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A three-dimensional magnetostatics computer code for insertion devices

JOURNAL OF SYNCHROTRON RADIATION 1998年第3期5卷 481-484页

作者： Chubar, O Elleaume, P Chavanne, J European Synchrotron Radiat Facil F-38043 Grenoble France

RADIA is a three-dimensional magnetostatics computer code optimized for the design of undulators and wigglers. It solves boundary magnetostatics problems with magnetized and current-carrying volumes using the boundary integral approach. The magnetized volumes can be arbitrary polyhedrons with non-linear (iron) or linear anisotropic (permanent magnet) characteristics. The current-carrying elements can be straight or curved blocks with rectangular cross sections. Boundary conditions are simulated by the technique of mirroring. Analytical formulae used for the computation of the field produced by a magnetized volume of a polyhedron shape are detailed. The RADIA code is written in object-oriented C++ and interfaced to Mathematica [Mathematica is a registered trademark of Wolfram Research, Inc.]. The code outperforms currently available finite-element packages with respect to the CPU time of the solver and accuracy of the field integral estimations. An application of the code to the case of a wedge-pole undulator is presented.

关键词： insertion devices undulators wigglers computer code

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The IRAC code system to calculate activation and transmutation in the TIARA facility

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Journal of Nuclear Science and Technology 2000年 37卷 840-844页

作者： Tanaka, Susumu Fukuda, Mitsuhiro Nishimura, Koichi Hosono, Masakazu Watanabe, Hiromasa Yamano, Naoki Japan Atomic Energy Research Ins tilute Gunma 370-1292 1233 Watanuki Takasaki Japan SumitomoAtomic Energy Industries Ltd. Tokyo 130-0026 2-10-14 Ryogokll Sumida-kll Japan

A computer code system IRAC has been updated to calculate nuclide transmutation and induced radioactivity for incident particles of neutron, proton, deuteron, alpha up to 150 MeV, and ^l2C, ^l4N, ^l6O,²⁰Ne, ⁴⁰Ar up to 500 MeV in arbitrary multi-layer target system of three-dimensional geometry. The system provides four libraries that consist of ACSELA for isotope production cross-sections, DECAYLIB for decay data, GAMMALIB for photon emission data, and MASWPS for atomic mass data. Features and the system performance have been presented. To validate accuracy of the code system, radioactivities of residual isotopes for incident 260-MeV ²⁰Ne ions on cobalt target were measured by using gamma-spectrometry. Comparison of radioactivity has been shown between calculated results and experimental data for each cooling time. © 2014 Atomic Energy Society of Japan. All rights reserved.

关键词： Accelerator Facility Activation Cobalt computer code Induced Radioactivity Ions IRAC code System Isotope Production Neon Neutrons Transmutation Validation

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On Application of Multicriterial Optimization in Partical Beam Control Problem

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IFAC Proceedings Volumes 2000年第16期33卷 385-389页

作者： Ludmila V. Vladimirova St. Petersburg State University Faculty of Applied Mathematics and Control Processes Petrodvorets Bibliotechnaja pl. 2 St.-Petersburg 198904 Russia

In this paper multicriterial approach to solving problem of particle beam dynamics control is considered. Numerical algorithm of approximate Pareto set constructing is worked out. This algorithm is realized in compute... 详细信息

关键词： parameters optimization effective point optimal control criteria vector computer code

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Radiation shielding for fission reactors

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Journal of Nuclear Science and Technology 2000年 37卷 1-10页

作者： Oka, Yoshiaki Nuclear Engineering Research Laboratory The University oF Tokyo Tokyo 113-8656 7-3-1 Hongo Bunkyo-ku Japan

Radiation shielding aspects relating fission reactors have been reviewed. Domestic activities in the past five years have been mainly described concerning nuclear data, calculation methods, shielding and skyshine experiments, Advanced Boiling Water Reactor (ABWR), Advanced Pressurized Water Reactor (APWR), High Temperature Engineering Test Reactor (HTTR), Experimental and Prototype Fast Reactors (JOYO, MONJU), Demonstration FBR, core shroud replacement of BWR, and spent fuel transportation cask and vessel. These studies have valuable information in safety and cost reduction issues of fission reactor design for not only existing reactors but also new reactor concepts in the next century. It has been concluded that we should maintain existing shielding technologies and improve these data and methods for coming generations in the next millennium. © 2000 Atomic Energy Society of Japan. All rights reserved.

关键词： ABWR APWR Cask computer code Core Shroud Cross Sections Experiment FBR Fission Reactors HTTR JASPER JOYO Library Life Extension MONJU Monte Carlo. Deterministic Nuclear Data Review Shielding Skyshine Vessel

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Overview of fluence to dose conversion coefficients for high energy radiations: -Calculational methods and results of two kinds of effective dose per unit particle fluence

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Journal of Nuclear Science and Technology 2000年 37卷 855-860页

作者： Iwai, Satoshi Uehara, Takashi Sato, Osamu Yoshizawa, Nobuaki Furihata, Shiori Takagi, Syunji Tanaka, Syun-Ichi Sakamoto, Yukio Nuclear Development Corporation Saitama-ken 330-0835 1-297 Kitabukuro-cho Omiya-shi Japan Mitsubishi Research Institute Inc. Tokyo 100-8141 2-3-6 Otemachi Chiyoda-ku Japan Japan Atomic Energy Research Institute Ibaraki-ken 319-1195 Tokai-mura Naka-gun Japan

In and around high energy accelerator facilities, high energy radiation dose estimation is important to the radiological protection of workers. The contribution of high energy protons and neutrons to the radiation exposure of astronauts can not also be disregarded inside a spacecraft. Effective doses are adequate radiological quantities to evaluate radiation risks of individuals, exposed to radiation not only of low and intermediate energy but also of high energy. The conversion coefficients of fluence to effective doses have not been specified for all kinds of radiation in any publications of ICRP or ICRU, except the data up to 180 MeV of neutron in the ICRP74. This report provides an overview of calculational methodologies and estimated results of dose conversion coefficient of effective doses for high energy radiation such as photon, neutron, proton, alpha particle, electron, muon and pion in five groups: USA group (LAHET code), German group (LAHET), Japanese group (HERMES), Italian group (FLUKA) and Russian group (HADRON). This overview also summarized the calculational conditions and data such as computer codes, data libraries, mathematical phantoms and energy range of each particle for estimating dose conversion coefficient for each particle. Comparison of effective dose conversion coefficients of neutron, photon and proton in five groups were performed up to 10 GeV. These results will be helpful to establish an authoritative dose equivalent conversion coefficient for high energy radiation up to 10 GeV. © 2014 Atomic Energy Society of Japan. All rights reserved.

关键词： Accelerator facilities Calculational methodologies computer code Conversion coefficients of fluence to effective doses up to lOGeV Date library Effective dose Effective dose equivalent High energy radiation Mathematical phantom Quality factor Radiation weighting factor Spacecraft

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T/BEST: A computer code for assessing the benefits of advanced aerospace technologies

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ADVANCES IN ENGINEERING SOFTWARE 1997年第4期28卷 231-238页

作者： Abumeri, GH Chamis, CC NASA LEWIS RES CTRCLEVELANDOH 44135 NYMA INC BROOKPARKOH 44142

A computer software named T/BEST, Technology Benefit Estimator, has been developed to provide a formal method to assess advanced aerospace technologies and quantify the benefit contributions for prioritization. An open-ended, modular approach is used to allow for upgrade and insertion of advanced technology modules. T/BEST's software framework, beginner-to-expert operation, interface architecture and key analysis modules are discussed. fn this paper, selected features and applications of T/BEST are demonstrated. Sample cases pertaining to structural analysis of fan and compressor blades made of titanium and composite are presented. The performance of hot and cold composite fan blades is also discussed. The cost required to manufacture titanium and composite fan blades is estimated. (C) 1997 Elsevier Science Limited.

关键词： advanced composites computer code multidisciplinary stage performance structural analysis manufacturing process cost assessment

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The general particle tracer code applied to the fusion free-electron maser

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NUCLEAR INSTRUMENTS & METHODS IN PHYSICS RESEARCH SECTION B-BEAM INTERACTIONS WITH MATERIALS AND ATOMS 1998年第1-4期139卷 481-486页

作者： de Loos, MJ van der Geer, SB van der Geer, CAJ Verhoeven, AGA Urbanus, WH Pulsar Phys NL-3582 SX Utrecht Netherlands EURATOM FOM Inst Plasma Phys Rijnhuizen NL-3430 BE Nieuwegein Netherlands

The fusion Free-Electron Maser (FEM) is the prototype of a high power, electrostatic mm-wave source, tunable in the range 130-260 GHz. In order to achieve a high overall efficiency, the charge and energy of the spent electron beam, i.e. the beam which leaves the undulator after interaction with the EM-wave, has to be recovered. A 50% overall efficiency is achieved, even for the maximum energy spread of 320 keV generated in the undulator, using a collection system consisting of a decelerator and a depressed collector. The General Particle Tracer code (GPT) is being used as the major design tool for the whole fusion FEM beam line: from the accelerator to the depressed collector. The high accuracy, ability to include FEL interaction and full 3D treatment make GPT the ideal choice for such a project. An overview of the separate sections and the use of GPT for each part of the FEM is presented. GPT is currently being applied to the design of the energy recovery system of the fusion FEM, The first simulation results, including a 3D off-axis bending scheme and scattered incident electrons, are shown. (C) 1998 Elsevier Science B.V.

关键词： beam line design particle tracking free-electron laser space-charge beam dynamics computer code

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Circuit optimization via sequential computer experiments: design of an output buffer

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JOURNAL OF THE ROYAL STATISTICAL SOCIETY SERIES C-APPLIED STATISTICS 1998年第1期47卷 31-48页

作者： Aslett, R Buck, RJ Duvall, SG Sacks, J Welch, WJ Univ Waterloo Dept Stat & Actuarial Sci Waterloo ON N2L 3G1 Canada Intel Corp Hillsboro OR 97124 USA Western Michigan Univ Kalamazoo MI 49008 USA Intel Corp Santa Clara CA 95051 USA Natl Inst Stat Sci Res Triangle Pk NC USA

In electrical engineering, circuit designs are now often optimized via circuit simulation computer models. Typically, many response variables characterize the circuit's performance. Each response is a function of many input variables, including factors that can be set in the engineering design and noise factors representing manufacturing conditions. We describe a modelling approach which is appropriate for the simulator's deterministic input-output relationships. Non-linearities and interactions are identified without explicit assumptions about the functional form. These models lead to predictors to guide the reduction of the ranges of the designable factors in a sequence of experiments. Ultimately, the predictors are used to optimize the engineering design. We also show how a visualization of the fitted relationships facilitates an understanding of the engineering tradeoffs between responses. The example used to demonstrate these methods, the design of a buffer circuit, has multiple targets for the responses, representing different trade-offs between the key performance measures.

关键词： circuit simulator computer code computer model engineering design parameter design stochastic process visualization

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How the code analyst can influence the calculation and use good practices to minimize pitfalls

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NUCLEAR TECHNOLOGY 1997年第1期117卷 40-48页

作者： Shotkin, LM U.S. Nuclear Regulatory Commission Mail Stop T-10E37 Washington DC 20555 United States Harvard University United States U.S. Nucl. Regulatory Commn. Off. R. United States

Almost 20 yr ago, the U.S. Nuclear Regulatory Commission (NRC) thermal-hydraulic code development effort made the transition from a homogeneous equilibrium formulation to a two-fluid formulation. The objective was to introduce a more physically based model so that the code analyst would have to make fewer choices in the input deck to simulate expected phenomena. There were still several options left open to the user, especially the noding for the simulation. Recent experience with NRC analyses, as well as with International Standard Problems, has shown that there can still be a considerable ''user effect'' in the use of even the two-fluid codes. Two specific examples are given. Using the RELAP5 code as a specific prototype to focus the discussion, examples are given of the choices currently available to the analyst. Similar choices are available in almost all thermal-hydraulic system codes. These example choices serve to show that even though thermal-hydraulic system codes are reaching a certain state of maturity, the user must still make many choices in setting up an input deck or in running a calculation. There are several pitfalls that the user can encounter, and there are good practices that can avoid many of these pitfalls. Specific examples of current practices for minimizing pitfalls and increasing good practices are discussed. They apply to any thermal-hydraulic system code. These include the following: training and mentoring for the code analyst;user guidelines documentation;internal review and quality assurance of the input deck by knowledgeable individuals;use of standard noding;and fixed noding for each test facility and reactor system. Guidance for the average code analyst is provided in terms of common pitfalls typically encountered and suggestions for good practices in choosing input deck options.

关键词： code analyst code user computer code

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

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JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 1997年第3期34卷 229-239页

作者： Honda, T Bartels, HW Uckan, NA Seki, Y Okazaki, T ITER JOINT CENT TEAM SAN DIEGO CA USA OAK RIDGE NATL LAB OAK RIDGE TN 37831 USA JAPAN ATOM ENERGY RES INST NAKA FUS RES ESTAB NAKA IBARAKI 31101 JAPAN

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).

关键词： ITER plasma anomaly events safety analysis computer code plasma dynamics thermal analysis beta limit density limit confinement transition in-vessel components

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