examines the types of batteries used f...

Development of a new EMP code at LANL

Science.gov (United States)

Colman, J. J.; Roussel-Dupré, R. A.; Symbalisty, E. M.; Triplett, L. A.; Travis, B. J.

2006-05-01

A new code for modeling the generation of an electromagnetic pulse (EMP) by a nuclear explosion in the atmosphere is being developed. The source of the EMP is the Compton current produced by the prompt radiation (γ-rays, X-rays, and neutrons) of the detonation. As a first step in building a multi- dimensional EMP code we have written three kinetic codes, Plume, Swarm, and Rad. Plume models the transport of energetic electrons in air. The Plume code solves the relativistic Fokker-Planck equation over a specified energy range that can include ~ 3 keV to 50 MeV and computes the resulting electron distribution function at each cell in a two dimensional spatial grid. The energetic electrons are allowed to transport, scatter, and experience Coulombic drag. Swarm models the transport of lower energy electrons in air, spanning 0.005 eV to 30 keV. The swarm code performs a full 2-D solution to the Boltzmann equation for electrons in the presence of an applied electric field. Over this energy range the relevant processes to be tracked are elastic scattering, three body attachment, two body attachment, rotational excitation, vibrational excitation, electronic excitation, and ionization. All of these occur due to collisions between the electrons and neutral bodies in air. The Rad code solves the full radiation transfer equation in the energy range of 1 keV to 100 MeV. It includes effects of photo-absorption, Compton scattering, and pair-production. All of these codes employ a spherical coordinate system in momentum space and a cylindrical coordinate system in configuration space. The "z" axis of the momentum and configuration spaces is assumed to be parallel and we are currently also assuming complete spatial symmetry around the "z" axis. Benchmarking for each of these codes will be discussed as well as the way forward towards an integrated modern EMP code.

Recent developments in seismic analysis in the code Aster

International Nuclear Information System (INIS)

Guihot, P.; Devesa, G.; Dumond, A.; Panet, M.; Waeckel, F.

1996-01-01

Progress in the field of seismic qualification and design methods made these last few years allows physical phenomena actually in play to be better considered, while cutting down the conservatism associated with some simplified design methods. So following the change in methods and developing the most advantageous ones among them contributes to the process of the seismic margins assessment and the preparation of new design tools for future series. In this paper, the main developments and improvements in methods which have been made these last two years in the Code Aster, in order to improve seismic calculation methods and seismic margin assessment are presented. The first development relates to making the MISS3D soil structure interaction code available, thanks to an interface made with the Code Aster. The second relates to the possibility of making modal basis time calculations on multi-supported structures by considering local non linearities like impact, friction or squeeze fluid forces. Recent developments in random dynamics and postprocessing devoted to earthquake designs are then mentioned. Three applications of these developments are then ut forward. The first application relates to a test case for soil structure interaction design using MISS3D-Aster coupling. The second is a test case for a multi-supported structure. The last application, more for manufacturing, refers to seismic qualification of Main Live Steam stop valves. First results of the independent validation of the Code Aster seismic design functionalities, which provide and improve the quality of software, are also recalled. (authors)

Development of probabilistic fracture mechanics code PASCAL and user's manual

Energy Technology Data Exchange (ETDEWEB)

Shibata, Katsuyuki; Onizawa, Kunio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Li, Yinsheng; Kato, Daisuke [Fuji Research Institute Corporation, Tokyo (Japan)

2001-03-01

As a part of the aging and structural integrity research for LWR components, a new PFM (Probabilistic Fracture Mechanics) code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed since FY1996. This code evaluates the failure probability of an aged reactor pressure vessel subjected to transient loading such as PTS (Pressurized Thermal Shock). The development of the code has been aimed to improve the accuracy and reliability of analysis by introducing new analysis methodologies and algorithms considering the recent development in the fracture mechanics methodologies and computer performance. The code has some new functions in optimized sampling and cell dividing procedure in stratified Monte Carlo simulation, elastic-plastic fracture criterion of R6 method, extension analysis models in semi-elliptical crack, evaluation of effect of thermal annealing and etc. In addition, an input data generator of temperature and stress distribution time histories was also prepared in the code. Functions and performance of the code have been confirmed based on the verification analyses and some case studies on the influence parameters. The present phase of the development will be completed in FY2000. Thus this report provides the user's manual and theoretical background of the code. (author)

Progress in ICRH and lower hybrid launcher development

International Nuclear Information System (INIS)

Kaye, A.S.

1993-01-01

Radio frequency methods of heating and non-inductive current drive have become well established and are likely to be part of any next-step Tokamak programme. The present state of development of antennae for ion cyclotron heating and recent developments to enhance the effectiveness of fast wave current drive systems are reviewed. The performance achieved by present systems enables the provision of an ICRH system for next step devices within the existing technology limits. The main Lower Hybrid current drive systems are also reviewed. Present operating limits suggest that the design power density at the grill in large multijunction launchers must be somewhat reduced due to peaking of the electric field. The resulting launcher for a next step machine based on present technology is a large and highly complex device. Development of recent proposals such as the rod array or the hyperguide, in parallel with necessary improvements in the current drive efficiency, would make Lower Hybrid a more attractive method of non-inductive current drive for next step machines. (Author)

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.

Development of statistical analysis code for meteorological data (W-View)

Energy Technology Data Exchange (ETDEWEB)

Tachibana, Haruo; Sekita, Tsutomu; Yamaguchi, Takenori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2003-03-01

A computer code (W-View: Weather View) was developed to analyze the meteorological data statistically based on 'the guideline of meteorological statistics for the safety analysis of nuclear power reactor' (Nuclear Safety Commission on January 28, 1982; revised on March 29, 2001). The code gives statistical meteorological data to assess the public dose in case of normal operation and severe accident to get the license of nuclear reactor operation. This code was revised from the original code used in a large office computer code to enable a personal computer user to analyze the meteorological data simply and conveniently and to make the statistical data tables and figures of meteorology. (author)

Development of Monte Carlo-based pebble bed reactor fuel management code

International Nuclear Information System (INIS)

Setiadipura, Topan; Obara, Toru

2014-01-01

Highlights: • A new Monte Carlo-based fuel management code for OTTO cycle pebble bed reactor was developed. • The double-heterogeneity was modeled using statistical method in MVP-BURN code. • The code can perform analysis of equilibrium and non-equilibrium phase. • Code-to-code comparisons for Once-Through-Then-Out case were investigated. • Ability of the code to accommodate the void cavity was confirmed. - Abstract: A fuel management code for pebble bed reactors (PBRs) based on the Monte Carlo method has been developed in this study. The code, named Monte Carlo burnup analysis code for PBR (MCPBR), enables a simulation of the Once-Through-Then-Out (OTTO) cycle of a PBR from the running-in phase to the equilibrium condition. In MCPBR, a burnup calculation based on a continuous-energy Monte Carlo code, MVP-BURN, is coupled with an additional utility code to be able to simulate the OTTO cycle of PBR. MCPBR has several advantages in modeling PBRs, namely its Monte Carlo neutron transport modeling, its capability of explicitly modeling the double heterogeneity of the PBR core, and its ability to model different axial fuel speeds in the PBR core. Analysis at the equilibrium condition of the simplified PBR was used as the validation test of MCPBR. The calculation results of the code were compared with the results of diffusion-based fuel management PBR codes, namely the VSOP and PEBBED codes. Using JENDL-4.0 nuclide library, MCPBR gave a 4.15% and 3.32% lower k eff value compared to VSOP and PEBBED, respectively. While using JENDL-3.3, MCPBR gave a 2.22% and 3.11% higher k eff value compared to VSOP and PEBBED, respectively. The ability of MCPBR to analyze neutron transport in the top void of the PBR core and its effects was also confirmed

Windows user-friendly code package development for operation of research reactors

International Nuclear Information System (INIS)

Hoang Anh Tuan

1998-01-01

The content of the project was to developed: 1. MS Windows interface to spectral codes like THERMOS, PEACO-COLLIS, GRACE and burn-up code. 2. MS Windows C-language burn-up diffusion hexagonal lattice code. The overall scope of the project was to develop a PC-based MS Windows code package for operation of Dalat research reactor. Various problems relating to neutronic physics like thermalization, resonance treatment, fast spectral treatment, change of isotopic concentration during burn-up time as well as burn-up distribution in the reactor core are considered in parallel to application of informatics technique. The developing process is a subject of the concept of user-friendly interface between end-users and the code package. High level input features through system of icon, menu, dialog box with regard to Common User Access (CUA) convention and sophisticated graphical output in MS Windows environment was used. The user-computer interface is also enhanced by using both keyboard and mouse, which creates a very natural manner for end-user. (author)

Developing multiplatform apps : using hybrid solutions

OpenAIRE

Kristbergsson, Ellert

2015-01-01

The aim of this thesis is to describe the process of developing multiplatform apps for mobile phones by discussing and comparing the different methods of developing apps for all platforms. The purpose is to ease decision making of smaller companies that are starting to develop their own product for smartphones. The paper analyses tools based on the Cordova framework and the Qt framework. The difference between how the frameworks deliver the code to the machine, Cordova compiles JavaScript...

The newly developed Toyota plug-in hybrid system

Energy Technology Data Exchange (ETDEWEB)

Takaoka, Toshifumi; Ichinose, Hiroki [Toyota Motor Corporation (Japan)

2010-07-01

Toyota has been introducing several hybrid vehicles (HV) as a countermeasure to the automobile's concerns, like CO2 reduction, energy security, and emission reduction in urban areas. A next step towards an even more effective solution for these concerns is a plug-in hybrid vehicle (PHV). This vehicle combines the advantages of electric vehicles (EV), which use clean electric energy, and HV, with it's high environmental potential and user- friendliness comparable to conventional vehicles, such as a long cruising range. This paper describes a newly developed plug-in hybrid system and its vehicle performance. This system uses a Li-ion battery with high energy density and has an affordable EV range without sacrificing cabin space. The vehicle achieves a CO2 emission of 59g/km and meets the most stringent emission regulations in the world. The new PHV is a forerunner of the large-scale mass production PHV two years later. PHVs have the potential to become popular as a realistic solution towards sustainable mobility by renewable electricity usage in the future. (orig.)

A restructuring proposal based on MELCOR for severe accident analysis code development

Energy Technology Data Exchange (ETDEWEB)

Park, Sun Hee; Song, Y. M.; Kim, D. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-03-01

In order to develop a template based on existing MELCOR code, current data saving and transferring methods used in MELCOR are addressed first. Then a naming convention for the constructed module is suggested and an automatic program to convert old variables into new derived type variables has been developed. Finally, a restructured module for the SPR package has been developed to be applied to MELCOR. The current MELCOR code ensures a fixed-size storage for four different data types, and manages the variable-sized data within the storage limit by storing the data on the stacked packages. It uses pointer to identify the variables between the packages. This technique causes a difficult grasping of the meaning of the variables as well as memory waste. New features of FORTRAN90, however, make it possible to allocate the storage dynamically, and to use the user-defined data type which lead to a restructured module development for the SPR package. An efficient memory treatment and as easy understanding of the code are allowed in this developed module. The validation of the template has been done by comparing the results of the modified code with those from the existing code, and it is confirmed that the results are the same. The template for the SPR package suggested in this report hints the extension of the template to the entire code. It is expected that the template will accelerate the code domestication thanks to direct understanding of each variable and easy implementation of modified or newly developed models. 3 refs., 15 figs., 16 tabs. (Author)

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

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.

Near term hybrid passenger vehicle development program, phase 1

Science.gov (United States)

1980-01-01

Missions for hybrid vehicles that promise to yield high petroleum impact were identified and a preliminary design, was developed that satisfies the mission requirements and performance specifications. Technologies that are critical to successful vehicle design, development and fabrication were determined. Trade-off studies to maximize fuel savings were used to develop initial design specifications of the near term hybrid vehicle. Various designs were "driven" through detailed computer simulations which calculate the petroleum consumption in standard driving cycles, the petroleum and electricity consumptions over the specified missions, and the vehicle's life cycle costs over a 10 year vehicle lifetime. Particular attention was given to the selection of the electric motor, heat engine, drivetrain, battery pack and control system. The preliminary design reflects a modified current compact car powered by a currently available turbocharged diesel engine and a 24 kW (peak) compound dc electric motor.

Sub-channel/system coupled code development and its application to SCWR-FQT loop

International Nuclear Information System (INIS)

Liu, X.J.; Cheng, X.

2015-01-01

Highlights: • A coupled code is developed for SCWR accident simulation. • The feasibility of the code is shown by application to SCWR-FQT loop. • Some measures are selected by sensitivity analysis. • The peak cladding temperature can be reduced effectively by the proposed measures. - Abstract: In the frame of Super-Critical Reactor In Pipe Test Preparation (SCRIPT) project in China, one of the challenge tasks is to predict the transient performance of SuperCritical Water Reactor-Fuel Qualification Test (SCWR-FQT) loop under some accident conditions. Several thermal–hydraulic codes (system code, sub-channel code) are selected to perform the safety analysis. However, the system code cannot simulate the local behavior of the test bundle, and the sub-channel code is incapable of calculating the whole system behavior of the test loop. Therefore, to combine the merits of both codes, and minimizes their shortcomings, a coupled sub-channel and system code system is developed in this paper. Both of the sub-channel code COBRA-SC and system code ATHLET-SC are adapted to transient analysis of SCWR. Two codes are coupled by data transfer and data adaptation at the interface. In the new developed coupled code, the whole system behavior including safety system characteristic is analyzed by system code ATHLET-SC, whereas the local thermal–hydraulic parameters are predicted by the sub-channel code COBRA-SC. The codes are utilized to get the local thermal–hydraulic parameters in the SCWR-FQT fuel bundle under some accident case (e.g. a flow blockage during LOCA). Some measures to mitigate the accident consequence are proposed by the sensitivity study and trialed to demonstrate their effectiveness in the coupled simulation. The results indicate that the new developed code has good feasibility to transient analysis of supercritical water-cooled test. And the peak cladding temperature caused by blockage in the fuel bundle can be reduced effectively by the safety measures

Sub-channel/system coupled code development and its application to SCWR-FQT loop

Energy Technology Data Exchange (ETDEWEB)

Liu, X.J., E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Cheng, X. [Institute of Fusion and Reactor Technology, Karlsruhe Institute of Technology, Vincenz-Prießnitz-Str. 3, 76131 Karlsruhe (Germany)

2015-04-15

Highlights: • A coupled code is developed for SCWR accident simulation. • The feasibility of the code is shown by application to SCWR-FQT loop. • Some measures are selected by sensitivity analysis. • The peak cladding temperature can be reduced effectively by the proposed measures. - Abstract: In the frame of Super-Critical Reactor In Pipe Test Preparation (SCRIPT) project in China, one of the challenge tasks is to predict the transient performance of SuperCritical Water Reactor-Fuel Qualification Test (SCWR-FQT) loop under some accident conditions. Several thermal–hydraulic codes (system code, sub-channel code) are selected to perform the safety analysis. However, the system code cannot simulate the local behavior of the test bundle, and the sub-channel code is incapable of calculating the whole system behavior of the test loop. Therefore, to combine the merits of both codes, and minimizes their shortcomings, a coupled sub-channel and system code system is developed in this paper. Both of the sub-channel code COBRA-SC and system code ATHLET-SC are adapted to transient analysis of SCWR. Two codes are coupled by data transfer and data adaptation at the interface. In the new developed coupled code, the whole system behavior including safety system characteristic is analyzed by system code ATHLET-SC, whereas the local thermal–hydraulic parameters are predicted by the sub-channel code COBRA-SC. The codes are utilized to get the local thermal–hydraulic parameters in the SCWR-FQT fuel bundle under some accident case (e.g. a flow blockage during LOCA). Some measures to mitigate the accident consequence are proposed by the sensitivity study and trialed to demonstrate their effectiveness in the coupled simulation. The results indicate that the new developed code has good feasibility to transient analysis of supercritical water-cooled test. And the peak cladding temperature caused by blockage in the fuel bundle can be reduced effectively by the safety measures

The development, qualification and availability of AECL analytical, scientific and design codes

International Nuclear Information System (INIS)

Kupferschmidt, W.C.H.; Fehrenbach, P.J.; Wolgemuth, G.A.; McDonald, B.H.; Snell, V.G.

2001-01-01

Over the past several years, AECL has embarked on a comprehensive program to develop, qualify and support its key safety and licensing codes, and to make executable versions of these codes available to the international nuclear community. To this end, we have instituted a company-wide Software Quality Assurance (SQA) Program for Analytical, Scientific and Design Computer Programs to ensure that the design, development, maintenance, modification, procurement and use of computer codes within AECL is consistent with today's quality assurance standards. In addition, we have established a comprehensive Code Validation Project (CVP) with the goal of qualifying AECL's 'front-line' safety and licensing codes by 2001 December. The outcome of this initiative will be qualified codes, which are properly verified and validated for the expected range of applications, with associated statements of accuracy and uncertainty for each application. The code qualification program, based on the CSA N286.7 standard, is intended to ensure (1) that errors are not introduced into safety analyses because of deficiencies in the software, (2) that an auditable documentation base is assembled that demonstrates to the regulator that the codes are of acceptable quality, and (3) that these codes are formally qualified for their intended applications. Because AECL and the Canadian nuclear utilities (i.e., Ontario Power Generation, Bruce Power, Hydro Quebec and New Brunswick Power) generally use the same safety and licensing codes, the nuclear industry in Canada has agreed to work cooperatively together towards the development, qualification and maintenance of a common set of analysis tools, referred to as the Industry Standard Toolset (IST). This paper provides an overview of the AECL Software Quality Assurance Program and the Code Validation Project, and their associated linkages to the Canadian nuclear community's Industry Standard Toolset initiative to cooperatively qualify and support commonly

Development of the integrated system reliability analysis code MODULE

International Nuclear Information System (INIS)

Han, S.H.; Yoo, K.J.; Kim, T.W.

1987-01-01

The major components in a system reliability analysis are the determination of cut sets, importance measure, and uncertainty analysis. Various computer codes have been used for these purposes. For example, SETS and FTAP are used to determine cut sets; Importance for importance calculations; and Sample, CONINT, and MOCUP for uncertainty analysis. There have been problems when the codes run each other and the input and output are not linked, which could result in errors when preparing input for each code. The code MODULE was developed to carry out the above calculations simultaneously without linking input and outputs to other codes. MODULE can also prepare input for SETS for the case of a large fault tree that cannot be handled by MODULE. The flow diagram of the MODULE code is shown. To verify the MODULE code, two examples are selected and the results and computation times are compared with those of SETS, FTAP, CONINT, and MOCUP on both Cyber 170-875 and IBM PC/AT. Two examples are fault trees of the auxiliary feedwater system (AFWS) of Korea Nuclear Units (KNU)-1 and -2, which have 54 gates and 115 events, 39 gates and 92 events, respectively. The MODULE code has the advantage that it can calculate the cut sets, importances, and uncertainties in a single run with little increase in computing time over other codes and that it can be used in personal computers

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)

Agronomic potentials of quality protein maize hybrids developed in ...

African Journals Online (AJOL)

Agronomic potentials of quality protein maize hybrids developed in Ghana. ... Ghana Journal of Agricultural Science ... and Wheat Improvement (CIMMYT) were evaluated on research stations and in farmers\\' fields in Ghana from 1995 to 1996.

Development of coupling systems at the hybrid frequency for the non-inductive current generation inside a tokamak; Developpement de coupleurs a la frequence hybride pour la generation non inductive du courant dans un tokamak

Energy Technology Data Exchange (ETDEWEB)

Berio, S. [Association Euratom-CEA, Centre d`Etudes Nucleaires de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee]|[Aix-Marseille-1 Univ., 13 - Marseille (France)

1996-12-31

Used at its first time as an heating method in order to reach the temperature requisite for the fusion of a thermonuclear plasma, the hybrid waves has shown that they were the more efficient method for non-inductive current drive in a tokamak. The size and the objectives of a next machine such as ITER lead to the design of new antennae (in process of realisation on Tore Supra) made of oversized waveguides. This new concept of antenna will be more simple, more robust and will be able to transmit the same if not much power than the present antennae. This thesis contribute to the development of a new code called ALOHA (for `Advanced LOwer Hybrid Antenna`) which, at the end, will be able to give the characteristics and the behaviours of this new oversized antennae in front of a tokamak plasma. This thesis is also a first step in the interpretation of some experimental data concerning the measurement of coupling, absorption and current drive of the actual hybrid wave launched by a grill with rectangular waveguides. Moreover, this thesis lay some foundations of the study of these new antennae in front of a on-parallel confinement magnetic field and/or in front of poloidal inhomogeneities of plasma. (authors) 53 refs.

Object-Oriented Programming in the Development of Containment Analysis Code

International Nuclear Information System (INIS)

Han, Tae Young; Hong, Soon Joon; Hwang, Su Hyun; Lee, Byung Chul; Byun, Choong Sup

2009-01-01

After the mid 1980s, the new programming concept, Object-Oriented Programming (OOP), was introduced and designed, which has the features such as the information hiding, encapsulation, modularity and inheritance. These offered much more convenient programming paradigm to code developers. The OOP concept was readily developed into the programming language as like C++ in the 1990s and is being widely used in the modern software industry. In this paper, we show that the OOP concept is successfully applicable to the development of safety analysis code for containment and propose the more explicit and easy OOP design for developers

Development and application of a hybrid transport methodology for active interrogation systems

Energy Technology Data Exchange (ETDEWEB)

Royston, K.; Walters, W.; Haghighat, A. [Nuclear Engineering Program, Department of Mechanical Engineering, Virginia Tech., 900 N Glebe Rd., Arlington, VA 22203 (United States); Yi, C.; Sjoden, G. [Nuclear and Radiological Engineering, Georgia Tech, 801 Ferst Drive, Atlanta, GA 30332 (United States)

2013-07-01

A hybrid Monte Carlo and deterministic methodology has been developed for application to active interrogation systems. The methodology consists of four steps: i) neutron flux distribution due to neutron source transport and subcritical multiplication; ii) generation of gamma source distribution from (n, 7) interactions; iii) determination of gamma current at a detector window; iv) detection of gammas by the detector. This paper discusses the theory and results of the first three steps for the case of a cargo container with a sphere of HEU in third-density water cargo. To complete the first step, a response-function formulation has been developed to calculate the subcritical multiplication and neutron flux distribution. Response coefficients are pre-calculated using the MCNP5 Monte Carlo code. The second step uses the calculated neutron flux distribution and Bugle-96 (n, 7) cross sections to find the resulting gamma source distribution. In the third step the gamma source distribution is coupled with a pre-calculated adjoint function to determine the gamma current at a detector window. The AIMS (Active Interrogation for Monitoring Special-Nuclear-Materials) software has been written to output the gamma current for a source-detector assembly scanning across a cargo container using the pre-calculated values and taking significantly less time than a reference MCNP5 calculation. (authors)

Multi-dimensional Code Development for Safety Analysis of LMR

International Nuclear Information System (INIS)

Ha, K. S.; Jeong, H. Y.; Kwon, Y. M.; Lee, Y. B.

2006-08-01

A liquid metal reactor loaded a metallic fuel has the inherent safety mechanism due to the several negative reactivity feedback. Although this feature demonstrated through experiments in the EBR-II, any of the computer programs until now did not exactly analyze it because of the complexity of the reactivity feedback mechanism. A multi-dimensional detail program was developed through the International Nuclear Energy Research Initiative(INERI) from 2003 to 2005. This report includes the numerical coupling the multi-dimensional program and SSC-K code which is used to the safety analysis of liquid metal reactors in KAERI. The coupled code has been proved by comparing the analysis results using the code with the results using SAS-SASSYS code of ANL for the UTOP, ULOF, and ULOHS applied to the safety analysis for KALIMER-150

Development of the CRIPTE Code for Electromagnetic Coupling

National Research Council Canada - National Science Library

Parmantier, Jean-Philippe

2005-01-01

.... This code was originally developed as part of an experiment performed under the joint US-France international data exchange program on the atmospheric electricity/aircraft interactions, DEA-AF-79-7336...

THE STAGES OF HETEROTIC HYBRIDS F1 DEVELOPMENT IN EUROPEAN RADISH

Directory of Open Access Journals (Sweden)

M. A. Kosenko

2017-01-01

Full Text Available The scheme  of  development  of  two-line  of  hybrids  F1  in European radish based on self-incompatibility includes five stages, as follows: 1 – selection of self-incompatible lines, common and specified combining ability estimation; 2 – inbreeding and selection to make the lines homozygous for morphological traits, common and specified combining ability estimation; maintenance and reproduction of self-incompatible lines; 4 – production of hybrid seeds. The research work on assessment of hybrid F1 that were obtained from cross of eight self-incompatible lines of European winter radish by the Griffing’s method was carried out in 2016. The assessment of length, diameter and yield of radish root was performed. According to the root shape the heterotic hybrids F1 were divided into three groups: rounded-flat,  48.2%; round, 50.0%; and flatten-round,  1.8%. The level of root marketability of hybrids F1 reached 100%. As a result of the work the promising hybrid combination distinguished by high uniformity, marketability and high yield were selected out.

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.

Safety analysis code SCTRAN development for SCWR and its application to CGNPC SCWR

International Nuclear Information System (INIS)

Wu, Pan; Gou, Junli; Shan, Jianqiang; Jiang, Yang; Yang, Jue; Zhang, Bo

2013-01-01

Highlights: ► A new safety analysis code named SCTRAN is developed for SCWRs. ► Capability of SCTRAN is verified by comparing with code APROS and RELAP5-3D. ► A new passive safety system is proposed for CGNPC SCWR and analyzed with SCTRAN. ► CGNPC SCWR is able to cope with two critical accidents for SCWRs, LOFA and LOCA. - Abstract: Design analysis is one of the main difficulties during the research and design of SCWRs. Currently, the development of safety analysis code for SCWR is still in its infancy all around the world, and very few computer codes could carry out the trans-critical calculations where significant changes in water properties would take place. In this paper, a safety analysis code SCTRAN for SCWRs has been developed based on code RETRAN-02, the best estimate code used for safety analysis of light water reactors. The ability of SCTRAN code to simulate transients where both supercritical and subcritical regimes are encountered has been verified by comparing with APROS and RELAP5-3D codes. Furthermore, the LOFA and LOCA transients for the CGNPC SCWR design were analyzed with SCTRAN code. The characteristics and performance of the passive safety systems applied to CGNPC SCWR were evaluated. The results show that: (1) The SCTRAN computer code developed in this study is capable to perform design analysis for SCWRs; (2) During LOFA and LOCA accidents in a CGNPC SCWR, the passive safety systems would significantly mitigate the consequences of these transients and enhance the inherent safety

Hybrid rendering of the chest and virtual bronchoscopy [corrected].

Science.gov (United States)

Seemann, M D; Seemann, O; Luboldt, W; Gebicke, K; Prime, G; Claussen, C D

2000-10-30

Thin-section spiral computed tomography was used to acquire the volume data sets of the thorax. The tracheobronchial system and pathological changes of the chest were visualized using a color-coded surface rendering method. The structures of interest were then superimposed on a volume rendering of the other thoracic structures, thus producing a hybrid rendering. The hybrid rendering technique exploit the advantages of both rendering methods and enable virtual bronchoscopic examinations using different representation models. Virtual bronchoscopic examinations with a transparent color-coded shaded-surface model enables the simultaneous visualization of both the airways and the adjacent structures behind of the tracheobronchial wall and therefore, offers a practical alternative to fiberoptic bronchoscopy. Hybrid rendering and virtual endoscopy obviate the need for time consuming detailed analysis and presentation of axial source images.

BBU code development for high-power microwave generators

International Nuclear Information System (INIS)

Houck, T.L.; Westenskow, G.A.; Yu, S.S.

1992-01-01

We are developing a two-dimensional, time-dependent computer code for the simulation of transverse instabilities in support of relativistic klystron-two beam accelerator research at LLNL. The code addresses transient effects as well as both cumulative and regenerative beam breakup modes. Although designed specifically for the transport of high current (kA) beams through traveling-wave structures, it is applicable to devices consisting of multiple combinations of standing-wave, traveling-wave, and induction accelerator structures. In this paper we compare code simulations to analytical solutions for the case where there is no rf coupling between cavities, to theoretical scaling parameters for coupled cavity structures, and to experimental data involving beam breakup in the two traveling-wave output structure of our microwave generator. (Author) 4 figs., tab., 5 refs

Efficient DS-UWB MUD Algorithm Using Code Mapping and RVM

Directory of Open Access Journals (Sweden)

Pingyan Shi

2016-01-01

Full Text Available A hybrid multiuser detection (MUD using code mapping and a wrong code recognition based on relevance vector machine (RVM for direct sequence ultra wide band (DS-UWB system is developed to cope with the multiple access interference (MAI and the computational efficiency. A new MAI suppression mechanism is studied in the following steps: firstly, code mapping, an optimal decision function, is constructed and the output candidate code of the matched filter is mapped to a feature space by the function. In the feature space, simulation results show that the error codes caused by MAI and the single user mapped codes can be classified by a threshold which is related to SNR of the receiver. Then, on the base of code mapping, use RVM to distinguish the wrong codes from the right ones and finally correct them. Compared with the traditional MUD approaches, the proposed method can considerably improve the bit error ratio (BER performance due to its special MAI suppression mechanism. Simulation results also show that the proposed method can approximately achieve the BER performance of optimal multiuser detection (OMD and the computational complexity approximately equals the matched filter. Moreover, the proposed method is less sensitive to the number of users.

Twelve gordian knots when developing an organizational code of ethics

NARCIS (Netherlands)

Kaptein, Muel; Wempe, Johan

1998-01-01

Following the example of the many organizations in the United States which have a code of ethics, an increasing interest on the part of companies, trade organizations, (semi-)governmental organizations and professions in the Netherlands to develop codes of ethics can be witnessed. We have been able

TRAC code development status and plans

International Nuclear Information System (INIS)

Spore, J.W.; Liles, D.R.; Nelson, R.A.

1986-01-01

This report summarizes the characteristics and current status of the TRAC-PF1/MOD1 computer code. Recent error corrections and user-convenience features are described, and several user enhancements are identified. Current plans for the release of the TRAC-PF1/MOD2 computer code and some preliminary MOD2 results are presented. This new version of the TRAC code implements stability-enhancing two-step numerics into the 3-D vessel, using partial vectorization to obtain a code that has run 400% faster than the MOD1 code

Development of bio-hybrid material based on Salmonella ...

African Journals Online (AJOL)

Teodoro

2016-07-13

Jul 13, 2016 ... Full Length Research Paper. Development of bio-hybrid material based on. Salmonella Typhimurium and layered double hydroxides. Slah Hidouri .... the LDH with co-precipitation synthesis method was successfully done according the study given by Hidouri et al. (2011), Abdelkader et al. (2011), Hidouri et ...

Development of a coupling code for PWR reactor cavity radiation streaming calculation

International Nuclear Information System (INIS)

Zheng, Z.; Wu, H.; Cao, L.; Zheng, Y.; Zhang, H.; Wang, M.

2012-01-01

PWR reactor cavity radiation streaming is important for the safe of the personnel and equipment, thus calculation has to be performed to evaluate the neutron flux distribution around the reactor. For this calculation, the deterministic codes have difficulties in fine geometrical modeling and need huge computer resource; and the Monte Carlo codes require very long sampling time to obtain results with acceptable precision. Therefore, a coupling method has been developed to eliminate the two problems mentioned above in each code. In this study, we develop a coupling code named DORT2MCNP to link the Sn code DORT and Monte Carlo code MCNP. DORT2MCNP is used to produce a combined surface source containing top, bottom and side surface simultaneously. Because SDEF card is unsuitable for the combined surface source, we modify the SOURCE subroutine of MCNP and compile MCNP for this application. Numerical results demonstrate the correctness of the coupling code DORT2MCNP and show reasonable agreement between the coupling method and the other two codes (DORT and MCNP). (authors)

Development of hybrid artificial intelligent based handover decision algorithm

Directory of Open Access Journals (Sweden)

A.M. Aibinu

2017-04-01

Full Text Available The possibility of seamless handover remains a mirage despite the plethora of existing handover algorithms. The underlying factor responsible for this has been traced to the Handover decision module in the Handover process. Hence, in this paper, the development of novel hybrid artificial intelligent handover decision algorithm has been developed. The developed model is made up of hybrid of Artificial Neural Network (ANN based prediction model and Fuzzy Logic. On accessing the network, the Received Signal Strength (RSS was acquired over a period of time to form a time series data. The data was then fed to the newly proposed k-step ahead ANN-based RSS prediction system for estimation of prediction model coefficients. The synaptic weights and adaptive coefficients of the trained ANN was then used to compute the k-step ahead ANN based RSS prediction model coefficients. The predicted RSS value was later codified as Fuzzy sets and in conjunction with other measured network parameters were fed into the Fuzzy logic controller in order to finalize handover decision process. The performance of the newly developed k-step ahead ANN based RSS prediction algorithm was evaluated using simulated and real data acquired from available mobile communication networks. Results obtained in both cases shows that the proposed algorithm is capable of predicting ahead the RSS value to about ±0.0002 dB. Also, the cascaded effect of the complete handover decision module was also evaluated. Results obtained show that the newly proposed hybrid approach was able to reduce ping-pong effect associated with other handover techniques.

Development of Visual CINDER Code with Visual C⧣.NET

International Nuclear Information System (INIS)

Kim, Oyeon

2016-01-01

CINDER code, CINDER' 90 or CINDER2008 that is integrated with the Monte Carlo code, MCNPX, is widely used to calculate the inventory of nuclides in irradiated materials. The MCNPX code provides decay processes to the particle transport scheme that traditionally only covered prompt processes. The integration schemes serve not only the reactor community (MCNPX burnup) but also the accelerator community as well (residual production information). The big benefit for providing these options lies in the easy cross comparison of the transmutation codes since the calculations are based on exactly the same material, neutron flux and isotope production/destruction inputs. However, it is just frustratingly cumbersome to use. In addition, multiple human interventions may increase the possibility of making errors. The number of significant digits in the input data varies in steps, which may cause big errors for highly nonlinear problems. Thus, it is worthwhile to find a new way to wrap all the codes and procedures in one consistent package which can provide ease of use. The visual CINDER code development is underway with visual C .NET framework. It provides a few benefits for the atomic transmutation simulation with CINDER code. A few interesting and useful properties of visual C .NET framework are introduced. We also showed that the wrapper could make the simulation accurate for highly nonlinear transmutation problems and also increase the possibility of direct combination a radiation transport code MCNPX with CINDER code. Direct combination of CINDER with MCNPX in a wrapper will provide more functionalities for the radiation shielding and prevention study

Development of Visual CINDER Code with Visual C⧣.NET

Energy Technology Data Exchange (ETDEWEB)

Kim, Oyeon [Institute for Modeling and Simulation Convergence, Daegu (Korea, Republic of)

2016-10-15

CINDER code, CINDER' 90 or CINDER2008 that is integrated with the Monte Carlo code, MCNPX, is widely used to calculate the inventory of nuclides in irradiated materials. The MCNPX code provides decay processes to the particle transport scheme that traditionally only covered prompt processes. The integration schemes serve not only the reactor community (MCNPX burnup) but also the accelerator community as well (residual production information). The big benefit for providing these options lies in the easy cross comparison of the transmutation codes since the calculations are based on exactly the same material, neutron flux and isotope production/destruction inputs. However, it is just frustratingly cumbersome to use. In addition, multiple human interventions may increase the possibility of making errors. The number of significant digits in the input data varies in steps, which may cause big errors for highly nonlinear problems. Thus, it is worthwhile to find a new way to wrap all the codes and procedures in one consistent package which can provide ease of use. The visual CINDER code development is underway with visual C .NET framework. It provides a few benefits for the atomic transmutation simulation with CINDER code. A few interesting and useful properties of visual C .NET framework are introduced. We also showed that the wrapper could make the simulation accurate for highly nonlinear transmutation problems and also increase the possibility of direct combination a radiation transport code MCNPX with CINDER code. Direct combination of CINDER with MCNPX in a wrapper will provide more functionalities for the radiation shielding and prevention study.

The in-core fuel management code system for VVER reactors

International Nuclear Information System (INIS)

Cada, R.; Krysl, V.; Mikolas, P.; Sustek, J.; Svarny, J.

2004-01-01

The structure and methodology of a fuel management system for NPP VVER 1000 (NPP Temelin) and VVER 440 (NPP Dukovany) is described. It is under development in SKODA JS a.s. and is followed by practical applications. The general objectives of the system are maximization of end of cycle reactivity, the minimization of fresh fuel inventory for the minimization of fed enrichment and minimization of burnable poisons (BPs) inventory. They are also safety related constraints in witch minimization of power peaking plays a dominant role. General structure of the system consists in preparation of input data for macrocode calculation, algorithms (codes) for optimization of fuel loading, calculation of fuel enrichment and BPs assignment. At present core loading can be calculated (optimized) by Tabu search algorithm (code ATHENA), genetic algorithm (code Gen1) and hybrid algorithm - simplex procedure with application of Tabu search algorithm on binary shuffling (code OPAL B ). Enrichment search is realized by the application of simplex algorithm (OPAL B code) and BPs assignment by module BPASS and simplex algorithm in OPAL B code. Calculations of the real core loadings are presented and a comparison of different optimization methods is provided. (author)

Development of general-purpose particle and heavy ion transport monte carlo code

International Nuclear Information System (INIS)

Iwase, Hiroshi; Nakamura, Takashi; Niita, Koji

2002-01-01

The high-energy particle transport code NMTC/JAM, which has been developed at JAERI, was improved for the high-energy heavy ion transport calculation by incorporating the JQMD code, the SPAR code and the Shen formula. The new NMTC/JAM named PHITS (Particle and Heavy-Ion Transport code System) is the first general-purpose heavy ion transport Monte Carlo code over the incident energies from several MeV/nucleon to several GeV/nucleon. (author)

Uranium resources and their implications for fission breeder and fusion hybrid development

International Nuclear Information System (INIS)

Max, C.E.

1984-01-01

Present estimates of uranium resources and reserves in the US and the non-Communist world are reviewed. The resulting implications are considered for two proposed breeder technologies: the liquid metal fast breeder reactor (LMFBR) and the fusion hybrid reactor. Using both simple arguments and detailed scenarios from the published literature, conditions are explored under which the LMFBR and fusion hybrid could respectively have the most impact, considering both fuel-supply and economic factors. The conclusions emphasize strong potential advantages of the fusion hybrid, due to its inherently large breeding rate. A discussion is presented of proposed US development strategies for the fusion hybrid, which at present is far behind the LMFBR in its practical application and maturity

Performance Evaluation of HARQ Technique with UMTS Turbo Code

Directory of Open Access Journals (Sweden)

S. S. Brkić

2011-11-01

Full Text Available The hybrid automatic repeat request technique (HARQ represents the error control principle which combines an error correcting code and automatic repeat request procedure (ARQ, within the same transmission system. In this paper, using Monte Carlo simulation process, the characteristics of HARQ technique are determined, for the case of the Universal Mobile Telecommunication System (UMTS turbo code.

The development of hybrid vehicles for the Indian market

Energy Technology Data Exchange (ETDEWEB)

Wadhera, R.; Srinivas, A. [Mahindra and Mahindra Ltd., Mumbai (India); Ellinger, R.; Sattler, M. [AVL List GmbH, Graz (Austria)

2012-11-01

The passenger vehicle market in India continues to witness a steady growth rate and expected to do so in the future. While the growth rate is primarily driven by the conventional Internal Combustion based engines, Indian automotive industry is exploring various technological options to comply imminent regulations, remain competitive, reduce emissions and ownership cost. Based on the road trials conducted in various parts of India and subsequent analysis of the traffic characteristics, Hybrid Electric Vehicle promises to be appropriate technology for the near future so as to improve fuel efficiency, performance and reduced emissions. Mahindra and Mahindra has established a clear roadmap for development of HEV. Hybrid Electric Vehicles (HEV's) are seen as a major technology route in order to fulfill future automotive demands with regard to fuel efficiency, emissions, performance and drivability. Mahindra and Mahindra has therefore set-up a clear strategy and roadmap for the development and implementation of HEV's in their product portfolio. The development of hybrid electrical vehicle is also confronted by numerous technical challenges such as Thermal Management, Safety Concept, durability requirements and Noise vibration and Harshness. Also the cost-effectiveness needs to be ensured so as to have a bright market penetration and a viable business. This lecture covers the passenger vehicle sales statistics in India, prevailing traffic characteristics in various types of roads, environmental boundaries and challenges encountered for Indian Market. (orig.)

Hybrid Software and System Development in Practice: Waterfall, Scrum, and Beyond

DEFF Research Database (Denmark)

Kuhrmann, Marco; Diebold, Philipp; Münch, Jürgen

2017-01-01

Software and system development faces numerous challenges of rapidly changing markets. To address such challenges, companies and projects design and adopt specific development approaches by combining well-structured comprehensive methods and flexible agile practices. Yet, the number of methods...... and practices is large, and available studies argue that the actual process composition is carried out in a fairly ad-hoc manner. The present paper reports on a survey on hybrid software development approaches. We study which approaches are used in practice, how different approaches are combined, and what...... contextual factors influence the use and combination of hybrid software development approaches. Our results from 69 study participants show a variety of development approaches used and combined in practice. We show that most combinations follow a pattern in which a traditional process model serves...

Simulations for the transmutation of nuclear wastes with hybrid reactors

International Nuclear Information System (INIS)

Vuillier, St.

1998-06-01

A Monte Carlo simulation, devoted to the spallation, has been built in the framework of the hybrid systems proposed for the nuclear wastes incineration. This system GSPARTE, described the reactions evolution. It takes into account and improves the nuclear codes and the low and high energy particles transport in the GEANT code environment, adapted to the geometry of the hybrid reactors. Many applications and abacus useful for the wastes transmutation, have been realized with this system: production of thick target neutrons, source definition, material damages. (A.L.B.)

The development of depletion program coupled with Monte Carlo computer code

International Nuclear Information System (INIS)

Nguyen Kien Cuong; Huynh Ton Nghiem; Vuong Huu Tan

2015-01-01

The paper presents the development of depletion code for light water reactor coupled with MCNP5 code called the MCDL code (Monte Carlo Depletion for Light Water Reactor). The first order differential depletion system equations of 21 actinide isotopes and 50 fission product isotopes are solved by the Radau IIA Implicit Runge Kutta (IRK) method after receiving neutron flux, reaction rates in one group energy and multiplication factors for fuel pin, fuel assembly or whole reactor core from the calculation results of the MCNP5 code. The calculation for beryllium poisoning and cooling time is also integrated in the code. To verify and validate the MCDL code, high enriched uranium (HEU) and low enriched uranium (LEU) fuel assemblies VVR-M2 types and 89 fresh HEU fuel assemblies, 92 LEU fresh fuel assemblies cores of the Dalat Nuclear Research Reactor (DNRR) have been investigated and compared with the results calculated by the SRAC code and the MCNP R EBUS linkage system code. The results show good agreement between calculated data of the MCDL code and reference codes. (author)

Development of Probabilistic Internal Dosimetry Computer Code

Energy Technology Data Exchange (ETDEWEB)

Noh, Siwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kwon, Tae-Eun [Korea Institute of Radiological and Medical Sciences, Seoul (Korea, Republic of); Lee, Jai-Ki [Korean Association for Radiation Protection, Seoul (Korea, Republic of)

2017-02-15

Internal radiation dose assessment involves biokinetic models, the corresponding parameters, measured data, and many assumptions. Every component considered in the internal dose assessment has its own uncertainty, which is propagated in the intake activity and internal dose estimates. For research or scientific purposes, and for retrospective dose reconstruction for accident scenarios occurring in workplaces having a large quantity of unsealed radionuclides, such as nuclear power plants, nuclear fuel cycle facilities, and facilities in which nuclear medicine is practiced, a quantitative uncertainty assessment of the internal dose is often required. However, no calculation tools or computer codes that incorporate all the relevant processes and their corresponding uncertainties, i.e., from the measured data to the committed dose, are available. Thus, the objective of the present study is to develop an integrated probabilistic internal-dose-assessment computer code. First, the uncertainty components in internal dosimetry are identified, and quantitative uncertainty data are collected. Then, an uncertainty database is established for each component. In order to propagate these uncertainties in an internal dose assessment, a probabilistic internal-dose-assessment system that employs the Bayesian and Monte Carlo methods. Based on the developed system, we developed a probabilistic internal-dose-assessment code by using MATLAB so as to estimate the dose distributions from the measured data with uncertainty. Using the developed code, we calculated the internal dose distribution and statistical values (e.g. the 2.5{sup th}, 5{sup th}, median, 95{sup th}, and 97.5{sup th} percentiles) for three sample scenarios. On the basis of the distributions, we performed a sensitivity analysis to determine the influence of each component on the resulting dose in order to identify the major component of the uncertainty in a bioassay. The results of this study can be applied to various

Development of Probabilistic Internal Dosimetry Computer Code

International Nuclear Information System (INIS)

Noh, Siwan; Kwon, Tae-Eun; Lee, Jai-Ki

2017-01-01

Internal radiation dose assessment involves biokinetic models, the corresponding parameters, measured data, and many assumptions. Every component considered in the internal dose assessment has its own uncertainty, which is propagated in the intake activity and internal dose estimates. For research or scientific purposes, and for retrospective dose reconstruction for accident scenarios occurring in workplaces having a large quantity of unsealed radionuclides, such as nuclear power plants, nuclear fuel cycle facilities, and facilities in which nuclear medicine is practiced, a quantitative uncertainty assessment of the internal dose is often required. However, no calculation tools or computer codes that incorporate all the relevant processes and their corresponding uncertainties, i.e., from the measured data to the committed dose, are available. Thus, the objective of the present study is to develop an integrated probabilistic internal-dose-assessment computer code. First, the uncertainty components in internal dosimetry are identified, and quantitative uncertainty data are collected. Then, an uncertainty database is established for each component. In order to propagate these uncertainties in an internal dose assessment, a probabilistic internal-dose-assessment system that employs the Bayesian and Monte Carlo methods. Based on the developed system, we developed a probabilistic internal-dose-assessment code by using MATLAB so as to estimate the dose distributions from the measured data with uncertainty. Using the developed code, we calculated the internal dose distribution and statistical values (e.g. the 2.5 th , 5 th , median, 95 th , and 97.5 th percentiles) for three sample scenarios. On the basis of the distributions, we performed a sensitivity analysis to determine the influence of each component on the resulting dose in order to identify the major component of the uncertainty in a bioassay. The results of this study can be applied to various situations. In cases

A MHD equilibrium code 'EQUCIR version 2' applicable to up-down asymmetric toroidal plasma

International Nuclear Information System (INIS)

Shinya, Kichiro; Ninomiya, Hiromasa

1981-01-01

Computer code EQUCIR version 2, which can analyse tokamak plasma equilibrium without assuming up-down symmetry with respect to the mid-plane, has been developed. This code is essentially the same as EQUCIR version 1 which has already been reported and can deal with only symmetrical plasma with respect to the mid-plane. Because data input stream is slightly different from version 1 physical background of the change and the method of calculation are explained. Data input manual for the different points is also summarized. The code has been applied to the analysis of INTOR single-null divertor plasmas and to the design of hybrid poloidal coils resulting in useful and powerful means for the design. (author)

Development of 3-dimensional neutronics kinetics analysis code for CANDU-PHWR

International Nuclear Information System (INIS)

Kim, M. W.; Kim, C. H.; Hong, I. S.

2005-02-01

The followings are the major contents and scope of the research : development of kinetics power calculation module, formulation of space-dependent neutron transient analysis - implementation of 3-D and 2-G unified nodal method, verification of the kinetics module by benchmark problem - 3-D PHWR kinetics benchmark problem suggested by AECL, reactor trip simulation by shutdown system 1 in Wolsong unit 2. Development of a dynamic linked library code, SCAN D LL, for the coupled calculation with RELAP-CANDU : modeling of shutdown system 1, development of automatic shutdown module - automatic trip module based on rate log power control logic, automatic insertion of shutdown system 1. Development of a link code for coupled calculation - development of SCAN D LL(windows version), verification of coupled code by - 40% reactor inlet header break LOCA power pulse, 100% reactor outlet header break LOCA power pulse, 50% pump suction break LOCA power pulse

Development Perspective of Regulatory Audit Code System for SFR Nuclear Safety Evaluation

Energy Technology Data Exchange (ETDEWEB)

Bae, Moo Hoon; Lee, Gil Soo; Shin, An Dong; Suh, Nam Duk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2012-05-15

A sodium-cooled fast reactor (SFR) in Korea is based on the KALIMER-600 concept developed by KAERI. Based on 'Long-term R and D Plan for Future Reactor Systems' which was approved by the Korea Atomic Energy Commission in 2008, the KAERI designer is scheduled to apply the design certification of the prototype SFR in 2017. In order to establish regulatory infrastructure for the licensing of a prototype SFR, KINS has develop the regulatory requirements for the demonstration SFR since 2010, and are scheduled to develop the regulatory audit code systems in regard to core, fuel, and system, etc. since 2012. In this study, the domestic code systems used for core design and safety evaluation of PWRs and the nuclear physics and code system for SFRs were briefly reviewed, and the development perspective of regulatory audit code system for SFR nuclear safety evaluation were derived

Software Certification - Coding, Code, and Coders

Science.gov (United States)

Havelund, Klaus; Holzmann, Gerard J.

2011-01-01

We describe a certification approach for software development that has been adopted at our organization. JPL develops robotic spacecraft for the exploration of the solar system. The flight software that controls these spacecraft is considered to be mission critical. We argue that the goal of a software certification process cannot be the development of "perfect" software, i.e., software that can be formally proven to be correct under all imaginable and unimaginable circumstances. More realistically, the goal is to guarantee a software development process that is conducted by knowledgeable engineers, who follow generally accepted procedures to control known risks, while meeting agreed upon standards of workmanship. We target three specific issues that must be addressed in such a certification procedure: the coding process, the code that is developed, and the skills of the coders. The coding process is driven by standards (e.g., a coding standard) and tools. The code is mechanically checked against the standard with the help of state-of-the-art static source code analyzers. The coders, finally, are certified in on-site training courses that include formal exams.

On the development of an intrinsic hybrid composite

International Nuclear Information System (INIS)

Kießling, R; Ihlemann, J; Riemer, M; Drossel, W-G; Scharf, I; Lampke, T; Sharafiev, S; Pouya, M; Wagner, M F-X

2016-01-01

Hybrid parts, which combine low weight with high strength, are moving into the focus of the automotive industry, due to their high potential for usage in the field of crash-relevant structures. In this contribution, the development of an intrinsic hybrid composite is presented, with a focus on the manufacturing process, complex simulations of the material behaviour and material testing. The hybrid composite is made up of a continuous fibre- reinforced plastic (FRP), in which a metallic insert is integrated. The mechanical behaviour of the individual components is characterised. For material modelling, an approach is pointed out that enables modelling at large strains by directly connected rheological elements. The connection between the FRP and the metallic insert is realised by a combination of form fit and adhesive bonds. On the one hand, adhesive bonds are generated within a sol gel process. On the other hand, local form elements of the metallic insert are pressed into the FRP. We show how these form elements are generated during the macroscopic forming process. In addition, the applied sol gel process is explained. Finally, we consider design concepts for a specimen type for high strain testing of the resulting interfaces. (paper)

Development of burnup methods and capabilities in Monte Carlo code RMC

International Nuclear Information System (INIS)

She, Ding; Liu, Yuxuan; Wang, Kan; Yu, Ganglin; Forget, Benoit; Romano, Paul K.; Smith, Kord

2013-01-01

Highlights: ► The RMC code has been developed aiming at large-scale burnup calculations. ► Matrix exponential methods are employed to solve the depletion equations. ► The Energy-Bin method reduces the time expense of treating ACE libraries. ► The Cell-Mapping method is efficient to handle massive amounts of tally cells. ► Parallelized depletion is necessary for massive amounts of burnup regions. -- Abstract: The Monte Carlo burnup calculation has always been a challenging problem because of its large time consumption when applied to full-scale assembly or core calculations, and thus its application in routine analysis is limited. Most existing MC burnup codes are usually external wrappers between a MC code, e.g. MCNP, and a depletion code, e.g. ORIGEN. The code RMC is a newly developed MC code with an embedded depletion module aimed at performing burnup calculations of large-scale problems with high efficiency. Several measures have been taken to strengthen the burnup capabilities of RMC. Firstly, an accurate and efficient depletion module called DEPTH has been developed and built in, which employs the rational approximation and polynomial approximation methods. Secondly, the Energy-Bin method and the Cell-Mapping method are implemented to speed up the transport calculations with large numbers of nuclides and tally cells. Thirdly, the batch tally method and the parallelized depletion module have been utilized to better handle cases with massive amounts of burnup regions in parallel calculations. Burnup cases including a PWR pin and a 5 × 5 assembly group are calculated, thereby demonstrating the burnup capabilities of the RMC code. In addition, the computational time and memory requirements of RMC are compared with other MC burnup codes.

Developments in batteries and fuel cells for electric and hybrid electric vehicles

International Nuclear Information System (INIS)

Ahmed, R.

2013-01-01

Due to ever increasing threats of climate change, urban air pollution and costly and depleting oil and gas sources a lot of work is being done for the development of electric vehicles. Hybrid electric vehicles, plug-in hybrid electric vehicles and all electric vehicles are powered by batteries or by hydrogen and fuel cells are the main types of vehicles being developed. Main types of batteries which can be used for electric vehicles are lead-acid, Ni-Cd, Nickel-Metal-Hybrid ( NiMH) and Lithium-ion (Li-ion) batteries which are discussed and compared. Lithium ion battery is the mostly used battery. Developments in the lithium ion batteries are discussed and reviewed. Redox flow batteries are also potential candidates for electric vehicles and are described. Hybrid electric vehicles can reduce fuel consumption considerably and is a good midterm solution. Electric and hybrid electric vehicles are discussed. Electric vehicles are necessary to mitigate the effects of pollution and dependence on oil. For all the electric vehicles there are two options: batteries and fuel Cells. Batteries are useful for small vehicles and shorter distances but for vehicle range greater than 150 km fuel cells are superior to batteries in terms of cost, efficiency and durability even using natural gas and other fuels in addition to hydrogen. Ultimate solution for electric vehicles are hydrogen and fuel cells and this opinion is also shared by most of the automobile manufacturers. Developments in fuel cells and their applications for automobiles are described and reviewed. Comparisons have been done in the literature between batteries and fuel cells and are described. (author)

Analyses to support development of risk-informed separation distances for hydrogen codes and standards.

Energy Technology Data Exchange (ETDEWEB)

LaChance, Jeffrey L.; Houf, William G. (Sandia National Laboratories, Livermore, CA); Fluer, Inc., Paso Robels, CA; Fluer, Larry (Fluer, Inc., Paso Robels, CA); Middleton, Bobby

2009-03-01

The development of a set of safety codes and standards for hydrogen facilities is necessary to ensure they are designed and operated safely. To help ensure that a hydrogen facility meets an acceptable level of risk, code and standard development organizations are tilizing risk-informed concepts in developing hydrogen codes and standards.

Application of software quality assurance to a specific scientific code development task

International Nuclear Information System (INIS)

Dronkers, J.J.

1986-03-01

This paper describes an application of software quality assurance to a specific scientific code development program. The software quality assurance program consists of three major components: administrative control, configuration management, and user documentation. The program attempts to be consistent with existing local traditions of scientific code development while at the same time providing a controlled process of development

Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

Energy Technology Data Exchange (ETDEWEB)

Shi, Chengbin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Cheng, Maosong, E-mail: mscheng@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Liu, Guimin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China)

2016-08-15

Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center

Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

International Nuclear Information System (INIS)

Shi, Chengbin; Cheng, Maosong; Liu, Guimin

2016-01-01

Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center

3D-TRANS-2003, Workshop on Common Tools and Interfaces for Radiation Transport Codes

International Nuclear Information System (INIS)

2004-01-01

Description: Contents proceedings of Workshop on Common Tools and Interfaces for Deterministic Radiation Transport, for Monte Carlo and Hybrid Codes with a proposal to develop the following: GERALD - A General Environment for Radiation Analysis and Design. GERALD intends to create a unifying software environment where the user can define, solve and analyse a nuclear radiation transport problem using available numerical tools seamlessly. This environment will serve many purposes: teaching, research, industrial needs. It will also help to preserve the existing analytical and numerical knowledge base. This could represent a significant step towards solving the legacy problem. This activity should contribute to attracting young engineers to nuclear science and engineering and contribute to competence and knowledge preservation and management. This proposal was made at the on Workshop on C ommon Tools and Interfaces for Deterministic Radiation Transport, for Monte Carlo and Hybrid Codes , held from 25-26 September 2003 in connection with the conference SNA-2003. A first success with the development of such tools was achieved with the BOT3P2.0 and 3.0 codes providing an easy procedure and mechanism for defining and displaying 3D geometries and materials both in the form of refineable meshes for deterministic codes or Monte Carlo geometries consistent with deterministic models. Advanced SUSD: Improved tools for Sensitivity/Uncertainty Analysis. The development of tools for the analysis and estimation of sensitivities and uncertainties in calculations, or their propagation through complex computational schemes, in the field of neutronics, thermal hydraulics and also thermo-mechanics is of increasing importance for research and engineering applications. These tools allow establishing better margins for engineering designs and for the safe operation of nuclear facilities. Such tools are not sufficiently developed, but their need is increasingly evident in many activities

Sex-specific asymmetry in eye development in interspecific hybrids ...

Indian Academy of Sciences (India)

Home; Journals; Journal of Genetics; Volume 94; Issue 3. Sex-specific asymmetry in eye development in interspecific hybrids in the Drosophila bipectinata species complex. Bashisth N. Singh Parul Banerjee. Research Note Volume 94 Issue 3 September 2015 pp 493-495 ...

Development Of A Navier-Stokes Computer Code

Science.gov (United States)

Yoon, Seokkwan; Kwak, Dochan

1993-01-01

Report discusses aspects of development of CENS3D computer code, solving three-dimensional Navier-Stokes equations of compressible, viscous, unsteady flow. Implements implicit finite-difference or finite-volume numerical-integration scheme, called "lower-upper symmetric-Gauss-Seidel" (LU-SGS), offering potential for very low computer time per iteration and for fast convergence.

Development of An Automatic Verification Program for Thermal-hydraulic System Codes

Energy Technology Data Exchange (ETDEWEB)

Lee, J. Y.; Ahn, K. T.; Ko, S. H.; Kim, Y. S.; Kim, D. W. [Pusan National University, Busan (Korea, Republic of); Suh, J. S.; Cho, Y. S.; Jeong, J. J. [System Engineering and Technology Co., Daejeon (Korea, Republic of)

2012-05-15

As a project activity of the capstone design competitive exhibition, supported by the Education Center for Green Industry-friendly Fusion Technology (GIFT), we have developed a computer program which can automatically perform non-regression test, which is needed repeatedly during a developmental process of a thermal-hydraulic system code, such as the SPACE code. A non-regression test (NRT) is an approach to software testing. The purpose of the non-regression testing is to verify whether, after updating a given software application (in this case, the code), previous software functions have not been compromised. The goal is to prevent software regression, whereby adding new features results in software bugs. As the NRT is performed repeatedly, a lot of time and human resources will be needed during the development period of a code. It may cause development period delay. To reduce the cost and the human resources and to prevent wasting time, non-regression tests need to be automatized. As a tool to develop an automatic verification program, we have used Visual Basic for Application (VBA). VBA is an implementation of Microsoft's event-driven programming language Visual Basic 6 and its associated integrated development environment, which are built into most Microsoft Office applications (In this case, Excel)

Development of An Automatic Verification Program for Thermal-hydraulic System Codes

International Nuclear Information System (INIS)

Lee, J. Y.; Ahn, K. T.; Ko, S. H.; Kim, Y. S.; Kim, D. W.; Suh, J. S.; Cho, Y. S.; Jeong, J. J.

2012-01-01

As a project activity of the capstone design competitive exhibition, supported by the Education Center for Green Industry-friendly Fusion Technology (GIFT), we have developed a computer program which can automatically perform non-regression test, which is needed repeatedly during a developmental process of a thermal-hydraulic system code, such as the SPACE code. A non-regression test (NRT) is an approach to software testing. The purpose of the non-regression testing is to verify whether, after updating a given software application (in this case, the code), previous software functions have not been compromised. The goal is to prevent software regression, whereby adding new features results in software bugs. As the NRT is performed repeatedly, a lot of time and human resources will be needed during the development period of a code. It may cause development period delay. To reduce the cost and the human resources and to prevent wasting time, non-regression tests need to be automatized. As a tool to develop an automatic verification program, we have used Visual Basic for Application (VBA). VBA is an implementation of Microsoft's event-driven programming language Visual Basic 6 and its associated integrated development environment, which are built into most Microsoft Office applications (In this case, Excel)

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

Joint Coding/Decoding for Multi-message HARQ

OpenAIRE

Benyouss , Abdellatif; Jabi , Mohammed; Le Treust , Maël; Szczecinski , Leszek

2016-01-01

International audience; In this work, we propose and investigate a new coding strategy devised to increase the throughput of hybrid ARQ (HARQ) transmission over block fading channel. In our proposition, the transmitter jointly encodes a variable number of bits for each round of HARQ. The parameters (rates) of this joint coding can vary and may be based on the negative acknowledgment (NACK) signals provided by the receiver or, on the past (outdated) information about the channel states. The re...

Application of software engineering to development of reactor safety codes

International Nuclear Information System (INIS)

Wilburn, N.P.; Niccoli, L.G.

1981-01-01

Software Engineering, which is a systematic methodology by which a large scale software development project is partitioned into manageable pieces, has been applied to the development of LMFBR safety codes. The techniques have been applied extensively in the business and aerospace communities and have provided an answer to the drastically increasing cost of developing and maintaining software. The five phases of software engineering (Survey, Analysis, Design, Implementation, and Testing) were applied in turn to development of these codes, along with Walkthroughs (peer review) at each stage. The application of these techniques has resulted in SUPERIOR SOFTWARE which is well documented, thoroughly tested, easy to modify, easier to use and maintain. The development projects have resulted in lower overall cost. (orig.) [de

Development of the criticality accident analysis code, AGNES

International Nuclear Information System (INIS)

Nakajima, Ken

1989-01-01

In the design works for the facilities which handle nuclear fuel, the evaluation of criticality accidents cannot be avoided even if their possibility is as small as negligible. In particular in the system using solution fuel like uranyl nitrate, solution has the property easily becoming dangerous form, and all the past criticality accidents occurred in the case of solution, therefore, the evaluation of criticality accidents becomes the most important item of safety analysis. When a criticality accident occurred in a solution fuel system, due to the generation and movement of radiolysis gas voids, the oscillation of power output and pressure pulses are observed. In order to evaluate the effect of criticality accidents, these output oscillation and pressure pulses must be calculated accurately. For this purpose, the development of the dynamic characteristic code AGNES (Accidentally Generated Nuclear Excursion Simulation code) was carried out. The AGNES is the reactor dynamic characteristic code having two independent void models. Modified energy model and pressure model, and as the benchmark calculation of the AGNES code, the results of the experimental analysis on the CRAC experiment are reported. (K.I.)

Technology Development and Design of a Hybrid Mars Ascent Vehicle Concept

Science.gov (United States)

Karp, Ashley C.; Redmond, Matt; Nakazono, Barry; Vaughan, David; Shotwell, Robert; Story, George; Jackson, Dale; Young, David

2016-01-01

Hybrid propulsion has been investigated as an enhancing technology for a Mars Ascent Vehicle (MAV) concept as part of potential Mars Sample Return (MSR) because of its high specific impulse, restartability, and the ability to operate and survive at extremely low temperatures. A new wax-based hybrid fuel formulation has been developed that could withstand the harsh and variable Mars environment protected solely by a minimal layer of passive insulation. This formulation could provide substantial energy savings for a notional lander and is critical for rover mobility. Preliminary thermal cycle testing has determined that the formulation can survive the expected temperature extremes and lifetime thermal testing is currently underway. A complete preliminary design using this new fuel formulation combined with a low temperature oxidizer such as Mixed Oxides of Nitrogen (MON30) is presented. Several key features associated with a complete hybrid MAV concept are investigated to determine their mission suitability (e.g. Thrust Vector Control and restartable ignition options). Potential challenges along a path towards developing such a system are outlined and future work is suggested as a means of technology maturation. The hybrid design presented here was the lowest Gross Lift Off Mass (GLOM) result of a 2015 Jet Propulsion Laboratory (JPL) led MAV concept study.

Development of a hybrid pneumatic-power vehicle

International Nuclear Information System (INIS)

Huang, K. David; Tzeng, S.-C.

2005-01-01

Many complex technologies have been developed and applied to improve the energy efficiency and exhaust emission of an engine under different driving conditions. The overall thermal efficiency of an internal-combustion engine, however, can be maintained at only about 20-30%, with aggravated problems in the design and development, such as overall difficulty, excessive time consumption or excessively high cost. For electric cars, there is still no major technological breakthrough for the rapid recharging of a large capacity battery and detection of remaining power in it. Although all currently available hybrid-power engines are able to lower the amount of exhaust emissions and the fuel consumption of the engine, they are still unable to achieve a stable and optimal running condition immediately after ignition; hence the engine's thermal-efficiency remains low. To solve the aforementioned problems, an innovative concept - a hybrid pneumatic power-system (HPPS), which stores 'flow work' instead of storing electrochemical energy of the battery - is introduced. This innovative power system not only ensures that the internal-combustion ensures optimally but also recycles the exhaust flow to propel the vehicle. The optimization of the internal-combustion and recycling of the exhaust energy can increase the vehicle's efficiency from an original 15% to 33%, an overall increase of 18%

Development of an integral computer code for simulation of heat exchangers

International Nuclear Information System (INIS)

Horvat, A.; Catton, I.

2001-01-01

Heat exchangers are one of the basic installations in power and process industries. The present guidelines provide an ad-hoc solution to certain design problems. A unified approach based on simultaneous modeling of thermal-hydraulics and structural behavior does not exist. The present paper describes the development of integral numerical code for simulation of heat exchangers. The code is based on Volume Averaging Technique (VAT) for porous media flow modeling. The calculated values of the whole-section drag and heat transfer coefficients show an excellent agreement with already published values. The matching results prove the correctness of the selected approach and verify the developed numerical code used for this calculation.(author)

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.

Software requirements specification document for the AREST code development

International Nuclear Information System (INIS)

Engel, D.W.; McGrail, B.P.; Whitney, P.D.; Gray, W.J.; Williford, R.E.; White, M.D.; Eslinger, P.W.; Altenhofen, M.K.

1993-11-01

The Analysis of the Repository Source Term (AREST) computer code was selected in 1992 by the U.S. Department of Energy. The AREST code will be used to analyze the performance of an underground high level nuclear waste repository. The AREST code is being modified by the Pacific Northwest Laboratory (PNL) in order to evaluate the engineered barrier and waste package designs, model regulatory compliance, analyze sensitivities, and support total systems performance assessment modeling. The current version of the AREST code was developed to be a very useful tool for analyzing model uncertainties and sensitivities to input parameters. The code has also been used successfully in supplying source-terms that were used in a total systems performance assessment. The current version, however, has been found to be inadequate for the comparison and selection of a design for the waste package. This is due to the assumptions and simplifications made in the selection of the process and system models. Thus, the new version of the AREST code will be designed to focus on the details of the individual processes and implementation of more realistic models. This document describes the requirements of the new models that will be implemented. Included in this document is a section describing the near-field environmental conditions for this waste package modeling, description of the new process models that will be implemented, and a description of the computer requirements for the new version of the AREST code

Seismic Analysis Code (SAC): Development, porting, and maintenance within a legacy code base

Science.gov (United States)

Savage, B.; Snoke, J. A.

2017-12-01

The Seismic Analysis Code (SAC) is the result of toil of many developers over almost a 40-year history. Initially a Fortran-based code, it has undergone major transitions in underlying bit size from 16 to 32, in the 1980s, and 32 to 64 in 2009; as well as a change in language from Fortran to C in the late 1990s. Maintenance of SAC, the program and its associated libraries, have tracked changes in hardware and operating systems including the advent of Linux in the early 1990, the emergence and demise of Sun/Solaris, variants of OSX processors (PowerPC and x86), and Windows (Cygwin). Traces of these systems are still visible in source code and associated comments. A major concern while improving and maintaining a routinely used, legacy code is a fear of introducing bugs or inadvertently removing favorite features of long-time users. Prior to 2004, SAC was maintained and distributed by LLNL (Lawrence Livermore National Lab). In that year, the license was transferred from LLNL to IRIS (Incorporated Research Institutions for Seismology), but the license is not open source. However, there have been thousands of downloads a year of the package, either source code or binaries for specific system. Starting in 2004, the co-authors have maintained the SAC package for IRIS. In our updates, we fixed bugs, incorporated newly introduced seismic analysis procedures (such as EVALRESP), added new, accessible features (plotting and parsing), and improved the documentation (now in HTML and PDF formats). Moreover, we have added modern software engineering practices to the development of SAC including use of recent source control systems, high-level tests, and scripted, virtualized environments for rapid testing and building. Finally, a "sac-help" listserv (administered by IRIS) was setup for SAC-related issues and is the primary avenue for users seeking advice and reporting bugs. Attempts are always made to respond to issues and bugs in a timely fashion. For the past thirty-plus years

PCR-free quantitative detection of genetically modified organism from raw materials. An electrochemiluminescence-based bio bar code method.

Science.gov (United States)

Zhu, Debin; Tang, Yabing; Xing, Da; Chen, Wei R

2008-05-15

A bio bar code assay based on oligonucleotide-modified gold nanoparticles (Au-NPs) provides a PCR-free method for quantitative detection of nucleic acid targets. However, the current bio bar code assay requires lengthy experimental procedures including the preparation and release of bar code DNA probes from the target-nanoparticle complex and immobilization and hybridization of the probes for quantification. Herein, we report a novel PCR-free electrochemiluminescence (ECL)-based bio bar code assay for the quantitative detection of genetically modified organism (GMO) from raw materials. It consists of tris-(2,2'-bipyridyl) ruthenium (TBR)-labeled bar code DNA, nucleic acid hybridization using Au-NPs and biotin-labeled probes, and selective capture of the hybridization complex by streptavidin-coated paramagnetic beads. The detection of target DNA is realized by direct measurement of ECL emission of TBR. It can quantitatively detect target nucleic acids with high speed and sensitivity. This method can be used to quantitatively detect GMO fragments from real GMO products.

Development of a detailed core flow analysis code for prismatic fuel reactors

International Nuclear Information System (INIS)

Bennett, R.G.

1990-01-01

The detailed analysis of the core flow distribution in prismatic fuel reactors is of interest for modular high-temperature gas-cooled reactor (MHTGR) design and safety analyses. Such analyses involve the steady-state flow of helium through highly cross-connected flow paths in and around the prismatic fuel elements. Several computer codes have been developed for this purpose. However, since they are proprietary codes, they are not generally available for independent MHTGR design confirmation. The previously developed codes do not consider the exchange or diversion of flow between individual bypass gaps with much detail. Such a capability could be important in the analysis of potential fuel block motion, such as occurred in the Fort St. Vrain reactor, or for the analysis of the conditions around a flow blockage or misloaded fuel block. This work develops a computer code with fairly general-purpose capabilities for modeling the flow in regions of prismatic fuel cores. The code, called BYPASS solves a finite difference control volume formulation of the compressible, steady-state fluid flow in highly cross-connected flow paths typical of the MHTGR

The development of a transient neutron flux solution in the PANTHER code

International Nuclear Information System (INIS)

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

1990-01-01

In the United Kingdom a new three-dimensional, two-group, homogeneous reactor diffusion code, PANTHER, has been developed for the analysis of pressurized water reactors (PWRs) and advanced gas-cooled reactors (AGRs). The code can perform a comprehensive range of calculations, steady state, depletion, and transient with either a finite difference or analytic nodal flux solution. The nodal solution allows the representation of within-node burnup variation and pin-power reconstruction in either steady-state or transient mode. Specific steady-state and transient thermal feedback modules are included for both PWRs and AGRs. The code is being developed to perform a complete range of reactor calculations from online operational support to fuel management and fault transient analysis. In the area of transient analysis, the code is currently being used for a number of PWR fault transient assessments, including rod ejection and steam-line break. In addition, work is proceeding to incorporate the PANTHER 3D nodal transient solution in the TRAC-P code. This paper outlines the development of the transient flux solutions within PANTHER

Scalable Simulation of Electromagnetic Hybrid Codes

International Nuclear Information System (INIS)

Perumalla, Kalyan S.; Fujimoto, Richard; Karimabadi, Dr. Homa

2006-01-01

New discrete-event formulations of physics simulation models are emerging that can outperform models based on traditional time-stepped techniques. Detailed simulation of the Earth's magnetosphere, for example, requires execution of sub-models that are at widely differing timescales. In contrast to time-stepped simulation which requires tightly coupled updates to entire system state at regular time intervals, the new discrete event simulation (DES) approaches help evolve the states of sub-models on relatively independent timescales. However, parallel execution of DES-based models raises challenges with respect to their scalability and performance. One of the key challenges is to improve the computation granularity to offset synchronization and communication overheads within and across processors. Our previous work was limited in scalability and runtime performance due to the parallelization challenges. Here we report on optimizations we performed on DES-based plasma simulation models to improve parallel performance. The net result is the capability to simulate hybrid particle-in-cell (PIC) models with over 2 billion ion particles using 512 processors on supercomputing platforms

Methodology, status, and plans for development and assessment of the RELAP5 code

Energy Technology Data Exchange (ETDEWEB)

Johnson, G.W.; Riemke, R.A. [Idaho National Engineering Laboratory, Idaho Falls, ID (United States)

1997-07-01

RELAP/MOD3 is a computer code used for the simulation of transients and accidents in light-water nuclear power plants. The objective of the program to develop and maintain RELAP5 was and is to provide the U.S. Nuclear Regulatory Commission with an independent tool for assessing reactor safety. This paper describes code requirements, models, solution scheme, language and structure, user interface validation, and documentation. The paper also describes the current and near term development program and provides an assessment of the code`s strengths and limitations.

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.

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

Development and validation of ALEPH Monte Carlo burn-up code

International Nuclear Information System (INIS)

Stankovskiy, A.; Van den Eynde, G.; Vidmar, T.

2011-01-01

The Monte-Carlo burn-up code ALEPH is being developed in SCK-CEN since 2004. Belonging to the category of shells coupling Monte Carlo transport (MCNP or MCNPX) and 'deterministic' depletion codes (ORIGEN-2.2), ALEPH possess some unique features that distinguish it from other codes. The most important feature is full data consistency between steady-state Monte Carlo and time-dependent depletion calculations. Recent improvements of ALEPH concern full implementation of general-purpose nuclear data libraries (JEFF-3.1.1, ENDF/B-VII, JENDL-3.3). The upgraded version of the code is capable to treat isomeric branching ratios, neutron induced fission product yields, spontaneous fission yields and energy release per fission recorded in ENDF-formatted data files. The alternative algorithm for time evolution of nuclide concentrations is added. A predictor-corrector mechanism and the calculation of nuclear heating are available as well. The validation of the code on REBUS experimental programme results has been performed. The upgraded version of ALEPH has shown better agreement with measured data than other codes, including previous version of ALEPH. (authors)

Development of three dimensional transient analysis code STTA for SCWR core

International Nuclear Information System (INIS)

Wang, Lianjie; Zhao, Wenbo; Chen, Bingde; Yao, Dong; Yang, Ping

2015-01-01

Highlights: • A coupled three dimensional neutronics/thermal-hydraulics code STTA is developed for SCWR core transient analysis. • The Dynamic Link Libraries method is adopted for coupling computation for SCWR multi-flow core transient analysis. • The NEACRP-L-335 PWR benchmark problems are studied to verify STTA. • The SCWR rod ejection problems are studied to verify STTA. • STTA meets what is expected from a code for SCWR core 3-D transient preliminary analysis. - Abstract: A coupled three dimensional neutronics/thermal-hydraulics code STTA (SCWR Three dimensional Transient Analysis code) is developed for SCWR core transient analysis. Nodal Green’s Function Method based on the second boundary condition (NGFMN-K) is used for solving transient neutron diffusion equation. The SCWR sub-channel code ATHAS is integrated into NGFMN-K through the serial integration coupling approach. The NEACRP-L-335 PWR benchmark problem and SCWR rod ejection problems are studied to verify STTA. Numerical results show that the PWR solution of STTA agrees well with reference solutions and the SCWR solution is reasonable. The coupled code can be well applied to the core transients and accidents analysis with 3-D core model during both subcritical pressure and supercritical pressure operation

Development of the three dimensional flow model in the SPACE code

International Nuclear Information System (INIS)

Oh, Myung Taek; Park, Chan Eok; Kim, Shin Whan

2014-01-01

SPACE (Safety and Performance Analysis CodE) is a nuclear plant safety analysis code, which has been developed in the Republic of Korea through a joint research between the Korean nuclear industry and research institutes. The SPACE code has been developed with multi-dimensional capabilities as a requirement of the next generation safety code. It allows users to more accurately model the multi-dimensional flow behavior that can be exhibited in components such as the core, lower plenum, upper plenum and downcomer region. Based on generalized models, the code can model any configuration or type of fluid system. All the geometric quantities of mesh are described in terms of cell volume, centroid, face area, and face center, so that it can naturally represent not only the one dimensional (1D) or three dimensional (3D) Cartesian system, but also the cylindrical mesh system. It is possible to simulate large and complex domains by modelling the complex parts with a 3D approach and the rest of the system with a 1D approach. By 1D/3D co-simulation, more realistic conditions and component models can be obtained, providing a deeper understanding of complex systems, and it is expected to overcome the shortcomings of 1D system codes. (author)

Health effects estimation code development for accident consequence analysis

International Nuclear Information System (INIS)

Togawa, O.; Homma, T.

1992-01-01

As part of a computer code system for nuclear reactor accident consequence analysis, two computer codes have been developed for estimating health effects expected to occur following an accident. Health effects models used in the codes are based on the models of NUREG/CR-4214 and are revised for the Japanese population on the basis of the data from the reassessment of the radiation dosimetry and information derived from epidemiological studies on atomic bomb survivors of Hiroshima and Nagasaki. The health effects models include early and continuing effects, late somatic effects and genetic effects. The values of some model parameters are revised for early mortality. The models are modified for predicting late somatic effects such as leukemia and various kinds of cancers. The models for genetic effects are the same as those of NUREG. In order to test the performance of one of these codes, it is applied to the U.S. and Japanese populations. This paper provides descriptions of health effects models used in the two codes and gives comparisons of the mortality risks from each type of cancer for the two populations. (author)

Recent development and application of a new safety analysis code for fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Merrill, Brad J., E-mail: Brad.Merrill@inl.gov; Humrickhouse, Paul W.; Shimada, Masashi

2016-11-01

Highlights: • This paper presents recent code development activities for the MELCOR for fusion and Tritium Migration Analysis Program computer codes at the Idaho National Engineering Laboratory. • The capabilities of these computer codes are being merged into a single safety analysis tool for fusion reactor accidents. • The result of benchmarking these codes against previous code versions is presented by the authors of this paper. • This new capability is applied to study the tritium inventory and permeation rate for a water cold tungsten divertor that has neutron damage at 0.3 dpa. - Abstract: This paper describes the recent progress made in the development of two codes for fusion reactor safety assessments at the Idaho National Laboratory (INL): MELCOR for fusion and the Tritium Migration Analysis Program (TMAP). During the ITER engineering design activity (EDA), the INL Fusion Safety Program (FSP) modified the MELCOR 1.8.2 code for fusion applications to perform ITER thermal hydraulic safety analyses. Because MELCOR has undergone many improvements at SNL-NM since version 1.8.2 was released, the INL FSP recently imported these same fusion modifications into the MELCOR 1.8.6 code, along with the multiple fluids modifications of MELCOR 1.8.5 for fusion used in US advanced fusion reactor design studies. TMAP has also been under development for several decades at the INL by the FSP. TMAP treats multi-specie surface absorption and diffusion in composite materials with dislocation traps, plus the movement of these species from room to room by fluid flow within a given facility. Recently, TMAP was updated to consider multiple trap site types to allow the simulation of experimental data from neutron irradiated tungsten. The natural development path for both of these codes is to merge their capabilities into one computer code to provide a more comprehensive safety tool for analyzing accidents in fusion reactors. In this paper we detail recent developments in this

Recent development and application of a new safety analysis code for fusion reactors

International Nuclear Information System (INIS)

Merrill, Brad J.; Humrickhouse, Paul W.; Shimada, Masashi

2016-01-01

Highlights: • This paper presents recent code development activities for the MELCOR for fusion and Tritium Migration Analysis Program computer codes at the Idaho National Engineering Laboratory. • The capabilities of these computer codes are being merged into a single safety analysis tool for fusion reactor accidents. • The result of benchmarking these codes against previous code versions is presented by the authors of this paper. • This new capability is applied to study the tritium inventory and permeation rate for a water cold tungsten divertor that has neutron damage at 0.3 dpa. - Abstract: This paper describes the recent progress made in the development of two codes for fusion reactor safety assessments at the Idaho National Laboratory (INL): MELCOR for fusion and the Tritium Migration Analysis Program (TMAP). During the ITER engineering design activity (EDA), the INL Fusion Safety Program (FSP) modified the MELCOR 1.8.2 code for fusion applications to perform ITER thermal hydraulic safety analyses. Because MELCOR has undergone many improvements at SNL-NM since version 1.8.2 was released, the INL FSP recently imported these same fusion modifications into the MELCOR 1.8.6 code, along with the multiple fluids modifications of MELCOR 1.8.5 for fusion used in US advanced fusion reactor design studies. TMAP has also been under development for several decades at the INL by the FSP. TMAP treats multi-specie surface absorption and diffusion in composite materials with dislocation traps, plus the movement of these species from room to room by fluid flow within a given facility. Recently, TMAP was updated to consider multiple trap site types to allow the simulation of experimental data from neutron irradiated tungsten. The natural development path for both of these codes is to merge their capabilities into one computer code to provide a more comprehensive safety tool for analyzing accidents in fusion reactors. In this paper we detail recent developments in this

A framework for developing finite element codes for multi-disciplinary applications.

OpenAIRE

Dadvand, Pooyan

2007-01-01

The world of computing simulation has experienced great progresses in recent years and requires more exigent multidisciplinary challenges to satisfy the new upcoming demands. Increasing the importance of solving multi-disciplinary problems makes developers put more attention to these problems and deal with difficulties involved in developing software in this area. Conventional finite element codes have several difficulties in dealing with multi-disciplinary problems. Many of these codes are d...

Development of a general coupling interface for the fuel performance code transuranus tested with the reactor dynamic code DYN3D

International Nuclear Information System (INIS)

Holt, L.; Rohde, U.; Seidl, M.; Schubert, A.; Van Uffelen, P.

2013-01-01

Several institutions plan to couple the fuel performance code TRANSURANUS developed by the European Institute for Transuranium Elements with their own codes. One of these codes is the reactor dynamic code DYN3D maintained by the Helmholtz-Zentrum Dresden - Rossendorf. DYN3D was developed originally for VVER type reactors and was extended later to western type reactors. Usually, the fuel rod behavior is modeled in thermal hydraulics and neutronic codes in a simplified manner. The main idea of this coupling is to describe the fuel rod behavior in the frame of core safety analysis in a more detailed way, e.g. including the influence of the high burn-up structure, geometry changes and fission gas release. It allows to take benefit from the improved computational power and software achieved over the last two decades. The coupling interface was developed in a general way from the beginning. Thence it can be easily used also by other codes for a coupling with TRANSURANUS. The user can choose between a one-way as well as a two-way online coupling option. For a one-way online coupling, DYN3D provides only the time-dependent rod power and thermal hydraulics conditions to TRANSURANUS, but the fuel performance code doesn’t transfer any variable back to DYN3D. In a two-way online coupling, TRANSURANUS in addition transfers parameters like fuel temperature and cladding temperature back to DYN3D. This list of variables can be extended easily by geometric and further variables of interest. First results of the code system DYN3D-TRANSURANUS will be presented for a control rod ejection transient in a modern western type reactor. Pre-analyses show already that a detailed fuel rod behavior modeling will influence the thermal hydraulics and thence also the neutronics due to the Doppler reactivity effect of the fuel temperature. The coupled code system has therefore a potential to improve the assessment of safety criteria. The developed code system DYN3D-TRANSURANUS can be used also

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.

Development of EASYQAD version β: A Visualization Code System for QAD-CGGP-A Gamma and Neutron Shielding Calculation Code

International Nuclear Information System (INIS)

Kim, Jae Cheon; Lee, Hwan Soo; Ha, Pham Nhu Viet; Kim, Soon Young; Shin, Chang Ho; Kim, Jong Kyung

2007-01-01

EASYQAD had been previously developed by using MATLAB GUI (Graphical User Interface) in order to perform conveniently gamma and neutron shielding calculations at Hanyang University. It had been completed as version α of radiation shielding analysis code. In this study, EASYQAD was upgraded to version β with many additional functions and more user-friendly graphical interfaces. For general users to run it on Windows XP environment without any MATLAB installation, this version was developed into a standalone code system

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.

Hybrid Propulsion Demonstration Program 250K Hybrid Motor

Science.gov (United States)

Story, George; Zoladz, Tom; Arves, Joe; Kearney, Darren; Abel, Terry; Park, O.

2003-01-01

The Hybrid Propulsion Demonstration Program (HPDP) program was formed to mature hybrid propulsion technology to a readiness level sufficient to enable commercialization for various space launch applications. The goal of the HPDP was to develop and test a 250,000 pound vacuum thrust hybrid booster in order to demonstrate hybrid propulsion technology and enable manufacturing of large hybrid boosters for current and future space launch vehicles. The HPDP has successfully conducted four tests of the 250,000 pound thrust hybrid rocket motor at NASA's Stennis Space Center. This paper documents the test series.

Status of development and verification of the CTFD code FLUBOX

International Nuclear Information System (INIS)

Graf, U.; Paradimitriou, P.

2004-01-01

The Computational Two-Fluid Dynamics (CTFD) code FLUBOX is developed at GRS for the multidimensional simulation of two-phase flows. FLUBOX will also be used as a multidimensional module for the German system code ATHLET. The Benchmark test cases of the European ASTAR project were used to verify the ability of the code FLUBOX to calculate typical two-phase flow phenomena and conditions: void and pressure wave propagation, phase transitions, countercurrent flows, sharp interface movements, compressible (vapour) and nearly incompressible (water) conditions, thermal and mechanical non-equilibrium, stiff source terms due to mass and heat transfer between the phases. Realistic simulations of two-phase require beside the pure conservation equations additional transport equations for the interfacial area, turbulent energy and dissipation. A transport equation for the interfacial area density covering the whole two-phase flow range is in development. First validation calculations are presented in the paper. Turbulent shear stress for two-phase flows will be modelled by the development of transport equations for the turbulent kinetic energy and the turbulent dissipation rate. The development of the transport equations is mainly based on first principles on bubbles or drops and is largely free from empiricism. (author)

Development of RESRAD probabilistic computer codes for NRC decommissioning and license termination applications

International Nuclear Information System (INIS)

Chen, S. Y.; Yu, C.; Mo, T.; Trottier, C.

2000-01-01

In 1999, the US Nuclear Regulatory Commission (NRC) tasked Argonne National Laboratory to modify the existing RESRAD and RESRAD-BUILD codes to perform probabilistic, site-specific dose analysis for use with the NRC's Standard Review Plan for demonstrating compliance with the license termination rule. The RESRAD codes have been developed by Argonne to support the US Department of Energy's (DOEs) cleanup efforts. Through more than a decade of application, the codes already have established a large user base in the nation and a rigorous QA support. The primary objectives of the NRC task are to: (1) extend the codes' capabilities to include probabilistic analysis, and (2) develop parameter distribution functions and perform probabilistic analysis with the codes. The new codes also contain user-friendly features specially designed with graphic-user interface. In October 2000, the revised RESRAD (version 6.0) and RESRAD-BUILD (version 3.0), together with the user's guide and relevant parameter information, have been developed and are made available to the general public via the Internet for use

Development of M3C code for Monte Carlo reactor physics criticality calculations

International Nuclear Information System (INIS)

Kumar, Anek; Kannan, Umasankari; Krishanani, P.D.

2015-06-01

The development of Monte Carlo code (M3C) for reactor design entails use of continuous energy nuclear data and Monte Carlo simulations for each of the neutron interaction processes. BARC has started a concentrated effort for developing a new general geometry continuous energy Monte Carlo code for reactor physics calculation indigenously. The code development required a comprehensive understanding of the basic continuous energy cross section sets. The important features of this code are treatment of heterogeneous lattices by general geometry, use of point cross sections along with unionized energy grid approach, thermal scattering model for low energy treatment, capability of handling the microscopic fuel particles dispersed randomly. The capability of handling the randomly dispersed microscopic fuel particles which is very useful for the modeling of High-Temperature Gas-Cooled reactor fuels which are composed of thousands of microscopic fuel particle (TRISO fuel particle), randomly dispersed in a graphite matrix. The Monte Carlo code for criticality calculation is a pioneering effort and has been used to study several types of lattices including cluster geometries. The code has been verified for its accuracy against more than 60 sample problems covering a wide range from simple (like spherical) to complex geometry (like PHWR lattice). Benchmark results show that the code performs quite well for the criticality calculation of the system. In this report, the current status of the code, features of the code, some of the benchmark results for the testing of the code and input preparation etc. are discussed. (author)

Development and Validation of A Nuclear Fuel Cycle Analysis Tool: A FUTURE Code

Energy Technology Data Exchange (ETDEWEB)

Kim, S. K.; Ko, W. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Yoon Hee [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2013-10-15

This paper presents the development and validation methods of the FUTURE (FUel cycle analysis Tool for nUcleaR Energy) code, which was developed for a dynamic material flow evaluation and economic analysis of the nuclear fuel cycle. This code enables an evaluation of a nuclear material flow and its economy for diverse nuclear fuel cycles based on a predictable scenario. The most notable virtue of this FUTURE code, which was developed using C and MICROSOFT SQL DBMS, is that a program user can design a nuclear fuel cycle process easily using a standard process on the canvas screen through a drag-and-drop method. From the user's point of view, this code is very easy to use thanks to its high flexibility. In addition, the new code also enables the maintenance of data integrity by constructing a database environment of the results of the nuclear fuel cycle analyses.

Development of a solar-hydrogen hybrid energy system

International Nuclear Information System (INIS)

Sebastian, P.J.; Gamboa, S.A.; Vejar, Set; Campos, J.

2009-01-01

Full text: The details of the development of a PV-hydrogen hybrid energy system is presented. An arrangement of photovoltaic modules (125 W/module) was established to provide 9 kW installed power in a three-phase configuration at 127 Vrms/phase. A 5 kW fuel cell system (hydrogen/oxygen) operate as a dynamic backup of the photovoltaic system. The autonomous operation of the hybrid power system implies the production of hydrogen by electrolysis. The hydrogen is produced by water electrolysis using an electrolyzer of 1 kW power. The electrical energy used to produce hydrogen is supplied from solar panels by using 1kW of photovoltaic modules. The photovoltaic modules are installed in a sun-tracker arrangement for increasing the energy conversion efficiency. The hydrogen is stored in solar to electric commercial metal hydride based containers and supplied to the fuel cell. The hybrid system is monitored by internet and some dynamic characteristics such as demanding power, energy and power factor could be analyzed independently from the system. Some energy saving recommendations has been implemented as a pilot program at CIE-UNAM to improve the efficient use of clean energy in normal operating conditions in offices and laboratories. (author)

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)

Development of MATRA-LMR code α-version for LMR subchannel analysis

International Nuclear Information System (INIS)

Kim, Won Seok; Kim, Young Gyun; Kim, Young Gin

1998-05-01

Since the sodium boiling point is very high, maximum cladding and pin temperature are used for design limit condition in sodium cooled liquid metal reactor. It is necessary to predict accurately the core temperature distribution to increase the sodium coolant efficiency. Based on the MATRA code, which is developed for PWR analysis, MATRA-LMR is being developed for LMR. The major modification are as follows : A) The sodium properties table is implemented as subprogram in the code. B) Heat transfer coefficients are changed for LMR C) The pressure drop correlations are changed for more accurate calculations, which are Novendstern, Chiu-Rohsenow-Todreas, and Cheng-Todreas correlations. To assess the development status of MATRA-LMR code, calculations have been performed for ORNL 19 pin and EBR-II 61 pin tests. MATRA-LMR calculation results are also compared with the results obtained by the ALTHEN code, which uses more simplied thermal hydraulic model. The MATRA-LMR predictions are found to agree well to the measured values. The differences in results between MATRA-LMR and SLTHEN have occurred because SLTHEN code uses the very simplied thermal-hydraulic model to reduce computing time. MATRA-LMR can be used only for single assembly analysis, but it is planned to extend for multi-assembly calculation. (author). 18 refs., 8 tabs., 14 figs

The development of fluid codes for the laser compression of plasma

International Nuclear Information System (INIS)

Nicholas, D.J.

1982-08-01

Notes are given on the construction and use of simulation codes in plasma physics requiring only a limited background knowledge in numerical analysis and finite-difference techniques. The development of a 1-D Eulerian codes to source form is followed as an example. (U.K.)

Development of long-lived radionuclide transmutation technology - Development of a code system for core analysis of the transmutation reactors

Energy Technology Data Exchange (ETDEWEB)

Cho, Nam Zin; Kim, Yong Hee; Kim, Tae Hyung; Jo, Chang Keun; Park, Chang Je [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1996-07-01

The objective of this study is to develop a code system for core analysis= of the critical transmutation reactors utilizing fast neutrons. Core characteristics of the transmutation reactors were identified and four codes, HANCELL for pincell calculation, PRISM and AFEN-H3D for core calculation, and MA{sub B}URN for depletion calculation, were developed. The pincell calculation code is based on one-dimensional collision probability method and may provide homogenized/condensed parameters of a pincell and also can homogenize the control assembly via a nonlinear iterative method. The core calculation codes, PRISM and AFEN-H3D, solve the multi-group, multi-dimensional neutron diffusion equations for a hexagonal geometry and they are based on the finite difference method and analytic function expansion nodal (AFEN) method, respectively. The MA{sub B}URN code san analyze the behavior of actinides and fission products in a reactor core. Through benchmarking, we confirmed that the newly developed codes provide accurate solutions. 30 refs., 10 tabs., 8 figs. (author)

Development of an object oriented lattice QCD code ''Bridge++''

International Nuclear Information System (INIS)

Ueda, S; Aoki, S; Aoyama, T; Kanaya, K; Taniguchi, Y; Matsufuru, H; Motoki, S; Namekawa, Y; Nemura, H; Ukita, N

2014-01-01

We are developing a new lattice QCD code set ''Bridge++'' aiming at extensible, readable, and portable workbench for QCD simulations, while keeping a high performance at the same time. Bridge++ covers conventional lattice actions and numerical algorithms. The code set is constructed in C++ with an object oriented programming. In this paper we describe fundamental ingredients of the code and the current status of development

Code Development of Radioactive Aerosol Scrubbing in Pool-Injection Zone

Energy Technology Data Exchange (ETDEWEB)

Jo, Hyun Joung; Ha, Kwang Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Jang, Dong Soon [Chungnam National University, Daejeon (Korea, Republic of)

2015-10-15

The pool scrubbing models were reviewed and an aerosol scrubbing code has been prepared to calculate decontamination factor through the injection zone. The developed code has been verified using the experimental results and evaluated parametrically on the input variables. In injection zone, the initial steam condensation was most effective mechanism for the aerosol removal, and the steam fraction and pool temperature were highly affected on the decontamination factor by initial steam condensation. The aerosol scrubbing code will be updated to evaluate the decontamination factor at rise zone and finally whole pool scrubber phenomena. If a severe accident occurs in a nuclear power plant (NPP), the aerosol and gaseous fission products might be produced in the reactor vessel, and then released to the environment after the containment failure. FCVS (Filtered Containment Venting System) is one of the severe accident mitigation systems for retaining the containment integrity by discharging the high-temperature and high-pressure fission products to the environment after passing through the filtration system. In general, the FCVS is categorized into two types, wet and dry types. The scrubbing pool could play an important role in the wet type FCVS because a large amount of aerosol is captured in the water pool. The pool scrubbing phenomena have been modelled and embedded in several computer codes, such as SPARC (Suppression Pool Aerosol Removal Code), BUSCA (BUbble Scrubbing Algorithm) and SUPRA (Suppression Pool Retention Analysis). These codes aim at simulating the pool scrubbing process and estimating the decontamination factors (DFs) of the radioactive aerosol and iodine gas in the water pool, which is defined as the ratio of initial mass of the specific radioactive material to final massy after passing through the water pool. The pool scrubbing models were reviewed and an aerosol scrubbing code has been prepared to calculate decontamination factor through the injection

CODE SWITCHING AND THE DEVELOPMENT OF LINGUISTIC SYSTEM OF SIMULTANEOUS BILINGUAL CHILDREN

Directory of Open Access Journals (Sweden)

Leni Amelia Suek

2017-11-01

Full Text Available Code switching and code mixing are the phenomena commonly seen done by a bilingual. This behavior is influenced by several aspects such as the linguistic system, sociolinguistics, pragmatics, and language competence of the bilingual. If children are able to distinguish two different languages since early age, they will be considered simultaneous bilinguals. They show that they develop multiple, rather than single, linguistic systems. However, it was understood that code switching and code mixing were due to the failure in using proper words, language features, and sociolinguistic competence. Yet, recent studies have shown that bilingual children are able to use both languages proficiently with no signs of confusion or failure in language use. This ability also does not hinder their cognitive development.

Recent Developments in Abrasive Hybrid Manufacturing Processes

Directory of Open Access Journals (Sweden)

Ruszaj Adam

2017-06-01

Full Text Available Recent dynamic development of abrasive hybrid manufacturing processes results from application of a new difficult for machining materials and improvement of technological indicators of manufacturing processes already applied in practice. This tendency also occurs in abrasive machining processes which are often supported by ultrasonic vibrations, electrochemical dissolution or by electrical discharges. In the paper we present the review of new results of investigations and new practical applications of Abrasive Electrodischarge (AEDM and Electrochemical (AECM Machining.

Development of GUI systems for the MIDAS code

International Nuclear Information System (INIS)

Kim, K.R.; Park, S.H.; Kim, D.H.

2004-01-01

MIDAS is being developed at KAERI based on MELCOR as an integrated severe accident analysis code with existing model modification and new model addition. MIDAS was restructured to avoid the pointer based variable referencing style of MELCOR, and enhanced the memory effectiveness using the dynamic allocation method of Fortran 90. This paper describes recent activities of developing the GUI environments for MIDAS code at KAERI. Up to now, we have developed the four PC-based subsystems, which are IEDIT, IPLOT, SATS and HyperKAMG. IEDIT is an input management system that can read MELCOR input files and display its information in the Window panels. Users can modify each item in the panel and the input file will be modified according to that changes. IPLOT is a simple plotting system that can draw MIDAS plot variables trend graphs. SATS is developed as a severe accident training simulator that can display nuclear plant behavior graphically. Moreover SATS provides several controllable pumps and valves which appeared in the severe accidence. Together with SATS and the online severe accident guidance HyperKAMG, combined properly, severe accident mitigation scenarios could be presented graphically and dramatically without any change of MELCOR inputs. GUI development as a part of a severe accident management program package, MIDAS. (author)

DEVELOPMENT OF SALES APPLICATION OF PREPAID ELECTRICITY VOUCHER BASED ON ANFROID PLATFORM USING QUICK RESPONSE CODE (QR CODE

Directory of Open Access Journals (Sweden)

Ricky Akbar

2017-09-01

Full Text Available Perusahaan Listrik Negara (PLN has implemented a smart electricity system or prepaid electricity. The customers pay the electricity voucher first before use the electricity. The token contained in electricity voucher that has been purchased by the customer is inserted into the Meter Prabayar (MPB installed in the location of customers. When a customer purchases a voucher, it will get a receipt that contains all of the customer's identity and the 20-digit of voucher code (token to be entered into MPB as a substitute for electrical energy credit. Receipts obtained by the customer is certainly vulnerable to loss, or hijacked by unresponsible parties. In this study, authors designed and develop an android based application by utilizing QR code technology as a replacement for the receipt of prepaid electricity credit which contains the identity of the customer and the 20-digit voucher code. The application is developed by implemented waterfall methodology. The implementation process of the waterfall methods used, are (1 analysis of functional requirement of the system by conducting a preliminary study and data collection based on field studies and literature, (2 system design by using UML diagrams and Business Process Model Notation (BPMN and Entity Relationship diagram (ERD, (3 design implementation by using OOP (Object Oriented programming technique. Web application is developed by using laravel PHP framework and database MySQL while mobile application is developed by using B4A (4 developed system is tested by using blackbox method testing. Final result of this research is a Web and mobile applications for the sale of electricityvoucher by QR Code technology.

Process algebras for hybrid systems : comparison and development

NARCIS (Netherlands)

Khadim, U.

2008-01-01

Our research is about formal speci¯cation and analysis of hybrid systems. The formalism used is process algebra. Hybrid systems are systems that exhibit both discrete and continuous behaviour. An example of a hybrid system is a digital controller controlling a physical device such as present in

Hybrid pump for Pazflor deepwater development

Energy Technology Data Exchange (ETDEWEB)

Bibet, Pierre-Jean

2010-07-01

Subsea Processing gives engineers permanent technical challenges due to the comprehensive specifications which often stretch the proven limits of technology. It is particularly true for pumps that have to boost a multiphase effluent, because they must, often at the same time, be tolerant to free gas, able to handle very viscous, generate a high ?P, and be efficient. The Pazflor project decided to base the development on full utilization of subsea pumps. This paper presents the state of the art of the pumps that are designed to be installed on the sea floor. It further describes why existing technology could not match the technical requirements for the Pazflor project. It finally presents the pump specifically developed for this world first full field's development application, the so called 'Hybrid Pump', with the description of the associated qualification program and results. (Author)

Development of OCDMA system based on Flexible Cross Correlation (FCC) code with OFDM modulation

Science.gov (United States)

Aldhaibani, A. O.; Aljunid, S. A.; Anuar, M. S.; Arief, A. R.; Rashidi, C. B. M.

2015-03-01

The performance of the OCDMA systems is governed by numerous quantitative parameters such as the data rate, simultaneous number of users, the powers of transmitter and receiver, and the type of codes. This paper analyzes the performance of the OCDMA system using OFDM technique to enhance the channel data rate, to save power and increase the number of user of OSCDMA systems compared with previous hybrid subcarrier multiplexing/optical spectrum code division multiplexing (SCM/OSCDM) system. The average received signal to noise ratio (SNR) with the nonlinearity of subcarriers is derived. The theoretical results have been evaluated based on BER and number of users as well as amount of power saved. The proposed system gave better performance and save around -6 dBm of the power as well as increase the number of users twice compare to SCM/OCDMA system. In addition it is robust against interference and much more spectrally efficient than SCM/OCDMA system. The system was designed based on Flexible Cross Correlation (FCC) code which is easier construction, less complexity of encoder/decoder design and flexible in-phase cross-correlation for uncomplicated to implement using Fiber Bragg Gratings (FBGs) for the OCDMA systems for any number of users and weights. The OCDMA-FCC_OFDM improves the number of users (cardinality) 108% compare to SCM/ODCMA-FCC system.

Development of best estimate auditing code for CANDU thermal-hydraulic safety analysis

Energy Technology Data Exchange (ETDEWEB)

Chung, Bub Dong; Lee, Won Jae; Hwang, Moon Kyu; Lim, Hong Sik [Korea Atomic Energy Research Institute, Taejeon (Korea)

2002-04-01

The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3.The study was performed by reconsideration of the previous code assessment works and phenomena identification for essential accident scenario. Improvement areas of model development for auditing tool were identified based on the code comparison and PIRT results. Nine models have been improved significantly for the analysis of LOCA and Mon LOCA event. Conceptual problem or separate effect assessment have been performed to verify the model improvement. The linking calculation with CONTAIN 2.0 has been also enabled to establish the unified auditing code system. Analysis for the CANDU plant real transient and hypothetical LOCA bas been performed using the improved version. It has been concluded that the developed version can be utilized for the auditing analysis of LOCA and non-LOCA event for the CANDU reactor. 25 refs., 84 figs., 36 tabs. (Author)

DEVELOPMENT AND VALIDATION OF A NUCLEAR FUEL CYCLE ANALYSIS TOOL: A FUTURE CODE

Directory of Open Access Journals (Sweden)

S.K. KIM

2013-10-01

Full Text Available This paper presents the development and validation methods of the FUTURE (FUel cycle analysis Tool for nUcleaR Energy code, which was developed for a dynamic material flow evaluation and economic analysis of the nuclear fuel cycle. This code enables an evaluation of a nuclear material flow and its economy for diverse nuclear fuel cycles based on a predictable scenario. The most notable virtue of this FUTURE code, which was developed using C# and MICROSOFT SQL DBMS, is that a program user can design a nuclear fuel cycle process easily using a standard process on the canvas screen through a drag-and-drop method. From the user's point of view, this code is very easy to use thanks to its high flexibility. In addition, the new code also enables the maintenance of data integrity by constructing a database environment of the results of the nuclear fuel cycle analyses.

Numerical modeling of lower hybrid heating and current drive

International Nuclear Information System (INIS)

Valeo, E.J.; Eder, D.C.

1986-03-01

The generation of currents in toroidal plasma by application of waves in the lower hybrid frequency range involves the interplay of several physical phenomena which include: wave propagation in toroidal geometry, absorption via wave-particle resonances, the quasilinear generation of strongly nonequilibrium electron and ion distribution functions, and the self-consistent evolution of the current density in such a nonequilibrium plasma. We describe a code, LHMOD, which we have developed to treat these aspects of current drive and heating in tokamaks. We present results obtained by applying the code to a computation of current ramp-up and to an investigation of the possible importance of minority hydrogen absorption in a deuterium plasma as the ''density limit'' to current drive is approached

The development and application of a sub-channel code in ocean environment

International Nuclear Information System (INIS)

Wu, Pan; Shan, Jianqiang; Xiang, Xiong; Zhang, Bo; Gou, Junli; Zhang, Bin

2016-01-01

Highlights: • A sub-channel code named ATHAS/OE is developed for nuclear reactors in ocean environment. • ATHAS/OE is verified by another modified sub-channel code based on COBRA-IV. • ATHAS/OE is used to analyze thermal hydraulic of a typical SMR in heaving and rolling motion. • Calculation results show that ocean condition affect the thermal hydraulic of a reactor significantly. - Abstract: An upgraded version of ATHAS sub-channel code ATHAS/OE is developed for the investigation of the thermal hydraulic behavior of nuclear reactor core in ocean environment with consideration of heaving and rolling motion effect. The code is verified by another modified sub-channel code based on COBRA-IV and used to analyze the thermal hydraulic characteristics of a typical SMR under heaving and rolling motion condition. The calculation results show that the heaving and rolling motion affect the thermal hydraulic behavior of a reactor significantly.

Development of seismic analysis model for HTGR core on commercial FEM code

International Nuclear Information System (INIS)

Tsuji, Nobumasa; Ohashi, Kazutaka

2015-01-01

The aftermath of the Great East Japan Earthquake prods to revise the design basis earthquake intensity severely. In aseismic design of block-type HTGR, the securement of structural integrity of core blocks and other structures which are made of graphite become more important. For the aseismic design of block-type HTGR, it is necessary to predict the motion of core blocks which are collided with adjacent blocks. Some seismic analysis codes have been developed in 1970s, but these codes are special purpose-built codes and have poor collaboration with other structural analysis code. We develop the vertical 2 dimensional analytical model on multi-purpose commercial FEM code, which take into account the multiple impacts and friction between block interfaces and rocking motion on contact with dowel pins of the HTGR core by using contact elements. This model is verified by comparison with the experimental results of 12 column vertical slice vibration test. (author)

Develop a Hybrid Coordinate Ocean Model with Data Assimilation Capabilities

National Research Council Canada - National Science Library

Thacker, W. C

2003-01-01

.... The objectives of the research are as follows: (1) to develop a methodology for assimilating temperature and salinity profiles from XBT, CTD, and ARGO float data that accommodates the peculiarities of HYCOM's hybrid vertical coordinates, allowing...

Development of non-linear vibration analysis code for CANDU fuelling machine

International Nuclear Information System (INIS)

Murakami, Hajime; Hirai, Takeshi; Horikoshi, Kiyomi; Mizukoshi, Kaoru; Takenaka, Yasuo; Suzuki, Norio.

1988-01-01

This paper describes the development of a non-linear, dynamic analysis code for the CANDU 600 fuelling machine (F-M), which includes a number of non-linearities such as gap with or without Coulomb friction, special multi-linear spring connections, etc. The capabilities and features of the code and the mathematical treatment for the non-linearities are explained. The modeling and numerical methodology for the non-linearities employed in the code are verified experimentally. Finally, the simulation analyses for the full-scale F-M vibration testing are carried out, and the applicability of the code to such multi-degree of freedom systems as F-M is demonstrated. (author)

Development of a PC code package for the analysis of research and power reactors

International Nuclear Information System (INIS)

Urli, N.

1992-06-01

Computer codes available for performing reactor physics calculations for nuclear research reactors and power reactors are normally suited for running on mainframe computers. With the fast development in speed and memory of the PCs and affordable prices it became feasible to develop PC versions of commonly used codes. The present work performed under an IAEA sponsored research contract has successfully developed a code package for running on a PC. This package includes a cross-section generating code PSU-LEOPARD and 2D and 1D spatial diffusion codes, MCRAC and MCYC 1D. For adapting PSU-LEOPARD for a PC, the binary library has been reorganized to decimal form, upgraded to FORTRAN-77 standard and arrays and subroutines reorganized to conform to PC compiler. Similarly PC version of MCRAC for FORTRAN-77 and 1D code MCYC 1D have been developed. Tests, verification and bench mark results show excellent agreement with the results obtained from mainframe calculations. The execution speeds are also very satisfactory. 12 refs, 4 figs, 3 tabs

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

Energy Technology Data Exchange (ETDEWEB)

Page, R.; Jones, J.R.

1997-07-01

Ensuring that safety analysis needs are met in the future is likely to lead to the development of new codes and the further development of existing codes. It is therefore advantageous to define standards for data interfaces and to develop software interfacing techniques which can readily accommodate changes when they are made. Defining interface standards is beneficial but is necessarily restricted in application if future requirements are not known in detail. Code interfacing methods are of particular relevance with the move towards automatic grid frequency response operation where the integration of plant dynamic, core follow and fault study calculation tools is considered advantageous. This paper describes the background and features of a new code TALINK (Transient Analysis code LINKage program) used to provide a flexible interface to link the RELAP5 thermal hydraulics code with the PANTHER neutron kinetics and the SIBDYM whole plant dynamic modelling codes used by Nuclear Electric. The complete package enables the codes to be executed in parallel and provides an integrated whole plant thermal-hydraulics and neutron kinetics model. In addition the paper discusses the capabilities and pedigree of the component codes used to form the integrated transient analysis package and the details of the calculation of a postulated Sizewell `B` Loss of offsite power fault transient.

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

International Nuclear Information System (INIS)

Page, R.; Jones, J.R.

1997-01-01

Ensuring that safety analysis needs are met in the future is likely to lead to the development of new codes and the further development of existing codes. It is therefore advantageous to define standards for data interfaces and to develop software interfacing techniques which can readily accommodate changes when they are made. Defining interface standards is beneficial but is necessarily restricted in application if future requirements are not known in detail. Code interfacing methods are of particular relevance with the move towards automatic grid frequency response operation where the integration of plant dynamic, core follow and fault study calculation tools is considered advantageous. This paper describes the background and features of a new code TALINK (Transient Analysis code LINKage program) used to provide a flexible interface to link the RELAP5 thermal hydraulics code with the PANTHER neutron kinetics and the SIBDYM whole plant dynamic modelling codes used by Nuclear Electric. The complete package enables the codes to be executed in parallel and provides an integrated whole plant thermal-hydraulics and neutron kinetics model. In addition the paper discusses the capabilities and pedigree of the component codes used to form the integrated transient analysis package and the details of the calculation of a postulated Sizewell 'B' Loss of offsite power fault transient

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

Energy Technology Data Exchange (ETDEWEB)

Teschendorff, V.; Austregesilo, H.; Lerchl, G. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH Forschungsgelaende, Garching (Germany)

1997-07-01

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

Energy-Efficient Underwater Surveillance by Means of Hybrid Aquacopters

Science.gov (United States)

2014-12-01

method,” Acta Physica Polonica A , vol. 123, no. 6, 2013, pp. 1090–1093. [11] E. Dahlberg, A . Lauberts, R. K. Lennartsson, M. J. Levonen and L. Persson...those of the author and do not reflect the official policy or position of the Department of Defense or the U.S. Govemment. IRB protocol mnnber __ N/ A ...DISTRIBUTION CODE A This thesis develops algorithms in support of a prototype hybrid air-water quadcopter platform: the "AquaQuad." We consider the

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)