WorldWideScience

Sample records for nuclear engineering codes

  1. Overview of codes and tools for nuclear engineering education

    Science.gov (United States)

    Yakovlev, D.; Pryakhin, A.; Medvedeva, L.

    2017-01-01

    The recent world trends in nuclear education have been developed in the direction of social education, networking, virtual tools and codes. MEPhI as a global leader on the world education market implements new advanced technologies for the distance and online learning and for student research work. MEPhI produced special codes, tools and web resources based on the internet platform to support education in the field of nuclear technology. At the same time, MEPhI actively uses codes and tools from the third parties. Several types of the tools are considered: calculation codes, nuclear data visualization tools, virtual labs, PC-based educational simulators for nuclear power plants (NPP), CLP4NET, education web-platforms, distance courses (MOOCs and controlled and managed content systems). The university pays special attention to integrated products such as CLP4NET, which is not a learning course, but serves to automate the process of learning through distance technologies. CLP4NET organizes all tools in the same information space. Up to now, MEPhI has achieved significant results in the field of distance education and online system implementation.

  2. Dictionary of nuclear engineering

    Energy Technology Data Exchange (ETDEWEB)

    Sube, R.

    1985-01-01

    Ralf Sube, an experienced compiler of three wellknown four-language reference works has now prepared this glossary of nuclear engineering terms in English, German, French and Russian. Based on the proven lexicography of the Technik-Worterbuch series, it comprises about 30,000 terms in each language covering the following: Nuclear and Atomic Physics; Nuclear Radiation and Isotopes; Nuclear Materials; Nuclear Facilties; Nuclear Power Industry; Nuclear Weapons.

  3. Nuclear Reactor Engineering Analysis Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Carlos Chavez-Mercado; Jaime B. Morales-Sandoval; Benjamin E. Zayas-Perez

    1998-12-31

    The Nuclear Reactor Engineering Analysis Laboratory (NREAL) is a sophisticated computer system with state-of-the-art analytical tools and technology for analysis of light water reactors. Multiple application software tools can be activated to carry out different analyses and studies such as nuclear fuel reload evaluation, safety operation margin measurement, transient and severe accident analysis, nuclear reactor instability, operator training, normal and emergency procedures optimization, and human factors engineering studies. An advanced graphic interface, driven through touch-sensitive screens, provides the means to interact with specialized software and nuclear codes. The interface allows the visualization and control of all observable variables in a nuclear power plant (NPP), as well as a selected set of nonobservable or not directly controllable variables from conventional control panels.

  4. Nuclear engineering vocabulary

    Energy Technology Data Exchange (ETDEWEB)

    Dumont, X. [FRAMATOME, Dept. Corporate R and D, 92 - Paris-La-Defence (France); Andrieux, C. [CEA Saclay, Direction des Technologies de l' Information, DTI, 91 - Gif sur Yvette (France)

    2001-07-01

    The members of the CSTNIN - the Special Commission for Nuclear Engineering Terminology and Neology - have just produced a Nuclear Engineering Vocabulary, published by SFEN. A 120-page document which, to date, includes 400 nuclear engineering terms or expressions. For each term or expression, this Glossary gives: the primary and secondary subject field in which it is applied, a possible abbreviation, its definition, a synonym if appropriate, any relevant comments, any associated word(s), the English equivalent, its status on the date of publication of the Glossary. (author)

  5. Rocketdyne/Westinghouse nuclear thermal rocket engine modeling

    Science.gov (United States)

    Glass, James F.

    1993-01-01

    The topics are presented in viewgraph form and include the following: systems approach needed for nuclear thermal rocket (NTR) design optimization; generic NTR engine power balance codes; rocketdyne nuclear thermal system code; software capabilities; steady state model; NTR engine optimizer code-logic; reactor power calculation logic; sample multi-component configuration; NTR design code output; generic NTR code at Rocketdyne; Rocketdyne NTR model; and nuclear thermal rocket modeling directions.

  6. Space Nuclear Reactor Engineering

    Energy Technology Data Exchange (ETDEWEB)

    Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-03-06

    We needed to find a space reactor concept that could be attractive to NASA for flight and proven with a rapid turnaround, low-cost nuclear test. Heat-pipe-cooled reactors coupled to Stirling engines long identified as the easiest path to near-term, low-cost concept.

  7. Job Prospects for Nuclear Engineers.

    Science.gov (United States)

    Basta, Nicholas

    1985-01-01

    As the debate over nuclear safety continues, the job market remains healthy for nuclear engineers. The average salary offered to new nuclear engineers with bachelor's degrees is $27,400. Salary averages and increases compare favorably with other engineering disciplines. Various job sources in the field are noted. (JN)

  8. Earthquake engineering for nuclear facilities

    CERN Document Server

    Kuno, Michiya

    2017-01-01

    This book is a comprehensive compilation of earthquake- and tsunami-related technologies and knowledge for the design and construction of nuclear facilities. As such, it covers a wide range of fields including civil engineering, architecture, geotechnical engineering, mechanical engineering, and nuclear engineering, for the development of new technologies providing greater resistance against earthquakes and tsunamis. It is crucial both for students of nuclear energy courses and for young engineers in nuclear power generation industries to understand the basics and principles of earthquake- and tsunami-resistant design of nuclear facilities. In Part I, "Seismic Design of Nuclear Power Plants", the design of nuclear power plants to withstand earthquakes and tsunamis is explained, focusing on buildings, equipment's, and civil engineering structures. In Part II, "Basics of Earthquake Engineering", fundamental knowledge of earthquakes and tsunamis as well as the dynamic response of structures and foundation ground...

  9. Nuclear Rocket Engine Reactor

    CERN Document Server

    Lanin, Anatoly

    2013-01-01

    The development of a nuclear rocket engine reactor (NRER ) is presented in this book. The working capacity of an active zone NRER under mechanical and thermal load, intensive neutron fluxes, high energy generation (up to 30 MBT/l) in a working medium (hydrogen) at temperatures up to 3100 K is displayed. Design principles and bearing capacity of reactors area discussed on the basis of simulation experiments and test data of a prototype reactor. Property data of dense constructional, porous thermal insulating and fuel materials like carbide and uranium carbide compounds in the temperatures interval 300 - 3000 K are presented. Technological aspects of strength and thermal strength resistance of materials are considered. The design procedure of possible emergency processes in the NRER is developed and risks for their origination are evaluated. Prospects of the NRER development for pilotless space devices and piloted interplanetary ships are viewed.

  10. 78 FR 37885 - Approval of American Society of Mechanical Engineers' Code Cases

    Science.gov (United States)

    2013-06-24

    ... Mechanical Engineers' Code Cases; Proposed Rule #0;#0;Federal Register / Vol. 78, No. 121 / Monday, June 24... American Society of Mechanical Engineers' Code Cases AGENCY: Nuclear Regulatory Commission. ACTION... revised Code Cases published by the American Society of Mechanical Engineers (ASME). This proposed...

  11. Survey of nuclear fuel-cycle codes

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, C.R.; de Saussure, G.; Marable, J.H.

    1981-04-01

    A two-month survey of nuclear fuel-cycle models was undertaken. This report presents the information forthcoming from the survey. Of the nearly thirty codes reviewed in the survey, fifteen of these codes have been identified as potentially useful in fulfilling the tasks of the Nuclear Energy Analysis Division (NEAD) as defined in their FY 1981-1982 Program Plan. Six of the fifteen codes are given individual reviews. The individual reviews address such items as the funding agency, the author and organization, the date of completion of the code, adequacy of documentation, computer requirements, history of use, variables that are input and forecast, type of reactors considered, part of fuel cycle modeled and scope of the code (international or domestic, long-term or short-term, regional or national). The report recommends that the Model Evaluation Team perform an evaluation of the EUREKA uranium mining and milling code.

  12. ABB Combustion Engineering nuclear technology

    Energy Technology Data Exchange (ETDEWEB)

    Matzie, R.A.

    1994-12-31

    The activities of ABB Combustion Engineering in the design and construction of nuclear systems and components are briefly reviewed. ABB Construction Engineering continues to improve the design and design process for nuclear generating stations. Potential improvements are evaluated to meet new requirements both of the public and the regulator, so that the designs meet the highest standards worldwide. Advancements necessary to meet market needs and to ensure the highest level of performance in the future will be made.

  13. Coded-aperture imaging in nuclear medicine

    Science.gov (United States)

    Smith, Warren E.; Barrett, Harrison H.; Aarsvold, John N.

    1989-01-01

    Coded-aperture imaging is a technique for imaging sources that emit high-energy radiation. This type of imaging involves shadow casting and not reflection or refraction. High-energy sources exist in x ray and gamma-ray astronomy, nuclear reactor fuel-rod imaging, and nuclear medicine. Of these three areas nuclear medicine is perhaps the most challenging because of the limited amount of radiation available and because a three-dimensional source distribution is to be determined. In nuclear medicine a radioactive pharmaceutical is administered to a patient. The pharmaceutical is designed to be taken up by a particular organ of interest, and its distribution provides clinical information about the function of the organ, or the presence of lesions within the organ. This distribution is determined from spatial measurements of the radiation emitted by the radiopharmaceutical. The principles of imaging radiopharmaceutical distributions with coded apertures are reviewed. Included is a discussion of linear shift-variant projection operators and the associated inverse problem. A system developed at the University of Arizona in Tucson consisting of small modular gamma-ray cameras fitted with coded apertures is described.

  14. Overcoming Challenges in Engineering the Genetic Code.

    Science.gov (United States)

    Lajoie, M J; Söll, D; Church, G M

    2016-02-27

    Withstanding 3.5 billion years of genetic drift, the canonical genetic code remains such a fundamental foundation for the complexity of life that it is highly conserved across all three phylogenetic domains. Genome engineering technologies are now making it possible to rationally change the genetic code, offering resistance to viruses, genetic isolation from horizontal gene transfer, and prevention of environmental escape by genetically modified organisms. We discuss the biochemical, genetic, and technological challenges that must be overcome in order to engineer the genetic code.

  15. Development of nuclear power plant real-time engineering simulator

    Institute of Scientific and Technical Information of China (English)

    LIN Meng; YANG Yan-Hua; ZHANG Rong-Hua; HU Rui

    2005-01-01

    A nuclear power plant real-time engineering simulator was developed based on general-purpose thermal-hydraulic system simulation code RELAP5. It main1y consists of three parts: improved thermal-hydraulic system simulation code RELAP5, control and protection system and human-machine interface. A normal transient of CHASHMA nuclear power plant turbine step load change from 100% to 90% of full power, was simulated by the engineering simulator as an application example. This paper presents structure and main features of the engineering simulator, and application results are shown and discussed.

  16. Final Technical Report; NUCLEAR ENGINEERING RECRUITMENT EFFORT

    Energy Technology Data Exchange (ETDEWEB)

    Kerrick, Sharon S.; Vincent, Charles D.

    2007-07-02

    This report provides the summary of a project whose purpose was to support the costs of developing a nuclear engineering awareness program, an instruction program for teachers to integrate lessons on nuclear science and technology into their existing curricula, and web sites for the exchange of nuclear engineering career information and classroom materials. The specific objectives of the program were as follows: OBJECTIVE 1: INCREASE AWARENESS AND INTEREST OF NUCLEAR ENGINEERING; OBJECTIVE 2: INSTRUCT TEACHERS ON NUCLEAR TOPICS; OBJECTIVE 3: NUCLEAR EDUCATION PROGRAMS WEB-SITE; OBJECTIVE 4: SUPPORT TO UNIVERSITY/INDUSTRY MATCHING GRANTS AND REACTOR SHARING; OBJECTIVE 5: PILOT PROJECT; OBJECTIVE 6: NUCLEAR ENGINEERING ENROLLMENT SURVEY AT UNIVERSITIES

  17. 78 FR 37721 - Approval of American Society of Mechanical Engineers' Code Cases

    Science.gov (United States)

    2013-06-24

    ... Engineers' Code Cases AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guides; request for... regulatory guides (DG), DG-1230, ``Design, Fabrication and Materials Code Case Acceptability, ASME Section III''; DG-1231, ``Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1''; and...

  18. Binary Code Disassembly for Reverse Engineering

    Directory of Open Access Journals (Sweden)

    Marius Popa

    2013-01-01

    Full Text Available The disassembly of binary file is used to restore the software application code in a readable and understandable format for humans. Further, the assembly code file can be used in reverse engineering processes to establish the logical flows of the computer program or its vulnerabilities in real-world running environment. The paper highlights the features of the binary executable files under the x86 architecture and portable format, presents issues of disassembly process of a machine code file and intermediate code, disassembly algorithms which can be applied to a correct and complete reconstruction of the source file written in assembly language, and techniques and tools used in binary code disassembly.

  19. Nuclear Engine System Simulation (NESS) version 2.0

    Science.gov (United States)

    Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman J.

    1993-01-01

    The topics are presented in viewgraph form and include the following; nuclear thermal propulsion (NTP) engine system analysis program development; nuclear thermal propulsion engine analysis capability requirements; team resources used to support NESS development; expanded liquid engine simulations (ELES) computer model; ELES verification examples; NESS program development evolution; past NTP ELES analysis code modifications and verifications; general NTP engine system features modeled by NESS; representative NTP expander, gas generator, and bleed engine system cycles modeled by NESS; NESS program overview; NESS program flow logic; enabler (NERVA type) nuclear thermal rocket engine; prismatic fuel elements and supports; reactor fuel and support element parameters; reactor parameters as a function of thrust level; internal shield sizing; and reactor thermal model.

  20. Unique nuclear thermal rocket engine

    Energy Technology Data Exchange (ETDEWEB)

    Culver, D.W. (Aerojet Propulsion Division, P.O. Box 13222, Sacramento, California 95813-6000 (United States)); Rochow, R. (Babcock Wilcox Space Nuclear Systems, P.O. Box 11165, Lynchburg, Virginia 24506-1165 (United States))

    1993-01-15

    Earlier this year Aerojet Propulsion Division (APD) introduced a new, advanced nuclear thermal rocket engine (NTRE) concept intended for manned missions to the moon and to Mars. This NTRE promises to be both shorter and lighter in weight than conventionally designed engines, because its forward flowing reactor is located within an expansion-deflection (E-D) rocket nozzle. The concept has matured during the year, and this paper discusses a nearer term version that resolves four open issues identified in the initial concept: (1)Reactor design and cooling scheme simplification while retaining a high pressure power balance option; (2)Eliminate need for a new, uncooled nozzle throat material suitable for long life application; (3)Practical provision for reactor power control; and (4)Use near term, long life turbopumps.

  1. Unique nuclear thermal rocket engine

    Science.gov (United States)

    Culver, Donald W.; Rochow, Richard

    1993-06-01

    In January, 1992, a new, advanced nuclear thermal rocket engine (NTRE) concept intended for manned missions to the moon and to Mars was introduced (Culver, 1992). This NTRE promises to be both shorter and lighter in weight than conventionally designed engines, because its forward flowing reactor is located within an expansion-deflection rocket nozzle. The concept has matured during the year, and this paper discusses a nearer term version that resolves four open issues identified in the initial concept: (1) the reactor design and cooling scheme simplification while retaining a high pressure power balance option; (2) elimination need for a new, uncooled nozzle throat material suitable for long life application; (3) a practical provision for reactor power control; and (4) use of near-term, long-life turbopumps.

  2. US nuclear engineering education: Status and prospects

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    This study, conducted under the auspices of the Energy Engineering Board of the National Research Council, examines the status of and outlook for nuclear engineering education in the United States. The study resulted from a widely felt concern about the downward trends in student enrollments in nuclear engineering, in both graduate and undergraduate programs. Concerns have also been expressed about the declining number of US university nuclear engineering departments and programs, the aging of their faculties, the appropriateness of their curricula and research funding for industry and government needs, the availability of scholarships and research funding, and the increasing ratio of foreign to US graduate students. A fundamental issue is whether the supply of nuclear engineering graduates will be adequate for the future. Although such issues are more general, pertaining to all areas of US science and engineering education, they are especially acute for nuclear engineering education. 30 refs., 12 figs., 20 tabs.

  3. Nuclear reactions in Monte Carlo codes.

    Science.gov (United States)

    Ferrari, A; Sala, P R

    2002-01-01

    The physics foundations of hadronic interactions as implemented in most Monte Carlo codes are presented together with a few practical examples. The description of the relevant physics is presented schematically split into the major steps in order to stress the different approaches required for the full understanding of nuclear reactions at intermediate and high energies. Due to the complexity of the problem, only a few semi-qualitative arguments are developed in this paper. The description will be necessarily schematic and somewhat incomplete, but hopefully it will be useful for a first introduction into this topic. Examples are shown mostly for the high energy regime, where all mechanisms mentioned in the paper are at work and to which perhaps most of the readers are less accustomed. Examples for lower energies can be found in the references.

  4. Integration of the DRAGON5/DONJON5 codes in the SALOME platform for performing multi-physics calculations in nuclear engineering

    Science.gov (United States)

    Hébert, Alain

    2014-06-01

    We are presenting the computer science techniques involved in the integration of codes DRAGON5 and DONJON5 in the SALOME platform. This integration brings new capabilities in designing multi-physics computational schemes, with the possibility to couple our reactor physics codes with thermal-hydraulics or thermo-mechanics codes from other organizations. A demonstration is presented where two code components are coupled using the YACS module of SALOME, based on the CORBA protocol. The first component is a full-core 3D steady-state neuronic calculation in a PWR performed using DONJON5. The second component implement a set of 1D thermal-hydraulics calculations, each performed over a single assembly.

  5. 4{sup +} Dimensional nuclear systems engineering

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Kune Y. [PHILOSOPHIA, Seoul (Korea, Republic of)

    2009-04-15

    Nuclear power plants (NPPs) require massive quantity of data during the design, construction, operation, maintenance and decommissioning stages because of their special features like size, cost, radioactivity, and so forth. The system engineering thus calls for a fully integrated way of managing the information flow spanning their life cycle. This paper proposes digital systems engineering anchored in three dimensional (3D) computer aided design (CAD) models. The signature in the proposal lies with the four plus dimensional (4{sup +}D) Technology{sup TM}, a critical know how for digital management. ESSE (Engineering Super Simulation Emulation) features a 4{sup +}D Technology{sup TM}for nuclear energy systems engineering. The technology proposed in the 3D space and time plus cost coordinates, i.e. 4{sup +}D, is the backbone of digital engineering in the nuclear systems design and management. Dased on an integrated 3D configuration management system, ESSE consists of solutions JANUS (Junctional Analysis Neodynamic Unit SoftPower), EURUS (Engineering Utilities Research Unit SoftPower), NOTUS (Neosystemic Optimization Technical Unit SoftPower), VENUS (Virtual Engineering Neocybernetic Unit SoftPower) and INUUS (Informative Neographic Utilities Unit SoftPower). NOTUS contributes to reducing the construction cost of the NPPs by optimizing the component manufacturing procedure and the plant construction process. Planning and scheduling construction projects can thus benefit greatly by integrating traditional management techniques with digital process simulation visualization. The 3D visualization of construction processes and the resulting products intrinsically afford most of the advantages realized by incorporating a purely schedule level detail based the 4{sup +}D system. Problems with equipment positioning and manpower congestion in certain areas can be visualized prior to the actual operation, thus preventing accidents and safety problems such as collision between

  6. Nuclear Engineering Enrollments and Degrees, 1982.

    Science.gov (United States)

    Sweeney, Deborah H.; And Others

    This report presents data on the number of students enrolled and the number of bachelor's, master's, and doctoral degrees awarded in academic year 1981-82 from 72 United States institutions offering degree programs in nuclear engineering or nuclear options within other engineering fields. Presented as well are historical data for the last decade…

  7. Xenomicrobiology: a roadmap for genetic code engineering.

    Science.gov (United States)

    Acevedo-Rocha, Carlos G; Budisa, Nediljko

    2016-09-01

    Biology is an analytical and informational science that is becoming increasingly dependent on chemical synthesis. One example is the high-throughput and low-cost synthesis of DNA, which is a foundation for the research field of synthetic biology (SB). The aim of SB is to provide biotechnological solutions to health, energy and environmental issues as well as unsustainable manufacturing processes in the frame of naturally existing chemical building blocks. Xenobiology (XB) goes a step further by implementing non-natural building blocks in living cells. In this context, genetic code engineering respectively enables the re-design of genes/genomes and proteins/proteomes with non-canonical nucleic (XNAs) and amino (ncAAs) acids. Besides studying information flow and evolutionary innovation in living systems, XB allows the development of new-to-nature therapeutic proteins/peptides, new biocatalysts for potential applications in synthetic organic chemistry and biocontainment strategies for enhanced biosafety. In this perspective, we provide a brief history and evolution of the genetic code in the context of XB. We then discuss the latest efforts and challenges ahead for engineering the genetic code with focus on substitutions and additions of ncAAs as well as standard amino acid reductions. Finally, we present a roadmap for the directed evolution of artificial microbes for emancipating rare sense codons that could be used to introduce novel building blocks. The development of such xenomicroorganisms endowed with a 'genetic firewall' will also allow to study and understand the relation between code evolution and horizontal gene transfer. © 2016 The Authors. Microbial Biotechnology published by John Wiley & Sons Ltd and Society for Applied Microbiology.

  8. Alya: Towards Exascale for Engineering Simulation Codes

    CERN Document Server

    Vazquez, Mariano; Koric, Seid; Artigues, Antoni; Aguado-Sierra, Jazmin; Aris, Ruth; Mira, Daniel; Calmet, Hadrien; Cucchietti, Fernando; Owen, Herbert; Taha, Ahmed; Cela, Jose Maria

    2014-01-01

    Alya is the BSC in-house HPC-based multi-physics simulation code. It is designed from scratch to run efficiently in parallel supercomputers, solving coupled problems. The target domain is engineering, with all its particular features: complex geome- tries and unstructured meshes, coupled multi-physics with exotic coupling schemes and Physical models, ill-posed problems, flexibility needs for rapidly including new models, etc. Since its conception in 2004, Alya has shown scaling behaviour in an increasing number of cores. In this paper, we present its performance up to 100.000 cores in Blue Waters, the NCSA supercomputer. The selected tests are representative of the engineering world, all the problematic features included: incompressible flow in a hu- man respiratory system, low Mach combustion problem in a kiln furnace and coupled electro-mechanical problem in a heart. We show scalability plots for all cases, discussing all the aspects of such kind of simulations, including solvers convergence.

  9. Application of thermal-hydraulic codes in the nuclear sector; Aplicaciones de los codigos termo-hidraulicos en el sector nuclear espanol

    Energy Technology Data Exchange (ETDEWEB)

    Queral, C.; Coriso, M.; Garcia Sedano, P. J.; Ruiz, J. A.; Posada, J. M.; Jimenez Varas, G.; Sol, I.; Herranz, L. E.

    2011-07-01

    Use of thermal-hydraulic codes is extended all over many different aspects of nuclear engineering. This article groups and briefly describes the main features of some of the well known codes as an introduction to their recent applications in the Spain nuclear sector. the broad range and quality of applications highlight the maturity achieved both in industry and research organizations and universities within the Spanish nuclear sector. (Author)

  10. CAC - NUCLEAR THERMAL ROCKET CORE ANALYSIS CODE

    Science.gov (United States)

    Clark, J. S.

    1994-01-01

    One of the most important factors in the development of nuclear rocket engine designs is to be able to accurately predict temperatures and pressures throughout a fission nuclear reactor core with axial hydrogen flow through circular coolant passages. CAC is an analytical prediction program to study the heat transfer and fluid flow characteristics of a circular coolant passage. CAC predicts as a function of time axial and radial fluid conditions, passage wall temperatures, flow rates in each coolant passage, and approximate maximum material temperatures. CAC incorporates the hydrogen properties model STATE to provide fluid-state relations, thermodynamic properties, and transport properties of molecular hydrogen in any fixed ortho-para combination. The program requires the general core geometry, the core material properties as a function of temperature, the core power profile, and the core inlet conditions as function of time. Although CAC was originally developed in FORTRAN IV for use on an IBM 7094, this version is written in ANSI standard FORTRAN 77 and is designed to be machine independent. It has been successfully compiled on IBM PC series and compatible computers running MS-DOS with Lahey F77L, a Sun4 series computer running SunOS 4.1.1, and a VAX series computer running VMS 5.4-3. CAC requires 300K of RAM under MS-DOS, 422K of RAM under SunOS, and 220K of RAM under VMS. No sample executable is provided on the distribution medium. Sample input and output data are included. The standard distribution medium for this program is a 5.25 inch 360K MS-DOS format diskette. CAC was developed in 1966, and this machine independent version was released in 1992. IBM-PC and IBM are registered trademarks of International Business Machines. Lahey F77L is a registered trademark of Lahey Computer Systems, Inc. SunOS is a trademark of Sun Microsystems, Inc. VMS is a trademark of Digital Equipment Corporation. MS-DOS is a registered trademark of Microsoft Corporation.

  11. Nuclear Engineering Technologists in the Nuclear Power Era

    Science.gov (United States)

    Wang, C. H.; And Others

    1974-01-01

    Describes manpower needs in nuclear engineering in the areas of research and development, architectural engineering and construction supervision, power reactor operations, and regulatory tasks. Outlines a suitable curriculum to prepare students for the tasks related to construction and operation of power reactors. (GS)

  12. The MCEF code for nuclear evaporation and fission calculations

    Energy Technology Data Exchange (ETDEWEB)

    Deppman, A.; Pina, S.R. de; Likhachev, V.P.; Mesa, J. [Sao Paulo Univ., SP (Brazil). Inst. de Fisica; Tavares, O.A.P.; Duarte, S.B.; Oliveira, E.C. de [Centro Brasileiro de Pesquisas Fisicas (CBPF), Rio de Janeiro, RJ (Brazil); Arruda-Neto, J.D.T. [Universidade Santo Amaro (UNISA), SP (Brazil); Rodriguez, O. [Instituto Superior de Ciencias y Tecnologia Nucleares, La Habana (Cuba); Goncalves, M. [Instituto de Radioprotecao e Dosimetria (IRD), Rio de Janeiro, RJ (Brazil)

    2001-11-01

    We present an object oriented algorithm, written in the Java programming language, which performs a Monte Carlo calculation of the evaporation-fission process taking place inside an excited nucleus. We show that this nuclear physics problem is very suited for the object oriented programming by constructing two simple objects: one that handles all nuclear properties and another that takes care of the nuclear reaction. The MCEF code was used to calculate important results for nuclear reactions, and here we show examples of possible uses for this code. (author)

  13. Monte Carlo simulation of nuclear energy study (II). Annual report on Nuclear Code Evaluation Committee

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-01-01

    In the report, research results discussed in 1999 fiscal year at Nuclear Code Evaluation Committee of Nuclear Code Research Committee were summarized. Present status of Monte Carlo simulation on nuclear energy study was described. Especially, besides of criticality, shielding and core analyses, present status of applications to risk and radiation damage analyses, high energy transport and nuclear theory calculations of Monte Carlo Method was described. The 18 papers are indexed individually. (J.P.N.)

  14. Nuclear Engineering at Colleges and Universities

    Science.gov (United States)

    Duffey, Dick

    1973-01-01

    Presents five statistical tables relating to nuclear engineering education, namely, course offerings by U.S. and Canadian schools; degrees and enrollment; enrollment, courses, and staff by schools; degrees granted by schools; and research contributions to the American Nuclear Society meetings. (CC)

  15. Current challenges for education of nuclear engineers. Beyond nuclear basics

    Energy Technology Data Exchange (ETDEWEB)

    Schoenfelder, Christian [AREVA GmbH, Offenbach (Germany). Training Center

    2014-07-15

    In past decades, curricula for the education of nuclear engineers (either as a major or minor subject) have been well established all over the world. However, from the point of view of a nuclear supplier, recent experiences in large and complex new build as well as modernization projects have shown that important competences required in these projects were not addressed during the education of young graduates. Consequently, in the past nuclear industry has been obliged to either accept long periods for job familiarization, or to develop and implement various dedicated internal training measures. Although the topics normally addressed in nuclear engineering education (like neutron and reactor physics, nuclear materials or thermohydraulics and the associated calculation methods) build up important competences, this paper shows that the current status of nuclear applications requires adaptations of educational curricula. As a conclusion, when academic nuclear engineering curricula start taking into account current competence needs in nuclear industry, it will be for the benefit of the current and future generation of nuclear engineers. They will be better prepared for their future job positions and career perspectives, especially on an international level. The recommendations presented should not only be of importance for the nuclear fission field, but also for the fusion community. Here, the Horizon 2020 Roadmap to Fusion as published in 2012 now is focusing on ITER and on a longer-term development of fusion technology for a future demonstration reactor DEMO. The very challenging work program is leading to a strong need for exactly those skills that are described in this article.

  16. Decoding the function of nuclear long non-coding RNAs.

    Science.gov (United States)

    Chen, Ling-Ling; Carmichael, Gordon G

    2010-06-01

    Long non-coding RNAs (lncRNAs) are mRNA-like, non-protein-coding RNAs that are pervasively transcribed throughout eukaryotic genomes. Rather than silently accumulating in the nucleus, many of these are now known or suspected to play important roles in nuclear architecture or in the regulation of gene expression. In this review, we highlight some recent progress in how lncRNAs regulate these important nuclear processes at the molecular level. Copyright 2010 Elsevier Ltd. All rights reserved.

  17. Nuclear corrosion science and engineering

    CERN Document Server

    2012-01-01

    Understanding corrosion mechanisms, the systems and materials they affect, and the methods necessary for accurately measuring their incidence is of critical importance to the nuclear industry for the safe, economic and competitive running of its plants. This book reviews the fundamentals of nuclear corrosion. Corrosion of nuclear materials, i.e. the interaction between these materials and their environments, is a major issue for plant safety as well as for operation and economic competitiveness. Understanding these corrosion mechanisms, the systems and materials they affect, and the methods to accurately measure their incidence is of critical importance to the nuclear industry. Combining assessment techniques and analytical models into this understanding allows operators to predict the service life of corrosion-affected nuclear plant materials, and to apply the most appropriate maintenance and mitigation options to ensure safe long term operation. This book critically reviews the fundamental corrosion mechani...

  18. ABB Combustion Engineering`s nuclear experience and technologies

    Energy Technology Data Exchange (ETDEWEB)

    Matzie, R.A.

    1994-12-31

    ABB Combustion Engineering`s nuclear experience and technologies are outlined. The following topics are discussed: evolutionary approach using proven technology, substantial improvement to plant safety, utility perspective up front in developing design, integrated design, competitive plant cost, operability and maintainability, standardization, and completion of US NRC technical review.

  19. Vectorization, parallelization and porting of nuclear codes (porting). Progress report fiscal 1998

    Energy Technology Data Exchange (ETDEWEB)

    Nemoto, Toshiyuki; Kawai, Wataru; Ishizuki, Shigeru [Fujitsu Ltd., Tokyo (Japan); Kawasaki, Nobuo; Kume, Etsuo; Adachi, Masaaki; Ogasawara, Shinobu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Yatake, Yo-ichi [Hitachi Ltd., Tokyo (Japan)

    2000-03-01

    Several computer codes in the nuclear field have been vectorized, parallelized and transported on the FUJITSU VPP500 system, the AP3000 system and the Paragon system at Center for Promotion of Computational Science and Engineering in Japan Atomic Energy Research Institute. We dealt with 12 codes in fiscal 1998. These results are reported in 3 parts, i.e., the vectorization and parallelization on vector processors part, the parallelization on scalar processors part and the porting part. In this report, we describe the porting. In this porting part, the porting of Monte Carlo N-Particle Transport code MCNP4B2 and Reactor Safety Analysis code RELAP5 on the AP3000 are described. In the vectorization and parallelization on vector processors part, the vectorization of General Tokamak Circuit Simulation Program code GTCSP, the vectorization and parallelization of Molecular Dynamics Ntv Simulation code MSP2, Eddy Current Analysis code EDDYCAL, Thermal Analysis Code for Test of Passive Cooling System by HENDEL T2 code THANPACST2 and MHD Equilibrium code SELENEJ on the VPP500 are described. In the parallelization on scalar processors part, the parallelization of Monte Carlo N-Particle Transport code MCNP4B2, Plasma Hydrodynamics code using Cubic Interpolated propagation Method PHCIP and Vectorized Monte Carlo code (continuous energy model/multi-group model) MVP/GMVP on the Paragon are described. (author)

  20. 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.

  1. Nuclear thermal rocket engine operation and control

    Science.gov (United States)

    Gunn, Stanley V.; Savoie, Margarita T.; Hundal, Rolv

    1993-06-01

    The operation of a typical Rover/Nerva-derived nuclear thermal rocket (NTR) engine is characterized and the control requirements of the NTR are defined. A rationale for the selection of a candidate diverse redundant NTR engine control system is presented and the projected component operating requirements are related to the state of the art of candidate components and subsystems. The projected operational capabilities of the candidate system are delineated for the startup, full-thrust, shutdown, and decay heat removal phases of the engine operation.

  2. Efforts and Challenges in Engineering the Genetic Code.

    Science.gov (United States)

    Lin, Xiao; Yu, Allen Chi Shing; Chan, Ting Fung

    2017-03-14

    This year marks the 48th anniversary of Francis Crick's seminal work on the origin of the genetic code, in which he first proposed the "frozen accident" hypothesis to describe evolutionary selection against changes to the genetic code that cause devastating global proteome modification. However, numerous efforts have demonstrated the viability of both natural and artificial genetic code variations. Recent advances in genetic engineering allow the creation of synthetic organisms that incorporate noncanonical, or even unnatural, amino acids into the proteome. Currently, successful genetic code engineering is mainly achieved by creating orthogonal aminoacyl-tRNA/synthetase pairs to repurpose stop and rare codons or to induce quadruplet codons. In this review, we summarize the current progress in genetic code engineering and discuss the challenges, current understanding, and future perspectives regarding genetic code modification.

  3. Efforts and Challenges in Engineering the Genetic Code

    Directory of Open Access Journals (Sweden)

    Xiao Lin

    2017-03-01

    Full Text Available This year marks the 48th anniversary of Francis Crick’s seminal work on the origin of the genetic code, in which he first proposed the “frozen accident” hypothesis to describe evolutionary selection against changes to the genetic code that cause devastating global proteome modification. However, numerous efforts have demonstrated the viability of both natural and artificial genetic code variations. Recent advances in genetic engineering allow the creation of synthetic organisms that incorporate noncanonical, or even unnatural, amino acids into the proteome. Currently, successful genetic code engineering is mainly achieved by creating orthogonal aminoacyl-tRNA/synthetase pairs to repurpose stop and rare codons or to induce quadruplet codons. In this review, we summarize the current progress in genetic code engineering and discuss the challenges, current understanding, and future perspectives regarding genetic code modification.

  4. Scientific codes developed and used at GRS. Nuclear simulation chain

    Energy Technology Data Exchange (ETDEWEB)

    Schaffrath, Andreas; Sonnenkalb, Martin; Sievers, Juergen; Luther, Wolfgang; Velkov, Kiril [Gesellschaft fuer Anlagen und Reaktorsicherheit (GRS) gGmbH, Garching/Muenchen (Germany). Forschungszentrum

    2016-05-15

    Over 60 technical experts of the reactor safety research division of the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH are developing and validating reliable methods and computer codes - summarized under the term nuclear simulation chain - for the safety-related assessment for all types of nuclear power plants (NPP) and other nuclear facilities considering the current state of science and technology. This nuclear simulation chain has to be able to simulate and assess all relevant physical processes and phenomena for all operating states and (severe) accidents. In the present contribution, the nuclear simulation chain developed and applied by GRS as well as selected examples of its application are presented. The latter demonstrate impressively the width of its scope and its performance. The GRS codes can be passed on request to other (national as well as international) organizations. This contributes to a worldwide increase of the nuclear safety standards. The code transfer is especially important for developing and emerging countries lacking the financial means and/or the necessary know-how for this purpose. At the end of this contribution, the respective course of action is described.

  5. Nuclear Targeting Terms for Engineers and Scientists

    Energy Technology Data Exchange (ETDEWEB)

    St Ledger, John W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-01

    The Department of Defense has a methodology for targeting nuclear weapons, and a jargon that is used to communicate between the analysts, planners, aircrews, and missile crews. The typical engineer or scientist in the Department of Energy may not have been exposed to the nuclear weapons targeting terms and methods. This report provides an introduction to the terms and methodologies used for nuclear targeting. Its purpose is to prepare engineers and scientists to participate in wargames, exercises, and discussions with the Department of Defense. Terms such as Circular Error Probable, probability of hit and damage, damage expectancy, and the physical vulnerability system are discussed. Methods for compounding damage from multiple weapons applied to one target are presented.

  6. Midwest Nuclear Science and Engineering Consortium

    Energy Technology Data Exchange (ETDEWEB)

    Dr. Wynn Volkert; Dr. Arvind Kumar; Dr. Bryan Becker; Dr. Victor Schwinke; Dr. Angel Gonzalez; Dr. DOuglas McGregor

    2010-12-08

    The objective of the Midwest Nuclear Science and Engineering Consortium (MNSEC) is to enhance the scope, quality and integration of educational and research capabilities of nuclear sciences and engineering (NS/E) programs at partner schools in support of the U.S. nuclear industry (including DOE laboratories). With INIE support, MNSEC had a productive seven years and made impressive progress in achieving these goals. Since the past three years have been no-cost-extension periods, limited -- but notable -- progress has been made in FY10. Existing programs continue to be strengthened and broadened at Consortium partner institutions. The enthusiasm generated by the academic, state, federal, and industrial communities for the MNSEC activities is reflected in the significant leveraging that has occurred for our programs.

  7. Computer aided power flow software engineering and code generation

    Energy Technology Data Exchange (ETDEWEB)

    Bacher, R. [Swiss Federal Inst. of Tech., Zuerich (Switzerland)

    1996-02-01

    In this paper a software engineering concept is described which permits the automatic solution of a non-linear set of network equations. The power flow equation set can be seen as a defined subset of a network equation set. The automated solution process is the numerical Newton-Raphson solution process of the power flow equations where the key code parts are the numeric mismatch and the numeric Jacobian term computation. It is shown that both the Jacobian and the mismatch term source code can be automatically generated in a conventional language such as Fortran or C. Thereby one starts from a high level, symbolic language with automatic differentiation and code generation facilities. As a result of this software engineering process an efficient, very high quality newton-Raphson solution code is generated which allows easier implementation of network equation model enhancements and easier code maintenance as compared to hand-coded Fortran or C code.

  8. Computer aided power flow software engineering and code generation

    Energy Technology Data Exchange (ETDEWEB)

    Bacher, R. [Swiss Federal Inst. of Tech., Zuerich (Switzerland)

    1995-12-31

    In this paper a software engineering concept is described which permits the automatic solution of a non-linear set of network equations. The power flow equation set can be seen as a defined subset of a network equation set. The automated solution process is the numerical Newton-Raphson solution process of the power flow equations where the key code parts are the numeric mismatch and the numeric Jacobian term computation. It is shown that both the Jacobian and the mismatch term source code can be automatically generated in a conventional language such as Fortran or C. Thereby one starts from a high level, symbolic language with automatic differentiation and code generation facilities. As a result of this software engineering process an efficient, very high quality Newton-Raphson solution code is generated which allows easier implementation of network equation model enhancements and easier code maintenance as compared to hand-coded Fortran or C code.

  9. Vectorization, parallelization and porting of nuclear codes. Vectorization and parallelization. Progress report fiscal 1999

    Energy Technology Data Exchange (ETDEWEB)

    Adachi, Masaaki; Ogasawara, Shinobu; Kume, Etsuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ishizuki, Shigeru; Nemoto, Toshiyuki; Kawasaki, Nobuo; Kawai, Wataru [Fujitsu Ltd., Tokyo (Japan); Yatake, Yo-ichi [Hitachi Ltd., Tokyo (Japan)

    2001-02-01

    Several computer codes in the nuclear field have been vectorized, parallelized and trans-ported on the FUJITSU VPP500 system, the AP3000 system, the SX-4 system and the Paragon system at Center for Promotion of Computational Science and Engineering in Japan Atomic Energy Research Institute. We dealt with 18 codes in fiscal 1999. These results are reported in 3 parts, i.e., the vectorization and the parallelization part on vector processors, the parallelization part on scalar processors and the porting part. In this report, we describe the vectorization and parallelization on vector processors. In this vectorization and parallelization on vector processors part, the vectorization of Relativistic Molecular Orbital Calculation code RSCAT, a microscopic transport code for high energy nuclear collisions code JAM, three-dimensional non-steady thermal-fluid analysis code STREAM, Relativistic Density Functional Theory code RDFT and High Speed Three-Dimensional Nodal Diffusion code MOSRA-Light on the VPP500 system and the SX-4 system are described. (author)

  10. 2009 UK/US Nuclear Engineering Workshop Report

    Energy Technology Data Exchange (ETDEWEB)

    Richard Rankin

    2009-04-01

    This report summarizes the 2009 UK/US Nuclear Engineering Workshop held April 20-21, 2010, in Washington, D.C. to discuss opportunities for nuclear engineering collaboration between researchers in the United States and the United Kingdom.

  11. Transport code and nuclear data in intermediate energy region

    Energy Technology Data Exchange (ETDEWEB)

    Hasegawa, Akira; Odama, Naomitsu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Maekawa, F.; Ueki, K.; Kosaka, K.; Oyama, Y.

    1998-11-01

    We briefly reviewed the problems of intermediate energy nuclear data file and transport codes in connection with processing of the data. This is a summary of our group in the task force on JENDL High Energy File Integral Evaluation (JHEFIE). In this article we stress the necessity of the production of intermediate evaluated nuclear data file up to 3 GeV for the application of accelerator driven transmutation (ADT) system. And also we state the necessity of having our own transport code system to calculate the radiation fields using these evaluated files from the strategic points of view to keep our development of the ADT technology completely free from other conditions outside of our own such as imported codes and data with poor maintenance or unknown accuracy. (author)

  12. 76 FR 36231 - American Society of Mechanical Engineers (ASME) Codes and New and Revised ASME Code Cases

    Science.gov (United States)

    2011-06-21

    ... Engineers (ASME) Codes and New and Revised ASME Code Cases; Final Rule #0;#0;Federal Register / Vol. 76 , No... 50 RIN 3150-AI35 American Society of Mechanical Engineers (ASME) Codes and New and Revised ASME Code... 2004 ASME Boiler and Pressure Vessel Code, Section III, Division 1; 2007 ASME Boiler and...

  13. Usage of burnt fuel isotopic compositions from engineering codes in Monte-Carlo code calculations

    Energy Technology Data Exchange (ETDEWEB)

    Aleshin, Sergey S.; Gorodkov, Sergey S.; Shcherenko, Anna I. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2015-09-15

    A burn-up calculation of VVER's cores by Monte-Carlo code is complex process and requires large computational costs. This fact makes Monte-Carlo codes usage complicated for project and operating calculations. Previously prepared isotopic compositions are proposed to use for the Monte-Carlo code (MCU) calculations of different states of VVER's core with burnt fuel. Isotopic compositions are proposed to calculate by an approximation method. The approximation method is based on usage of a spectral functionality and reference isotopic compositions, that are calculated by engineering codes (TVS-M, PERMAK-A). The multiplication factors and power distributions of FA and VVER with infinite height are calculated in this work by the Monte-Carlo code MCU using earlier prepared isotopic compositions. The MCU calculation data were compared with the data which were obtained by engineering codes.

  14. Vectorization, parallelization and porting of nuclear codes (porting). Progress report fiscal 1999

    Energy Technology Data Exchange (ETDEWEB)

    Kawasaki, Nobuo; Nemoto, Toshiyuki; Kawai, Wataru; Ishizuki, Shigeru [Fujitsu Ltd., Tokyo (Japan); Ogasawara, Shinobu; Kume, Etsuo; Adachi, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Yatake, Yo-ichi [Hitachi Ltd., Tokyo (Japan)

    2001-01-01

    Several computer codes in the nuclear field have been vectorized, parallelized and transported on the FUJITSU VPP500 system, the AP3000 system, the SX-4 system and the Paragon system at Center for Promotion of Computational Science and Engineering in Japan Atomic Energy Research Institute. We dealt with 18 codes in fiscal 1999. These results are reported in 3 parts, i.e., the vectorization and the parallelization part on vector processors, the parallelization port on scalar processors and the porting part. In this report, we describe the porting. In this porting part, the porting of Assisted Model Building with Energy Refinement code version 5 (AMBER5), general purpose Monte Carlo codes far neutron and photon transport calculations based on continuous energy and multigroup methods (MVP/GMVP), automatic editing system for MCNP library code (autonj), neutron damage calculations for materials irradiations and neutron damage calculations for compounds code (SPECTER/SPECOMP), severe accident analysis code (MELCOR) and COolant Boiling in Rod Arrays, Two-Fluid code (COBRA-TF) on the VPP500 system and/or the AP3000 system are described. (author)

  15. Methods and computer codes for nuclear systems calculations

    Indian Academy of Sciences (India)

    B P Kochurov; A P Knyazev; A Yu Kwaretzkheli

    2007-02-01

    Some numerical methods for reactor cell, sub-critical systems and 3D models of nuclear reactors are presented. The methods are developed for steady states and space–time calculations. Computer code TRIFON solves space-energy problem in (, ) systems of finite height and calculates heterogeneous few-group matrix parameters of reactor cells. These parameters are used as input data in the computer code SHERHAN solving the 3D heterogeneous reactor equation for steady states and 3D space–time neutron processes simulation. Modification of TRIFON was developed for the simulation of space–time processes in sub-critical systems with external sources. An option of SHERHAN code for the system with external sources is under development.

  16. Summary of aerospace and nuclear engineering activities

    Science.gov (United States)

    1988-01-01

    The Texas A&M Nuclear and Aerospace engineering departments have worked on five different projects for the NASA/USRA Advanced Design Program during the 1987/88 year. The aerospace department worked on two types of lunar tunnelers that would create habitable space. The first design used a heated cone to melt the lunar regolith, and the second used a conventional drill to bore its way through the crust. Both used a dump truck to get rid of waste heat from the reactor as well as excess regolith from the tunneling operation. The nuclear engineering department worked on three separate projects. The NEPTUNE system is a manned, outer-planetary explorer designed with Jupiter exploration as the baseline mission. The lifetime requirement for both reactor and power-conversion systems was twenty years. The second project undertaken for the power supply was a Mars Sample Return Mission power supply. This was designed to produce 2 kW of electrical power for seven years. The design consisted of a General Purpose Heat Source (GPHS) utilizing a Stirling engine as the power conversion unit. A mass optimization was performed to aid in overall design. The last design was a reactor to provide power for propulsion to Mars and power on the surface. The requirements of 300 kW of electrical power output and a mass of less than 10,000 Rg were set. This allowed the reactor and power conversion unit to fit within the Space Shuttle cargo bay.

  17. Genetic code expansion for multiprotein complex engineering.

    Science.gov (United States)

    Koehler, Christine; Sauter, Paul F; Wawryszyn, Mirella; Girona, Gemma Estrada; Gupta, Kapil; Landry, Jonathan J M; Fritz, Markus Hsi-Yang; Radic, Ksenija; Hoffmann, Jan-Erik; Chen, Zhuo A; Zou, Juan; Tan, Piau Siong; Galik, Bence; Junttila, Sini; Stolt-Bergner, Peggy; Pruneri, Giancarlo; Gyenesei, Attila; Schultz, Carsten; Biskup, Moritz Bosse; Besir, Hueseyin; Benes, Vladimir; Rappsilber, Juri; Jechlinger, Martin; Korbel, Jan O; Berger, Imre; Braese, Stefan; Lemke, Edward A

    2016-12-01

    We present a baculovirus-based protein engineering method that enables site-specific introduction of unique functionalities in a eukaryotic protein complex recombinantly produced in insect cells. We demonstrate the versatility of this efficient and robust protein production platform, 'MultiBacTAG', (i) for the fluorescent labeling of target proteins and biologics using click chemistries, (ii) for glycoengineering of antibodies, and (iii) for structure-function studies of novel eukaryotic complexes using single-molecule Förster resonance energy transfer as well as site-specific crosslinking strategies.

  18. HINCOF-1: a Code for Hail Ingestion in Engine Inlets

    Science.gov (United States)

    Gopalaswamy, N.; Murthy, S. N. B.

    1995-01-01

    One of the major concerns during hail ingestion into an engine is the resulting amount and space- and time-wise distribution of hail at the engine face for a given geometry of inlet and set of atmospheric and flight conditions. The appearance of hail in the capture streamtube is invariably random in space and time, with respect to size and momentum. During the motion of a hailstone through an inlet, a hailstone undergoes several processes, namely impact with other hailstones and material surfaces of the inlet and spinner, rolling and rebound following impact; heat and mass transfer; phase change; and shattering, the latter three due to friction and impact. Taking all of these factors into account, a numerical code, designated HINCOF-I, has been developed for determining the motion hailstones from the atmosphere, through an inlet, and up to the engine face. The numerical procedure is based on the Monte-Carlo method. The report presents a description of the code, along with several illustrative cases. The code can be utilized to relate the spinner geometry - conical or, more effective, elliptical - to the possible diversion of hail at the engine face into the bypass stream. The code is also useful for assessing the influence of various hail characteristics on the ingestion and distribution of hailstones over the engine face.

  19. FinalReferencing within Code in Software Engineering Education!

    Institute of Scientific and Technical Information of China (English)

    HILL Gary; TURNER Scott

    2012-01-01

    Traditionally computer sciences courses will assess software code. It is common and accepted good practice (as in written reports) to reference other sources of appropriate material. However there appears to be no explicit method, recommendation or advice available to computer science tutors and students on a referencing approach! This paper aims to stimulate discussion from peers involved in software engineering education. By discussing the apparent lack of "referencing within code" advice to students and proposing suggestions for appropriate solutions. This will be based on the authors' experience of assessing code and the current advice given to their students.

  20. Software Design Document for the AMP Nuclear Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Philip, Bobby [ORNL; Clarno, Kevin T [ORNL; Cochran, Bill [ORNL

    2010-03-01

    The purpose of this document is to describe the design of the AMP nuclear fuel performance code. It provides an overview of the decomposition into separable components, an overview of what those components will do, and the strategic basis for the design. The primary components of a computational physics code include a user interface, physics packages, material properties, mathematics solvers, and computational infrastructure. Some capability from established off-the-shelf (OTS) packages will be leveraged in the development of AMP, but the primary physics components will be entirely new. The material properties required by these physics operators include many highly non-linear properties, which will be replicated from FRAPCON and LIFE where applicable, as well as some computationally-intensive operations, such as gap conductance, which depends upon the plenum pressure. Because there is extensive capability in off-the-shelf leadership class computational solvers, AMP will leverage the Trilinos, PETSc, and SUNDIALS packages. The computational infrastructure includes a build system, mesh database, and other building blocks of a computational physics package. The user interface will be developed through a collaborative effort with the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Capability Transfer program element as much as possible and will be discussed in detail in a future document.

  1. EMPIRE: Nuclear Reaction Model Code System for Data Evaluation

    Science.gov (United States)

    Herman, M.; Capote, R.; Carlson, B. V.; Obložinský, P.; Sin, M.; Trkov, A.; Wienke, H.; Zerkin, V.

    2007-12-01

    EMPIRE is a modular system of nuclear reaction codes, comprising various nuclear models, and designed for calculations over a broad range of energies and incident particles. A projectile can be a neutron, proton, any ion (including heavy-ions) or a photon. The energy range extends from the beginning of the unresolved resonance region for neutron-induced reactions (∽ keV) and goes up to several hundred MeV for heavy-ion induced reactions. The code accounts for the major nuclear reaction mechanisms, including direct, pre-equilibrium and compound nucleus ones. Direct reactions are described by a generalized optical model (ECIS03) or by the simplified coupled-channels approach (CCFUS). The pre-equilibrium mechanism can be treated by a deformation dependent multi-step direct (ORION + TRISTAN) model, by a NVWY multi-step compound one or by either a pre-equilibrium exciton model with cluster emission (PCROSS) or by another with full angular momentum coupling (DEGAS). Finally, the compound nucleus decay is described by the full featured Hauser-Feshbach model with γ-cascade and width-fluctuations. Advanced treatment of the fission channel takes into account transmission through a multiple-humped fission barrier with absorption in the wells. The fission probability is derived in the WKB approximation within the optical model of fission. Several options for nuclear level densities include the EMPIRE-specific approach, which accounts for the effects of the dynamic deformation of a fast rotating nucleus, the classical Gilbert-Cameron approach and pre-calculated tables obtained with a microscopic model based on HFB single-particle level schemes with collective enhancement. A comprehensive library of input parameters covers nuclear masses, optical model parameters, ground state deformations, discrete levels and decay schemes, level densities, fission barriers, moments of inertia and γ-ray strength functions. The results can be converted into ENDF-6 formatted files using the

  2. Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

    Energy Technology Data Exchange (ETDEWEB)

    Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E. [Sandia National Labs., Albuquerque, NM (United States); Tills, J. [J. Tills and Associates, Inc., Sandia Park, NM (United States)

    1997-12-01

    The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.

  3. Fire-safety engineering and performance-based codes

    DEFF Research Database (Denmark)

    Sørensen, Lars Schiøtt

    Fire-safety Engineering is written as a textbook for Engineering students at universities and other institutions of higher education that teach in the area of fire. The book can also be used as a work of reference for consulting engineers, Building product manufacturers, contractors, building...... project administrators, etc. The book deals with the following topics: • Historical presentation on the subject of fire • Legislation and building project administration • European fire standardization • Passive and active fire protection • Performance-based Codes • Fire-safety Engineering • Fundamental...... and respiratory physiology • Combustion and natural fires • Explosion theory • Fire Chemistry • Fire Extinction Chemistry and Physics • Evacuation and human behaviour during a fire • Sensitivity and risk analysis • Fire Models • Emission and Radiation Theory...

  4. Anti-seismic research on nuclear engineering siting

    Institute of Scientific and Technical Information of China (English)

    Li CHEN; Lei NIE; Jijiang LI; Delong WANG; Xiangyu REN

    2006-01-01

    Nuclear engineering belongs to significant project; there is higher requirement on sitings. The study has discussed basic factors of selecting sites, anti-seismic research on sitings including the seismic ground motion, probability methods of seismic hazard analysis as well as interaction about structure and foundation, meanwhile provide the reason for nuclear engineering selecting sites.

  5. Teaching Problem-Solving Skills to Nuclear Engineering Students

    Science.gov (United States)

    Waller, E.; Kaye, M. H.

    2012-01-01

    Problem solving is an essential skill for nuclear engineering graduates entering the workforce. Training in qualitative and quantitative aspects of problem solving allows students to conceptualise and execute solutions to complex problems. Solutions to problems in high consequence fields of study such as nuclear engineering require rapid and…

  6. Teaching Problem-Solving Skills to Nuclear Engineering Students

    Science.gov (United States)

    Waller, E.; Kaye, M. H.

    2012-01-01

    Problem solving is an essential skill for nuclear engineering graduates entering the workforce. Training in qualitative and quantitative aspects of problem solving allows students to conceptualise and execute solutions to complex problems. Solutions to problems in high consequence fields of study such as nuclear engineering require rapid and…

  7. Reactor Fuel Isotopics and Code Validation for Nuclear Applications

    Energy Technology Data Exchange (ETDEWEB)

    Francis, Matthew W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Weber, Charles F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pigni, Marco T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-02-01

    Experimentally measured isotopic concentrations of well characterized spent nuclear fuel (SNF) samples have been collected and analyzed by previous researchers. These sets of experimental data have been used extensively to validate the accuracy of depletion code predictions for given sets of burnups, initial enrichments, and varying power histories for different reactor types. The purpose of this report is to present the diversity of data in a concise manner and summarize the current accuracy of depletion modeling. All calculations performed for this report were done using the Oak Ridge Isotope GENeration (ORIGEN) code, an internationally used irradiation and decay code solver within the SCALE comprehensive modeling and simulation code. The diversity of data given in this report includes key actinides, stable fission products, and radioactive fission products. In general, when using the current ENDF/B-VII.0 nuclear data libraries in SCALE, the major actinides are predicted to within 5% of the measured values. Large improvements were seen for several of the curium isotopes when using improved cross section data found in evaluated nuclear data file ENDF/B-VII.0 as compared to ENDF/B-V-based results. The impact of the flux spectrum on the plutonium isotope concentrations as a function of burnup was also shown. The general accuracy noted for the actinide samples for reactor types with burnups greater than 5,000 MWd/MTU was not observed for the low-burnup Hanford B samples. More work is needed in understanding these large discrepancies. The stable neodymium and samarium isotopes were predicted to within a few percent of the measured values. Large improvements were seen in prediction for a few of the samarium isotopes when using the ENDF/B-VII.0 libraries compared to results obtained with ENDF/B-V libraries. Very accurate predictions were obtained for 133Cs and 153Eu. However, the predicted values for the stable ruthenium and rhodium isotopes varied

  8. Evaluation of radiological dispersion/consequence codes supporting DOE nuclear facility SARs

    Energy Technology Data Exchange (ETDEWEB)

    O`Kula, K.R.; Paik, I.K. [Westinghouse Savannah River Site, Aiken, SC (United States); Chung, D.Y. [Dept. of Energy, Germantown, MD (United States)

    1996-12-31

    Since the early 1990s, the authorization basis documentation of many U.S. Department of Energy (DOE) nuclear facilities has been upgraded to comply with DOE orders and standards. In this process, many safety analyses have been revised. Unfortunately, there has been nonuniform application of software, and the most appropriate computer and engineering methodologies often are not applied. A DOE Accident Phenomenology and Consequence (APAC) Methodology Evaluation Program was originated at the request of DOE Defense Programs to evaluate the safety analysis methodologies used in nuclear facility authorization basis documentation and to define future cost-effective support and development initiatives. Six areas, including source term development (fire, spills, and explosion analysis), in-facility transport, and dispersion/ consequence analysis (chemical and radiological) are contained in the APAC program. The evaluation process, codes considered, key results, and recommendations for future model and software development of the Radiological Dispersion/Consequence Working Group are summarized in this paper.

  9. Vectorization, parallelization and porting of nuclear codes on the VPP500 system (vectorization). Progress report fiscal 1996

    Energy Technology Data Exchange (ETDEWEB)

    Nemoto, Toshiyuki; Kawai, Wataru [Fujitsu Ltd., Tokyo (Japan); Kawasaki, Nobuo [and others

    1997-12-01

    Several computer codes in the nuclear field have been vectorized, parallelized and transported on the FUJITSU VPP500 system at Center for Promotion of Computational Science and Engineering in Japan Atomic Energy Research Institute. These results are reported in 3 parts, i.e., the vectorization part, the parallelization part and the porting part. In this report, we describe the vectorization. In this vectorization part, the vectorization of two and three dimensional discrete ordinates simulation code DORT-TORT, gas dynamics analysis code FLOWGR and relativistic Boltzmann-Uehling-Uhlenbeck simulation code RBUU are described. In the parallelization part, the parallelization of 2-Dimensional relativistic electromagnetic particle code EM2D, Cylindrical Direct Numerical Simulation code CYLDNS and molecular dynamics code for simulating radiation damages in diamond crystals DGR are described. And then, in the porting part, the porting of reactor safety analysis code RELAP5/MOD3.2 and RELAP5/MOD3.2.1.2, nuclear data processing system NJOY and 2-D multigroup discrete ordinate transport code TWOTRAN-II are described. And also, a survey for the porting of command-driven interactive data analysis plotting program IPLOT are described. (author)

  10. Vectorization, parallelization and porting of nuclear codes on the VPP500 system (porting). Progress report fiscal 1996

    Energy Technology Data Exchange (ETDEWEB)

    Nemoto, Toshiyuki [Fujitsu Ltd., Tokyo (Japan); Kawasaki, Nobuo; Tanabe, Hidenobu [and others

    1998-01-01

    Several computer codes in the nuclear field have been vectorized, parallelized and transported on the FUJITSU VPP500 system at Center for Promotion of Computational Science and Engineering in Japan Atomic Energy Research Institute. These results are reported in 3 parts, i.e., the vectorization part, the parallelization part and the porting part. In this report, we describe the porting. In this porting part, the porting of reactor safety analysis code RELAP5/MOD3.2 and RELAP5/MOD3.2.1.2, nuclear data processing system NJOY and 2-D multigroup discrete ordinate transport code TWOTRAN-II are described. And also, a survey for the porting of command-driven interactive data analysis plotting program IPLOT are described. In the parallelization part, the parallelization of 2-Dimensional relativistic electromagnetic particle code EM2D, Cylindrical Direct Numerical Simulation code CYLDNS and molecular dynamics code for simulating radiation damages in diamond crystals DGR are described. And then, in the vectorization part, the vectorization of two and three dimensional discrete ordinates simulation code DORT-TORT, gas dynamics analysis code FLOWGR and relativistic Boltzmann-Uehling-Uhlenbeck simulation code RBUU are described. (author)

  11. Vectorization, parallelization and porting of nuclear codes on the VPP500 system (parallelization). Progress report fiscal 1996

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Hideo; Kawai, Wataru; Nemoto, Toshiyuki [Fujitsu Ltd., Tokyo (Japan)] [and others

    1997-12-01

    Several computer codes in the nuclear field have been vectorized, parallelized and transported on the FUJITSU VPP500 system at Center for Promotion of Computational Science and Engineering in Japan Atomic Energy Research Institute. These results are reported in 3 parts, i.e., the vectorization part, the parallelization part and the porting part. In this report, we describe the parallelization. In this parallelization part, the parallelization of 2-Dimensional relativistic electromagnetic particle code EM2D, Cylindrical Direct Numerical Simulation code CYLDNS and molecular dynamics code for simulating radiation damages in diamond crystals DGR are described. In the vectorization part, the vectorization of two and three dimensional discrete ordinates simulation code DORT-TORT, gas dynamics analysis code FLOWGR and relativistic Boltzmann-Uehling-Uhlenbeck simulation code RBUU are described. And then, in the porting part, the porting of reactor safety analysis code RELAP5/MOD3.2 and RELAP5/MOD3.2.1.2, nuclear data processing system NJOY and 2-D multigroup discrete ordinate transport code TWOTRAN-II are described. And also, a survey for the porting of command-driven interactive data analysis plotting program IPLOT are described. (author)

  12. Visualized kinematics code for two-body nuclear reactions

    Science.gov (United States)

    Lee, E. J.; Chae, K. Y.

    2016-05-01

    The one or few nucleon transfer reaction has been a great tool for investigating the single-particle properties of a nucleus. Both stable and exotic beams are utilized to study transfer reactions in normal and inverse kinematics, respectively. Because many energy levels of the heavy recoil from the two-body nuclear reaction can be populated by using a single beam energy, identifying each populated state, which is not often trivial owing to high level-density of the nucleus, is essential. For identification of the energy levels, a visualized kinematics code called VISKIN has been developed by utilizing the Java programming language. The development procedure, usage, and application of the VISKIN is reported.

  13. A field release of genetically engineered gypsy moth (Lymantria dispar L.) Nuclear Polyhedrosis Virus (LdNPV)

    Science.gov (United States)

    Vincent D' Amico; Joseph S. Elkinton; John D. Podgwaite; James M. Slavicek; Michael L. McManus; John P. Burand

    1999-01-01

    The gypsy moth (Lymantria dispar L.) nuclear polyhedrosis virus was genetically engineered for nonpersistence by removal of the gene coding for polyhedrin production and stabilized using a coocclusion process. A β-galactosidase marker gene was inserted into the genetically engineered virus (LdGEV) so that infected larvae could be tested for...

  14. Coded Aperture Nuclear Scintigraphy: A Novel Small Animal Imaging Technique

    Directory of Open Access Journals (Sweden)

    Dawid Schellingerhout

    2002-10-01

    Full Text Available We introduce and demonstrate the utility of coded aperture (CA nuclear scintigraphy for imaging small animals. CA imaging uses multiple pinholes in a carefully designed mask pattern, mounted on a conventional gamma camera. System performance was assessed using point sources and phantoms, while several animal experiments were performed to test the usefulness of the imaging system in vivo, with commonly used radiopharmaceuticals. The sensitivity of the CA system for 99mTc was 4.2 × 103 cps/Bq (9400 cpm/μCi, compared to 4.4 × 104 cps/Bq (990 cpm/μCi for a conventional collimator system. The system resolution was 1.7 mm, as compared to 4–6 mm for the conventional imaging system (using a high-sensitivity low-energy collimator. Animal imaging demonstrated artifact-free imaging with superior resolution and image quality compared to conventional collimator images in several mouse and rat models. We conclude that: (a CA imaging is a useful nuclear imaging technique for small animal imaging. The advantage in signal-to-noise can be traded to achieve higher resolution, decreased dose or reduced imaging time. (b CA imaging works best for images where activity is concentrated in small volumes; a low count outline may be better demonstrated using conventional collimator imaging. Thus, CA imaging should be viewed as a technique to complement rather than replace traditional nuclear imaging methods. (c CA hardware and software can be readily adapted to existing gamma cameras, making their implementation a relatively inexpensive retrofit to most systems.

  15. A Comparison of Source Code Plagiarism Detection Engines

    Science.gov (United States)

    Lancaster, Thomas; Culwin, Fintan

    2004-06-01

    Automated techniques for finding plagiarism in student source code submissions have been in use for over 20 years and there are many available engines and services. This paper reviews the literature on the major modern detection engines, providing a comparison of them based upon the metrics and techniques they deploy. Generally the most common and effective techniques are seen to involve tokenising student submissions then searching pairs of submissions for long common substrings, an example of what is defined to be a paired structural metric. Computing academics are recommended to use one of the two Web-based detection engines, MOSS and JPlag. It is shown that whilst detection is well established there are still places where further research would be useful, particularly where visual support of the investigation process is possible.

  16. Brief 70 Nuclear Engineering Enrollments and Degrees, 2011 Summary Information

    Energy Technology Data Exchange (ETDEWEB)

    Dr. Don Johnson

    2012-10-31

    The survey includes degrees granted between September 1, 2010 and August 31, 2011. Enrollment information refers to the fall term 2011. The enrollment and degree data include students majoring in nuclear engineering or in an option program equivalent to a major. Thirty-two academic programs reported having nuclear engineering programs during 2011, and data was received from all thirty-two programs. The data for two nuclear engineering programs include enrollments and degrees in health physics options that are also reported in the health physics enrollments and degrees data.

  17. Brief 74 Nuclear Engineering Enrollments and Degrees Survey, 2014 Data

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2015-03-15

    The 2014 survey includes degrees granted between September 1, 2013 and August 31, 2014, and enrollments for fall 2014. There are three academic programs new to this year's survey. Thirty-five academic programs reported having nuclear engineering programs during 2014, and data were provided by all thirty-five. The enrollments and degrees data include students majoring in nuclear engineering or in an option program equivalent to a major. Two nuclear engineering programs have indicated that health physics option enrollments and degrees are also reported in the health physics enrollments and degrees survey.

  18. Brief 76 Nuclear Engineering Enrollments and Degrees Survey, 2015 Data

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2016-03-15

    The 2015 Nuclear Engineering Enrollments and Degrees Survey reports degrees granted between September 1, 2014 and August 31, 2015. Enrollment information refers to the fall term 2015. The enrollments and degrees data comprises students majoring in nuclear engineering or in an option program equivalent to a major. Thirty-five academic programs reported having nuclear engineering programs during 2015, and data was received from all thirty-five programs. The report includes enrollment information on undergraduate students and graduate students and information by degree level for post-graduation plans.

  19. Evaluation of CFETR as a Fusion Nuclear Science Facility using multiple system codes

    Science.gov (United States)

    Chan, V. S.; Costley, A. E.; Wan, B. N.; Garofalo, A. M.; Leuer, J. A.

    2015-02-01

    This paper presents the results of a multi-system codes benchmarking study of the recently published China Fusion Engineering Test Reactor (CFETR) pre-conceptual design (Wan et al 2014 IEEE Trans. Plasma Sci. 42 495). Two system codes, General Atomics System Code (GASC) and Tokamak Energy System Code (TESC), using different methodologies to arrive at CFETR performance parameters under the same CFETR constraints show that the correlation between the physics performance and the fusion performance is consistent, and the computed parameters are in good agreement. Optimization of the first wall surface for tritium breeding and the minimization of the machine size are highly compatible. Variations of the plasma currents and profiles lead to changes in the required normalized physics performance, however, they do not significantly affect the optimized size of the machine. GASC and TESC have also been used to explore a lower aspect ratio, larger volume plasma taking advantage of the engineering flexibility in the CFETR design. Assuming the ITER steady-state scenario physics, the larger plasma together with a moderately higher BT and Ip can result in a high gain Qfus ˜ 12, Pfus ˜ 1 GW machine approaching DEMO-like performance. It is concluded that the CFETR baseline mode can meet the minimum goal of the Fusion Nuclear Science Facility (FNSF) mission and advanced physics will enable it to address comprehensively the outstanding critical technology gaps on the path to a demonstration reactor (DEMO). Before proceeding with CFETR construction steady-state operation has to be demonstrated, further development is needed to solve the divertor heat load issue, and blankets have to be designed with tritium breeding ratio (TBR) >1 as a target.

  20. National Electrical Code in Power Engineering Course for Electrical Engineering Curriculum

    Science.gov (United States)

    Azizur, Rahman M. M.

    2011-01-01

    In order to ensure the safety of their inhabitants and properties, the residential, industrial and business installations require complying with NEC (national electrical code) for electrical systems. Electrical design engineers and technicians rely heavily on these very important design guidelines. However, these design guidelines are not formally…

  1. Human factor engineering applied to nuclear power plant design

    Energy Technology Data Exchange (ETDEWEB)

    Manrique, A. [TECNATOM SA, BWR General Electric Business Manager, Madrid (Spain); Valdivia, J.C. [TECNATOM SA, Operation Engineering Project Manager, Madrid (Spain); Jimenez, A. [TECNATOM SA, Operation Engineering Div. Manager, Madrid (Spain)

    2001-07-01

    For the design and construction of new nuclear power plants as well as for maintenance and operation of the existing ones new man-machine interface designs and modifications are been produced. For these new designs Human Factor Engineering must be applied the same as for any other traditional engineering discipline. Advantages of implementing adequate Human Factor Engineering techniques in the design of nuclear reactors have become not only a fact recognized by the majority of engineers and operators but also an explicit requirement regulated and mandatory for the new designs of the so called advanced reactors. Additionally, the big saving achieved by a nuclear power plant having an operating methodology which significantly decreases the risk of operating errors makes it necessary and almost vital its implementation. The first step for this is preparing a plan to incorporate all the Human Factor Engineering principles and developing an integral design of the Instrumentation and Control and Man-machine interface systems. (author)

  2. GALILEE: A nuclear data processing system for transport, depletion and shielding codes

    Energy Technology Data Exchange (ETDEWEB)

    COSTE-DELCLAUX, Mireille [Commissariat a l' Energie Atomique, CEA Saclay, DEN/DANS/DM2S/SERMA/LLPR, 91191 Gif sur Yvette CEDEX (France)

    2008-07-01

    The Nuclear Data Processing System for Transport, Depletion and Shielding Codes GALILEE is part of a CEA global development program dedicated to fine modelling of nuclear systems. The other projects contributing to this aim are APOLLO3 inherited from DESCARTES (Calvin and Fedon-Magnaud, 2007) which treats deterministic transport, TRIPOLI-4 (Diop et al., 2006) which treats Monte Carlo transport and DARWIN3 (Tsilanizara et al., 1999) which solves all fuel cycle problems. GALILEE aims are: - To provide to application codes (deterministic or Monte Carlo transport codes, shielding codes or depletion codes), a tool-box allowing a consistent processing for nuclear data coming from any evaluation given in ENDF-6 format, - To carry out an automatic chain for creating application libraries, - To provide consistent application libraries for modelling a nuclear system. GALILEE project is carried out in synergy with application codes in order to be able to share 'objects' but also 'tools'. (author)

  3. Human Factors in Nuclear Power Engineering in Polish Conditions

    Directory of Open Access Journals (Sweden)

    Agnieszka Kaczmarek-Kacprzak

    2014-09-01

    Full Text Available The paper “Human factors in nuclear power engineering in Polish conditions” focuses on analysis of dynamics of preparing Polish society to build fi rst nuclear power plant in XXI century in Poland. Authors compare experience from constructing nuclear power plant Sizewell B (Great Britain and Sizewell C, which is in preparation phase with polish nuclear power program. Paper includes aspects e.g. of creating nuclear safety culture and social opinion about investment. Human factors in nuclear power engineering are as well important as relevant economical and technical factors, but very often negligible. In Poland where history about Czarnobyl is still alive, and social opinion is created on emotions after accident in Fukushima, human factors are crucial and should be under comprehensive consideration.

  4. Engineering possibilities versus practical implementation. Nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    2015-06-15

    Europe's energy transition has placed the nuclear sector at a crossroads, and members of POWER-GEN Europe's Advisory Board have considered the role of nuclear in Europe's drive towards energy decarbonisation, ahead of the conference. Simon Hobday, Jacob Klimstra, David Porter and Ulla Pettersson talked about the role of nuclear in Europe's energy decarbonisation, nations in Europe where new nuclear generation appears likely, and the future role of the European Commission.

  5. Evaluation of the analysis models in the ASTRA nuclear design code system

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Nam Jin; Park, Chang Jea; Kim, Do Sam; Lee, Kyeong Taek; Kim, Jong Woon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2000-11-15

    In the field of nuclear reactor design, main practice was the application of the improved design code systems. During the process, a lot of basis and knowledge were accumulated in processing input data, nuclear fuel reload design, production and analysis of design data, et al. However less efforts were done in the analysis of the methodology and in the development or improvement of those code systems. Recently, KEPO Nuclear Fuel Company (KNFC) developed the ASTRA (Advanced Static and Transient Reactor Analyzer) code system for the purpose of nuclear reactor design and analysis. In the code system, two group constants were generated from the CASMO-3 code system. The objective of this research is to analyze the analysis models used in the ASTRA/CASMO-3 code system. This evaluation requires indepth comprehension of the models, which is important so much as the development of the code system itself. Currently, most of the code systems used in domestic Nuclear Power Plant were imported, so it is very difficult to maintain and treat the change of the situation in the system. Therefore, the evaluation of analysis models in the ASTRA nuclear reactor design code system in very important.

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

    Energy Technology Data Exchange (ETDEWEB)

    Schultz, Peter Andrew

    2011-12-01

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

  7. Nuclear engineering in the National Polytechnic Institute; Ingenieria nuclear en el Instituto Politecnico Nacional

    Energy Technology Data Exchange (ETDEWEB)

    Del Valle G, E. [IPN, Unidad Profesional Adolfo Lopez Mateos, Anexo del Edif. 9, 1er. piso, Av. Instituto Politecnico Nacional s/n, Col. San Pedro Zacatenco, Mexico 07738 D. F. (Mexico)

    2008-12-15

    In the National Polytechnic Institute the bachelor degree in physics and mathematics, consists of 48 subjects in the common trunk. For the nuclear engineering option, from the fifth semester undergoing 9 specific areas within the Nuclear Engineering Department : introduction to nuclear engineering, power cycles thermodynamics, heat transfer, two courses of nuclear reactors theory, two of nuclear engineering, one course of laboratory and other of radiation protection. There is also a master in nuclear engineering aims train human resources in the area of power and research nuclear reactors to meet the needs of the nuclear industry in Mexico, as well as train highly qualified personnel in branches where are used equipment involving radiation and radioisotopes tale as Medicine, Agriculture and Industry. Among its compulsory subjects are: radiation interaction with the matter, measurements laboratory, reactor physics I and II, reactor engineering, reactor laboratory and thesis seminar. Optional, are: engineering of the radiation protection, computers in the nuclear engineering, nuclear systems dynamics, power plants safety, flow in two phases, reliability and risk analysis, nuclear power systems design, neutron transport theory. Many graduates of this degree have been and are involved in various phases of the nuclear project of Laguna Verde. The Nuclear Engineering Department has a subcritical nuclear reactor of light water and natural uranium and one isotopic source of Pu-Be neutrons of 5 Ci. It also has a multichannel analyzers, calibrated sources of alpha, beta and gamma radiation, a gamma spectrometer of high resolution and low background, a specialized library and one data processing center. In relation particularly to radiation protection, it is clear that there is a lack of specialists, as reflected in radiological control problems in areas such as medicine and industry. Given this situation, it is perceived to be required post-graduate studies at Master and Ph

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-10-01

    The current state of the art in nuclear fuel cycle (NFC) modeling is an eclectic mixture of codes with various levels of applicability, flexibility, and availability. In support of the advanced fuel cycle systems analyses, especially those by the Advanced Fuel Cycle Initiative (AFCI), Unviery of Cincinnati in collaboration with Idaho State University carried out a detailed review of the existing codes describing various aspects of the nuclear fuel cycle and identified the research and development needs required for a comprehensive model of the global nuclear energy infrastructure and the associated nuclear fuel cycles. Relevant information obtained on the NFC codes was compiled into a relational database that allows easy access to various codes' properties. Additionally, the research analyzed the gaps in the NFC computer codes with respect to their potential integration into programs that perform comprehensive NFC analysis.

  9. Major Cyber threat on Nuclear Facility and Key Entry Points of Malicious Codes

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Ickhyun; Kwon, Kookheui [Korea Institute of Nuclear Nonproliferation and Control, Daejeon (Korea, Republic of)

    2013-05-15

    Cyber security incident explicitly shows that the domestic intra net system which is not connected to the Internet can be compromised by the USB based mal ware which was developed by the state-sponsored group. It also tells that the actor for cyber-attack has been changed from script kiddies to state's governments and the target has been changed to nation's main infrastructures such as electricity, transportation and etc. Since the cyber sabotage on nuclear facility has been proven to be possible and can be replicated again with same method, the cyber security on nuclear facility must be strengthened. In this paper, it is explained why the malicious code is the one of the biggest cyber threat in nuclear facility's digital I and C(Instrumentation and Controls) system by analyzing recent cyber attacks and well-known malicious codes. And a feasible cyber attack scenario on nuclear facility's digital I and C system is suggested along with some security measures for prevention of malicious code. As experienced from the cyber sabotage on Iranian nuclear facility in 2010, cyber attack on nuclear facility can be replicated by infecting the computer network with malicious codes. One of the cyber attack scenario on nuclear digital I and C computer network with using malicious code was suggested to help security manager establishing cyber security plan for prevention of malicious code. And some security measures on prevention of malicious code are also provided for reference.

  10. Educating Next Generation Nuclear Criticality Safety Engineers at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    J. D. Bess; J. B. Briggs; A. S. Garcia

    2011-09-01

    One of the challenges in educating our next generation of nuclear safety engineers is the limitation of opportunities to receive significant experience or hands-on training prior to graduation. Such training is generally restricted to on-the-job-training before this new engineering workforce can adequately provide assessment of nuclear systems and establish safety guidelines. Participation in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) can provide students and young professionals the opportunity to gain experience and enhance critical engineering skills. The ICSBEP and IRPhEP publish annual handbooks that contain evaluations of experiments along with summarized experimental data and peer-reviewed benchmark specifications to support the validation of neutronics codes, nuclear cross-section data, and the validation of reactor designs. Participation in the benchmark process not only benefits those who use these Handbooks within the international community, but provides the individual with opportunities for professional development, networking with an international community of experts, and valuable experience to be used in future employment. Traditionally students have participated in benchmarking activities via internships at national laboratories, universities, or companies involved with the ICSBEP and IRPhEP programs. Additional programs have been developed to facilitate the nuclear education of students while participating in the benchmark projects. These programs include coordination with the Center for Space Nuclear Research (CSNR) Next Degree Program, the Collaboration with the Department of Energy Idaho Operations Office to train nuclear and criticality safety engineers, and student evaluations as the basis for their Master's thesis in nuclear engineering.

  11. An engineering code to analyze hypersonic thermal management systems

    Science.gov (United States)

    Vangriethuysen, Valerie J.; Wallace, Clark E.

    1993-01-01

    Thermal loads on current and future aircraft are increasing and as a result are stressing the energy collection, control, and dissipation capabilities of current thermal management systems and technology. The thermal loads for hypersonic vehicles will be no exception. In fact, with their projected high heat loads and fluxes, hypersonic vehicles are a prime example of systems that will require thermal management systems (TMS) that have been optimized and integrated with the entire vehicle to the maximum extent possible during the initial design stages. This will not only be to meet operational requirements, but also to fulfill weight and performance constraints in order for the vehicle to takeoff and complete its mission successfully. To meet this challenge, the TMS can no longer be two or more entirely independent systems, nor can thermal management be an after thought in the design process, the typical pervasive approach in the past. Instead, a TMS that was integrated throughout the entire vehicle and subsequently optimized will be required. To accomplish this, a method that iteratively optimizes the TMS throughout the vehicle will not only be highly desirable, but advantageous in order to reduce the manhours normally required to conduct the necessary tradeoff studies and comparisons. A thermal management engineering computer code that is under development and being managed at Wright Laboratory, Wright-Patterson AFB, is discussed. The primary goal of the code is to aid in the development of a hypersonic vehicle TMS that has been optimized and integrated on a total vehicle basis.

  12. EXTENSION OF THE NUCLEAR REACTION MODEL CODE EMPIRE TO ACTINIDES NUCLEAR DATA EVALUATION.

    Energy Technology Data Exchange (ETDEWEB)

    CAPOTE,R.; SIN, M.; TRKOV, A.; HERMAN, M.; CARLSON, B.V.; OBLOZINSKY, P.

    2007-04-22

    Recent extensions and improvements of the EMPIRE code system are outlined. They add new capabilities to the code, such as prompt fission neutron spectra calculations using Hauser-Feshbach plus pre-equilibrium pre-fission spectra, cross section covariance matrix calculations by Monte Carlo method, fitting of optical model parameters, extended set of optical model potentials including new dispersive coupled channel potentials, parity-dependent level densities and transmission through numerically defined fission barriers. These features, along with improved and validated ENDF formatting, exclusive/inclusive spectra, and recoils make the current EMPIRE release a complete and well validated tool for evaluation of nuclear data at incident energies above the resonance region. The current EMPIRE release has been used in evaluations of neutron induced reaction files for {sup 232}Th and {sup 231,233}Pa nuclei in the fast neutron region at IAEA. Triple-humped fission barriers and exclusive pre-fission neutron spectra were considered for the fission data evaluation. Total, fission, capture and neutron emission cross section, average resonance parameters and angular distributions of neutron scattering are in excellent agreement with the available experimental data.

  13. Near-term lunar nuclear thermal rocket engine options

    Science.gov (United States)

    Pelaccio, Dennis G.; Scheil, Christine M.; Collins, John T.

    1991-01-01

    The Nuclear Thermal Rocket (NTR) is an attractive candidate propulsion system option for manned planetary missions. Its high performance capability for such missions translates into a substantial reduction in low-earth-orbit (LEO) required mass and trip times with increased operational flexibility. This study examined NTR engine options that could support near-term lunar mission operations. Expander and gas generator cycle, solid-core NERVA derivative reactor-based NTR engines were investigated. Weight, size, operational characteristics, and design features for representative NTR engine concepts are presented. The impact of using these NTR engines for a typical lunar mission scenario is also examined.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-03-01

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

  15. Perspectives on Validation and Uncertainty Evaluation of SFR Nuclear Design Code

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Moohoon; Choi, Yong Won; Shin, Andong; Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2014-05-15

    Fast reactors such as PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor) developed by KAERI have fundamental differences in terms of core characteristics and associated fuel cycle compared to thermal reactors, which need specific new effort for code validation. In current PWRs, nuclear design code systems have been validated using numerous data accumulated by wide operating experience, and its uncertainty can be assessed by statistical methods. However, in order to validate code systems for SFRs with little operating experience, and particularly prototype reactor, new approaches are required. In this study, a current procedure for validation and uncertainty evaluation is reviewed in nuclear design code systems for PWRs, and global approaches for validation of SFR code systems are surveyed. Through these reviews, perspectives on nuclear design code validation for SFRs are identified. In case of neutronics code V and V, current procedure for PWRs and global approaches for SFRs were reviewed and surveyed. Though this review, perspectives on nuclear design code V and V and uncertainty evaluation for SFRs were identified. Further study will be implemented to obtain more insight on code validation.

  16. Ultra-relativistic nuclear collisions: event shape engineering

    CERN Document Server

    Schukraft, Jurgen; Voloshin, Sergei A

    2013-01-01

    The evolution of the system created in a high energy nuclear collision is very sensitive to the fluctuations in the initial geometry of the system. In this letter we show how one can utilize these large fluctuations to select events corresponding to a specific initial shape. Such an "event shape engineering" opens many new possibilities in quantitative test of the theory of high energy nuclear collisions and understanding the properties of high density hot QCD matter.

  17. Coaxial Ring Cyclotron as a Perspective Nuclear Power Engineering Machine

    OpenAIRE

    Tumanyan, A. R.; Simonyan, Kh. A.; Mkrtchyan, R. L.; Amatuni, A. Ts.; Avakyan, R. O.; Khudaverdyan, A. G.

    1995-01-01

    The circuit arrangement of the proposed coaxial ring cyclotron (CRC) is described, and its main advantages, such as simple injection technique, several injected beams summation option, high efficiency, are considered. The proposed proton accelerator is a perspective machine for the solution of the main problems of the present day nuclear power engineering as well as for the next-generation nuclear power plants, representing a combination of subcritical reactors and particle accelerators. The ...

  18. Nuclear electric propulsion mission engineering study. Volume 2: Final report

    Science.gov (United States)

    1973-01-01

    Results of a mission engineering analysis of nuclear-thermionic electric propulsion spacecraft for unmanned interplanetary and geocentric missions are summarized. Critical technologies associated with the development of nuclear electric propulsion (NEP) are assessed, along with the impact of its availability on future space programs. Outer planet and comet rendezvous mission analysis, NEP stage design for geocentric and interplanetary missions, NEP system development cost and unit costs, and technology requirements for NEP stage development are studied.

  19. The open-cycle gas-core nuclear rocket engine - Some engineering considerations.

    Science.gov (United States)

    Taylor, M. F.; Whitmarsh, C. L., Jr.; Sirocky, P. J., Jr.; Iwanczyk, L. C.

    1971-01-01

    A preliminary design study of a conceptual 6000-MW open-cycle gas-core nuclear rocket engine system was made. The engine has a thrust of 44,200 lb and a specific impulse of 4400 sec. The nuclear fuel is uranium-235 and the propellant is hydrogen. Critical fuel mass was calculated for several reactor configurations. Major components of the reactor (reflector, pressure vessel) and the waste heat rejection system were considered conceptually and were sized.

  20. Grooved Fuel Rings for Nuclear Thermal Rocket Engines

    Science.gov (United States)

    Emrich, William

    2009-01-01

    An alternative design concept for nuclear thermal rocket engines for interplanetary spacecraft calls for the use of grooved-ring fuel elements. Beyond spacecraft rocket engines, this concept also has potential for the design of terrestrial and spacecraft nuclear electric-power plants. The grooved ring fuel design attempts to retain the best features of the particle bed fuel element while eliminating most of its design deficiencies. In the grooved ring design, the hydrogen propellant enters the fuel element in a manner similar to that of the Particle Bed Reactor (PBR) fuel element.

  1. Nuclear engineering vocabulary; Vocabulaire de l'ingenierie nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-07-01

    The terms, expressions and definitions presented in this booklet come from the works carried out by the French specialized commission of nuclear engineering terminology and neology. This selection of terms cannot be found, in general, in classical dictionaries, or can be found but with a different meaning than the one used in nuclear engineering. All terms and expressions contained in this booklet have been already published in different issues of the Official Journal of the French Republic. This publication makes their use mandatory in replacement of foreign language equivalents inside all government services and public buildings. (J.S.)

  2. Nuclear Explosion Monitoring Research and Engineering Program - Strategic Plan

    Energy Technology Data Exchange (ETDEWEB)

    Casey, Leslie A. [DOE/NNSA

    2004-09-01

    The Department of Energy (DOE)/National Nuclear Security Administration (NNSA) Nuclear Explosion Monitoring Research and Engineering (NEM R&E) Program is dedicated to providing knowledge, technical expertise, and products to US agencies responsible for monitoring nuclear explosions in all environments and is successful in turning scientific breakthroughs into tools for use by operational monitoring agencies. To effectively address the rapidly evolving state of affairs, the NNSA NEM R&E program is structured around three program elements described within this strategic plan: Integration of New Monitoring Assets, Advanced Event Characterization, and Next-Generation Monitoring Systems. How the Program fits into the National effort and historical accomplishments are also addressed.

  3. Hyperthermal Environments Simulator for Nuclear Rocket Engine Development

    Science.gov (United States)

    Litchford, Ron J.; Foote, John P.; Clifton, W. B.; Hickman, Robert R.; Wang, Ten-See; Dobson, Christopher C.

    2011-01-01

    An arc-heater driven hyperthermal convective environments simulator was recently developed and commissioned for long duration hot hydrogen exposure of nuclear thermal rocket materials. This newly established non-nuclear testing capability uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce hightemperature pressurized hydrogen flows representative of nuclear reactor core environments, excepting radiation effects, and is intended to serve as a low-cost facility for supporting non-nuclear developmental testing of hightemperature fissile fuels and structural materials. The resulting reactor environments simulator represents a valuable addition to the available inventory of non-nuclear test facilities and is uniquely capable of investigating and characterizing candidate fuel/structural materials, improving associated processing/fabrication techniques, and simulating reactor thermal hydraulics. This paper summarizes facility design and engineering development efforts and reports baseline operational characteristics as determined from a series of performance mapping and long duration capability demonstration tests. Potential follow-on developmental strategies are also suggested in view of the technical and policy challenges ahead. Keywords: Nuclear Rocket Engine, Reactor Environments, Non-Nuclear Testing, Fissile Fuel Development.

  4. Assessement of Codes and Standards Applicable to a Hydrogen Production Plant Coupled to a Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    M. J. Russell

    2006-06-01

    This is an assessment of codes and standards applicable to a hydrogen production plant to be coupled to a nuclear reactor. The result of the assessment is a list of codes and standards that are expected to be applicable to the plant during its design and construction.

  5. Qualitative knowledge engineering for nuclear applications

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jae H.; Kim, Ko R.; Lee, Jae C.

    1996-01-01

    After the TMI nuclear power plant accident, the two topics of plant safety and operational efficiency became more important areas of artificial intelligence, which have difference characteristics. Qualitative deep model is the recently prospective technology of AI, that can overcome several handicaps of the existing expert systems such as lack of common sense reasoning. The application of AI to the large and complex system like nuclear power plants is typically and effectively done through a module-based hierarchical system. As each module has to be built with suitable AI system. Through the experiences of hierarchical system construction, we aimed to develop basic AI application schemes for the power plant safety and operational efficiency as well as basic technologies for autonomous power plants. The goal of the research is to develop qualitative reasoning technologies for nuclear power plants. For this purpose, the development of qualitative modeling technologies and qualitative behaviour prediction technologies of the power plant are accomplished. In addition, the feasibility of application of typical qualitative reasoning technologies to power plants is studied . The goal of the application is to develop intelligent control technologies of power plants, support technologies. For these purposes, we analyzed the operation of power plants according to its operation purpose: power generation operation, shut-down and start-up operation. As a result, qualitative model of basic components were sketched, including pipes, valves, pumps and heat exchangers. Finally, plant behaviour prediction technologies through qualitative plant heat transfer model and design support technologies through 2nd-order differential equation were developed. For the construction of AI system of power plants, we have studied on the mixed module based hierarchical software. As a testbed, we have considered the spent fuel system and the feedwater system. We also studied the integration

  6. Training of nuclear criticality safety engineers

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, R.G. [Oak Ridge Y-12 Plant, TN (United States)

    1997-06-01

    The site specific analysis of nuclear criticality training needs is very briefly described. Analysis indicated that the four major components required were analysis, surveillance, business practices or administration, and emergency preparedness. The analysis component was further divided into process analysis, accident analysis, and transportation analysis. Ten subject matter areas for the process analysis component were identified as candidates for class development. Training classes developed from the job content analysis have demonstrated that the specialized information can be successfully delivered to new entrants. 1 fig.

  7. Evolution of the CYCLE code for the system analysis of the nuclear fuel cycle

    Directory of Open Access Journals (Sweden)

    A.G. Kalashnikov

    2016-06-01

    Full Text Available The CYCLE code is intended to simulate mathematically the operation of a nuclear power system (NPS with thermal and fast reactors in an open or closed nuclear fuel cycle, to develop scenarios of efficient nuclear power evolution in Russia and to analyze trends in global nuclear power. The code is based on a well-known software program, WIMSD-5B, broadly used for the design of thermal reactor cells, and on a 2D multi-group software system, RZA, for the fast neutron reactor simulation. The CYCLE code was developed at IPPE in Obninsk. This paper presents a brief review of the capabilities and information on the current status of the CYCLE code. The code allows simulation of key facilities of the external fuel cycle (fuel fabrication and reprocessing facilities, SNF storage, uranium, plutonium, neptunium, americium and curium stores, RW long-term storage sites, nuclear reactors, including RBMK-1000 reactors, existing and advanced VVER reactors (using different fuel types, and fast reactors (both existing and innovative. As an important feature, the CYCLE code allows the evolution of the fuel's nuclide composition both in reactors and at the external fuel cycle phase to be considered in details. Offered as an extra option is the capability to calculate a variety of the nuclear fuel cycle cost parameters for nuclear power plants with thermal and fast reactors. For years, the code has been successfully used as part of INPRO, an international innovative nuclear reactor and fuel cycle project. The results of studies into the Russian NPS evolution scenarios were presented at Global 2011. Some other of the CYCLE-based simulation results were presented at Global 2015.

  8. Subgroup A : nuclear model codes report to the Sixteenth Meeting of the WPEC

    Energy Technology Data Exchange (ETDEWEB)

    Talou, P. (Patrick); Chadwick, M. B. (Mark B.); Dietrich, F. S.; Herman, M.; Kawano, T. (Toshihiko); Konig, A.; Obložinský, P.

    2004-01-01

    The Subgroup A activities focus on the development of nuclear reaction models and codes, used in evaluation work for nuclear reactions from the unresolved energy region up to the pion threshold production limit, and for target nuclides from the low teens and heavier. Much of the efforts are devoted by each participant to the continuing development of their own Institution codes. Progresses in this arena are reported in detail for each code in the present document. EMPIRE-II is of public access. The release of the TALYS code has been announced for the ND2004 Conference in Santa Fe, NM, October 2004. McGNASH is still under development and is not expected to be released in the very near future. In addition, Subgroup A members have demonstrated a growing interest in working on common modeling and codes capabilities, which would significantly reduce the amount of duplicate work, help manage efficiently the growing lines of existing codes, and render codes inter-comparison much easier. A recent and important activity of the Subgroup A has therefore been to develop the framework and the first bricks of the ModLib library, which is constituted of mostly independent pieces of codes written in Fortran 90 (and above) to be used in existing and future nuclear reaction codes. Significant progresses in the development of ModLib have been made during the past year. Several physics modules have been added to the library, and a few more have been planned in detail for the coming year.

  9. Codes of Practice related to Harbour and Coastal Engineering in Denmark

    DEFF Research Database (Denmark)

    Burcharth, H. F.

    2000-01-01

    Codes of practice for building and civil engineering works have been produced since 1893 by the "Danish Society of Engineers". Among the early codes are: Reinforces concrete structures (1908, 1943), calculation of reinforced concrete structures in harbour works (1926), Harbour Works (1927), Steel...

  10. Annual report of nuclear code evaluation committee for fiscal 2000 year

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-03-01

    In this report, research results discussed in fiscal 2000 year at Nuclear Code Evaluation Committee of Nuclear Code Research Committee were summarized. In 2000, papers mainly on the three topics of (1) present status of burnup credit evaluation methods, (2) issues concerning convergence of criticality calculation and (3) estimation methods for errors associated with criticality calculation based on nuclear data covariance file, are presented and discussed. These results are sorted to grasp the present status of related technology and described in this report. (author)

  11. The neutron's children nuclear engineers and the shaping of identity

    CERN Document Server

    Johnston, Sean

    2012-01-01

    This account tracks the Allied atomic energy experts who emerged from the Manhattan Project to explore optimistic but distinct paths in the USA, UK and Canada. Characterised successively as admired atomic scientists, mistrusted spies and heroic engineers, their identities were ultimately shaped by nuclear accidents.

  12. Nuclear magnetic resonance in environmental engineering: principles and applications.

    NARCIS (Netherlands)

    Lens, P.N.L.; Hemminga, M.A.

    1998-01-01

    This paper gives an introduction to nuclear magnetic resonance spectroscopy (NMR) and magnetic resonance imaging (MRI) in relation to applications in the field of environmental science and engineering. The underlying principles of high resolution solution and solid state NMR, relaxation time measure

  13. U.S. Nuclear Engineering Education: Status and Prospects.

    Science.gov (United States)

    National Academy of Sciences - National Research Council, Washington, DC. Commission on Engineering and Technical Systems.

    This study examines the status of and outlook for nuclear engineering (NE) in the United States. The study resulted from a concern about the downward trends in student enrollments in NE, in both graduate and undergraduate programs. Concerns have also been expressed about the declining number of U.S. university NE departments and programs, the…

  14. Evaluating Open-Source Full-Text Search Engines for Matching ICD-10 Codes.

    Science.gov (United States)

    Jurcău, Daniel-Alexandru; Stoicu-Tivadar, Vasile

    2016-01-01

    This research presents the results of evaluating multiple free, open-source engines on matching ICD-10 diagnostic codes via full-text searches. The study investigates what it takes to get an accurate match when searching for a specific diagnostic code. For each code the evaluation starts by extracting the words that make up its text and continues with building full-text search queries from the combinations of these words. The queries are then run against all the ICD-10 codes until a match indicates the code in question as a match with the highest relative score. This method identifies the minimum number of words that must be provided in order for the search engines choose the desired entry. The engines analyzed include a popular Java-based full-text search engine, a lightweight engine written in JavaScript which can even execute on the user's browser, and two popular open-source relational database management systems.

  15. An Interactive Reverse Engineering Environment for Large-Scale C++ Code

    NARCIS (Netherlands)

    Telea, Alexandru; Voinea, Lucian

    2008-01-01

    Few toolsets for reverse-engineering and understanding of C++ code provide parsing and fact extraction, querying, analysis and code metrics, navigation, and visualization of source-code-level facts in a way which is as easy-to-use as integrated development environments (IDEs) are for forward enginee

  16. An Interactive Reverse Engineering Environment for Large-Scale C plus plus Code

    NARCIS (Netherlands)

    Telea, Alexandru; Voinea, Lucian; Spencer, SN

    2008-01-01

    Few toolsets for reverse-engineering and understanding of C++ code provide parsing and fact extraction, querying, analysis and code metrics, navigation, and visualization of source-code-level facts in a way which is as easy-to-use as integrated development environments (IDEs) are for forward enginee

  17. An introduction to the engineering of fast nuclear reactors

    CERN Document Server

    Judd, Anthony M

    2014-01-01

    An invaluable resource for both graduate-level engineering students and practising nuclear engineers who want to expand their knowledge of fast nuclear reactors, the reactors of the future! This book is a concise yet comprehensive introduction to all aspects of fast reactor engineering. It covers topics including neutron physics; neutron flux spectra; flux distribution; Doppler and coolant temperature coefficients; the performance of ceramic and metal fuels under irradiation, structural changes, and fission-product migration; the effects of irradiation and corrosion on structural materials, irradiation swelling; heat transfer in the reactor core and its effect on core design; coolants including sodium and lead-bismuth alloy; coolant circuits; pumps; heat exchangers and steam generators; and plant control. The book includes new discussions on lead-alloy and gas coolants, metal fuel, the use of reactors to consume radioactive waste, and accelerator-driven subcritical systems.

  18. A Framework for Reverse Engineering Large C++ Code Bases

    NARCIS (Netherlands)

    Telea, Alexandru; Byelas, Heorhiy; Voinea, Lucian

    2008-01-01

    When assessing the quality and maintainability of large C++ code bases, tools are needed for extracting several facts from the source code, such as: architecture, structure, code smells, and quality metrics. Moreover, these facts should be presented in such ways so that one can correlate them and fi

  19. A Framework for Reverse Engineering Large C++ Code Bases

    NARCIS (Netherlands)

    Telea, Alexandru; Byelas, Heorhiy; Voinea, Lucian

    2009-01-01

    When assessing the quality and maintainability of large C++ code bases, tools are needed for extracting several facts from the source code, such as: architecture, structure, code smells, and quality metrics. Moreover, these facts should be presented in such ways so that one can correlate them and fi

  20. Modification of Neutron Kinetic Code for Plate Type Fuel Nuclear Reactor

    Directory of Open Access Journals (Sweden)

    Salah Ud-Din Khan

    2013-01-01

    Full Text Available The research is conducted on the modification of neutron kinetic code for the plate type fuel nuclear reactor. REMARK is a neutron kinetic code that works only for the cylindrical type fuel nuclear reactor. In this research, our main emphasis is on the modification of this code in order to be applicable for the plate type fuel nuclear reactor. For this purpose, detailed mathematical studies have been performed and are subjected to write the program in Fortran language. Since REMARK code is written in Fortran language, so we have developed the program in Fortran and then inserted it into the source library of the code. The main emphasis is on the modification of subroutine in the source library of the code for hexagonal fuel assemblies with plate type fuel elements in it. The number of steps involved in the modification of the code has been included in the paper. The verification studies were performed by considering the small modular reactor with hexagonal assemblies and plate type fuel in it to find out the power distribution of the reactor core. The purpose of the research is to make the code work for the hexagonal fuel assemblies with plate type fuel element.

  1. Computer simulation in nuclear science and engineering

    Energy Technology Data Exchange (ETDEWEB)

    Akiyama, Mamoru; Miya, Kenzo; Iwata, Shuichi; Yagawa, Genki; Kondo, Shusuke (Tokyo Univ. (Japan)); Hoshino, Tsutomu; Shimizu, Akinao; Takahashi, Hiroshi; Nakagawa, Masatoshi

    1992-03-01

    The numerical simulation technology used for the design of nuclear reactors includes the scientific fields of wide range, and is the cultivated technology which grew in the steady efforts to high calculation accuracy through safety examination, reliability verification test, the assessment of operation results and so on. Taking the opportunity of putting numerical simulation to practical use in wide fields, the numerical simulation of five basic equations which describe the natural world and the progress of its related technologies are reviewed. It is expected that numerical simulation technology contributes to not only the means of design study but also the progress of science and technology such as the construction of new innovative concept, the exploration of new mechanisms and substances, of which the models do not exist in the natural world. The development of atomic energy and the progress of computers, Boltzmann's transport equation and its periphery, Navier-Stokes' equation and its periphery, Maxwell's electromagnetic field equation and its periphery, Schroedinger wave equation and its periphery, computational solid mechanics and its periphery, and probabilistic risk assessment and its periphery are described. (K.I.).

  2. Training in nuclear engineering companies; La formacion en las empresas de ingenieria del ambito nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Perezagua, R. L.

    2013-03-01

    The importance of training is growing in all business areas and fields and especially in hi-tech companies like engineering firms. Nuclear projects are highly multidisciplinary and, even in the initial awarding and pre-construction phases, need to be staffed with personnel that is well-prepared and highly-qualified in areas that, in most cases, are not covered by university studies. This article examines the variables that influence the design of specific training for nuclear projects in engineering firms, along with new training technologies (e-learning) and new regulatory aspects (IS-12). (Author)

  3. 77 FR 3073 - American Society of Mechanical Engineers (ASME) Codes and New and Revised ASME Code Cases...

    Science.gov (United States)

    2012-01-23

    ... Maintenance of Nuclear Power Plants (OM Code). The final rule also incorporated by reference (with conditions... described in the GALL report, or propose alternatives (exceptions) for the NRC to review as part of a plant... acceptable approach for aging management--through inservice inspection--of PWR nickel-alloy upper vessel head...

  4. Engine System Model Development for Nuclear Thermal Propulsion

    Science.gov (United States)

    Nelson, Karl W.; Simpson, Steven P.

    2006-01-01

    In order to design, analyze, and evaluate conceptual Nuclear Thermal Propulsion (NTP) engine systems, an improved NTP design and analysis tool has been developed. The NTP tool utilizes the Rocket Engine Transient Simulation (ROCETS) system tool and many of the routines from the Enabler reactor model found in Nuclear Engine System Simulation (NESS). Improved non-nuclear component models and an external shield model were added to the tool. With the addition of a nearly complete system reliability model, the tool will provide performance, sizing, and reliability data for NERVA-Derived NTP engine systems. A new detailed reactor model is also being developed and will replace Enabler. The new model will allow more flexibility in reactor geometry and include detailed thermal hydraulics and neutronics models. A description of the reactor, component, and reliability models is provided. Another key feature of the modeling process is the use of comprehensive spreadsheets for each engine case. The spreadsheets include individual worksheets for each subsystem with data, plots, and scaled figures, making the output very useful to each engineering discipline. Sample performance and sizing results with the Enabler reactor model are provided including sensitivities. Before selecting an engine design, all figures of merit must be considered including the overall impacts on the vehicle and mission. Evaluations based on key figures of merit of these results and results with the new reactor model will be performed. The impacts of clustering and external shielding will also be addressed. Over time, the reactor model will be upgraded to design and analyze other NTP concepts with CERMET and carbide fuel cores.

  5. CTCN: Colloid transport code -- nuclear; A user`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Jain, R.

    1993-09-01

    This report describes the CTCN computer code, designed to solve the equations of transient colloidal transport of radionuclides in porous and fractured media. This Fortran 77 package solves systems of coupled nonlinear differential-algebraic equations with a wide range of boundary conditions. The package uses the Method of Lines technique with a special section which forms finite-difference discretizations in up to four spatial dimensions to automatically convert the system into a set of ordinary differential equations. The CTCN code then solves these equations using a robust, efficient ODE solver. Thus CTCN can be used to solve population balance equations along with the usual transport equations to model colloid transport processes or as a general problem solver to treat up to four-dimensional differential-algebraic systems.

  6. Colloid transport code-nuclear user`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Jain, R. [New Mexico Univ., Albuquerque, NM (United States)

    1992-04-03

    This report describes the CTCN computer code, designed to solve the equations of transient colloidal transport of radionuclides in porous and fractured media. This Fortran 77 package solves systems of coupled nonlinear differential equations with a wide range of boundary conditions. The package uses the Method of Lines technique with a special section which forms finite-difference discretizations in up to four spatial dimensions to automatically convert the system into a set of ordinary differential equations. The CTCN code then solves these equations using a robust, efficient ODE solver. Thus CTCN can be used to solve population balance equations along with the usual transport equations to model colloid transport processes or as a general problem solver to treat up to four-dimensional differential systems.

  7. A comparison of shuttle vernier engine plume contamination with CONTAM 3.4 code predictions

    Science.gov (United States)

    Maag, Carl R.; Jones, Thomas M.; Rao, Shankar M.; Linder, W. Kelly

    1992-01-01

    In 1985, using the CONTAM 3.2 code, it was predicted that the shuttle Primary Reaction Control System (PRCS) and Vernier Reaction Control System (VRCS) engines could be potential contamination sources to sensitive surfaces located within the shuttle payload bay. Spaceflight test data on these engines is quite limited. Shuttle mission STS-32, the Long Duration Exposure Facility retrieval mission, was instrumented with an experiment that provided the design engineer with evidence that contaminant species from the VRCS engines can enter the payload bay. More recently, the most recent version of the analysis code, CONTAM 3.4, has re-examined the contamination potential of these engines.

  8. γ-ray shielding behaviors of some nuclear engineering materials

    Energy Technology Data Exchange (ETDEWEB)

    Mann, Kulwinder Singh [Dept. of Physics, D.A.V. College, Punjab (India)

    2017-06-15

    The essential requirement of a material to be used for engineering purposes at nuclear establishments is its ability to attenuate the most penetrating ionizing radiations, gamma (γ)-rays. Mostly, high-Z materials such as heavy concrete, lead, mercury, and their mixtures or alloys have been used in the construction of nuclear establishments and thus termed as nuclear engineering materials (NEM). The NEM are classified into two categories, namely opaque and transparent, depending on their behavior towards the visible spectrum of EM waves. The majority of NEM are opaque. By contrast, various types of glass, which are transparent to visible light, are necessary at certain places in the nuclear establishments. In the present study, γ-ray shielding behaviors (GSB) of six glass samples (transparent NEM) were evaluated and compared with some opaque NEM in a wide range of energy (15 keV–15 MeV) and optical thickness (OT). The study was performed by computing various γ-ray shielding parameters (GSP) such as the mass attenuation coefficient, equivalent atomic number, and buildup factor. A self-designed and validated computer-program, the buildup factor-tool, was used for various computations. It has been established that some glass samples show good GSB, thus can safely be used in the construction of nuclear establishments in conjunction with the opaque NEM as well.

  9. γ-Ray Shielding Behaviors of Some Nuclear Engineering Materials

    Directory of Open Access Journals (Sweden)

    Kulwinder Singh Mann

    2017-06-01

    Full Text Available The essential requirement of a material to be used for engineering purposes at nuclear establishments is its ability to attenuate the most penetrating ionizing radiations, gamma (γ-rays. Mostly, high-Z materials such as heavy concrete, lead, mercury, and their mixtures or alloys have been used in the construction of nuclear establishments and thus termed as nuclear engineering materials (NEM. The NEM are classified into two categories, namely opaque and transparent, depending on their behavior towards the visible spectrum of EM waves. The majority of NEM are opaque. By contrast, various types of glass, which are transparent to visible light, are necessary at certain places in the nuclear establishments. In the present study, γ-ray shielding behaviors (GSB of six glass samples (transparent NEM were evaluated and compared with some opaque NEM in a wide range of energy (15 keV–15 MeV and optical thickness (OT. The study was performed by computing various γ-ray shielding parameters (GSP such as the mass attenuation coefficient, equivalent atomic number, and buildup factor. A self-designed and validated computer-program, the buildup factor-tool, was used for various computations. It has been established that some glass samples show good GSB, thus can safely be used in the construction of nuclear establishments in conjunction with the opaque NEM as well.

  10. Nuclear morphology and deformation in engineered cardiac myocytes and tissues.

    Science.gov (United States)

    Bray, Mark-Anthony P; Adams, William J; Geisse, Nicholas A; Feinberg, Adam W; Sheehy, Sean P; Parker, Kevin K

    2010-07-01

    Cardiac tissue engineering requires finely-tuned manipulation of the extracellular matrix (ECM) microenvironment to optimize internal myocardial organization. The myocyte nucleus is mechanically connected to the cell membrane via cytoskeletal elements, making it a target for the cellular response to perturbation of the ECM. However, the role of ECM spatial configuration and myocyte shape on nuclear location and morphology is unknown. In this study, printed ECM proteins were used to configure the geometry of cultured neonatal rat ventricular myocytes. Engineered one- and two-dimensional tissue constructs and single myocyte islands were assayed using live fluorescence imaging to examine nuclear position, morphology and motion as a function of the imposed ECM geometry during diastolic relaxation and systolic contraction. Image analysis showed that anisotropic tissue constructs cultured on microfabricated ECM lines possessed a high degree of nuclear alignment similar to that found in vivo; nuclei in isotropic tissues were polymorphic in shape with an apparently random orientation. Nuclear eccentricity was also increased for the anisotropic tissues, suggesting that intracellular forces deform the nucleus as the cell is spatially confined. During systole, nuclei experienced increasing spatial confinement in magnitude and direction of displacement as tissue anisotropy increased, yielding anisotropic deformation. Thus, the nature of nuclear displacement and deformation during systole appears to rely on a combination of the passive myofibril spatial organization and the active stress fields induced by contraction. Such findings have implications in understanding the genomic consequences and functional response of cardiac myocytes to their ECM surroundings under conditions of disease.

  11. Nuclear electric propulsion mission engineering study. Volume 1: Executive summary

    Science.gov (United States)

    1973-01-01

    Results of a mission engineering analysis of nuclear-thermionic electric propulsion spacecraft for unmanned interplanetary and geocentric missions are summarized. Critical technologies associated with the development of nuclear electric propulsion (NEP) are assessed. Outer planet and comet rendezvous mission analysis, NEP stage design for geocentric and interplanetary missions, NEP system development cost and unit costs, and technology requirements for NEP stage development are studied. The NEP stage design provides both inherent reliability and high payload mass capability. The NEP stage and payload integration was found to be compatible with the space shuttle.

  12. Handbook of nuclear engineering: vol 1: nuclear engineering fundamentals; vol 2: reactor design; vol 3: reactor analysis; vol 4: reactors of waste disposal and safeguards

    CERN Document Server

    2013-01-01

    The Handbook of Nuclear Engineering is an authoritative compilation of information regarding methods and data used in all phases of nuclear engineering. Addressing nuclear engineers and scientists at all academic levels, this five volume set provides the latest findings in nuclear data and experimental techniques, reactor physics, kinetics, dynamics and control. Readers will also find a detailed description of data assimilation, model validation and calibration, sensitivity and uncertainty analysis, fuel management and cycles, nuclear reactor types and radiation shielding. A discussion of radioactive waste disposal, safeguards and non-proliferation, and fuel processing with partitioning and transmutation is also included. As nuclear technology becomes an important resource of non-polluting sustainable energy in the future, The Handbook of Nuclear Engineering is an excellent reference for practicing engineers, researchers and professionals.

  13. Educational Innovation in the Design of an Online Nuclear Engineering Curriculum

    Science.gov (United States)

    Hall, Simin; Jones, Brett D.; Amelink, Catherine; Hu, Deyu

    2013-01-01

    The purpose of this paper is to describe the development and implementation phases of online graduate nuclear engineering courses that are part of the Graduate Nuclear Engineering Certificate program at Virginia Tech. Virginia Tech restarted its nuclear engineering program in the Fall of 2007 with 60 students, and by 2009, the enrollment had grown…

  14. Establishment of nuclear knowledge-information base; development of courseware on introductory nuclear engineering and establishment of digital education platform

    Energy Technology Data Exchange (ETDEWEB)

    Song, Jong Soon; Na, Mang Yun; Lee, Goung Jin; Yang, Won Sik [Chosun University, Gwangju (Korea)

    2002-01-01

    In this research, there are two major tasks. The first one is a development of digital course-ware program for introductory nuclear engineering. For this task, a development of lecture note is followed by lecture Slide files in html file format, which is based on web. For this purpose, following activities were performed; collection of related materials. planning of overall courseware, writing of lecture note and exercise plan, and securing the computer programs and codes needed. The second task of this research is to plan and install several hardwares in a multimedia class room as a digital education platform. The platform includes smart board with touch screen functionality, network server and personal computers. The digital education platform was established as a multimedia class room in the 2nd College of Engineering building, room 16210 by using the Server-Client environment and smart board, personal computer, and internet was connected by a TCP/IP way. For the courseware, hypertext was supported to be web-based, and photo, picture, data and related web links including text were developed in a close relation, it is possible for students to study big amounts of information in a systemized way and to maximize the learning efficiency. The whole range of introductory nuclear engineering course was divided into nuclear fuel cycle, reactor theory, heat transport, and reactor control, and digital contents were developed by each experts, but the final format of the courseware was maintained consistently for easy understanding . Also, the reactor experiment courseware developed by Kyunghee University can be utilized on this platform. 5 refs., 36 figs., 4 tabs. (Author)

  15. Structured system engineering methodologies used to develop a nuclear thermal propulsion engine

    Science.gov (United States)

    Corban, R.; Wagner, R.

    1993-01-01

    To facilitate the development of a space nuclear thermal propulsion engine for manned flights to Mars, requirements must be established early in the technology development cycle. The long lead times for the acquisition of the engine system and nuclear test facilities demands that the engine system size, performance and safety goals be defined at the earliest possible time. These systems are highly complex and require a large multidisciplinary systems engineering team to develop and track requirements, and to ensure that the as-built system reflects the intent of the mission. A methodology has been devised which uses sophisticated computer tools to effectively develop and interpret functional requirements, and furnish these to the specification level for implementation.

  16. Nuclear magnetic resonance in environmental engineering: principles and applications.

    Science.gov (United States)

    Lens, P N; Hemminga, M A

    1998-01-01

    This paper gives an introduction to nuclear magnetic resonance spectroscopy (NMR) and magnetic resonance imaging (MRI) in relation to applications in the field of environmental science and engineering. The underlying principles of high resolution solution and solid state NMR, relaxation time measurements and imaging are presented. Then, the use of NMR is illustrated and reviewed in studies of biodegradation and biotransformation of soluble and solid organic matter, removal of nutrients and xenobiotics, fate of heavy metal ions, and transport processes in bioreactor systems.

  17. Implementation of a tree algorithm in MCNP code for nuclear well logging applications.

    Science.gov (United States)

    Li, Fusheng; Han, Xiaogang

    2012-07-01

    The goal of this paper is to develop some modeling capabilities that are missing in the current MCNP code. Those missing capabilities can greatly help for some certain nuclear tools designs, such as a nuclear lithology/mineralogy spectroscopy tool. The new capabilities to be developed in this paper include the following: zone tally, neutron interaction tally, gamma rays index tally and enhanced pulse-height tally. The patched MCNP code also can be used to compute neutron slowing-down length and thermal neutron diffusion length.

  18. Study on the code system for the off-site consequences assessment of severe nuclear accident

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sora; Mn, Byung Il; Park, Ki Hyun; Yang, Byung Mo; Suh, Kyung Suk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    The importance of severe nuclear accidents and probabilistic safety assessment (PSA) were brought to international attention with the occurrence of severe nuclear accidents caused by the extreme natural disaster at Fukushima Daiichi nuclear power plant in Japan. In Korea, studies on level 3 PSA had made little progress until recently. The code systems of level 3 PSA, MACCS2 (MELCORE Accident Consequence Code System 2, US), COSYMA (COde SYstem from MAria, EU) and OSCAAR (Off-Site Consequence Analysis code for Atmospheric Releases in reactor accidents, JAPAN), were reviewed in this study, and the disadvantages and limitations of MACCS2 were also analyzed. Experts from Korea and abroad pointed out that the limitations of MACCS2 include the following: MACCS2 cannot simulate multi-unit accidents/release from spent fuel pools, and its atmospheric dispersion is based on a simple Gaussian plume model. Some of these limitations have been improved in the updated versions of MACCS2. The absence of a marine and aquatic dispersion model and the limited simulating range of food-chain and economic models are also important aspects that need to be improved. This paper is expected to be utilized as basic research material for developing a Korean code system for assessing off-site consequences of severe nuclear accidents.

  19. POPCYCLE: a computer code for calculating nuclear and fossil plant levelized life-cycle power costs

    Energy Technology Data Exchange (ETDEWEB)

    Hardie, R.W.

    1982-02-01

    POPCYCLE, a computer code designed to calculate levelized life-cycle power costs for nuclear and fossil electrical generating plants is described. Included are (1) derivations of the equations and a discussion of the methodology used by POPCYCLE, (2) a description of the input required by the code, (3) a listing of the input for a sample case, and (4) the output for a sample case.

  20. Human factors engineering plan for reviewing nuclear plant modernization programs

    Energy Technology Data Exchange (ETDEWEB)

    O' Hara, John; Higgins, James [Brookhaven National Laboratory, Upton, NY (United States)

    2004-12-01

    The Swedish Nuclear Power Inspectorate reviews the human factors engineering (HFE) aspects of nuclear power plants (NPPs) involved in the modernization of the plant systems and control rooms. The purpose of a HFE review is to help ensure personnel and public safety by verifying that accepted HFE practices and guidelines are incorporated into the program and nuclear power plant design. Such a review helps to ensure the HFE aspects of an NPP are developed, designed, and evaluated on the basis of a structured top-down system analysis using accepted HFE principles. The review addresses eleven HFE elements: HFE Program Management, Operating Experience Review, Functional Requirements Analysis and Allocation, Task Analysis, Staffing, Human Reliability Analysis, Human-System Interface Design, Procedure Development, Training Program Development, Human Factors Verification and Validation, and Design Implementation.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-02-01

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

  2. Computer code system for the R and D of nuclear fuel cycle with fast reactor. 5. Development and application of reactor analysis code system

    Energy Technology Data Exchange (ETDEWEB)

    Yokoyama, Kenji; Hazama, Taira; Chiba, Go; Ohki, Shigeo; Ishikawa, Makoto [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    2002-12-01

    In the core design of fast reactors (FRs), it is very important to improve the prediction accuracy of the nuclear characteristics for both reducing cost and ensuring reliability of FR plants. A nuclear reactor analysis code system for FRs has been developed by the Japan Nuclear Cycle Development Institute (JNC). This paper describes the outline of the calculation models and methods in the system consisting of several analysis codes, such as the cell calculation code CASUP, the core calculation code TRITAC and the sensitivity analysis code SAGEP. Some examples of verification results and improvement of the design accuracy are also introduced based on the measurement data from critical assemblies, e.g, the JUPITER experiment (USA/Japan), FCA (Japan), MASURCA (France), and BFS (Russia). Furthermore, application fields and future plans, such as the development of new generation nuclear constants and applications to MA{center_dot}FP transmutation, are described. (author)

  3. Engineering thinking in emergency situations: A new nuclear safety concept.

    Science.gov (United States)

    Guarnieri, Franck; Travadel, Sébastien

    2014-11-01

    The lessons learned from the Fukushima Daiichi accident have focused on preventive measures designed to protect nuclear reactors, and crisis management plans. Although there is still no end in sight to the accident that occurred on March 11, 2011, how engineers have handled the aftermath offers new insight into the capacity of organizations to adapt in situations that far exceed the scope of safety standards based on probabilistic risk assessment and on the comprehensive identification of disaster scenarios. Ongoing crises in which conventional resources are lacking, but societal expectations are high, call for "engineering thinking in emergency situations." This is a new concept that emphasizes adaptability and resilience within organizations-such as the ability to create temporary new organizational structures; to quickly switch from a normal state to an innovative mode; and to integrate a social dimension into engineering activities. In the future, nuclear safety oversight authorities should assess the ability of plant operators to create and implement effective engineering strategies on the fly, and should require that operators demonstrate the capability for resilience in the aftermath of an accident.

  4. Application of CFD Codes in Nuclear Reactor Safety Analysis

    Directory of Open Access Journals (Sweden)

    T. Höhne

    2010-01-01

    Full Text Available Computational Fluid Dynamics (CFD is increasingly being used in nuclear reactor safety (NRS analyses as a tool that enables safety relevant phenomena occurring in the reactor coolant system to be described in more detail. Numerical investigations on single phase coolant mixing in Pressurised Water Reactors (PWR have been performed at the FZD for almost a decade. The work is aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity. For the experimental investigation of horizontal two phase flows, different non pressurized channels and the TOPFLOW Hot Leg model in a pressure chamber was build and simulated with ANSYS CFX. In a common project between the University of Applied Sciences Zittau/Görlitz and FZD the behaviour of insulation material released by a LOCA released into the containment and might compromise the long term emergency cooling systems is investigated. Moreover, the actual capability of CFD is shown to contribute to fuel rod bundle design with a good CHF performance.

  5. Nuclear Engine System Simulation (NESS). Volume 1: Program user's guide

    Science.gov (United States)

    Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman J.

    1993-03-01

    A Nuclear Thermal Propulsion (NTP) engine system design analysis tool is required to support current and future Space Exploration Initiative (SEI) propulsion and vehicle design studies. Currently available NTP engine design models are those developed during the NERVA program in the 1960's and early 1970's and are highly unique to that design or are modifications of current liquid propulsion system design models. To date, NTP engine-based liquid design models lack integrated design of key NTP engine design features in the areas of reactor, shielding, multi-propellant capability, and multi-redundant pump feed fuel systems. Additionally, since the SEI effort is in the initial development stage, a robust, verified NTP analysis design tool could be of great use to the community. This effort developed an NTP engine system design analysis program (tool), known as the Nuclear Engine System Simulation (NESS) program, to support ongoing and future engine system and stage design study efforts. In this effort, Science Applications International Corporation's (SAIC) NTP version of the Expanded Liquid Engine Simulation (ELES) program was modified extensively to include Westinghouse Electric Corporation's near-term solid-core reactor design model. The ELES program has extensive capability to conduct preliminary system design analysis of liquid rocket systems and vehicles. The program is modular in nature and is versatile in terms of modeling state-of-the-art component and system options as discussed. The Westinghouse reactor design model, which was integrated in the NESS program, is based on the near-term solid-core ENABLER NTP reactor design concept. This program is now capable of accurately modeling (characterizing) a complete near-term solid-core NTP engine system in great detail, for a number of design options, in an efficient manner. The following discussion summarizes the overall analysis methodology, key assumptions, and capabilities associated with the NESS presents an

  6. Nuclear Engine System Simulation (NESS). Volume 1: Program user's guide

    Science.gov (United States)

    Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman J.

    1993-01-01

    A Nuclear Thermal Propulsion (NTP) engine system design analysis tool is required to support current and future Space Exploration Initiative (SEI) propulsion and vehicle design studies. Currently available NTP engine design models are those developed during the NERVA program in the 1960's and early 1970's and are highly unique to that design or are modifications of current liquid propulsion system design models. To date, NTP engine-based liquid design models lack integrated design of key NTP engine design features in the areas of reactor, shielding, multi-propellant capability, and multi-redundant pump feed fuel systems. Additionally, since the SEI effort is in the initial development stage, a robust, verified NTP analysis design tool could be of great use to the community. This effort developed an NTP engine system design analysis program (tool), known as the Nuclear Engine System Simulation (NESS) program, to support ongoing and future engine system and stage design study efforts. In this effort, Science Applications International Corporation's (SAIC) NTP version of the Expanded Liquid Engine Simulation (ELES) program was modified extensively to include Westinghouse Electric Corporation's near-term solid-core reactor design model. The ELES program has extensive capability to conduct preliminary system design analysis of liquid rocket systems and vehicles. The program is modular in nature and is versatile in terms of modeling state-of-the-art component and system options as discussed. The Westinghouse reactor design model, which was integrated in the NESS program, is based on the near-term solid-core ENABLER NTP reactor design concept. This program is now capable of accurately modeling (characterizing) a complete near-term solid-core NTP engine system in great detail, for a number of design options, in an efficient manner. The following discussion summarizes the overall analysis methodology, key assumptions, and capabilities associated with the NESS presents an

  7. Conceptual Engine System Design for NERVA derived 66.7KN and 111.2KN Thrust Nuclear Thermal Rockets

    Science.gov (United States)

    Fittje, James E.; Buehrle, Robert J.

    2006-01-01

    The Nuclear Thermal Rocket concept is being evaluated as an advanced propulsion concept for missions to the moon and Mars. A tremendous effort was undertaken during the 1960's and 1970's to develop and test NERVA derived Nuclear Thermal Rockets in the 111.2 KN to 1112 KN pound thrust class. NASA GRC is leveraging this past NTR investment in their vehicle concepts and mission analysis studies, and has been evaluating NERVA derived engines in the 66.7 KN to the 111.2 KN thrust range. The liquid hydrogen propellant feed system, including the turbopumps, is an essential component of the overall operation of this system. The NASA GRC team is evaluating numerous propellant feed system designs with both single and twin turbopumps. The Nuclear Engine System Simulation code is being exercised to analyze thermodynamic cycle points for these selected concepts. This paper will present propellant feed system concepts and the corresponding thermodynamic cycle points for 66.7 KN and 111.2 KN thrust NTR engine systems. A pump out condition for a twin turbopump concept will also be evaluated, and the NESS code will be assessed against the Small Nuclear Rocket Engine preliminary thermodynamic data.

  8. Sandia Engineering Analysis Code Access System v. 2.0.1

    Energy Technology Data Exchange (ETDEWEB)

    2017-10-30

    The Sandia Engineering Analysis Code Access System (SEACAS) is a suite of preprocessing, post processing, translation, visualization, and utility applications supporting finite element analysis software using the Exodus database file format.

  9. Lattice Boltzmann method fundamentals and engineering applications with computer codes

    CERN Document Server

    Mohamad, A A

    2014-01-01

    Introducing the Lattice Boltzmann Method in a readable manner, this book provides detailed examples with complete computer codes. It avoids the most complicated mathematics and physics without scarifying the basic fundamentals of the method.

  10. Hyperthermal Environments Simulator for Nuclear Rocket Engine Development

    Science.gov (United States)

    Litchford, R. J.; Foote, J. P.; Clifton, W. B.; Hickman, R. R.; Wang, T.-S.; Dobson, C. C.

    An arc-heater driven hyperthermal convective environments simulator was recently developed and commissioned for long duration hot hydrogen exposure of nuclear thermal rocket materials. This newly established non-nuclear testing capability uses a high-power, multi-gas, wall-stabilised constricted arc-heater to produce high-temperature pressurised hydrogen flows representative of nuclear reactor core environments, excepting radiation effects, and is intended to serve as a low-cost facility for supporting non-nuclear developmental testing of high-temperature fissile fuels and structural materials. The resulting reactor environments simulator represents a valuable addition to the available inventory of non-nuclear test facilities and is uniquely capable of investigating and characterising candidate fuel/structural materials, improving associated processing/ fabrication techniques, and simulating reactor thermal hydraulics. This paper summarizes facility design and engineering development efforts and reports baseline operational characteristics as determined from a series of performance mapping and long duration capability demonstration tests. Potential follow-on developmental strategies are also suggested in view of the technical and policy challenges ahead.

  11. Do we need a universal 'code of ethics' in nuclear medicine?

    Science.gov (United States)

    Ramesh, Chandakacharla N; Vinjamuri, Sobhan

    2010-06-01

    Recent years have seen huge advances in medicine and the science of medicine. Nuclear medicine has been no exception and there has been rapid acceptance of new concepts, new technologies and newer ways of working. Ethical principles have been traditionally considered as generic skills applicable to wide groups of scientists and doctors, with only token refinement at specialty level. Specialist bodies across the world representing wide groups of practitioners frequently have subgroups dealing exclusively with ethical issues. It could easily be argued that the basic principles of ethical practice adopted by specialist bodies closest to nuclear medicine practice, such as radiology and oncology, will also be applicable to nuclear medicine and that time and effort need not be spent on specifying a separate code for nuclear medicine. It could also be argued that nuclear medicine is an independent specialty and some (if not most) practitioners will not be aware of the guidelines adopted by other specialist societies, and that there is a need for re-iteration of ethical principles at the specialty level and on a worldwide scale.In this article we would like to present a brief history of medical ethics, discuss some of the advances in nuclear medicine and their associated ethical aspects, as well as list a framework of principles for consideration, should a specialist body deem it suitable to establish a 'code of ethics' for nuclear medicine.

  12. An Object-Oriented Computer Code for Aircraft Engine Weight Estimation

    Science.gov (United States)

    Tong, Michael T.; Naylor, Bret A.

    2009-01-01

    Reliable engine-weight estimation at the conceptual design stage is critical to the development of new aircraft engines. It helps to identify the best engine concept amongst several candidates. At NASA Glenn Research Center (GRC), the Weight Analysis of Turbine Engines (WATE) computer code, originally developed by Boeing Aircraft, has been used to estimate the engine weight of various conceptual engine designs. The code, written in FORTRAN, was originally developed for NASA in 1979. Since then, substantial improvements have been made to the code to improve the weight calculations for most of the engine components. Most recently, to improve the maintainability and extensibility of WATE, the FORTRAN code has been converted into an object-oriented version. The conversion was done within the NASA's NPSS (Numerical Propulsion System Simulation) framework. This enables WATE to interact seamlessly with the thermodynamic cycle model which provides component flow data such as airflows, temperatures, and pressures, etc., that are required for sizing the components and weight calculations. The tighter integration between the NPSS and WATE would greatly enhance system-level analysis and optimization capabilities. It also would facilitate the enhancement of the WATE code for next-generation aircraft and space propulsion systems. In this paper, the architecture of the object-oriented WATE code (or WATE++) is described. Both the FORTRAN and object-oriented versions of the code are employed to compute the dimensions and weight of a 300-passenger aircraft engine (GE90 class). Both versions of the code produce essentially identical results as should be the case.

  13. Use of numerical simulation computer codes to fire problems in nuclear power plants in Finland

    Energy Technology Data Exchange (ETDEWEB)

    Keski-Rahkonen, O.; Eloranta, E. (Valtion Teknillinen Tutkimuskeskus, Espoo (Finland). Fire Technology Lab.); Huhtanen, R. (Valtion Teknillinen Tutkimuskeskus, Helsinki (Finland). Nuclear Engineering Lab.)

    1991-03-01

    Zone and field model codes are used for fire simulations, including nuclear facilities, in Finland. Here two examples are described: (a) calculation of evaporation rate of a pool fire (8 MW) in a compartment using FIRST, and calculation of an oil spill fire (180 MW) in a turbine hall using PHOENICS. (orig.).

  14. Russian Nuclear Rocket Engine Design for Mars Exploration

    Institute of Scientific and Technical Information of China (English)

    Vadim Zakirov; Vladimir Pavshook

    2007-01-01

    This paper is to promote investigation into the nuclear rocket engine (NRE) propulsion option that is considered as a key technology for manned Mars exploration. Russian NRE developed since the 1950 s in the former Soviet Union to a full-scale prototype by the 1990 s is viewed as advantageous and the most suitable starting point concept for manned Mars mission application study. The main features of Russian heterogeneous core NRE design are described and the most valuable experimental performance results are summarized. These results have demonstrated the significant specific impulse performance advantage of the NRE over conventional liquid rocket engine (LRE) propulsion technologies. Based on past experience,the recent developments in the field of high-temperature nuclear fuels, and the latest conceptual studies, the developed NRE concept is suggested to be upgraded to the nuclear power and propulsion system (NPPS),more suitable for future manned Mars missions. Although the NRE still needs development for space application, the problems are solvable with additional effort and funding.

  15. ARC Code TI: E-Standards for Mass Properties Engineering

    Data.gov (United States)

    National Aeronautics and Space Administration — The purpose of this Opensource forum is to promote the development of a JAVA based Application Programming Interface for the field of Mass Properties Engineering.

  16. Development of design automation codes using software engineering methods

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R.J. II

    1976-10-31

    The Electrical Engineering Department of the Lawrence Livermore Laboratory (LLL) has recently formed a Design Automation (DA) Group responsible for development of new DA capabilities at the Laboratory. This paper briefly discusses the environment in which the software is being produced, and methodologies employed by the development team. The discussion of software engineering approaches should be of interest to small groups producing relatively large complex software systems. (auth)

  17. Successful Recovery of Nuclear Protein-Coding Genes from Small Insects in Museums Using Illumina Sequencing.

    Science.gov (United States)

    Kanda, Kojun; Pflug, James M; Sproul, John S; Dasenko, Mark A; Maddison, David R

    2015-01-01

    In this paper we explore high-throughput Illumina sequencing of nuclear protein-coding, ribosomal, and mitochondrial genes in small, dried insects stored in natural history collections. We sequenced one tenebrionid beetle and 12 carabid beetles ranging in size from 3.7 to 9.7 mm in length that have been stored in various museums for 4 to 84 years. Although we chose a number of old, small specimens for which we expected low sequence recovery, we successfully recovered at least some low-copy nuclear protein-coding genes from all specimens. For example, in one 56-year-old beetle, 4.4 mm in length, our de novo assembly recovered about 63% of approximately 41,900 nucleotides in a target suite of 67 nuclear protein-coding gene fragments, and 70% using a reference-based assembly. Even in the least successfully sequenced carabid specimen, reference-based assembly yielded fragments that were at least 50% of the target length for 34 of 67 nuclear protein-coding gene fragments. Exploration of alternative references for reference-based assembly revealed few signs of bias created by the reference. For all specimens we recovered almost complete copies of ribosomal and mitochondrial genes. We verified the general accuracy of the sequences through comparisons with sequences obtained from PCR and Sanger sequencing, including of conspecific, fresh specimens, and through phylogenetic analysis that tested the placement of sequences in predicted regions. A few possible inaccuracies in the sequences were detected, but these rarely affected the phylogenetic placement of the samples. Although our sample sizes are low, an exploratory regression study suggests that the dominant factor in predicting success at recovering nuclear protein-coding genes is a high number of Illumina reads, with success at PCR of COI and killing by immersion in ethanol being secondary factors; in analyses of only high-read samples, the primary significant explanatory variable was body length, with small beetles

  18. News from the Library: A new key reference work for the engineer: ASME's Boiler and Pressure Vessel Code at the CERN Library

    CERN Multimedia

    CERN Library

    2011-01-01

    The Library is aiming at offering a range of constantly updated reference books, to cover all areas of CERN activity. A recent addition to our collections strengthens our offer in the Engineering field.   The CERN Library now holds a copy of the complete ASME Boiler and Pressure Vessel Code, 2010 edition. This code establishes rules of safety governing the design, fabrication, and inspection of boilers and pressure vessels, and nuclear power plant components during construction. This document is considered worldwide as a reference for mechanical design and is therefore important for the CERN community. The Code published by ASME (American Society of Mechanical Engineers) is kept current by the Boiler and Pressure Committee, a volunteer group of more than 950 engineers worldwide. The Committee meets regularly to consider requests for interpretations, revision, and to develop new rules. The CERN Library receives updates and includes them in the volumes until the next edition, which is expected to ...

  19. Engineering autonomous error correction in stabilizer codes at finite temperature

    Science.gov (United States)

    Freeman, C. Daniel; Herdman, C. M.; Whaley, K. B.

    2017-07-01

    We present an error-correcting protocol that enhances the lifetime of stabilizer code-based qubits which are susceptible to the creation of pairs of localized defects (due to stringlike error operators) at finite temperature, such as the toric code. The primary tool employed is periodic application of a local, unitary operator, which exchanges defects and thereby translates localized excitations. Crucially, the protocol does not require any measurements of stabilizer operators and therefore can be used to enhance the lifetime of a qubit in the absence of such experimental resources.

  20. The Development of a Systematic Coding System for Elementary Students' Drawings of Engineers

    Science.gov (United States)

    Weber, Nicole; Duncan, Daphne; Dyehouse, Melissa; Strobel, Johannes; Diefes-Dux, Heidi A.

    2011-01-01

    The Draw an Engineer Test (DAET) is a common measure of students' perceptions of engineers. The coding systems currently used for K-12 research are general rubrics or checklists to capture the images presented in the drawing, which leave out some of the richness of students' perceptions, currently only captured with an accompanying student…

  1. Calculation code evaluating the confinement of a nuclear facility in case of fires

    Energy Technology Data Exchange (ETDEWEB)

    Laborde, J.C.; Prevost, C.; Vendel, J. [and others

    1995-02-01

    Accident events involving fire are quite frequent and could have a severe effect on the safety of nuclear facilities. As confinement must be maintained, the ventilation and filtration systems have to be designed to limit radioactive release to the environment. To determine and analyse the consequences of a fire on the contamination confinement, IPSN, COGEMA and SGN are participating in development of a calculation code based on introduction, in the SIMEVENT ventilation code, of various models associated to fire risk and mass transfer in the ventilation networks. This calculation code results from the coupling of the SIMEVENT code with several models describing the temperature in a room resulting of a fire, the temperatures along the ventilation ducts, the contamination transfers through out the ventilation equipments (ducts, dampers, valves, air cleaning systems) and the High Efficiency Particulate Air (HEPA) filters clogging. The paper proposed presents the current level of progress in development of this calculation code. It describes, in particular, the empirical model used for the clogging of HEPA filters by the aerosols derived from the combustion of standard materials used in the nuclear industry. It describes, also, the specific models used to take into account the mass transfers and resulting from the basic mechanisms of aerosols physics. In addition, an assessment of this code is given using the example of a simple laboratory installation.

  2. Progress of teaching and learning of nuclear engineering courses at College of Engineering, Universiti Tenaga Nasional (UNITEN)

    Energy Technology Data Exchange (ETDEWEB)

    Hamid, Nasri A., E-mail: Nasri@uniten.edu.my; Mohamed, Abdul Aziz; Yusoff, Mohd. Zamri [Nuclear Energy Center, College of Engineering, Universiti Tenaga Nasional, Jalan IKRAM-UNITEN, 43000 Kajang, Selangor (Malaysia)

    2015-04-29

    Developing human capital in nuclear with required nuclear background and professional qualifications is necessary to support the implementation of nuclear power projects in the near future. Sufficient educational and training skills are required to ensure that the human resources needed by the nuclear power industry meets its high standard. The Government of Malaysia has made the decision to include nuclear as one of the electricity generation option for the country, post 2020 in order to cater for the increasing energy demands of the country as well as to reduce CO{sub 2} emission. The commitment by the government has been made clearer with the inclusion of the development of first NPP by 2021 in the Economic Transformation Program (ETP) which was launched by the government in October 2010. The In tandem with the government initiative to promote nuclear energy, Center for Nuclear Energy, College of Engineering, Universiti Tenaga Nasional (UNITEN) is taking the responsibility in developing human capital in the area of nuclear power and technology. In the beginning, the College of Engineering has offered the Introduction to Nuclear Technology course as a technical elective course for all undergraduate engineering students. Gradually, other nuclear technical elective courses are offered such as Nuclear Policy, Security and Safeguards, Introduction to Nuclear Engineering, Radiation Detection and Nuclear Instrumentation, Introduction to Reactor Physics, Radiation Safety and Waste Management, and Nuclear Thermal-hydraulics. In addition, another course Advancement in Nuclear Energy is offered as one of the postgraduate elective courses. To enhance the capability of teaching staffs in nuclear areas at UNITEN, several junior lecturers are sent to pursue their postgraduate studies in the Republic of Korea, United States and the United Kingdom, while the others are participating in short courses and workshops in nuclear that are conducted locally and abroad. This paper

  3. Progress of teaching and learning of nuclear engineering courses at College of Engineering, Universiti Tenaga Nasional (UNITEN)

    Science.gov (United States)

    Hamid, Nasri A.; Mohamed, Abdul Aziz; Yusoff, Mohd. Zamri

    2015-04-01

    Developing human capital in nuclear with required nuclear background and professional qualifications is necessary to support the implementation of nuclear power projects in the near future. Sufficient educational and training skills are required to ensure that the human resources needed by the nuclear power industry meets its high standard. The Government of Malaysia has made the decision to include nuclear as one of the electricity generation option for the country, post 2020 in order to cater for the increasing energy demands of the country as well as to reduce CO2 emission. The commitment by the government has been made clearer with the inclusion of the development of first NPP by 2021 in the Economic Transformation Program (ETP) which was launched by the government in October 2010. The In tandem with the government initiative to promote nuclear energy, Center for Nuclear Energy, College of Engineering, Universiti Tenaga Nasional (UNITEN) is taking the responsibility in developing human capital in the area of nuclear power and technology. In the beginning, the College of Engineering has offered the Introduction to Nuclear Technology course as a technical elective course for all undergraduate engineering students. Gradually, other nuclear technical elective courses are offered such as Nuclear Policy, Security and Safeguards, Introduction to Nuclear Engineering, Radiation Detection and Nuclear Instrumentation, Introduction to Reactor Physics, Radiation Safety and Waste Management, and Nuclear Thermal-hydraulics. In addition, another course Advancement in Nuclear Energy is offered as one of the postgraduate elective courses. To enhance the capability of teaching staffs in nuclear areas at UNITEN, several junior lecturers are sent to pursue their postgraduate studies in the Republic of Korea, United States and the United Kingdom, while the others are participating in short courses and workshops in nuclear that are conducted locally and abroad. This paper describes

  4. SOC-DS computer code provides tool for design evaluation of homogeneous two-material nuclear shield

    Science.gov (United States)

    Disney, R. K.; Ricks, L. O.

    1967-01-01

    SOC-DS Code /Shield Optimization Code-Direc Search/, selects a nuclear shield material of optimum volume, weight, or cost to meet the requirments of a given radiation dose rate or energy transmission constraint. It is applicable to evaluating neutron and gamma ray shields for all nuclear reactors.

  5. Engineering for new-built nuclear power plant projects; Ingenieria para proyectos de nuevas centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez Lopez, A.

    2012-11-01

    This article reviews the opportunities existing in the market (electrical utilities and reactor vendors) for an engineering company with the profile of Empresarios Agrupados (EA) in new-built nuclear power plant projects. To do this, reference is made to some representative examples of projects in which EA has been participating recently. the article concludes sharing with the reader some lessons learned from this participation. (Author)

  6. Nuclear data libraries for Tripoli-3.5 code; Bibliotheques de donnees nucleaires pour le code tripoli-3.5

    Energy Technology Data Exchange (ETDEWEB)

    Vergnaud, Th

    2001-07-01

    The TRIPOLI-3 code uses multigroup nuclear data libraries generated using the NJOY-THEMIS suite of modules: for neutrons, they are produced from the ENDF/B-VI evaluations and cover the range between 20 MeV and 10{sup -5} eV, either in 315 groups and for one temperature, or in 3209 groups and for five temperatures; for gamma-rays, they are from JEF2 and are processed in groups between 14 MeV and keV. The probability tables used for the neutron transport calculations have been derived from the ENDF/B-VI evaluations using the CALENDF code. Cross sections for gamma production by neutron interaction (fission, capture or inelastic scattering) have been derived from ENDF/B-VI in 315 neutron groups and 75 gamma groups. The code also uses two response function libraries: for neutrons; based on several sources, in particular the dosimetry libraries IRDF/85 and IRDF/90; for gamma-rays it is based on the JEF2 evaluation and contains the kerma factors for all the elements and cross sections for all interactions. (author)

  7. Ground test facilities for evaluating nuclear thermal propulsion engines and fuel elements

    Science.gov (United States)

    Allen, G. C.; Beck, D. F.; Harmon, C. D.; Shipers, L. R.

    Interagency panels evaluating nuclear thermal propulsion development options have consistently recognized the need for constructing a major new ground test facility to support fuel element and engine testing. This paper summarizes the requirements, configuration, and design issues of a proposed ground test complex for evaluating nuclear thermal propulsion engines and fuel elements being developed for the Space Nuclear Thermal Propulsion (SNTP) program.

  8. Implanting a Discipline: The Academic Trajectory of Nuclear Engineering in the USA and UK

    Science.gov (United States)

    Johnston, Sean F.

    2009-01-01

    The nuclear engineer emerged as a new form of recognised technical professional between 1940 and the early 1960s as nuclear fission, the chain reaction and their applications were explored. The institutionalization of nuclear engineering--channelled into new national laboratories and corporate design offices during the decade after the war, and…

  9. 10 CFR Appendix S to Part 50 - Earthquake Engineering Criteria for Nuclear Power Plants

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Earthquake Engineering Criteria for Nuclear Power Plants S... FACILITIES Pt. 50, App. S Appendix S to Part 50—Earthquake Engineering Criteria for Nuclear Power Plants..., as specified in § 50.54(ff), nuclear power plants that have implemented the earthquake...

  10. Application of software engineering to development of reactor-safety codes

    Energy Technology Data Exchange (ETDEWEB)

    Wilburn, N P; Niccoli, L G

    1980-11-01

    As a result of the drastically increasing cost of software and the lack of an engineering approach, the technology of Software Engineering is being developed. Software Engineering provides an answer to the increasing cost of developing and maintaining software. It has been applied extensively in the business and aerospace communities and is just now being applied to the development of scientific software and, in particular, to the development of reactor safety codes at HEDL.

  11. Compilation of documented computer codes applicable to environmental assessment of radioactivity releases. [Nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, F. O.; Miller, C. W.; Shaeffer, D. L.; Garten, Jr., C. T.; Shor, R. W.; Ensminger, J. T.

    1977-04-01

    The objective of this paper is to present a compilation of computer codes for the assessment of accidental or routine releases of radioactivity to the environment from nuclear power facilities. The capabilities of 83 computer codes in the areas of environmental transport and radiation dosimetry are summarized in tabular form. This preliminary analysis clearly indicates that the initial efforts in assessment methodology development have concentrated on atmospheric dispersion, external dosimetry, and internal dosimetry via inhalation. The incorporation of terrestrial and aquatic food chain pathways has been a more recent development and reflects the current requirements of environmental legislation and the needs of regulatory agencies. The characteristics of the conceptual models employed by these codes are reviewed. The appendixes include abstracts of the codes and indexes by author, key words, publication description, and title.

  12. User input verification and test driven development in the NJOY21 nuclear data processing code

    Energy Technology Data Exchange (ETDEWEB)

    Trainer, Amelia Jo [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Conlin, Jeremy Lloyd [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); McCartney, Austin Paul [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-08-21

    Before physically-meaningful data can be used in nuclear simulation codes, the data must be interpreted and manipulated by a nuclear data processing code so as to extract the relevant quantities (e.g. cross sections and angular distributions). Perhaps the most popular and widely-trusted of these processing codes is NJOY, which has been developed and improved over the course of 10 major releases since its creation at Los Alamos National Laboratory in the mid-1970’s. The current phase of NJOY development is the creation of NJOY21, which will be a vast improvement from its predecessor, NJOY2016. Designed to be fast, intuitive, accessible, and capable of handling both established and modern formats of nuclear data, NJOY21 will address many issues that many NJOY users face, while remaining functional for those who prefer the existing format. Although early in its development, NJOY21 is quickly providing input validation to check user input. By providing rapid and helpful responses to users while writing input files, NJOY21 will prove to be more intuitive and easy to use than any of its predecessors. Furthermore, during its development, NJOY21 is subject to regular testing, such that its test coverage must strictly increase with the addition of any production code. This thorough testing will allow developers and NJOY users to establish confidence in NJOY21 as it gains functionality. This document serves as a discussion regarding the current state input checking and testing practices of NJOY21.

  13. Data Mining for Secure Software Engineering – Source Code Management Tool Case Study

    Directory of Open Access Journals (Sweden)

    A.V.Krishna Prasad,

    2010-07-01

    Full Text Available As Data Mining for Secure Software Engineering improves software productivity and quality, software engineers are increasingly applying data mining algorithms to various software engineering tasks. However mining software engineering data poses several challenges, requiring various algorithms to effectively mine sequences, graphs and text from such data. Software engineering data includes code bases, execution traces, historical code changes,mailing lists and bug data bases. They contains a wealth of information about a projects-status, progress and evolution. Using well established data mining techniques, practitioners and researchers can explore the potential of this valuable data in order to better manage their projects and do produce higher-quality software systems that are delivered on time and with in budget. Data mining can be used in gathering and extracting latent security requirements, extracting algorithms and business rules from code, mining legacy applications for requirements and business rules for new projects etc. Mining algorithms for software engineering falls into four main categories: Frequent pattern mining – finding commonly occurring patterns; Pattern matching – finding data instances for given patterns; Clustering – grouping data into clusters and Classification – predicting labels of data based on already labeled data. In this paper, we will discuss the overview of strategies for data mining for secure software engineering, with the implementation of a case study of text mining for source code management tool.

  14. Structured automated code checking through structural components and systems engineering

    NARCIS (Netherlands)

    Coenders, J.L.; Rolvink, A.

    2014-01-01

    This paper presents a proposal to employ the design computing methodology proposed as StructuralComponents (Rolvink et al [6] and van de Weerd et al [7]) as a method to perform a digital verification process to fulfil the requirements related to structural design and engineering as part of a buildin

  15. Virus-host co-evolution under a modified nuclear genetic code

    Directory of Open Access Journals (Sweden)

    Derek J. Taylor

    2013-03-01

    Full Text Available Among eukaryotes with modified nuclear genetic codes, viruses are unknown. However, here we provide evidence of an RNA virus that infects a fungal host (Scheffersomyces segobiensis with a derived nuclear genetic code where CUG codes for serine. The genomic architecture and phylogeny are consistent with infection by a double-stranded RNA virus of the genus Totivirus. We provide evidence of past or present infection with totiviruses in five species of yeasts with modified genetic codes. All but one of the CUG codons in the viral genome have been eliminated, suggesting that avoidance of the modified codon was important to viral adaptation. Our mass spectroscopy analysis indicates that a congener of the host species has co-opted and expresses a capsid gene from totiviruses as a cellular protein. Viral avoidance of the host’s modified codon and host co-option of a protein from totiviruses suggest that RNA viruses co-evolved with yeasts that underwent a major evolutionary transition from the standard genetic code.

  16. Machine Code and Metaphysics: A Perspective on Software Engineering

    OpenAIRE

    2015-01-01

    A major, but too-little-considered problem for Software Engineering (SE) is a lack of consensus concerning Computer Science (CS) and how this relates to developing unpredictable computing technology. We consider some implications for SE of computer systems differing scientific basis, exemplified with the International Standard Organisations Open Systems Interconnection (ISO-OSI) layered architectural model. An architectural view allows comparison of computing technology components facilitatin...

  17. Engineering Codes of Ethics and the Duty to Set a Moral Precedent.

    Science.gov (United States)

    Schlossberger, Eugene

    2016-10-01

    Each of the major engineering societies has its own code of ethics. Seven "common core" clauses and several code-specific clauses can be identified. The paper articulates objections to and rationales for two clauses that raise controversy: do engineers have a duty (a) to provide pro bono services and/or speak out on major issues, and (b) to associate only with reputable individuals and organizations? This latter "association clause" can be justified by the "proclamative principle," an alternative to Kant's universalizability requirement. At the heart of engineering codes of ethics, and implicit in what it is to be a moral agent, the "proclamative principle" asserts that one's life should proclaim one's moral stances (one's values, principles, perceptions, etc.). More specifically, it directs engineers to strive to insure that their actions, thoughts, and relationships be fit to offer to their communities as part of the body of moral precedents for how to be an engineer. Understanding codes of ethics as reflections of this principle casts light both on how to apply the codes and on the distinction between private and professional morality.

  18. The efficiency of the use of penetration nuclear logging in hydrogeology and engineering geology

    Energy Technology Data Exchange (ETDEWEB)

    Ferronsky, V.I. (AN SSSR, Moscow (USSR). Water Problems Inst.); Griaznov, T.A.; Selivanov, L.V. (All-Union Research Inst. for Hydrogeology and Engineering Geology, Moscow (USSR))

    1992-03-01

    The latest developments in equipment and techniques for nuclear and combined non-nuclear logging in friable unconsolidated deposits, including marine bottom sediments are described. The effectiveness of these techniques in hydrogeological and engineering geological investigations is discussed. (Author).

  19. Development of Nuclear Engineering Educational Program at Ibaraki University with Regional Collaboration

    Science.gov (United States)

    Matsumura, Kunihito; Kaminaga, Fumito; Kanto, Yasuhiro; Tanaka, Nobuatsu; Saigusa, Mikio; Kikuchi, Kenji; Kurumada, Akira

    The College of Engineering, Ibaraki University is located at the Hitachi city, in the north part of Ibaraki prefecture. Hitachi and Tokai areas are well known as concentration of advanced technology center of nuclear power research organizations. By considering these regional advantages, we developed a new nuclear engineering educational program for students in the Collage of Engineering and The Graduate School of Science and Engineering of Ibaraki University. The program is consisted of the fundamental lectures of nuclear engineering and nuclear engineering experiments. In addition, several observation learning programs by visiting cooperative organizations are also included in the curriculum. In this paper, we report about the progress of the new educational program for nuclear engineering in Ibaraki University.

  20. Composing Data Parallel Code for a SPARQL Graph Engine

    Energy Technology Data Exchange (ETDEWEB)

    Castellana, Vito G.; Tumeo, Antonino; Villa, Oreste; Haglin, David J.; Feo, John

    2013-09-08

    Big data analytics process large amount of data to extract knowledge from them. Semantic databases are big data applications that adopt the Resource Description Framework (RDF) to structure metadata through a graph-based representation. The graph based representation provides several benefits, such as the possibility to perform in memory processing with large amounts of parallelism. SPARQL is a language used to perform queries on RDF-structured data through graph matching. In this paper we present a tool that automatically translates SPARQL queries to parallel graph crawling and graph matching operations. The tool also supports complex SPARQL constructs, which requires more than basic graph matching for their implementation. The tool generates parallel code annotated with OpenMP pragmas for x86 Shared-memory Multiprocessors (SMPs). With respect to commercial database systems such as Virtuoso, our approach reduces memory occupation due to join operations and provides higher performance. We show the scaling of the automatically generated graph-matching code on a 48-core SMP.

  1. Implementation of a tree algorithm in MCNP code for nuclear well logging applications

    Energy Technology Data Exchange (ETDEWEB)

    Li Fusheng, E-mail: fusheng.li@bakerhughes.com [Baker Hughes Incorporated, 2001 Rankin Rd. Houston, TX 77073-5101 (United States); Han Xiaogang [Baker Hughes Incorporated, 2001 Rankin Rd. Houston, TX 77073-5101 (United States)

    2012-07-15

    The goal of this paper is to develop some modeling capabilities that are missing in the current MCNP code. Those missing capabilities can greatly help for some certain nuclear tools designs, such as a nuclear lithology/mineralogy spectroscopy tool. The new capabilities to be developed in this paper include the following: zone tally, neutron interaction tally, gamma rays index tally and enhanced pulse-height tally. The patched MCNP code also can be used to compute neutron slowing-down length and thermal neutron diffusion length. - Highlights: Black-Right-Pointing-Pointer Tree structure programming is suitable for Monte-Carlo based particle tracking. Black-Right-Pointing-Pointer Enhanced pulse height tally is developed for oilwell logging tool simulation. Black-Right-Pointing-Pointer Neutron interaction tally and gamma ray index tally for geochemical logging.

  2. Assessment of computer codes for VVER-440/213-type nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Szabados, L.; Ezsol, Gy.; Perneczky [Atomic Energy Research Institute, Budapest (Hungary)

    1995-09-01

    Nuclear power plant of VVER-440/213 designed by the former USSR have a number of special features. As a consequence of these features the transient behaviour of such a reactor system should be different from the PWR system behaviour. To study the transient behaviour of the Hungarian Paks Nuclear Power Plant of VVER-440/213-type both analytical and experimental activities have been performed. The experimental basis of the research in the PMK-2 integral-type test facility , which is a scaled down model of the plant. Experiments performed on this facility have been used to assess thermal-hydraulic system codes. Four tests were selected for {open_quotes}Standard Problem Exercises{close_quotes} of the International Atomic Energy Agency. Results of the 4th Exercise, of high international interest, are presented in the paper, focusing on the essential findings of the assessment of computer codes.

  3. Comparisons of hadrontherapy-relevant data to nuclear interaction codes in the Geant4 toolkit

    Science.gov (United States)

    Braunn, B.; Boudard, A.; Colin, J.; Cugnon, J.; Cussol, D.; David, J. C.; Kaitaniemi, P.; Labalme, M.; Leray, S.; Mancusi, D.

    2013-03-01

    Comparisons between experimental data, INCL and other nuclear models available in the Geant4 toolkit are presented. The data used for the comparisons come from a fragmentation experiment realised at GANIL facility. The main purpose of this experiment was to measure production rates and angular distributions of emitted particles from the collision of a 95.A MeV 12C beam and thick PMMA (plastic) targets. The latest version of the Intra Nuclear Cascade of Liege code extended to nucleus-nucleus collisions for ion beam therapy application will be described. This code as well as JQMD and the Geant4 binary cascade has been compared with these hadrontherapy-oriented experimental data. The results from the comparisons exhibit an overall qualitative agreement between the models and the experimental data. However, at a quantitative level, it has been shown that none of this three models manage to reproduce precisely all the data. The nucleus-nucleus extension of INCL, which is not predictive enough for ion beam therapy application yet, has nevertheless proven to be competitive with other nuclear collisions codes.

  4. Validation of a multidimensional deterministic nuclear data sensitivity and uncertainty code system: an application needing supercomputing

    Energy Technology Data Exchange (ETDEWEB)

    Bidaud, A.; Mastrangelo, V. [Conservatoire National des Arts et Metiers, Laboratoire de Physique (CNAM), 75 - Paris (France); Institut de Physique Nucleaire (IN2P3/CNRS) 91 - Orsay (France); Kodeli, I.; Sartori, E. [OECD NEA Data Bank, 92 - Issy les Moulineaux (France)

    2003-07-01

    The quality of nuclear core modelling is linked to the quality of basic nuclear data such as probability of reaction (i.e. cross sections) between neutrons and the nucleus of the core materials. Perturbation Theory, whose applications in nuclear science has been largely developed in the sixties provides tools for estimating the sensitivity of integral parameters such as k-eff, reaction rates, or breeding ratio to the cross sections. The computation with these tools requires approximations in the simulation of space, angles and energy dependent neutron transport. To minimise the impact of the geometry modelling approximations in the calculation, use of 3 dimensional multigroup transport codes is recommended. Sensitivity and uncertainty analyses are the tools needed to estimate the accuracy that a code system with data libraries can achieve. They can guide users as to the specific need for improved data to carry out reliable simulations. However, as full-scale models in 3 dimensions with refined descriptions of the phase-space are used, high performance computers and codes designed to run on parallel architectures are needed to obtain results within acceptable time limits.

  5. Applications of FLUKA Monte Carlo code for nuclear and accelerator physics

    CERN Document Server

    Battistoni, Giuseppe; Brugger, Markus; Campanella, Mauro; Carboni, Massimo; Empl, Anton; Fasso, Alberto; Gadioli, Ettore; Cerutti, Francesco; Ferrari, Alfredo; Ferrari, Anna; Lantz, Matthias; Mairani, Andrea; Margiotta, M; Morone, Christina; Muraro, Silvia; Parodi, Katerina; Patera, Vincenzo; Pelliccioni, Maurizio; Pinsky, Lawrence; Ranft, Johannes; Roesler, Stefan; Rollet, Sofia; Sala, Paola R; Santana, Mario; Sarchiapone, Lucia; Sioli, Maximiliano; Smirnov, George; Sommerer, Florian; Theis, Christian; Trovati, Stefania; Villari, R; Vincke, Heinz; Vincke, Helmut; Vlachoudis, Vasilis; Vollaire, Joachim; Zapp, Neil

    2011-01-01

    FLUKA is a general purpose Monte Carlo code capable of handling all radiation components from thermal energies (for neutrons) or 1keV (for all other particles) to cosmic ray energies and can be applied in many different fields. Presently the code is maintained on Linux. The validity of the physical models implemented in FLUKA has been benchmarked against a variety of experimental data over a wide energy range, from accelerator data to cosmic ray showers in the Earth atmosphere. FLUKA is widely used for studies related both to basic research and to applications in particle accelerators, radiation protection and dosimetry, including the specific issue of radiation damage in space missions, radiobiology (including radiotherapy) and cosmic ray calculations. After a short description of the main features that make FLUKA valuable for these topics, the present paper summarizes some of the recent applications of the FLUKA Monte Carlo code in the nuclear as well high energy physics. In particular it addresses such top...

  6. MCNP and other nuclear codes output graphical representation using python scripts; Representacion grafica de outputs de MCNP y codigos nucleares mediante el uso de scripts en python

    Energy Technology Data Exchange (ETDEWEB)

    Cadenas Mendicoa, A. M.

    2016-08-01

    Due to the lack of graphical representation capability of same nuclear codes like MCNP of GOTHIC, widely used in the industry, the following article describes the development of an interface to use a graphical representation open source (Paraview) with the outputs generated by the nuclear codes. Moreover, this article aims at describing the advantage of this type of visualization programs for the modeling and decision making in the calculation. (Author)

  7. Three-dimensional all-speed CFD code for safety analysis of nuclear reactor containment: Status of GASFLOW parallelization, model development, validation and application

    Energy Technology Data Exchange (ETDEWEB)

    Xiao, Jianjun, E-mail: jianjun.xiao@kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Travis, John R., E-mail: jack_travis@comcast.com [Engineering and Scientific Software Inc., 3010 Old Pecos Trail, Santa Fe, NM 87505 (United States); Royl, Peter, E-mail: peter.royl@partner.kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Necker, Gottfried, E-mail: gottfried.necker@partner.kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Svishchev, Anatoly, E-mail: anatoly.svishchev@kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Jordan, Thomas, E-mail: thomas.jordan@kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2016-05-15

    Highlights: • 3-D scalable semi-implicit pressure-based CFD code for containment safety analysis. • Robust solution algorithm valid for all-speed flows. • Well validated and widely used CFD code for hydrogen safety analysis. • Code applied in various types of nuclear reactor containments. • Parallelization enables high-fidelity models in large scale containment simulations. - Abstract: GASFLOW is a three dimensional semi-implicit all-speed CFD code which can be used to predict fluid dynamics, chemical kinetics, heat and mass transfer, aerosol transportation and other related phenomena involved in postulated accidents in nuclear reactor containments. The main purpose of the paper is to give a brief review on recent GASFLOW code development, validations and applications in the field of nuclear safety. GASFLOW code has been well validated by international experimental benchmarks, and has been widely applied to hydrogen safety analysis in various types of nuclear power plants in European and Asian countries, which have been summarized in this paper. Furthermore, four benchmark tests of a lid-driven cavity flow, low Mach number jet flow, 1-D shock tube and supersonic flow over a forward-facing step are presented in order to demonstrate the accuracy and wide-ranging capability of ICE’d ALE solution algorithm for all-speed flows. GASFLOW has been successfully parallelized using the paradigms of Message Passing Interface (MPI) and domain decomposition. The parallel version, GASFLOW-MPI, adds great value to large scale containment simulations by enabling high-fidelity models, including more geometric details and more complex physics. It will be helpful for the nuclear safety engineers to better understand the hydrogen safety related physical phenomena during the severe accident, to optimize the design of the hydrogen risk mitigation systems and to fulfill the licensing requirements by the nuclear regulatory authorities. GASFLOW-MPI is targeting a high

  8. Vectorization, parallelization and implementation of nuclear codes =MVP/GMVP, QMDRELP, EQMD, HSABC, CURBAL, STREAM V3.1, TOSCA, EDDYCAL, RELAP5/MOD2/C36-05, RELAP5/MOD3= on the VPP500 computer system. Progress report 1995 fiscal year

    Energy Technology Data Exchange (ETDEWEB)

    Nemoto, Toshiyuki; Watanabe, Hideo; Fujita, Toyozo [Fujitsu Ltd., Tokyo (Japan); Kawai, Wataru; Harada, Hiroo; Gorai, Kazuo; Yamasaki, Kazuhiko; Shoji, Makoto; Fujii, Minoru

    1996-06-01

    At Center for Promotion of Computational Science and Engineering, time consuming eight nuclear codes suggested by users have been vectorized, parallelized on the VPP500 computer system. In addition, two nuclear codes used on the VP2600 computer system were implemented on the VPP500 computer system. Neutron and photon transport calculation code MVP/GMVP and relativistic quantum molecular dynamics code QMDRELP have been parallelized. Extended quantum molecular dynamics code EQMD and adiabatic base calculation code HSABC have been parallelized and vectorized. Ballooning turbulence simulation code CURBAL, 3-D non-stationary compressible fluid dynamics code STREAM V3.1, operating plasma analysis code TOSCA and eddy current analysis code EDDYCAL have been vectorized. Reactor safety analysis code RELAP5/MOD2/C36-05 and RELAP5/MOD3 were implemented on the VPP500 computer system. (author)

  9. Nuclear engineering vocabulary; Vocabulaire de l'ingenierie nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Dumont, X. [FRAMATOME, 92 - Paris-La-Defense (France); Andrieux, C. [CEA/Saclay, Direction des Technologies de l' Information (DTI), 91 - Gif-sur-Yvette (France)] [and others

    2000-06-01

    The aim of this book is to bring together the French technical terms and expressions as defined by the specialized commission of terminology and neology of nuclear engineering (CSTNIN). For each term or expression is given: a possible abbreviation, its domain of use, its definition, sometimes a synonymous, eventually some notes, the related terms, the English equivalent, and its status at the date of publication of the book. This status comprises several steps: the eventual publication in the Journal Officiel de la Republique Francaise (official journal of the French republic, with its date), the related reference list and working group, the position of the enable authorities (CSTNIN, COGETERM, French Academy, minister), the last date of review, and some possible additional details. (J.S.)

  10. Producing hydrogen from water for nuclear hydrogen power engineering

    Energy Technology Data Exchange (ETDEWEB)

    Gorbachev, A.K.; Andryushchenko, F.K.; Bochin, V.P.; Ishchenko, L.I.; Nechiporenko, N.N.; Nikiforov, V.K.; Soldat, Ye.F.; Zadorozhnyy, P.S.; Zhuk, G.G.

    1980-01-01

    The prospects of production of H/sub 2/ for nuclear-hydrogen power engineering using the sulphuric acid cycle with depolarization of the anode by sulphur gas are demonstrated. The kinetics of oxidation of sulphur gas on porous activated graphite anodes depending on the concentration of H/sub 2/SO/sub 4/ and the presence of a homogeneous catalyst in the solution is examined. The results for kinetics of separation of H/sub 2/ in 5 M solution H/sub 2/SO/sub 4/ on titanium nitride, alloys E194, EP496, EP567, E159 and an alloy of titanium and miludimum 4201 are cited. Electroysis of aqueous solutions of H/sub 2/SO/sub 4/ with depolarization of the anode process can be carried out on technically accessible materials. The voltage on the electrolyzer is approximately 1 volt when carrying out the process on recommended materials.

  11. Selection of a computer code for Hanford low-level waste engineered-system performance assessment

    Energy Technology Data Exchange (ETDEWEB)

    McGrail, B.P.; Mahoney, L.A.

    1995-10-01

    Planned performance assessments for the proposed disposal of low-level waste (LLW) glass produced from remediation of wastes stored in underground tanks at Hanford, Washington will require calculations of radionuclide release rates from the subsurface disposal facility. These calculations will be done with the aid of computer codes. Currently available computer codes were ranked in terms of the feature sets implemented in the code that match a set of physical, chemical, numerical, and functional capabilities needed to assess release rates from the engineered system. The needed capabilities were identified from an analysis of the important physical and chemical process expected to affect LLW glass corrosion and the mobility of radionuclides. The highest ranked computer code was found to be the ARES-CT code developed at PNL for the US Department of Energy for evaluation of and land disposal sites.

  12. Sustaining engineering codes of ethics for the twenty-first century.

    Science.gov (United States)

    Michelfelder, Diane; Jones, Sharon A

    2013-03-01

    How much responsibility ought a professional engineer to have with regard to supporting basic principles of sustainable development? While within the United States, professional engineering societies, as reflected in their codes of ethics, differ in their responses to this question, none of these professional societies has yet to put the engineer's responsibility toward sustainability on a par with commitments to public safety, health, and welfare. In this paper, we aim to suggest that sustainability should be included in the paramountcy clause because it is a necessary condition to ensure the safety, health, and welfare of the public. Part of our justification rests on the fact that to engineer sustainably means among many things to consider social justice, understood as the fair and equitable distribution of social goods, as a design constraint similar to technical, economic, and environmental constraints. This element of social justice is not explicit in the current paramountcy clause. Our argument rests on demonstrating that social justice in terms of both inter- and intra-generational equity is an important dimension of sustainability (and engineering). We also propose that embracing sustainability in the codes while recognizing the role that social justice plays may elevate the status of the engineer as public intellectual and agent of social good. This shift will then need to be incorporated in how we teach undergraduate engineering students about engineering ethics.

  13. NUSTART: A PC code for NUclear STructure And Radiative Transition analysis and supplementation

    Energy Technology Data Exchange (ETDEWEB)

    Larsen, G.L.; Gardner, D.G.; Gardner, M.A.

    1990-10-01

    NUSTART is a computer program for the IBM PC/At. It is designed for use with the nuclear reaction cross-section code STAPLUS, which is a STAPRE-based CRAY computer code that is being developed at Lawrence Livermore National Laboratory. The NUSTART code was developed to handle large sets of discrete nuclear levels and the multipole transitions among these levels; it operates in three modes. The Data File Error Analysis mode analyzes an existing STAPLUS input file containing the levels and their multipole transition branches for a number of physics and/or typographical errors. The Interactive Data File Generation mode allows the user to create input files of discrete levels and their branching fractions in the format required by STAPLUS, even though the user enters the information in the (different) format used by many people in the nuclear structure field. In the Branching Fractions Calculations mode, the discrete nuclear level set is read, and the multipole transitions among the levels are computed under one of two possible assumptions: (1) the levels have no collective character, or (2) the levels are all rotational band heads. Only E1, M1, and E2 transitions are considered, and the respective strength functions may be constants or, in the case of E1 transitions, the strength function may be energy dependent. The first option is used for nuclei closed shells; the bandhead option may be used to vary the E1, M1, and E2 strengths for interband transitions. K-quantum number selection rules may be invoked if desired. 19 refs.

  14. A Hydrogen Containment Process For Nuclear Thermal Engine Ground Testing

    Science.gov (United States)

    Wang, Ten-See; Stewart, Eric; Canabal, Francisco

    2016-01-01

    A hydrogen containment process was proposed for ground testing of a nuclear thermal engine. The hydrogen exhaust from the engine is contained in two unit operations: an oxygen-rich burner and a tubular heat exchanger. The burner burns off the majority of the hydrogen, and the remaining hydrogen is removed in the tubular heat exchanger through the species recombination mechanism. A multi-dimensional, pressure-based multiphase computational fluid dynamics methodology was used to conceptually sizing the oxygen-rich burner, while a one-dimensional thermal analysis methodology was used to conceptually sizing the heat exchanger. Subsequently, a steady-state operation of the entire hydrogen containment process, from pressure vessel, through nozzle, diffuser, burner and heat exchanger, was simulated numerically, with the afore-mentioned computational fluid dynamics methodology. The computational results show that 99% of hydrogen reduction is achieved at the end of the burner, and the rest of the hydrogen is removed to a trivial level in the heat exchanger. The computed flammability at the exit of the heat exchanger is less than the lower flammability limit, confirming the hydrogen containment capability of the proposed process.

  15. Severe accident analysis in a two-loop PWR nuclear power plant with the ASTEC code

    Energy Technology Data Exchange (ETDEWEB)

    Sadek, Sinisa; Amizic, Milan; Grgic, Davor [Zagreb Univ. (Croatia). Faculty of Electrical Engineering and Computing

    2013-12-15

    The ASTEC/V2.0 computer code was used to simulate a hypothetical severe accident sequence in the nuclear power plant Krsko, a 2-loop pressurized water reactor (PWR) plant. ASTEC is an integral code jointly developed by Institut de Radioprotection et de Surete Nucleaire (IRSN, France) and Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS, Germany) to assess nuclear power plant behaviour during a severe accident. The analysis was conducted in 2 steps. First, the steady state calculation was performed in order to confirm the applicability of the plant model and to obtain correct initial conditions for the accident analysis. The second step was the calculation of the station blackout accident with a leakage of the primary coolant through degraded reactor coolant pump seals, which was a small LOCA without makeup capability. Two scenarios were analyzed: one with and one without the auxiliary feedwater (AFW). The latter scenario, without the AFW, resulted in earlier core damage. In both cases, the accident ended with a core melt and a reactor pressure vessel failure with significant release of hydrogen. In addition, results of the ASTEC calculation were compared with results of the RELAP5/SCDAPSIM calculation for the same transient scenario. The results comparison showed a good agreement between predictions of those 2 codes. (orig.)

  16. Nuclear Reactor Component Code CUPID-I: Numerical Scheme and Preliminary Assessment Results

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Hyoung Kyu; Jeong, Jae Jun; Park, Ik Kyu; Kim, Jong Tae; Yoon, Han Young

    2007-12-15

    A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analysis of components of a nuclear reactor, such as reactor vessel, steam generator, containment, etc. It adopted three-dimensional, transient, two phase and three-field model. In order to develop the numerical schemes for the three-field model, various numerical schemes have been examined including the SMAC, semi-implicit ICE, SIMPLE, Row Scheme and so on. Among them, the ICE scheme for the three-field model was presented in the present report. The CUPID code is utilizing unstructured mesh for the simulation of complicated geometries of the nuclear reactor components. The conventional ICE scheme that was applied to RELAP5 and COBRA-TF, therefore, were modified for the application to the unstructured mesh. Preliminary calculations for the unstructured semi-implicit ICE scheme have been conducted for a verification of the numerical method from a qualitative point of view. The preliminary calculation results showed that the present numerical scheme is robust and efficient for the prediction of phase changes and flow transitions due to a boiling and a flashing. These calculation results also showed the strong coupling between the pressure and void fraction changes. Thus, it is believed that the semi-implicit ICE scheme can be utilized for transient two-phase flows in a component of a nuclear reactor.

  17. The signal sequence coding region promotes nuclear export of mRNA.

    Science.gov (United States)

    Palazzo, Alexander F; Springer, Michael; Shibata, Yoko; Lee, Chung-Sheng; Dias, Anusha P; Rapoport, Tom A

    2007-12-01

    In eukaryotic cells, most mRNAs are exported from the nucleus by the transcription export (TREX) complex, which is loaded onto mRNAs after their splicing and capping. We have studied in mammalian cells the nuclear export of mRNAs that code for secretory proteins, which are targeted to the endoplasmic reticulum membrane by hydrophobic signal sequences. The mRNAs were injected into the nucleus or synthesized from injected or transfected DNA, and their export was followed by fluorescent in situ hybridization. We made the surprising observation that the signal sequence coding region (SSCR) can serve as a nuclear export signal of an mRNA that lacks an intron or functional cap. Even the export of an intron-containing natural mRNA was enhanced by its SSCR. Like conventional export, the SSCR-dependent pathway required the factor TAP, but depletion of the TREX components had only moderate effects. The SSCR export signal appears to be characterized in vertebrates by a low content of adenines, as demonstrated by genome-wide sequence analysis and by the inhibitory effect of silent adenine mutations in SSCRs. The discovery of an SSCR-mediated pathway explains the previously noted amino acid bias in signal sequences and suggests a link between nuclear export and membrane targeting of mRNAs.

  18. Current state of nuclear fuel cycles in nuclear engineering and trends in their development according to the environmental safety requirements

    Science.gov (United States)

    Vislov, I. S.; Pischulin, V. P.; Kladiev, S. N.; Slobodyan, S. M.

    2016-08-01

    The state and trends in the development of nuclear fuel cycles in nuclear engineering, taking into account the ecological aspects of using nuclear power plants, are considered. An analysis of advantages and disadvantages of nuclear engineering, compared with thermal engineering based on organic fuel types, was carried out. Spent nuclear fuel (SNF) reprocessing is an important task in the nuclear industry, since fuel unloaded from modern reactors of any type contains a large amount of radioactive elements that are harmful to the environment. On the other hand, the newly generated isotopes of uranium and plutonium should be reused to fabricate new nuclear fuel. The spent nuclear fuel also includes other types of fission products. Conditions for SNF handling are determined by ecological and economic factors. When choosing a certain handling method, one should assess these factors at all stages of its implementation. There are two main methods of SNF handling: open nuclear fuel cycle, with spent nuclear fuel assemblies (NFAs) that are held in storage facilities with their consequent disposal, and closed nuclear fuel cycle, with separation of uranium and plutonium, their purification from fission products, and use for producing new fuel batches. The development of effective closed fuel cycles using mixed uranium-plutonium fuel can provide a successful development of the nuclear industry only under the conditions of implementation of novel effective technological treatment processes that meet strict requirements of environmental safety and reliability of process equipment being applied. The diversity of technological processes is determined by different types of NFA devices and construction materials being used, as well as by the composition that depends on nuclear fuel components and operational conditions for assemblies in the nuclear power reactor. This work provides an overview of technological processes of SNF treatment and methods of handling of nuclear fuel

  19. Development of undergraduate nuclear security curriculum at College of Engineering, Universiti Tenaga Nasional

    Science.gov (United States)

    Hamid, Nasri A.; Mujaini, Madihah; Mohamed, Abdul Aziz

    2017-01-01

    The Center for Nuclear Energy (CNE), College of Engineering, Universiti Tenaga Nasional (UNITEN) has a great responsibility to undertake educational activities that promote developing human capital in the area of nuclear engineering and technology. Developing human capital in nuclear through education programs is necessary to support the implementation of nuclear power projects in Malaysia in the near future. In addition, the educational program must also meet the nuclear power industry needs and requirements. In developing a certain curriculum, the contents must comply with the university's Outcomes Based Education (OBE) philosophy. One of the important courses in the nuclear curriculum is in the area of nuclear security. Basically the nuclear security course covers the current issues of law, politics, military strategy, and technology with regard to weapons of mass destruction and related topics in international security, and review legal regulations and political relationship that determine the state of nuclear security at the moment. In addition, the course looks into all aspects of the nuclear safeguards, builds basic knowledge and understanding of nuclear non-proliferation, nuclear forensics and nuclear safeguards in general. The course also discusses tools used to combat nuclear proliferation such as treaties, institutions, multilateral arrangements and technology controls. In this paper, we elaborate the development of undergraduate nuclear security course at the College of Engineering, Universiti Tenaga Nasional. Since the course is categorized as mechanical engineering subject, it must be developed in tandem with the program educational objectives (PEO) of the Bachelor of Mechanical Engineering program. The course outcomes (CO) and transferrable skills are also identified. Furthermore, in aligning the CO with program outcomes (PO), the PO elements need to be emphasized through the CO-PO mapping. As such, all assessments and distribution of Bloom Taxonomy

  20. Reflections on the Fukushima Daiichi nuclear accident toward social-scientific literacy and engineering resilience

    CERN Document Server

    Carson, Cathryn; Jensen, Mikael; Juraku, Kohta; Nagasaki, Shinya; Tanaka, Satoru

    2015-01-01

    This book focuses on nuclear engineering education in the post-Fukushima era. It was edited by the organizers of the summer school held in August 2011 in University of California, Berkeley, as part of a collaborative program between the University of Tokyo and UC Berkeley. Motivated by the particular relevance and importance of social-scientific approaches to various crucial aspects of nuclear technology, special emphasis was placed on integrating nuclear science and engineering with social science. The book consists of the lectures given in 2011 summer school and additional chapters that cover developments in the past three years since the accident. It provides an arena for discussions to find and create a renewed platform for engineering practices, and thus nuclear engineering education, which are essential in the post-Fukushima era for nurturing nuclear engineers who need to be both technically competent and trusted in society.

  1. Coupling methods for parallel running RELAPSim codes in nuclear power plant simulation

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yankai; Lin, Meng, E-mail: linmeng@sjtu.edu.cn; Yang, Yanhua

    2016-02-15

    When the plant is modeled detailedly for high precision, it is hard to achieve real-time calculation for one single RELAP5 in a large-scale simulation. To improve the speed and ensure the precision of simulation at the same time, coupling methods for parallel running RELAPSim codes were proposed in this study. Explicit coupling method via coupling boundaries was realized based on a data-exchange and procedure-control environment. Compromise of synchronization frequency was well considered to improve the precision of simulation and guarantee the real-time simulation at the same time. The coupling methods were assessed using both single-phase flow models and two-phase flow models and good agreements were obtained between the splitting–coupling models and the integrated model. The mitigation of SGTR was performed as an integral application of the coupling models. A large-scope NPP simulator was developed adopting six splitting–coupling models of RELAPSim and other simulation codes. The coupling models could improve the speed of simulation significantly and make it possible for real-time calculation. In this paper, the coupling of the models in the engineering simulator is taken as an example to expound the coupling methods, i.e., coupling between parallel running RELAPSim codes, and coupling between RELAPSim code and other types of simulation codes. However, the coupling methods are also referable in other simulator, for example, a simulator employing ATHLETE instead of RELAP5, other logic code instead of SIMULINK. It is believed the coupling method is commonly used for NPP simulator regardless of the specific codes chosen in this paper.

  2. Just sustainability? Sustainability and social justice in professional codes of ethics for engineers.

    Science.gov (United States)

    Brauer, Cletus S

    2013-09-01

    Should environmental, social, and economic sustainability be of primary concern to engineers? Should social justice be among these concerns? Although the deterioration of our natural environment and the increase in social injustices are among today's most pressing and important issues, engineering codes of ethics and their paramountcy clause, which contains those values most important to engineering and to what it means to be an engineer, do not yet put either concept on a par with the safety, health, and welfare of the public. This paper addresses a recent proposal by Michelfelder and Jones (2011) to include sustainability in the paramountcy clause as a way of rectifying the current disregard for social justice issues in the engineering codes. That proposal builds on a certain notion of sustainability that includes social justice as one of its dimensions and claims that social justice is a necessary condition for sustainability, not vice versa. The relationship between these concepts is discussed, and the original proposal is rejected. Drawing on insights developed throughout the paper, some suggestions are made as to how one should address the different requirements that theory and practice demand of the value taxonomy of professional codes of ethics.

  3. STEM Leader from the Roeper School: An Interview with Nuclear Engineer Clair J. Sullivan

    Science.gov (United States)

    Ambrose, Don

    2016-01-01

    Clair J. Sullivan is an assistant professor in the Department of Nuclear, Plasma and Radiological Engineering at the University of Illinois at Urbana-Champaign (UIUC). Her research interests include radiation detection and measurements; gamma-ray spectroscopy; automated isotope identification algorithms; nuclear forensics; nuclear security;…

  4. MINA 2008: an approach to professionalization in Nuclear Engineering; MINA 2008: una via de profesionalizacion en Ingenieria Nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L. E.; Garcia Cuesta, J. C.; Falcon, S.; Marco, M.

    2008-07-01

    At present there are evident signs around the world of what has been called the Nuclear Renaissance. The anticipated nuclear revival is coinciding in time with the retirement of a generation of professionals who have been responsible for the design, construction and operation of the second generation nuclear power plants and, with them, a large part of the experience gained. Whatever its future energy strategy may be, Spain is already recognizing this reality through projects that national companies, engineering firms and institutions are undertaking beyond the country borders. Consequently, it is expected that the nuclear sector will be demanding competent human capital. National support for nuclear power would increase this need even more. In this national and international panorama, CIEMAT and the UAM have recognized the opportunity to adopt a new concept, as ambitious as it is stimulating, for training of nuclear professionals: the Masters in Nuclear Engineering and Applications (MINA). CIEMAT long tradition of training in nuclear technology and the background of collaboration between CIEMAT and the UAM have been the fundamental basis for defining MINA-2008. Born with a multi-institutional vocation, MINA aims to address the needs of the nuclear sector and promote the interest of young people in nuclear technology. (Author) 3 refs.

  5. Safe, Compact Nuclear Propulsion: Solid Core Nuclear Propulsion Concept

    Science.gov (United States)

    1988-10-01

    analysis group developed ROM component cost estimates given in representative ranges. 4.1 Engine System A representative nuclear thermal rocket engine...nuclear thermal rocket engine cycle balance computer code. The design requirements for the engine were: Thrust : 15,000 lbf Champer Pressure 500 psia...advanced nuclear thermal rockets . Our analysis was based on an examination of presentation material provided by Martin, some independent calculations of

  6. Automatic code generation enables nuclear gradient computations for fully internally contracted multireference theory

    CERN Document Server

    MacLeod, Matthew K

    2015-01-01

    Analytical nuclear gradients for fully internally contracted complete active space second-order perturbation theory (CASPT2) are reported. This implementation has been realized by an automated code generator that can handle spin-free formulas for the CASPT2 energy and its derivatives with respect to variations of molecular orbitals and reference coefficients. The underlying complete active space self-consistent field and the so-called Z-vector equations are solved using density fitting. With full internal contraction the size of first-order wave functions scales polynomially with the number of active orbitals. The CASPT2 gradient program and the code generator are both publicly available. This work enables the CASPT2 geometry optimization of molecules as complex as those investigated by respective single-point calculations.

  7. The nuclear protein-coding gene ANKRD23 negatively regulates myoblast differentiation.

    Science.gov (United States)

    Wang, Xiaojing; Zeng, Rui; Xu, Haiyang; Xu, Zaiyan; Zuo, Bo

    2017-09-20

    Muscle fiber formation is a complex process and subject to fine regulation of a variety of protein-coding genes and non-coding RNA. In this study, we identified a nuclear protein-coding gene ANKRD23 which was highly expressed in muscle. Quantitative real-time PCR, western blotting and immunofluorescence were used to detect the expression change of myoblast differentiation marker genes after knockdown and overexpression of ANKRD23. The results showed that the expression of myoblast differentiation marker genes were increased by interference and reduced by ANKRD23 overexpression, indicating that ANKRD23 played a negative role in the myoblast differentiation. Interestingly, we discovered a long non-coding RNA-AK004293 which was overlapped with the 3'UTR of ANKRD23 gene. Then we detected the effect of AK004293 on the expression of ANKRD23 and myoblast differentiation marker genes in C2C12 myoblasts. The results showed that AK004293 had no significant effect on the expression of myoblast differentiation maker genes and ANKRD23. In conclusion, our results established the foundation for further studies about the regulation mechanism of ANKRD23 in muscle development. Copyright © 2017 Elsevier B.V. All rights reserved.

  8. Improvement of Level-1 PSA computer code package -A study for nuclear safety improvement-

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang Kyu; Kim, Tae Woon; Ha, Jae Joo; Han, Sang Hoon; Cho, Yeong Kyun; Jeong, Won Dae; Jang, Seung Cheol; Choi, Young; Seong, Tae Yong; Kang, Dae Il; Hwang, Mi Jeong; Choi, Seon Yeong; An, Kwang Il [Korea Atomic Energy Res. Inst., Taejon (Korea, Republic of)

    1994-07-01

    This year is the second year of the Government-sponsored Mid- and Long-Term Nuclear Power Technology Development Project. The scope of this subproject titled on `The Improvement of Level-1 PSA Computer Codes` is divided into three main activities : (1) Methodology development on the under-developed fields such as risk assessment technology for plant shutdown and external events, (2) Computer code package development for Level-1 PSA, (3) Applications of new technologies to reactor safety assessment. At first, in the area of PSA methodology development, foreign PSA reports on shutdown and external events have been reviewed and various PSA methodologies have been compared. Level-1 PSA code KIRAP and CCF analysis code COCOA are converted from KOS to Windows. Human reliability database has been also established in this year. In the area of new technology applications, fuzzy set theory and entropy theory are used to estimate component life and to develop a new measure of uncertainty importance. Finally, in the field of application study of PSA technique to reactor regulation, a strategic study to develop a dynamic risk management tool PEPSI and the determination of inspection and test priority of motor operated valves based on risk importance worths have been studied. (Author).

  9. Implementation of a multiple round quantum dense coding using nuclear magnetic resonance

    Institute of Scientific and Technical Information of China (English)

    ZHANG; Jingfu; XIE; Jingyi; WANG; Chuan; DENG; Zhiwei; LU

    2005-01-01

    A multiple round quantum dense coding scheme based on the quantum phase estimation algorithm is proposed and implemented in a three qubit nuclear magnetic resonance (NMR) quantum computer. Using an m + 1 qubit system, Bob can transmit one of 2m+1 messages to Alice, through manipulating only one qubit and exchanging it between Alice and Bob for m rounds. The information capacity is enhanced to m + 1 bits as compared to m bits in a classical scheme. The scheme has been demonstrated in NMR system, and the experimental results show a good agreement between theory and experiment.

  10. Communication: Automatic code generation enables nuclear gradient computations for fully internally contracted multireference theory

    Energy Technology Data Exchange (ETDEWEB)

    MacLeod, Matthew K.; Shiozaki, Toru [Department of Chemistry, Northwestern University, 2145 Sheridan Rd., Evanston, Illinois 60208 (United States)

    2015-02-07

    Analytical nuclear gradients for fully internally contracted complete active space second-order perturbation theory (CASPT2) are reported. This implementation has been realized by an automated code generator that can handle spin-free formulas for the CASPT2 energy and its derivatives with respect to variations of molecular orbitals and reference coefficients. The underlying complete active space self-consistent field and the so-called Z-vector equations are solved using density fitting. The implementation has been applied to the vertical and adiabatic ionization potentials of the porphin molecule to illustrate its capability.

  11. Nuclear numerical range and quantum error correction codes for non-unitary noise models

    Science.gov (United States)

    Lipka-Bartosik, Patryk; Życzkowski, Karol

    2017-01-01

    We introduce a notion of nuclear numerical range defined as the set of expectation values of a given operator A among normalized pure states, which belong to the nucleus of an auxiliary operator Z. This notion proves to be applicable to investigate models of quantum noise with block-diagonal structure of the corresponding Kraus operators. The problem of constructing a suitable quantum error correction code for this model can be restated as a geometric problem of finding intersection points of certain sets in the complex plane. This technique, worked out in the case of two-qubit systems, can be generalized for larger dimensions.

  12. Application of complex engineering solutions through advanced composite innovation (for repair of degraded buried pipe at Vandellos II Nuclear Power Plant); Reparacion de tuberias de un sistema de servicios no esenciales con recubrimiento interno de fibra de carbono

    Energy Technology Data Exchange (ETDEWEB)

    Bueno, J. M.; Raji, B. B.

    2011-07-01

    This technical presentation is focused on introducing an engineering solution approach and identification of sensitivity of applications of advanced carbon fiber in a pressurized wet environment: Engineering design, quality assurance of installation, inspection, and a comprehensive testing program to validate and bench mark the design data and compliance with code requirements in nuclear power plants.

  13. Bibliography of Connecticut Advanced Nuclear Engineering Laboratory reports

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1965-12-01

    This report, published in two, volumes, is a bibliography of the reports published at the Connecticut Advanced Nuclear Engineering Laboratory (CANEL). The reports cover the period 1952 through 1965 and include the Aircraft Nuclear Propulsion program, the Advanced Liquid Metal Cooled Reactor program, the Advanced Reactor Materials program and the SNAP-50 program. The bibliography contains the report number, title, author, date published, and classification. In some cases where the writing of a report was a group effort, and in some reports containing compilations of certain types of data, the author column is not applicable. This is indicated by a {open_quotes}n.a.{close_quotes} in the author column. The following types of reports are included: PWAC`s, TIM`s, CNLM`s. FXM`s and miscellaneous reports. PWAC and TIM reports conform to the requirements of AEC Manual Chapter 3202-041 and 3202-042, respectively. Most of the technical information of interest generated by this project is documented in these reports, CNLM and FXM reports were written primarily for internal distribution. However, these reports contain enough information of technical interest to warrant their inclusion. All CNLM`s and those FXM`s considered to be of interest are included in this bibliography. The MPR`s (Monthly Progress Reports) are the most important of the miscellaneous categories of reports. The other miscellaneous categories relate primarily to equipment and reactor specifications. The Division of Technical Information Extension (DTIE) at Oak Ridge, Tennessee has been designated as the primary recipient of the reports in the CANEL library. When more than one copy of a report was available, the additional copies were delivered to the Lawrence Radiation Laboratory, Livermore, California.

  14. Bibliography of Connecticut Advanced Nuclear Engineering Laboratory reports

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1965-12-01

    This report, published in two volumes, is a bibliography of the reports published at the Connecticut Advanced Nuclear Engineering Laboratory (CANEL). The reports cover the period 1952 through 1965 and include the Aircraft Nuclear Propulsion program, the Advanced Liquid Metal Cooled Reactor program, the Advanced Reactor Materials program and the SNAP-50 program. The bibliography contains the report number, title, author, date published, and classification. In some cases where the writing of a report was a group effort, and in some reports containing compilations of certain types of data, the author column is not applicable. This is indicated by a {open_quotes}n.a.{close_quotes} in the author column. The following types of reports are included: PWAC`s, TIM`s, CNLM`s, FXM`s and miscellaneous reports. PWAC and TIM reports conform to the requirements of AEC Manual Chapter 3202-041 and 3202-042, respectively. Most of the technical information of interest generated by this project is documented in these reports. CNLM and FXM reports were written primarily for internal distribution. However, these reports contain enough information of technical interest to warrant their inclusion. All CNLM`s and those FXM`s considered to be of interest are included in this bibliography. The MPR`s (Monthly Progress Reports) are the most important of the miscellaneous categories of reports. The other miscellaneous categories relate primarily to equipment and reactor specifications. The Division of Technical Information Extension (DTIE) at Oak Ridge, Tennessee has been designated as the primary recipient of the reports in the CANEL library. When more than one copy of a report was available, the additional copies were delivered to the Lawrence Radiation Laboratory, Livermore, California.

  15. Synthetic alienation of microbial organisms by using genetic code engineering: Why and how?

    Science.gov (United States)

    Kubyshkin, Vladimir; Budisa, Nediljko

    2017-08-01

    The main goal of synthetic biology (SB) is the creation of biodiversity applicable for biotechnological needs, while xenobiology (XB) aims to expand the framework of natural chemistries with the non-natural building blocks in living cells to accomplish artificial biodiversity. Protein and proteome engineering, which overcome limitation of the canonical amino acid repertoire of 20 (+2) prescribed by the genetic code by using non-canonic amino acids (ncAAs), is one of the main focuses of XB research. Ideally, estranging the genetic code from its current form via systematic introduction of ncAAs should enable the development of bio-containment mechanisms in synthetic cells potentially endowing them with a "genetic firewall" i.e. orthogonality which prevents genetic information transfer to natural systems. Despite rapid progress over the past two decades, it is not yet possible to completely alienate an organism that would use and maintain different genetic code associations permanently. In order to engineer robust bio-contained life forms, the chemical logic behind the amino acid repertoire establishment should be considered. Starting from recent proposal of Hartman and Smith about the genetic code establishment in the RNA world, here the authors mapped possible biotechnological invasion points for engineering of bio-contained synthetic cells equipped with non-canonical functionalities. Copyright © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  16. Engineering and maintenance applied to safety-related valves in nuclear power plants; Ingenieria y mantenimiento aplicado a valvulas relacionadas con la seguridad en centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Verdu, M. F.; Perez-Aranda, J.

    2014-04-01

    Nuclear Division in Iberdrola engineering and Construction has a team with extensive experience on engineering and services works related to valves. Also, this team is linked to UNESA as Technical support and Reference Center. Iberdrola engineering and construction experience in nuclear power plants valves, gives effective response to engineering and maintenance works that can be demanded in a nuclear power plant and it requires a high degree of qualification and knowledge both in Operation and Outages. (Author)

  17. Second International Workshop on Software Engineering and Code Design in Parallel Meteorological and Oceanographic Applications

    Science.gov (United States)

    OKeefe, Matthew (Editor); Kerr, Christopher L. (Editor)

    1998-01-01

    This report contains the abstracts and technical papers from the Second International Workshop on Software Engineering and Code Design in Parallel Meteorological and Oceanographic Applications, held June 15-18, 1998, in Scottsdale, Arizona. The purpose of the workshop is to bring together software developers in meteorology and oceanography to discuss software engineering and code design issues for parallel architectures, including Massively Parallel Processors (MPP's), Parallel Vector Processors (PVP's), Symmetric Multi-Processors (SMP's), Distributed Shared Memory (DSM) multi-processors, and clusters. Issues to be discussed include: (1) code architectures for current parallel models, including basic data structures, storage allocation, variable naming conventions, coding rules and styles, i/o and pre/post-processing of data; (2) designing modular code; (3) load balancing and domain decomposition; (4) techniques that exploit parallelism efficiently yet hide the machine-related details from the programmer; (5) tools for making the programmer more productive; and (6) the proliferation of programming models (F--, OpenMP, MPI, and HPF).

  18. Parameter Study of the LIFE Engine Nuclear Design

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, K J; Meier, W R; Latkowski, J F; Abbott, R P

    2009-07-10

    LLNL is developing the nuclear fusion based Laser Inertial Fusion Energy (LIFE) power plant concept. The baseline design uses a depleted uranium (DU) fission fuel blanket with a flowing molten salt coolant (flibe) that also breeds the tritium needed to sustain the fusion energy source. Indirect drive targets, similar to those that will be demonstrated on the National Ignition Facility (NIF), are ignited at {approx}13 Hz providing a 500 MW fusion source. The DU is in the form of a uranium oxycarbide kernel in modified TRISO-like fuel particles distributed in a carbon matrix forming 2-cm-diameter pebbles. The thermal power is held at 2000 MW by continuously varying the 6Li enrichment in the coolants. There are many options to be considered in the engine design including target yield, U-to-C ratio in the fuel, fission blanket thickness, etc. Here we report results of design variations and compare them in terms of various figures of merit such as time to reach a desired burnup, full-power years of operation, time and maximum burnup at power ramp down and the overall balance of plant utilization.

  19. Challenge problem and milestones for : Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC).

    Energy Technology Data Exchange (ETDEWEB)

    Freeze, Geoffrey A.; Wang, Yifeng; Howard, Robert; McNeish, Jerry A.; Schultz, Peter Andrew; Arguello, Jose Guadalupe, Jr.

    2010-09-01

    This report describes the specification of a challenge problem and associated challenge milestones for the Waste Integrated Performance and Safety Codes (IPSC) supporting the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The NEAMS challenge problems are designed to demonstrate proof of concept and progress towards IPSC goals. The goal of the Waste IPSC is to develop an integrated suite of modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with robust verification, validation, and software quality requirements. To demonstrate proof of concept and progress towards these goals and requirements, a Waste IPSC challenge problem is specified that includes coupled thermal-hydrologic-chemical-mechanical (THCM) processes that describe (1) the degradation of a borosilicate glass waste form and the corresponding mobilization of radionuclides (i.e., the processes that produce the radionuclide source term), (2) the associated near-field physical and chemical environment for waste emplacement within a salt formation, and (3) radionuclide transport in the near field (i.e., through the engineered components - waste form, waste package, and backfill - and the immediately adjacent salt). The initial details of a set of challenge milestones that collectively comprise the full challenge problem are also specified.

  20. Monte-Carlo code calculation of 3D reactor core model with usage of burnt fuel isotopic compositions, obtained by engineering codes

    Energy Technology Data Exchange (ETDEWEB)

    Aleshin, Sergey S.; Gorodkov, Sergey S.; Shcherenko, Anna I. [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation)

    2016-09-15

    A burn-up calculation of large systems by Monte-Carlo code (MCU) is complex process and it requires large computational costs. Previously prepared isotopic compositions are proposed to be used for the Monte-Carlo code calculations of different system states with burnt fuel. Isotopic compositions are calculated by an approximation method. The approximation method is based on usage of a spectral functionality and reference isotopic compositions, that are calculated by the engineering codes (TVS-M, BIPR-7A and PERMAK-A). The multiplication factors and power distributions of FAs from a 3-D reactor core are calculated in this work by the Monte-Carlo code MCU using earlier prepared isotopic compositions. The separate conditions of the burnt core are observed. The results of MCU calculations were compared with those that were obtained by engineering codes.

  1. Brief 72 Nuclear Engineering Enrollments and Degrees Survey, 2013 Data (2-14)

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2014-02-15

    The survey includes degrees granted between September 1, 2012 and August 31, 2013. Enrollment information refers to the fall term 2013. The enrollments and degrees data include students majoring in nuclear engineering or in an option program equivalent to a major. Thirty-two academic programs reported having nuclear engineering programs during 2013, and data was received from all thirty-two programs. The data for two nuclear engineering programs include enrollments and degrees in health physics options that are also reported in the health physics enrollments and degrees data.

  2. A mono-dimensional nuclear fuel performance analysis code, PUMA, development from a coupled approach

    Energy Technology Data Exchange (ETDEWEB)

    Cheon, J. S.; Lee, B. O.; Lee, C. B. [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong, Daejeon, 305-353 (Korea, Republic of); Yacout, A. M. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2013-07-01

    Multidimensional-multi-physical phenomena in nuclear fuels are treated as a set of mono-dimensional-coupled problems which encompass heat, displacement, fuel constituent redistribution, and fission gas release. Rather than uncoupling these coupled equations as in conventional fuel performance analysis codes, efforts are put into to obtain fully coupled solutions by relying on the recent advances of numerical analysis. Through this approach, a new SFR metal fuel performance analysis code, called PUMA (Performance of Uranium Metal fuel rod Analysis code) is under development. Although coupling between temperature and fuel constituent was made easily, the coupling between the mechanical equilibrium equation and a set of stiff kinetics equations for fission gas release is accomplished by introducing one-level Newton scheme through backward differentiation formula. Displacement equations from 1D finite element formulation of the mechanical equilibrium equation are solved simultaneously with stress equation, creep equation, swelling equation, and FGR equations. Calculations was made successfully such that the swelling and the hydrostatic pressure are interrelated each other. (authors)

  3. Computer simulation of Angra-2 PWR nuclear reactor core using MCNPX code

    Energy Technology Data Exchange (ETDEWEB)

    Medeiros, Marcos P.C. de; Rebello, Wilson F., E-mail: eng.cavaliere@ime.eb.br, E-mail: rebello@ime.eb.br [Instituto Militar de Engenharia - Secao de Engenharia Nuclear, Rio de Janeiro, RJ (Brazil); Oliveira, Claudio L. [Universidade Gama Filho, Departamento de Matematica, Rio de Janeiro, RJ (Brazil); Vellozo, Sergio O., E-mail: vellozo@cbpf.br [Centro Tecnologico do Exercito. Divisao de Defesa Quimica, Biologica e Nuclear, Rio de Janeiro, RJ (Brazil); Silva, Ademir X. da, E-mail: ademir@nuclear.ufrj.br [Coordenacao dos Programas de Pos Gaduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    In this work the MCNPX (Monte Carlo N-Particle Transport Code) code was used to develop a computerized model of the core of Angra 2 PWR (Pressurized Water Reactor) nuclear reactor. The model was created without any kind of homogenization, but using real geometric information and material composition of that reactor, obtained from the FSAR (Final Safety Analysis Report). The model is still being improved and the version presented in this work is validated by comparing values calculated by MCNPX with results calculated by others means and presented on FSAR. This paper shows the results already obtained to K{sub eff} and K{infinity}, general parameters of the core, considering the reactor operating under stationary conditions of initial testing and operation. Other stationary operation conditions have been simulated and, in all tested cases, there was a close agreement between values calculated computationally through this model and data presented on the FSAR, which were obtained by other codes. This model is expected to become a valuable tool for many future applications. (author)

  4. Research on the improvement of nuclear safety -Improvement of level 1 PSA computer code package-

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang Kyoo; Kim, Tae Woon; Kim, Kil Yoo; Han, Sang Hoon; Jung, Won Dae; Jang, Seung Chul; Yang, Joon Un; Choi, Yung; Sung, Tae Yong; Son, Yung Suk; Park, Won Suk; Jung, Kwang Sub; Kang Dae Il; Park, Jin Heui; Hwang, Mi Jung; Hah, Jae Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This year is the third year of the Government-sponsored mid- and long-term nuclear power technology development project. The scope of this sub project titled on `The improvement of level-1 PSA computer codes` is divided into three main activities : (1) Methodology development on the underdeveloped fields such as risk assessment technology for plant shutdown and low power situations, (2) Computer code package development for level-1 PSA, (3) Applications of new technologies to reactor safety assessment. At first, in this area of shutdown risk assessment technology development, plant outage experiences of domestic plants are reviewed and plant operating states (POS) are decided. A sample core damage frequency is estimated for over draining event in RCS low water inventory i.e. mid-loop operation. Human reliability analysis and thermal hydraulic support analysis are identified to be needed to reduce uncertainty. Two design improvement alternatives are evaluated using PSA technique for mid-loop operation situation: one is use of containment spray system as backup of shutdown cooling system and the other is installation of two independent level indication system. Procedure change is identified more preferable option to hardware modification in the core damage frequency point of view. Next, level-1 PSA code KIRAP is converted to PC-windows environment. For the improvement of efficiency in performing PSA, the fast cutest generation algorithm and an analytical technique for handling logical loop in fault tree modeling are developed. 48 figs, 15 tabs, 59 refs. (Author).

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

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

  6. Cerebellar Nuclear Neurons Use Time and Rate Coding to Transmit Purkinje Neuron Pauses.

    Science.gov (United States)

    Sudhakar, Shyam Kumar; Torben-Nielsen, Benjamin; De Schutter, Erik

    2015-12-01

    Neurons of the cerebellar nuclei convey the final output of the cerebellum to their targets in various parts of the brain. Within the cerebellum their direct upstream connections originate from inhibitory Purkinje neurons. Purkinje neurons have a complex firing pattern of regular spikes interrupted by intermittent pauses of variable length. How can the cerebellar nucleus process this complex input pattern? In this modeling study, we investigate different forms of Purkinje neuron simple spike pause synchrony and its influence on candidate coding strategies in the cerebellar nuclei. That is, we investigate how different alignments of synchronous pauses in synthetic Purkinje neuron spike trains affect either time-locking or rate-changes in the downstream nuclei. We find that Purkinje neuron synchrony is mainly represented by changes in the firing rate of cerebellar nuclei neurons. Pause beginning synchronization produced a unique effect on nuclei neuron firing, while the effect of pause ending and pause overlapping synchronization could not be distinguished from each other. Pause beginning synchronization produced better time-locking of nuclear neurons for short length pauses. We also characterize the effect of pause length and spike jitter on the nuclear neuron firing. Additionally, we find that the rate of rebound responses in nuclear neurons after a synchronous pause is controlled by the firing rate of Purkinje neurons preceding it.

  7. Validation of WIMS-SNAP code systems for calculations in TRIGA-MARK II type reactors; Validacion del sistema de codigos WIMS-SNAP para calculos en reactores nucleares tipo TRIGA-MARK II

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez Valle, S.; Lopez Aldama, D. [Centro de Investigaciones Nucleares, Tecnologicas y Ambientales, La Habana (Cuba). E-mail: svalle@ctn.isctn.edu.cu

    2000-07-01

    The following paper contributes to validate the Nuclear Engineering Department methods to carry out calculations in TRIGA reactors solving a Benchmark. The benchmark is analyzed with the WIMS-D/4-SNAP/3D code system and using the cross section library WIMS-TRIGA. A brief description of the DSN method is presented used in WIMS/d{sup 4} code and also the SNAP-3d code is shortly explained. The results are presented and compared with the experimental values. In other hand the possible error sources are analyzed. (author)

  8. Pre-engineering Spaceflight Validation of Environmental Models and the 2005 HZETRN Simulation Code

    Science.gov (United States)

    Nealy, John E.; Cucinotta, Francis A.; Wilson, John W.; Badavi, Francis F.; Dachev, Ts. P.; Tomov, B. T.; Walker, Steven A.; DeAngelis, Giovanni; Blattnig, Steve R.; Atwell, William

    2006-01-01

    The HZETRN code has been identified by NASA for engineering design in the next phase of space exploration highlighting a return to the Moon in preparation for a Mars mission. In response, a new series of algorithms beginning with 2005 HZETRN, will be issued by correcting some prior limitations and improving control of propagated errors along with established code verification processes. Code validation processes will use new/improved low Earth orbit (LEO) environmental models with a recently improved International Space Station (ISS) shield model to validate computational models and procedures using measured data aboard ISS. These validated models will provide a basis for flight-testing the designs of future space vehicles and systems of the Constellation program in the LEO environment.

  9. UMCP-BG and E collaboration in nuclear power engineering in the framework of DOE-Utility Nuclear Power Engineering Education Matching Grant Program

    Energy Technology Data Exchange (ETDEWEB)

    Wolfe, Lothar PhD

    2000-03-01

    The DOE-Utility Nuclear Power Engineering Education Matching Grant Program has been established to support the education of students in Nuclear Engineering Programs to maintain a knowledgeable workforce in the United States in order to keep nuclear power as a viable component in a mix of energy sources for the country. The involvement of the utility industry ensures that this grant program satisfies the needs and requirements of local nuclear energy producers and at the same time establishes a strong linkage between education and day-to-day nuclear power generation. As of 1997, seventeen pairs of university-utility partners existed. UMCP was never a member of that group of universities, but applied for the first time with a proposal to Baltimore Gas and Electric Company in January 1999 [1]. This proposal was generously granted by BG&E [2,3] in the form of a gift in the amount of $25,000 from BG&E's Corporate Contribution Program. Upon the arrival of a newly appointed Director of Administration in the Department of Materials and Nuclear Engineering, the BG&E check was deposited into the University's Maryland Foundation Fund. The receipt of the letter and the check enabled UMCP to apply for DOE's matching funds in the same amount by a proposal.

  10. A Report of the Nuclear Engineering Division Sessions at the 1971 ASEE Annual Conference

    Science.gov (United States)

    Eckley, Wayne; Nelson, George W.

    1972-01-01

    Summarizes the discussions at the conference under the topics, Objective Criteria for the Future" and Teaching Concepts Basic to Nuclear Engineering." Includes comments from personnel representing universities, industries, and government laboratories. (TS)

  11. Detection and reconstruction of error control codes for engineered and biological regulatory systems.

    Energy Technology Data Exchange (ETDEWEB)

    May, Elebeoba Eni; Rintoul, Mark Daniel; Johnston, Anna Marie; Pryor, Richard J.; Hart, William Eugene; Watson, Jean-Paul

    2003-10-01

    A fundamental challenge for all communication systems, engineered or living, is the problem of achieving efficient, secure, and error-free communication over noisy channels. Information theoretic principals have been used to develop effective coding theory algorithms to successfully transmit information in engineering systems. Living systems also successfully transmit biological information through genetic processes such as replication, transcription, and translation, where the genome of an organism is the contents of the transmission. Decoding of received bit streams is fairly straightforward when the channel encoding algorithms are efficient and known. If the encoding scheme is unknown or part of the data is missing or intercepted, how would one design a viable decoder for the received transmission? For such systems blind reconstruction of the encoding/decoding system would be a vital step in recovering the original message. Communication engineers may not frequently encounter this situation, but for computational biologists and biotechnologist this is an immediate challenge. The goal of this work is to develop methods for detecting and reconstructing the encoder/decoder system for engineered and biological data. Building on Sandia's strengths in discrete mathematics, algorithms, and communication theory, we use linear programming and will use evolutionary computing techniques to construct efficient algorithms for modeling the coding system for minimally errored engineered data stream and genomic regulatory DNA and RNA sequences. The objective for the initial phase of this project is to construct solid parallels between biological literature and fundamental elements of communication theory. In this light, the milestones for FY2003 were focused on defining genetic channel characteristics and providing an initial approximation for key parameters, including coding rate, memory length, and minimum distance values. A secondary objective addressed the question of

  12. Coding efficiency of AVS 2.0 for CBAC and CABAC engines

    Science.gov (United States)

    Cui, Jing; Choi, Youngkyu; Chae, Soo-Ik

    2015-12-01

    In this paper we compare the coding efficiency of AVS 2.0[1] for engines of the Context-based Binary Arithmetic Coding (CBAC)[2] in the AVS 2.0 and the Context-Adaptive Binary Arithmetic Coder (CABAC)[3] in the HEVC[4]. For fair comparison, the CABAC is embedded in the reference code RD10.1 because the CBAC is in the HEVC in our previous work[5]. The rate estimation table is employed only for RDOQ in the RD code. To reduce the computation complexity of the video encoder, therefore we modified the RD code so that the rate estimation table is employed for all RDO decision. Furthermore, we also simplify the complexity of rate estimation table by reducing the bit depth of its fractional part to 2 from 8. The simulation result shows that the CABAC has the BD-rate loss of about 0.7% compared to the CBAC. It seems that the CBAC is a little more efficient than that the CABAC in the AVS 2.0.

  13. Master on Nuclear Engineering and Applications (MINA): instrument of knowledge management in the nuclear sector; Master en Ingenieria Nuclear y Aplicaciones (MINA): instrumento de gestion del conocimiento en el sector nuclear espanol

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L. E.; Garcia Cuesta, J. C.; Falcon, S.; Casas, J. A.

    2013-03-01

    Knowledge Management in nuclear industry is indispensable to ensure excellence in performance and safety of nuclear installations. The Master on Nuclear Engineering and Applications (MINA) is a Spanish education venture which foundations and evolution have meant and adaptation to the European Education system and to the domestic and international changes occurred in the nuclear environment. This paper summarizes the most relevant aspects of such transformation, its motivation and the final outcome. Finally, it discusses the potential benefit of a closer collaboration among the existing national education ventures in the frame of Nuclear Engineering. (Author)

  14. Feasibility study of nuclear transmutation by negative muon capture reaction using the PHITS code

    Science.gov (United States)

    Abe, Shin-ichiro; Sato, Tatsuhiko

    2016-06-01

    Feasibility of nuclear transmutation of fission products in high-level radioactive waste by negative muon capture reaction is investigated using the Particle and Heave Ion Transport code System (PHITS). It is found that about 80 % of stopped negative muons contribute to transmute target nuclide into stable or short-lived nuclide in the case of 135Cs, which is one of the most important nuclide in the transmutation. The simulation result also indicates that the position of transmutation is controllable by changing the energy of incident negative muon. Based on our simulation, it takes approximately 8.5 × 108years to transmute 500 g of 135Cs by negative muon beam with the highest intensity currently available.

  15. Development of accident management technology and computer codes -A study for nuclear safety improvement-

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang Kyu; Jae, Moo Sung; Jo, Young Gyun; Park, Rae Jun; Kim, Jae Hwan; Ha, Jae Ju; Kang, Dae Il; Choi, Sun Young; Kim, Si Hwan [Korea Atomic Energy Res. Inst., Taejon (Korea, Republic of)

    1994-07-01

    We have surveyed new technologies and research results for the accident management of nuclear power plants. And, based on the concept of using the existing plant capabilities for accident management, both in-vessel and ex-vessel strategies were identified and analyzed. When assessing accident management strategies, their effectiveness, adverse effects, and their feasibility must be considered. We have developed a framework for assessing the strategies with these factors in mind. We have applied the developed framework to assessing the strategies, including the likelihood that the operator correctly diagnoses the situation and successfully implements the strategies. Finally, the cavity flooding strategy was assessed by applying it to the station blackout sequence, which have been identified as one of the major contributors to risk at the reference plant. The thermohydraulic analyses with sensitivity calculations have been performed using MAAP 4 computer code. (Author).

  16. Nonelastic nuclear reactions induced by light ions with the BRIEFF code

    CERN Document Server

    Duarte, H

    2010-01-01

    The intranuclear cascade (INC) code BRIC has been extended to compute nonelastic reactions induced by light ions on target nuclei. In our approach the nucleons of the incident light ion move freely inside the mean potential of the ion in its center-of-mass frame while the center-of-mass of the ion obeys to equations of motion dependant on the mean nuclear+Coulomb potential of the target nucleus. After transformation of the positions and momenta of the nucleons of the ion into the target nucleus frame, the collision term between the nucleons of the target and of the ion is computed taking into account the partial or total breakup of the ion. For reactions induced by low binding energy systems like deuteron, the Coulomb breakup of the ion at the surface of the target nucleus is an important feature. Preliminary results of nucleon production in light ion induced reactions are presented and discussed.

  17. Basal jawed vertebrate phylogeny inferred from multiple nuclear DNA-coded genes

    Directory of Open Access Journals (Sweden)

    Ishida Osamu

    2004-03-01

    Full Text Available Abstract Background Phylogenetic analyses of jawed vertebrates based on mitochondrial sequences often result in confusing inferences which are obviously inconsistent with generally accepted trees. In particular, in a hypothesis by Rasmussen and Arnason based on mitochondrial trees, cartilaginous fishes have a terminal position in a paraphyletic cluster of bony fishes. No previous analysis based on nuclear DNA-coded genes could significantly reject the mitochondrial trees of jawed vertebrates. Results We have cloned and sequenced seven nuclear DNA-coded genes from 13 vertebrate species. These sequences, together with sequences available from databases including 13 jawed vertebrates from eight major groups (cartilaginous fishes, bichir, chondrosteans, gar, bowfin, teleost fishes, lungfishes and tetrapods and an outgroup (a cyclostome and a lancelet, have been subjected to phylogenetic analyses based on the maximum likelihood method. Conclusion Cartilaginous fishes have been inferred to be basal to other jawed vertebrates, which is consistent with the generally accepted view. The minimum log-likelihood difference between the maximum likelihood tree and trees not supporting the basal position of cartilaginous fishes is 18.3 ± 13.1. The hypothesis by Rasmussen and Arnason has been significantly rejected with the minimum log-likelihood difference of 123 ± 23.3. Our tree has also shown that living holosteans, comprising bowfin and gar, form a monophyletic group which is the sister group to teleost fishes. This is consistent with a formerly prevalent view of vertebrate classification, although inconsistent with both of the current morphology-based and mitochondrial sequence-based trees. Furthermore, the bichir has been shown to be the basal ray-finned fish. Tetrapods and lungfish have formed a monophyletic cluster in the tree inferred from the concatenated alignment, being consistent with the currently prevalent view. It also remains possible that

  18. Calcined Waste Storage at the Idaho Nuclear Technology and Engineering Center

    Energy Technology Data Exchange (ETDEWEB)

    M. D. Staiger

    2007-06-01

    This report provides a quantitative inventory and composition (chemical and radioactivity) of calcined waste stored at the Idaho Nuclear Technology and Engineering Center. From December 1963 through May 2000, liquid radioactive wastes generated by spent nuclear fuel reprocessing were converted into a solid, granular form called calcine. This report also contains a description of the calcine storage bins.

  19. The application of the Monte-Carlo neutron transport code MCNP to a small "nuclear battery" system

    OpenAIRE

    Puigdellívol Sadurní, Roger

    2009-01-01

    The project consist in calculate the keff to a small nuclear battery. The code Monte- Carlo neutron transport code MCNP is used to calculate the keff. The calculations are done at the beginning of life to know the capacity of the core becomes critical in different conditions. These conditions are the study parameters that determine the criticality of the core. These parameters are the uranium enrichment, the coated particles (TRISO) packing factor and the size of the core. More...

  20. Reevaluation of the emergency planning zone for nuclear power plants in Taiwan using MACCS2 code.

    Science.gov (United States)

    Wu, Jay; Yang, Yung-Muh; Chen, Ing-Jane; Chen, Huan-Tong; Chuang, Keh-Shih

    2006-04-01

    According to government regulations, the emergency planning zone (EPZ) of a nuclear power plant (NPP) must be designated before operation and reevaluated every 5 years. Corresponding emergency response planning (ERP) has to be made in advance to guarantee that all necessary resources are available under accidental releases of radioisotope. In this study, the EPZ for each of the three operating NPPs, Chinshan, Kuosheng, and Maanshan, in Taiwan was reevaluated using the MELCOR Accident Consequence Code System 2 (MACCS2) developed by Sandia National Laboratory. Meteorological data around the nuclear power plant were collected during 2003. The source term data including inventory, sensible heat content, and timing duration, were based on previous PRA information of each plant. The effective dose equivalent and thyroid dose together with the related individual risk and societal risk were calculated. By comparing the results to the protective action guide and related safety criteria, 1.5, 1.5, and 4.5km were estimated for Chinshan, Kuosheng, and Maanshan NPPs, respectively. We suggest that a radius of 5.0km is a reasonably conservative value of EPZ for each of the three operating NPPs in Taiwan.

  1. Engineering Ethics In Islam: An Evaluative And Comparative Study Between Code Of Ethics Of Institution Of Engineers, Bangladesh (Ieb And Code Of Professional Conduct Of Board Of Engineers Malaysia (BEM

    Directory of Open Access Journals (Sweden)

    Muhammad Amanullah

    2012-01-01

    Full Text Available During the past two centuries or so a number of buildings and bridges had been structurally failed and collapsed all over the world. Some of these incidents caused a sizeable number of human casualties. For instance, collapse of Tay Bridge in 1879 killed at least sixty persons. Beside the problems related to their design and construction, probably the failure to follow engineering ethics properly was partially responsible for these incidents. Growing engineering professionalism during the nineteenth century gave rise to the development of a number of famous engineering societies, such as American Institute of Electrical Engineers (AIEE ( (1884, American Institute of Mining Engineers (AIME (1871, etc. On the other hand, responding to series of significant structural failures mentioned above, some engineering societies developed formal codes of ethics. Following these societies, engineers of Bangladesh (previous East Pakistan established Institution of Engineers, Bangladesh (IEB. Likewise, Malaysian engineers established Board of Engineers Malaysia (BEM. Both of these societies have their codes of ethics. Islam also has offered a number of ethics to be followed by the engineers. Analyzing the related verses of the Qur'an and ahadith of the Prophet (pbuh, this paper intends to highlight these Islamic ethics and in light of them, tries to evaluate the codes of ethics of these two societies and compare between them. The paper may conclude that although the codes of ethics of IEB and BEM are supported by Islamic ethics they require further modification.ABSTRAK - Sejak lebih kurang dua abad kebelakangan ini, banyak binaan yang gagal dari segi strukturnya dan juga jambatan yang runtuh di merata dunia. Sesetengah tragedi yang berlaku juga mengakibatkan kehilangan nyawa manusia. Contohnya, robohnya Jambatan Tay pada 1879 telah meragut nyawa lebih kurang enam puluh orang. Selain daripada masalah yang berkaitan dengan reka bentuk dan pembinaanya, mungkin

  2. A code for simulation of human failure events in nuclear power plants: SIMPROC

    Energy Technology Data Exchange (ETDEWEB)

    Gil, Jesus, E-mail: jesus.gil@indizen.co [Indizen Technologies S.L., Pablo Iglesias 2-3 oB-2, 28003 Madrid (Spain); Fernandez, Ivan, E-mail: ivan.fernandez@indizen.co [Indizen Technologies S.L., Pablo Iglesias 2-3 oB-2, 28003 Madrid (Spain); Murcia, Santiago, E-mail: santiago.murcia@indizen.co [Indizen Technologies S.L., Pablo Iglesias 2-3 oB-2, 28003 Madrid (Spain); Gomez, Javier, E-mail: javier.mundina@indizen.co [Indizen Technologies S.L., Pablo Iglesias 2-3 oB-2, 28003 Madrid (Spain); Marrao, Hugo, E-mail: hmarrao@indizen.co [Indizen Technologies S.L., Pablo Iglesias 2-3 oB-2, 28003 Madrid (Spain); Queral, Cesar, E-mail: cesar.queral@upm.e [ETSI Minas - Universidad Politecnica de Madrid, Alenza 4, 28003 Madrid (Spain); Exposito, Antonio, E-mail: antonio.exposito@upm.e [ETSI Minas - Universidad Politecnica de Madrid, Alenza 4, 28003 Madrid (Spain); Rodriguez, Gabriel, E-mail: gabriel.rodriguez.martin@upm.e [ETSI Minas - Universidad Politecnica de Madrid, Alenza 4, 28003 Madrid (Spain); Ibanez, Luisa, E-mail: luisa.ibanez@upm.e [ETSI Minas - Universidad Politecnica de Madrid, Alenza 4, 28003 Madrid (Spain); Hortal, Javier, E-mail: fjhr@csn.e [Consejo de Seguridad Nuclear, Justo Dorado 11, 28040 Madrid (Spain); Izquierdo, Jose M., E-mail: jmir@csn.e [Consejo de Seguridad Nuclear, Justo Dorado 11, 28040 Madrid (Spain); Sanchez, Miguel, E-mail: msp@csn.e [Consejo de Seguridad Nuclear, Justo Dorado 11, 28040 Madrid (Spain); Melendez, Enrique, E-mail: ema@csn.e [Consejo de Seguridad Nuclear, Justo Dorado 11, 28040 Madrid (Spain)

    2011-04-15

    Over the past years, many Nuclear Power Plant organizations have performed Probabilistic Safety Assessments to identify and understand key plant vulnerabilities. As part of enhancing the PSA quality, the Human Reliability Analysis is essential to make a realistic evaluation of safety and about the potential facility's weaknesses. Moreover, it has to be noted that HRA continues to be a large source of uncertainty in the PSAs. Within their current joint collaborative activities, Indizen, Universidad Politecnica de Madrid and Consejo de Seguridad Nuclear have developed the so-called SIMulator of PROCedures (SIMPROC), a tool aiming at simulate events related with human actions and able to interact with a plant simulation model. The tool helps the analyst to quantify the importance of human actions in the final plant state. Among others, the main goal of SIMPROC is to check the Emergency Operating Procedures being used by operating crew in order to lead the plant to a safe shutdown plant state. Currently SIMPROC is coupled with the SCAIS software package, but the tool is flexible enough to be linked to other plant simulation codes. SIMPROC-SCAIS applications are shown in the present article to illustrate the tool performance. The applications were developed in the framework of the Nuclear Energy Agency project on Safety Margin Assessment and Applications (SM2A). First an introductory example was performed to obtain the damage domain boundary of a selected sequence from a SBLOCA. Secondly, the damage domain area of a selected sequence from a loss of Component Cooling Water with a subsequent seal LOCA was calculated. SIMPROC simulates the corresponding human actions in both cases. The results achieved shown how the system can be adapted to a wide range of purposes such as Dynamic Event Tree delineation, Emergency Operating Procedures and damage domain search.

  3. Geoethics: what can we learn from existing bio-, ecological, and engineering ethics codes?

    Science.gov (United States)

    Kieffer, Susan W.; Palka, John

    2014-05-01

    Many scientific disciplines are concerned about ethics, and codes of ethics for these professions exist, generally through the professional scientific societies such as the American Geophysical Union (AGU), American Geological Institute (AGI), American Association of Petroleum Engineers (AAPE), National Society of Professional Engineers (NSPE), Ecological Society of America (ESA), and many others worldwide. These vary considerably in depth and specificity. In this poster, we review existing codes with the goal of extracting fundamentals that should/can be broadly applied to all geo-disciplines. Most of these codes elucidate a set of principles that cover practical issues such as avoiding conflict of interest, avoiding plagiarism, not permitting illegitimate use of intellectual products, enhancing the prestige of the profession, acknowledging an obligation to perform services only in areas of competence, issuing public statements only in an objective manner, holding paramount the welfare of the public, and in general conducting oneself honorably, responsibly, and lawfully. It is striking that, given that the work of these societies and their members is relevant to the future of the earth, few discuss in any detail ethical obligations regarding our relation to the planet itself. The AGU code, for example, only states that "Members have an ethical obligation to weigh the societal benefits of their research against the costs and risks to human and animal welfare and impacts on the environment and society." The NSPE and AGI codes go somewhat further: "Engineers are encouraged to adhere to the principles of sustainable development in order to protect the environment for future generations," and "Geoscientists should strive to protect our natural environment. They should understand and anticipate the environmental consequences of their work and should disclose the consequences of recommended actions. They should acknowledge that resource extraction and use are necessary

  4. Validation of a new library of nuclear constants of the WIMS code; Validacion de una nueva biblioteca de constantes nucleares del Codigo WIMS

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F. [Departamento de Experimentacion, Gerencia del Reactor, ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    1991-10-15

    The objective of the present work is to reproduce the experimental results of the thermal reference problems (benchmarks) TRX-1, TRX-2 and BAPL-1 to BAPL-3 with the WIMS code. It was proceeded in two stages, the first one consisted on using the original library of the code, while in the second one, a library that only contains the present elements in the benchmarks: H{sup 1}, O{sup 16}, Al{sup 27}, U{sup 235} and U{sup 238} was generated. To generate the present nuclear data in the WIMS library, it was used the ENDF/B-IV database and the Data processing system of Nuclear Data NJOY, the library was generated using the FIXER code. (Author)

  5. Validation of a new library of nuclear constants of the WIMS code; Validacion de una nueva biblioteca de constantes nucleares del Codigo WIMS

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F. [Departamento de Experimentacion, Gerencia del Reactor, ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    1991-10-15

    The objective of the present work is to reproduce the experimental results of the thermal reference problems (benchmarks) TRX-1, TRX-2 and BAPL-1 to BAPL-3 with the WIMS code. It was proceeded in two stages, the first one consisted on using the original library of the code, while in the second one, a library that only contains the present elements in the benchmarks: H{sup 1}, O{sup 16}, Al{sup 27}, U{sup 235} and U{sup 238} was generated. To generate the present nuclear data in the WIMS library, it was used the ENDF/B-IV database and the Data processing system of Nuclear Data NJOY, the library was generated using the FIXER code. (Author)

  6. EELQMS - the European quality management system for engine lubricants - the ATC Code of Practice; EELQMS - Das Europaeische Qualitaets-Management System fuer Motorenoele - der ATC Code of Practice

    Energy Technology Data Exchange (ETDEWEB)

    Raddatz, J.H.; Eberan-Eberhorst, C.G.A. von

    1998-01-01

    In 1995 the ATC developed a Code of Practice which, in conjunction with the ATIEL Code of Practice, represents the basis for the European Engine Lubricant Quality Management System (EELQMS). Compliance with the requirements of this system is a prerequisite for performance claims made by engine oil marketers regarding the European ACEA Engine Oil Sequences. TAD, the German section of the Technical Committee of Petroleum Additive Manufacturers in Europe (ATC), has prepared this presentation in order to promote the dialogue between the industries concerned and to provide information on EELQMS and the ATC Code of Practice to a broader audience. Key elements of the paper are: - What is EELQMS? - How does EELQMS work? - What is the role of the ATC Code of Practice in EELQMS? - What are the most important rules of the ATC Code of Practice? - What benefits do EELQMS and the ATC Code of Practice offer to the end-user? - What is the current status of EELQMS? We hope that this presentation will help to promote a better understanding and acceptance of EELQMS on a broad basis. (orig.) [Deutsch] Im Jahre 1995 hat der ATC eine Code of Practice entwickelt, der in Verbindung mit dem ATIEL Code of Practice die Grundlage des Europaeischen Qualitaets-Management-Systems fuer Motoroele (European Engine Lubricant Quality Management System=EELQMS) ist. Die Einhaltung der in diesem System spezifizierten Regeln ist Voraussetzung fuer die Erfuellung der ACEA-Richtlinien und der entsprechenden Performance-Aussagen nach den jeweiligen europaeischen ACEA-Motorenoelsequenzen. Zur Vertiefung des Dialogs zwischen den beteiligten Industrien und zur Verbreitung der Kenntnisse ueber EELQMS und den ATC Code of Practice hat die TAD, die deutsche nationale Organisation innerhalb des europaeischen Dachverbandes der Additivindustrie (ATC), folgende Praesentation ausgearbeitet. Wesentliche Elemente der Praesentation sind: - Was ist EELQMS? - Wie funktioniert EELQMS? - Welche Rolle spielt der ATC Code of

  7. A Programmatic and Engineering Approach to the Development of a Nuclear Thermal Rocket for Space Exploration

    Science.gov (United States)

    Bordelon, Wayne J., Jr.; Ballard, Rick O.; Gerrish, Harold P., Jr.

    2006-01-01

    With the announcement of the Vision for Space Exploration on January 14, 2004, there has been a renewed interest in nuclear thermal propulsion. Nuclear thermal propulsion is a leading candidate for in-space propulsion for human Mars missions; however, the cost to develop a nuclear thermal rocket engine system is uncertain. Key to determining the engine development cost will be the engine requirements, the technology used in the development and the development approach. The engine requirements and technology selection have not been defined and are awaiting definition of the Mars architecture and vehicle definitions. The paper discusses an engine development approach in light of top-level strategic questions and considerations for nuclear thermal propulsion and provides a suggested approach based on work conducted at the NASA Marshall Space Flight Center to support planning and requirements for the Prometheus Power and Propulsion Office. This work is intended to help support the development of a comprehensive strategy for nuclear thermal propulsion, to help reduce the uncertainty in the development cost estimate, and to help assess the potential value of and need for nuclear thermal propulsion for a human Mars mission.

  8. SEACC: the systems engineering and analysis computer code for small wind systems

    Energy Technology Data Exchange (ETDEWEB)

    Tu, P.K.C.; Kertesz, V.

    1983-03-01

    The systems engineering and analysis (SEA) computer program (code) evaluates complete horizontal-axis SWECS performance. Rotor power output as a function of wind speed and energy production at various wind regions are predicted by the code. Efficiencies of components such as gearbox, electric generators, rectifiers, electronic inverters, and batteries can be included in the evaluation process to reflect the complete system performance. Parametric studies can be carried out for blade design characteristics such as airfoil series, taper rate, twist degrees and pitch setting; and for geometry such as rotor radius, hub radius, number of blades, coning angle, rotor rpm, etc. Design tradeoffs can also be performed to optimize system configurations for constant rpm, constant tip speed ratio and rpm-specific rotors. SWECS energy supply as compared to the load demand for each hour of the day and during each session of the year can be assessed by the code if the diurnal wind and load distributions are known. Also available during each run of the code is blade aerodynamic loading information.

  9. PWR circuit contamination assessment tool. Use of OSCAR code for engineering studies at EDF

    Directory of Open Access Journals (Sweden)

    Benfarah Moez

    2016-01-01

    Full Text Available Normal operation of PWR generates corrosion and wear products in the primary circuit which are activated in the core and constitute the major source of the radiation field. In addition, cases of fuel failure and alpha emitter dissemination in the coolant system could represent a significant radiological risk. Radiation field and alpha risks are the main constraints to carry out maintenance and to handle effluents. To minimize these risks and constraints, it is essential to understand the behavior of corrosion products and actinides and to carry out the appropriate measurements in PWR circuits and loop experiments. As a matter of fact, it is more than necessary to develop and use a reactor contamination assessment code in order to take into account the chemical and physical mechanisms in different situations in operating reactors or at design stage. OSCAR code has actually been developed and used for this aim. It is presented in this paper, as well as its use in the engineering studies at EDF. To begin with, the code structure is described, including the physical, chemical and transport phenomena considered for the simulation of the mechanisms regarding PWR contamination. Then, the use of OSCAR is illustrated with two examples from our engineering studies. The first example of OSCAR engineering studies is linked to the behavior of the activated corrosion products. The selected example carefully explores the impact of the restart conditions following a reactor mid-cycle shutdown on circuit contamination. The second example of OSCAR use concerns fission products and disseminated fissile material behavior in the primary coolant. This example is a parametric study of the correlation between the quantity of disseminated fuel and the variation of Iodine 134 in the primary coolant.

  10. Draft fracture mechanics code case for American Society of Mechanical Engineers NUPACK rules

    Energy Technology Data Exchange (ETDEWEB)

    McConnell, P.; Sorenson, K. [Sandia National Labs., Albuquerque (United States); Nickell, R. [Applied Science and Technology, Poway (United States); Saegusa, T. [Central Research Inst. for Electric Power Industry, Abiko (Japan)

    2004-07-01

    The containment boundaries of most spent-fuel casks certified for use in the United States by the Nuclear Regulatory Commission are constructed with stainless steel, a material that is ductile in an engineering sense at all temperatures and for which, therefore, fracture mechanics principles are not relevant for the containment application. Ferritic materials may fail in a nonductile manner at sufficiently low temperatures, so fracture mechanics principles may be applied to preclude nonductile fracture. Because of the need to transport and store spent nuclear fuel safely in all types of climatic conditions, these vessels have regulatory lowest service temperatures that range down to -40 C (-40 F) for transport application. Such low service temperatures represent a severe challenge in terms of fracture toughness to many ferritic materials. Linear-elastic and elastic-plastic fracture mechanics principles provide a methodology for evaluating ferritic materials under such conditions.

  11. A preliminary neutronic evaluation of high temperature engineering test reactor using the SCALE6 code

    Science.gov (United States)

    Tanure, L. P. A. R.; Sousa, R. V.; Costa, D. F.; Cardoso, F.; Veloso, M. A. F.; Pereira, C.

    2014-02-01

    Neutronic parameters of some fourth generation nuclear reactors have been investigated at the Departamento de Engenharia Nuclear/UFMG. Previous studies show the possibility to increase the transmutation capabilities of these fourth generation systems to achieve significant reduction concerning transuranic elements in spent fuel. To validate the studies, a benchmark on core physics analysis, related to initial testing of the High Temperature Engineering Test Reactor and provided by International Atomic Energy Agency (IAEA) was simulated using the Standardized Computer Analysis for Licensing Evaluation (SCALE). The CSAS6/KENO-VI control sequence and the 44-group ENDF/B-V 0 cross-section neutron library were used to evaluate the keff (effective multiplication factor) and the result presents good agreement with experimental value.

  12. Appliance of software engineering in development of nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Y. W.; Kim, H. C.; Yun, C. [Chungnam National Univ., Taejon (Korea, Republic of); Kim, B. R. [KINS, Taejon (Korea, Republic of)

    1999-10-01

    Application of computer technology in nuclear power plant is also a necessary transformation as in other industry fields. But until now, application of software technology was not wide-spread because of its potential effect to safety in nuclear field. It is an urgent theme to develop evaluation guide and regulation techniques to guarantee safety, reliability and quality assurance. To meet these changes, techniques for development and operation should be enhanced to ensure the quality of software systems. In this study, we show the difference between waterfall model and software life-cycle needed in development of nuclear power plant and propose the consistent framework needed in development of instrumentation and control system of nuclear power plant.

  13. Application of flow network models of SINDA/FLUINT{sup TM} to a nuclear power plant system thermal hydraulic code

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Ji Bum [Institute for Advanced Engineering, Yongin (Korea, Republic of); Park, Jong Woon [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    In order to enhance the dynamic and interactive simulation capability of a system thermal hydraulic code for nuclear power plant, applicability of flow network models in SINDA/FLUINT{sup TM} has been tested by modeling feedwater system and coupling to DSNP which is one of a system thermal hydraulic simulation code for a pressurized heavy water reactor. The feedwater system is selected since it is one of the most important balance of plant systems with a potential to greatly affect the behavior of nuclear steam supply system. The flow network model of this feedwater system consists of condenser, condensate pumps, low and high pressure heaters, deaerator, feedwater pumps, and control valves. This complicated flow network is modeled and coupled to DSNP and it is tested for several normal and abnormal transient conditions such turbine load maneuvering, turbine trip, and loss of class IV power. The results show reasonable behavior of the coupled code and also gives a good dynamic and interactive simulation capabilities for the several mild transient conditions. It has been found that coupling system thermal hydraulic code with a flow network code is a proper way of upgrading simulation capability of DSNP to mature nuclear plant analyzer (NPA). 5 refs., 10 figs. (Author)

  14. Neutron Transport and Nuclear Burnup Analysis for the Laser Inertial Confinement Fusion-Fission Energy (LIFE) Engine

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, K J; Latkowski, J F; Abbott, R P; Boyd, J K; Powers, J J; Seifried, J E

    2008-10-24

    Lawrence Livermore National Laboratory is currently developing a hybrid fusion-fission nuclear energy system, called LIFE, to generate power and burn nuclear waste. We utilize inertial confinement fusion to drive a subcritical fission blanket surrounding the fusion chamber. It is composed of TRISO-based fuel cooled by the molten salt flibe. Low-yield (37.5 MJ) targets and a repetition rate of 13.3 Hz produce a 500 MW fusion source that is coupled to the subcritical blanket, which provides an additional gain of 4-8, depending on the fuel. In the present work, we describe the neutron transport and nuclear burnup analysis. We utilize standard analysis tools including, the Monte Carlo N-Particle (MCNP) transport code, ORIGEN2 and Monteburns to perform the nuclear design. These analyses focus primarily on a fuel composed of depleted uranium not requiring chemical reprocessing or enrichment. However, other fuels such as weapons grade plutonium and highly-enriched uranium are also under consideration. In addition, we have developed a methodology using {sup 6}Li as a burnable poison to replace the tritium burned in the fusion targets and to maintain constant power over the lifetime of the engine. The results from depleted uranium analyses suggest up to 99% burnup of actinides is attainable while maintaining full power at 2GW for more than five decades.

  15. Considerations on fatigue stress range calculations in nuclear power plants using on-line monitoring systems and the ASME Code

    Energy Technology Data Exchange (ETDEWEB)

    Cicero, R., E-mail: ciceror@unican.e [INESCO INGENIEROS S.L., Santander (Spain); Departamento de Ciencia e Ingenieria del Terreno y los Materiales, Universidad de Cantabria, Santander (Spain); Cicero, S. [Departamento de Ciencia e Ingenieria del Terreno y los Materiales, Universidad de Cantabria, Santander (Spain); Gorrochategui, I. [Centro Tecnologico de Componentes, Santander (Spain); Lacalle, R. [INESCO INGENIEROS S.L., Santander (Spain); Departamento de Ciencia e Ingenieria del Terreno y los Materiales, Universidad de Cantabria, Santander (Spain)

    2010-01-15

    Nuclear power plants are generally designed and inspected according to the ASME Code. This code indicates stress intensity (S{sub INT}) as the parameter to be used in the stress analysis of components. One of the particularities of S{sub INT} is that it always takes positive values, independently of the nature of the stress (tensile or compressive). This circumstance is relevant in the Fatigue Monitoring Systems used in nuclear power plants, due to the manner in which the different variable stresses are combined in order to obtain the final total stress range. This paper describes some situations derived from the application of the ASME Code, shows different ways of dealing with them and illustrates their influence on the evaluation of the fatigue usage factor through a case study.

  16. Spent nuclear fuel project systems engineering management plan

    Energy Technology Data Exchange (ETDEWEB)

    Womack, J.C., Westinghouse Hanford

    1996-07-19

    The purpose of this document is to describe the systems engineering approach and methods that will be integrated with established WHC engineering practices. The methodology promotes and ensures sound management of the SNF Project. The scope of the document encompasses the efforts needed to manage the WHC implementation of systems engineering on the SNF Project including risk management process, design authority/design agent concept, and documentation responsibilities. This implementation applies to, and is tailored to the needs of the SNF Project and all its Subprojects, including all current and future Subprojects.

  17. Test results of a 40 kW Stirling engine and comparison with the NASA-Lewis computer code predictions

    Science.gov (United States)

    Allen, D.; Cairelli, J.

    1985-01-01

    A Stirling engine was tested without auxiliaries at NASA-Lewis. Three different regenerator configurations were tested with hydrogen. The test objectives were (1) to obtain steady-state and dynamic engine data, including indicated power, for validation of an existing computer model for this engine; and (2) to evaluate structurally the use of silicon carbide regenerators. This paper presents comparisons of the measured brake performance, indicated mean effective pressure, and cyclic pressure variations with those predicted by the code. The measured data tended to be lower than the computer code predictions. The silicon carbide foam regenerators appear to be structurally suitable, but the foam matrix tested severely reduced performance.

  18. [Occupational medicine in nuclear industry and power engineering].

    Science.gov (United States)

    Gus'kova, A K

    2004-01-01

    The author analysed results of medical service in atomic industry and power engineering over 50 years. Those results are beneficial for management in occupational medicine for any new complicated and potentially dangerous technology and activity.

  19. A Hydrogen Containment Process for Nuclear Thermal Engine Ground testing

    Science.gov (United States)

    Wang, Ten-See; Stewart, Eric; Canabal, Francisco

    2016-01-01

    The objective of this study is to propose a new total hydrogen containment process to enable the testing required for NTP engine development. This H2 removal process comprises of two unit operations: an oxygen-rich burner and a shell-and-tube type of heat exchanger. This new process is demonstrated by simulation of the steady state operation of the engine firing at nominal conditions.

  20. Voluntary Code of Conduct for Nuclear Export Control by Private Sector

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Seong Youn; Shin, Dong Hoon [Korea Institute of Nuclear Non-Prolferation and Control Daejeon (Korea, Republic of)

    2011-10-15

    The nuclear renaissance offered a good opportunity to overcome energy crisis that we are bound to confront in the near future. At the opposite end, however, our world might be exposed to a new threat which could be resulted from the boost of nuclear power. The expansion of nuclear power plants worldwide inevitably makes the countries involve more transaction in nuclear items, which could lead to increase in nuclear proliferation risk. The international efforts to prevent proliferation attempts have been mainly being focused on limitation of nuclear sensitive technology and nuclear material. Considering the governments of major nuclear suppliers' key roles in the stage, it should be notable that a new initiative by private nuclear companies has been raised for the purpose of strengthening nuclear export control

  1. Manipulation of dynamic nuclear spin polarization in single quantum dots by photonic environment engineering

    Science.gov (United States)

    Fong, C. F.; Ota, Y.; Iwamoto, S.; Arakawa, Y.

    2017-06-01

    Optically induced dynamic nuclear spin polarization (DNP) in a semiconductor quantum dot (QD) requires many cycles of excitation of spin polarized carriers and carrier recombination. As such, the radiative lifetime of the exciton containing the electron becomes one of the limiting factors of DNP. In principle, changing the radiative lifetime of the exciton will affect DNP and thus the nuclear spin polarization. Here, we demonstrate the manipulation of DNP in single QDs through the engineering of the photonic environment using two-dimensional photonic crystals. We find that the achievable degree of nuclear spin polarization can be controlled through the modification of exciton radiative lifetime. Our results show the promise of achieving a higher degree of nuclear spin polarization via photonic environment engineering, with implications on spin-based quantum information processing.

  2. Simulation of a nuclear densimeter using the Monte Carlo MCNP-4C code; Simulacao de um densimetro nuclear utilizando o codigo Monte Carlo MCNP-4C

    Energy Technology Data Exchange (ETDEWEB)

    Penna, Rodrigo [UNI-BH, Belo Horizonte, MG (Brazil). Dept. de Ciencias Biologicas, Ambientais e da Saude (DCBAS/DCET); Silva, Clemente Jose Gusmao Carneiro da [Universidade Estadual de Santa Cruz, UESC, Ilheus, BA (Brazil); Gomes, Paulo Mauricio Costa [Universidade FUMEC, Belo Horizonte, MG (Brazil)

    2008-07-01

    Viability of building a nuclear wood densimeter based on low energy photons Compton scattering was done using Monte Carlo code (MCNP- 4C). It is simulated a collimated 60 keV beam of gamma rays emitted by {sup 241}Am source reaching wood blocks. Backscattered radiation by these blocks was calculated. Photons scattered were correlated with blocks of different wood densities. Results showed a linear relationship on wood density and scattered photons, therefore the viability of this wood densimeter. (author)

  3. An overview of the activities of the OECD/NEA Task Force on adapting computer codes in nuclear applications to parallel architectures

    Energy Technology Data Exchange (ETDEWEB)

    Kirk, B.L. [Oak Ridge National Lab., TN (United States); Sartori, E. [OCDE/OECD NEA Data Bank, Issy-les-Moulineaux (France); Viedma, L.G. de [Consejo de Seguridad Nuclear, Madrid (Spain)

    1997-06-01

    Subsequent to the introduction of High Performance Computing in the developed countries, the Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA) created the Task Force on Adapting Computer Codes in Nuclear Applications to Parallel Architectures (under the guidance of the Nuclear Science Committee`s Working Party on Advanced Computing) to study the growth area in supercomputing and its applicability to the nuclear community`s computer codes. The result has been four years of investigation for the Task Force in different subject fields - deterministic and Monte Carlo radiation transport, computational mechanics and fluid dynamics, nuclear safety, atmospheric models and waste management.

  4. Computer codes in nuclear safety, radiation transport and dosimetry; Les codes de calcul en radioprotection, radiophysique et dosimetrie

    Energy Technology Data Exchange (ETDEWEB)

    Bordy, J.M.; Kodeli, I.; Menard, St.; Bouchet, J.L.; Renard, F.; Martin, E.; Blazy, L.; Voros, S.; Bochud, F.; Laedermann, J.P.; Beaugelin, K.; Makovicka, L.; Quiot, A.; Vermeersch, F.; Roche, H.; Perrin, M.C.; Laye, F.; Bardies, M.; Struelens, L.; Vanhavere, F.; Gschwind, R.; Fernandez, F.; Quesne, B.; Fritsch, P.; Lamart, St.; Crovisier, Ph.; Leservot, A.; Antoni, R.; Huet, Ch.; Thiam, Ch.; Donadille, L.; Monfort, M.; Diop, Ch.; Ricard, M

    2006-07-01

    The purpose of this conference was to describe the present state of computer codes dedicated to radiation transport or radiation source assessment or dosimetry. The presentations have been parted into 2 sessions: 1) methodology and 2) uses in industrial or medical or research domains. It appears that 2 different calculation strategies are prevailing, both are based on preliminary Monte-Carlo calculations with data storage. First, quick simulations made from a database of particle histories built though a previous Monte-Carlo simulation and secondly, a neuronal approach involving a learning platform generated through a previous Monte-Carlo simulation. This document gathers the slides of the presentations.

  5. Engineering the Genetic Code in Cells and Animals: Biological Considerations and Impacts.

    Science.gov (United States)

    Wang, Lei

    2017-10-06

    Expansion of the genetic code allows unnatural amino acids (Uaas) to be site-specifically incorporated into proteins in live biological systems, thus enabling novel properties selectively introduced into target proteins in vivo for basic biological studies and for engineering of novel biological functions. Orthogonal components including tRNA and aminoacyl-tRNA synthetase (aaRS) are expressed in live cells to decode a unique codon (often the amber stop codon UAG) as the desired Uaa. Initially developed in E. coli, this methodology has now been expanded in multiple eukaryotic cells and animals. In this Account, we focus on addressing various biological challenges for rewriting the genetic code, describing impacts of code expansion on cell physiology and discussing implications for fundamental studies of code evolution. Specifically, a general method using the type-3 polymerase III promoter was developed to efficiently express prokaryotic tRNAs as orthogonal tRNAs and a transfer strategy was devised to generate Uaa-specific aaRS for use in eukaryotic cells and animals. The aaRSs have been found to be highly amenable for engineering substrate specificity toward Uaas that are structurally far deviating from the native amino acid, dramatically increasing the stereochemical diversity of Uaas accessible. Preparation of the Uaa in ester or dipeptide format markedly increases the bioavailability of Uaas to cells and animals. Nonsense-mediated mRNA decay (NMD), an mRNA surveillance mechanism of eukaryotic cells, degrades mRNA containing a premature stop codon. Inhibition of NMD increases Uaa incorporation efficiency in yeast and Caenorhabditis elegans. In bacteria, release factor one (RF1) competes with the orthogonal tRNA for the amber stop codon to terminate protein translation, leading to low Uaa incorporation efficiency. Contradictory to the paradigm that RF1 is essential, it is discovered that RF1 is actually nonessential in E. coli. Knockout of RF1 dramatically

  6. Study on fault diagnosis method for nuclear power plant based on hadamard error-correcting output code

    Science.gov (United States)

    Mu, Y.; Sheng, G. M.; Sun, P. N.

    2017-05-01

    The technology of real-time fault diagnosis for nuclear power plants(NPP) has great significance to improve the safety and economy of reactor. The failure samples of nuclear power plants are difficult to obtain, and support vector machine is an effective algorithm for small sample problem. NPP is a very complex system, so in fact the type of NPP failure may occur very much. ECOC is constructed by the Hadamard error correction code, and the decoding method is Hamming distance method. The base models are established by lib-SVM algorithm. The result shows that this method can diagnose the faults of the NPP effectively.

  7. Gas core nuclear thermal rocket engine research and development in the former USSR

    Energy Technology Data Exchange (ETDEWEB)

    Koehlinger, M.W.; Bennett, R.G.; Motloch, C.G. [eds.; Gurfink, M.M.

    1992-09-01

    Beginning in 1957 and continuing into the mid 1970s, the USSR conducted an extensive investigation into the use of both solid and gas core nuclear thermal rocket engines for space missions. During this time the scientific and engineering. problems associated with the development of a solid core engine were resolved. At the same time research was undertaken on a gas core engine, and some of the basic engineering problems associated with the concept were investigated. At the conclusion of the program, the basic principles of the solid core concept were established. However, a prototype solid core engine was not built because no established mission required such an engine. For the gas core concept, some of the basic physical processes involved were studied both theoretically and experimentally. However, no simple method of conducting proof-of-principle tests in a neutron flux was devised. This report focuses primarily on the development of the. gas core concept in the former USSR. A variety of gas core engine system parameters and designs are presented, along with a summary discussion of the basic physical principles and limitations involved in their design. The parallel development of the solid core concept is briefly described to provide an overall perspective of the magnitude of the nuclear thermal propulsion program and a technical comparison with the gas core concept.

  8. Nuclear Deterrence in the 21st Century: The Role of Science and Engineering

    Energy Technology Data Exchange (ETDEWEB)

    Martz, Joseph C [Los Alamos National Laboratory; Ventura, Jonathan S [Los Alamos National Laboratory

    2008-01-01

    Twenty-first century security challenges are multi-polar and asymmetric. A few nations have substantial nuclear arsenals and active nuclear weapons programs that still threaten vital US national security directly or by supporting proliferation. Maintaining a credible US nuclear deterrent and containing further proliferation will continue to be critical to US national security. Overlaid against this security backdrop, the rising worldwide population and its effects on global climate, food, and energy resources are greatly complicating the degree and number of security challenges before policy makers.This new paradigm requires new ways to assure allies that the United States remains a trusted security partner and to deter potential adversaries from aggressive actions that threaten global stability. Every U.S. President since Truman has affirmed the role of nuclear weapons as a supreme deterrent and protector of last resort of U.S. national security interests. Recently, President Bush called for a nuclear deterrent consistent with the 'lowest number of nuclear weapons' that still protects U.S. interests. How can this be achieved? And how can we continue on a path of nuclear reductions while retaining the security benefits of nuclear deterrence? Science and engineering have a key role to play in a potential new paradigm for nuclear deterrence, a concept known as 'capability-based deterrence.'

  9. High performance computing in Nuclear Engineering Institute, IEN, Rio de Janeiro, Brazil; Computacao de alto desempenho no Instituto de Engenharia Nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Sampaio, Paulo A.B. de; Lapa, Celso M.F.; Jospin, Reinaldo J.; Pereira, Claudio M.N.A.; Moreira, Maria de L.; Lapa, Nelbia da Silva; Nery, Domingos E. de Sa; Mol, Antonio C. [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil)

    2000-07-01

    Recently, advanced computational methods have been used for resolution of nuclear engineering problems. Examples of this are the utilization of computational simulation methods in fluid dynamics for thermal stratification problems analysis in PWR and the utilization of genetic algorithms for nuclear reactor core optimization, as well as, preventive maintenance scheduling optimization. The objective of this article is presents to national nuclear community the project of Parallel Computing Laboratory of IEN, and also discuss its aspects related to parallel processing philosophy and the new possibilities for nuclear engineering applications.

  10. Research on the improvement of nuclear safety -Development of a nuclear power plant system analysis code TASS (Transient and setpoint simulation)

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Suk Koo; Jang, Won Pyo; Kim, Heui Chul; Kim, Kyung Doo; Lee, Sung Jae; Hah, Kyooi Suk; Song, Soon Jah; Um, Kil Sub; Yoon, Han Yung; Kim, Doo Il; Yoo, Hyung Keun; Choi, Jae Don; Lee, Byung Il; Kim, Jung Jin [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-07-01

    During the third year of the project the development of TASS 1.0 code has been completed and validated its capability in applying for the licensing transient analyses of the Westinghouse and CE type operating reactors as well as the PWR reactors under construction in Korea. The validation of the TASS 1.0 code has been achieved through the comparison calculations of the YGN-3/4 FSAR transients, Kori-3 loss of AC power transient, plant data, Kori-4 load rejection and YGN-3 startup test data as well as the BETHSY loop steam generator tube rupture test data. TASS 1.0 calculation agrees well with the best estimate RELAP5/MOD 3.1 calculation for the YGN-3/4 FASR transients and shows its capability in simulating plant transient and startup data as well as the thermal hydraulic transient test data. Topical reports on TASS 1.0 code have been prepared and will be submitted to Korea Institute of Nuclear Safety for its licensing application to Westinghouse and CE type PWR transient analyses. The development of TASS 2.0 code has been head started in this year to timely utilize the TASS 2.0 code for the KNGR design certification. 65 figs, 30 tabs, 44 refs. (Author).

  11. Radiological effluents released from nuclear rocket and ramjet engine tests at the Nevada Test Site 1959 through 1969: Fact Book

    Energy Technology Data Exchange (ETDEWEB)

    Friesen, H.N.

    1995-06-01

    Nuclear rocket and ramjet engine tests were conducted on the Nevada Test Site (NTS) in Area 25 and Area 26, about 80 miles northwest of Las Vegas, Nevada, from July 1959 through September 1969. This document presents a brief history of the nuclear rocket engine tests, information on the off-site radiological monitoring, and descriptions of the tests.

  12. Effluent Scrubbing of Engine Exhaust of a Nuclear Thermal Propulsion Engine Project

    Data.gov (United States)

    National Aeronautics and Space Administration — This project identified specific knowledge and expertise in radioactive hydrogen effluent filter technology, so that internal resources on NTP engine exhaust...

  13. Systematic Approach to Training for System Engineers in Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kwak, Jeong-keun [Korea Hydro and Nuclear Power Co., Ulsan (Korea, Republic of)

    2015-10-15

    In my paper, comprehensive preparations, tangible applications, and final establishments of training for system engineers are described using practical materials in KHNP. The purpose of this paper is to formulate SAT based training in KHNP, especially for system engineers. Hence, to achieve this goal, over one year study was performed considering voluminous materials and working experiences. Through the process, SAT based training package for system engineers was finished, in the end. In terms of training in NPPs, SAT methodology is the unwavering trend in South Korea since NPPs export to UAE. Therefore, materialization of SAT based training for system engineers from the origin of SAT to the finalization of SAT should not be overlooked. A variety of accident preventive approaches have been adopted since the first commercial NPP operation in Calder Hall, United Kingdom. Among diverse event preventive ways, training has played an important role for the improvement of NPPs reliability and safety. This is reason why nuclear industry in every country has established and maintained own training institutes and methods. Since the Three Mile Island (TMI) accident, United States Nuclear Regulatory Commission (USNRC) recommended many betterment plans to US nuclear industry for the elevation of NPPs safety. In the suggested considerations, systematic approach to training, so called SAT appeared in the world. Basically, SAT is composed of five stages, what is called ADDIE. Hence, through ADDIE process, holistic and trustworthy training could be realized in the actual NPPs operation and maintenance. For this reason, SAT is the representative training methodology in the US nuclear business.

  14. Enhancement of Teaching and Learning of the Fundamentals of Nuclear Engineering Using Multimedia Courseware.

    Science.gov (United States)

    Keyvan, Shahla A.; Pickard, Rodney; Song, Xiaolong

    1997-01-01

    Computer-aided instruction incorporating interactive multimedia and network technologies can boost teaching effectiveness and student learning. This article describes the development and implementation of network server-based interactive multimedia courseware for a fundamental course in nuclear engineering. A student survey determined that 80% of…

  15. Nuclear Engineering Enrollment and Degree Survey: Enrollments - Fall 1972; Degrees Granted - July 1965-June 1972.

    Science.gov (United States)

    Chewning, June S.

    The Atomic Energy Commission's survey of nuclear engineering degrees granted during the 1971-72 academic year shows a continuing increase in bachelor's recipients, a slight increase in the number of master's, but a continuing decline in new Ph.D.'s. If the present rate of decline persists, by 1974 the number of new Ph.D.'s in the field will be…

  16. "PROCESS": a systems code for fusion power plants - Part 2:Engineering

    CERN Document Server

    Kovari, M; Harrington, C; Kembleton, R; Knight, P; Lux, H; Morris, J

    2016-01-01

    PROCESS is a reactor systems code - it assesses the engineering and economic viability of a hypothetical fusion power station using simple models of all parts of a reactor system. PROCESS allows the user to choose which constraints to impose and which to ignore, so when evaluating the results it is vital to study the list of constraints used. New algorithms submitted by collaborators can be incorporated - for example safety, first wall erosion, and fatigue life will be crucial and are not yet taken into account. This paper describes algorithms relating to the engineering aspects of the plant. The toroidal field (TF) coils and the central solenoid are assumed by default to be wound from niobium-tin superconductor with the same properties as the ITER conductors. The winding temperature and induced voltage during a quench provide a limit on the current density in the TF coils. Upper limits are placed on the stresses in the structural materials of the TF coil, using a simple two-layer model of the inboard leg of ...

  17. “PROCESS”: A systems code for fusion power plants – Part 2: Engineering

    Energy Technology Data Exchange (ETDEWEB)

    Kovari, M., E-mail: michael.kovari@ccfe.ac.uk; Fox, F.; Harrington, C.; Kembleton, R.; Knight, P.; Lux, H.; Morris, J.

    2016-03-15

    Highlights: • PROCESS is an optimising systems code for fusion reactors. • It allows the user to choose which constraints to impose and which to ignore. • Multiple constraints greatly restrict the parameter space of the optimised model. • For example, when coil current is increased greatly, major radius hardly changes. - Abstract: PROCESS is a reactor systems code – it assesses the engineering and economic viability of a hypothetical fusion power station using simple models of all parts of a reactor system. PROCESS allows the user to choose which constraints to impose and which to ignore, so when evaluating the results it is vital to study the list of constraints used. New algorithms submitted by collaborators can be incorporated – for example safety, first wall erosion, and fatigue life will be crucial and are not yet taken into account. This paper describes algorithms relating to the engineering aspects of the plant. The toroidal field (TF) coils and the central solenoid are assumed by default to be wound from niobium-tin superconductor with the same properties as the ITER conductors. The winding temperature and induced voltage during a quench provide a limit on the current density in the TF coils. Upper limits are placed on the stresses in the structural materials of the TF coil, using a simple two-layer model of the inboard leg of the coil. The thermal efficiency of the plant can be estimated using the maximum coolant temperature, and the capacity factor is derived from estimates of the planned and unplanned downtime, and the duty cycle if the reactor is pulsed. An example of a pulsed power plant is given. The need for a large central solenoid to induce most of the plasma current, and physics assumptions that are conservative compared to some other studies, result in a large machine, with a cryostat 36 m in diameter. Multiple constraints, working together, restrict the parameter space of the optimised model. For example, even when the ratio of

  18. Applications of nuclear magnetic resonance imaging in process engineering

    Science.gov (United States)

    Gladden, Lynn F.; Alexander, Paul

    1996-03-01

    During the past decade, the application of nuclear magnetic resonance (NMR) imaging techniques to problems of relevance to the process industries has been identified. The particular strengths of NMR techniques are their ability to distinguish between different chemical species and to yield information simultaneously on the structure, concentration distribution and flow processes occurring within a given process unit. In this paper, examples of specific applications in the areas of materials and food processing, transport in reactors and two-phase flow are discussed. One specific study, that of the internal structure of a packed column, is considered in detail. This example is reported to illustrate the extent of new, quantitative information of generic importance to many processing operations that can be obtained using NMR imaging in combination with image analysis.

  19. Environmental Degradation of Materials for Nuclear Waste Repositories Engineered Barriers

    Energy Technology Data Exchange (ETDEWEB)

    Rebak, R B

    2006-12-24

    Several countries are considering geological repositories for the storage of nuclear waste. Most of the environments for these repositories will be reducing in nature, except for the repository in the US, which is going to be oxidizing. For the reducing repositories, alloys such as carbon steel, copper, stainless steels and titanium are being evaluated. For the repository in the US, some of the most corrosion resistant commercially available alloys are being investigated. This paper presents a summary of the behavior of the different materials under consideration for the repositories and the current understanding of the degradation modes of the proposed alloys in ground water environments from the point of view of general corrosion, localized corrosion and environmentally assisted cracking.

  20. Groundwater Waves in a Coastal Fractured Aquifer of the Third Phase Qinshan Nuclear Power Engineering Field

    Institute of Scientific and Technical Information of China (English)

    ZHOU Nian-qing; TANG Yi-qun; TANG He-ping

    2005-01-01

    Tidal fluctuations of Hangzhou Bay produce progressive pressure waves in adjacent field fractured aquifers, as the pressure waves propagate, groundwater levels and hydraulic gradients continuously fluctuate. The effect of tidal fluctuations on groundwater flow can be determined using the mean hydraulic gradient that can be calculated by comparing mean ground and surface water elevations. Tidal fluctuation is shown to affect the piezometer readings taken in a nearshore fractured aquifer around the nuclear power engineering field. Continuous monitoring of a network of seven piezometers provided relations between the tidal cycle and the piezometer readings. The relations can be expressed in times of a time and amplitude scaling factor. The time lag and the tidal effi ciency factor and wavelength are calculated using these parameters. It provides significant scientific basis to prevent tide and groundwater for the nuclear power engineering construction and safety run of nuclear power station in the future.

  1. Engineering factors influencing Corbicula fouling in nuclear-service water systems

    Energy Technology Data Exchange (ETDEWEB)

    Henager, C.H.; Johnson, K.I.; Page, T.L.

    1983-06-01

    Corbicula fouling is a continuing problem in nuclear-service water systems. More knowledge of biological and engineering factors is needed to develop effective detection and control methods. A data base on Corbicula fouling was compiled from nuclear and non-nuclear power stations and industries using raw water. This data base was used in an analysis to identify systems and components which are conducive to fouling by Corbicula. Bounds on several engineering parameters such as velocity and temperature which support Corbicula growth are given. Service water systems found in BWR and PWR reactors are listed and those that show fouling are identified. Possible safety implications of Corbicula fouling are discussed for specific service water systems. Several effective control methods in current use include backflushing with heated water, centrifugal strainers, and continuous chlorination during spawning seasons.

  2. Recommendations to the NRC on human engineering guidelines for nuclear power plant maintainability

    Energy Technology Data Exchange (ETDEWEB)

    Badalamente, R.V.; Fecht, B.A.; Blahnik, D.E.; Eklund, J.D.; Hartley, C.S.

    1986-03-01

    This document contains human engineering guidelines which can enhance the maintainability of nuclear power plants. The guidelines have been derived from general human engineering design principles, criteria, and data. The guidelines may be applied to existing plants as well as to plants under construction. They apply to nuclear power plant systems, equipment and facilities, as well as to maintenance tools and equipment. The guidelines are grouped into seven categories: accessibility and workspace, physical environment, loads and forces, maintenance facilities, maintenance tools and equipment, operating equipment design, and information needs. Each chapter of the document details specific maintainability problems encountered at nuclear power plants, the safety impact of these problems, and the specific maintainability design guidelines whose application can serve to avoid these problems in new or existing plants.

  3. Nuclear and plastid genetic engineering of plants: comparison of opportunities and challenges.

    Science.gov (United States)

    Meyers, Benjamin; Zaltsman, Adi; Lacroix, Benoît; Kozlovsky, Stanislav V; Krichevsky, Alexander

    2010-01-01

    Plant genetic engineering is one of the key technologies for crop improvement as well as an emerging approach for producing recombinant proteins in plants. Both plant nuclear and plastid genomes can be genetically modified, yet fundamental functional differences between the eukaryotic genome of the plant cell nucleus and the prokaryotic-like genome of the plastid will have an impact on key characteristics of the resulting transgenic organism. So, which genome, nuclear or plastid, to transform for the desired transgenic phenotype? In this review we compare the advantages and drawbacks of engineering plant nuclear and plastid genomes to generate transgenic plants with the traits of interest, and evaluate the pros and cons of their use for different biotechnology and basic research applications, ranging from generation of commercial crops with valuable new phenotypes to 'bioreactor' plants for large-scale production of recombinant proteins to research model plants expressing various reporter proteins.

  4. Design of a requirements system for decommissioning of a nuclear power plant based on systems engineering

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hee Seong; Park, Seung Kook; Jin, Hyung Gon; Song, Chan Ho; Choi, Jong won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The nuclear industry has required an advanced system that can manage decommissioning information ever since the Korean government decide to decommission the Gori No.1 nuclear power plant. The D and D division at KAERI has been developing a system that can secure the reliability and sustainability of the decommissioning project based on the engineering system of the KRR-2 (Korean Research Reactor-2). To establish a decommissioning information system, a WBS that needs to be managed for the decommissioning of an NPP has been extracted, and requirements management research composed of system engineering technology has progressed. This paper propose a new type of system based on systems engineering technology. Even though a decommissioning engineering system was developed through the KRR-2, we are now developing an advanced decommissioning information system because it is not easy to apply this system to a commercial nuclear power plant. An NPP decommissioning is a project requiring a high degree of safety and economic feasibility. Therefore, we have to use a systematic project management at the initial phase of the decommissioning. An advanced system can manage the decommissioning information from preparation to remediation by applying a previous system to the systems engineering technology that has been widely used in large-scale government projects. The first phase of the system has progressed the requirements needed for a decommissioning project for a full life cycle. The defined requirements will be used in various types of documents during the decommissioning preparation phase.

  5. Calcine Waste Storage at the Idaho Nuclear Technology and Engineering Center

    Energy Technology Data Exchange (ETDEWEB)

    Staiger, Merle Daniel; M. C. Swenson

    2005-01-01

    This report documents an inventory of calcined waste produced at the Idaho Nuclear Technology and Engineering Center during the period from December 1963 to May 2000. The report was prepared based on calciner runs, operation of the calcined solids storage facilities, and miscellaneous operational information that establishes the range of chemical compositions of calcined waste stored at Idaho Nuclear Technology and Engineering Center. The report will be used to support obtaining permits for the calcined solids storage facilities, possible treatment of the calcined waste at the Idaho National Engineering and Environmental Laboratory, and to ship the waste to an off-site facility including a geologic repository. The information in this report was compiled from calciner operating data, waste solution analyses and volumes calcined, calciner operating schedules, calcine temperature monitoring records, and facility design of the calcined solids storage facilities. A compact disk copy of this report is provided to facilitate future data manipulations and analysis.

  6. Test results of a 40-kW Stirling engine and comparison with the NASA Lewis computer code predictions

    Science.gov (United States)

    Allen, David J.; Cairelli, James E.

    1988-01-01

    A Stirling engine was tested without auxiliaries at Nasa-Lewis. Three different regenerator configurations were tested with hydrogen. The test objectives were: (1) to obtain steady-state and dynamic engine data, including indicated power, for validation of an existing computer model for this engine; and (2) to evaluate structurally the use of silicon carbide regenerators. This paper presents comparisons of the measured brake performance, indicated mean effective pressure, and cyclic pressure variations from those predicted by the code. The silicon carbide foam generators appear to be structurally suitable, but the foam matrix showed severely reduced performance.

  7. Model of fracture for the Zry cladding of nuclear fuel rods included in the code DIONISIO 1.0

    Energy Technology Data Exchange (ETDEWEB)

    Soba, Alejandro [Departamento Combustibles Nucleares, Comision Nacional de Energia Atomica, Av. del Libertador 8250, 1429 Buenos Aires (Argentina)], E-mail: soba@cnea.gov.ar; Denis, Alicia [Departamento Combustibles Nucleares, Comision Nacional de Energia Atomica, Av. del Libertador 8250, 1429 Buenos Aires (Argentina)], E-mail: denis@cnea.gov.ar

    2008-12-15

    The DIONISIO code describes most of the main phenomena occurring in a fuel rod during normal operation of a nuclear power reactor. Starting from the irradiation history, the code predicts the temperature distribution, elastic and plastic stress and strain, creep, swelling and densification, release of fission gases, caesium and iodine to the internal rod volume, gas mixing, pressure increase, irradiation growth of the cladding, development of an oxide layer on its surface and hydrogen uptake, restructuring and grain growth in the pellet. This work presents the model of Zircaloy fracture included in the code DIONISIO 1.0. The model of pellet-cladding mechanical interaction (PCMI) provides the forces caused by the solid-solid contact which add to the changing internal pressure and to the constant external pressure. Besides, the program evaluates the effects of a corrosive atmosphere (stress corrosion cracking, SCC) internal or external. With these data, the code calculates the J integral around the tip of an initiated crack, and proceeds to analyze, according to the quantity of corrosive substance dissolved and the cladding stress field, if the crack remains unchanged, if it grows due to the I-SCC mechanism, or if propagation is ductile, following the R curve of the material. Results corresponding to different PHWR and PWR reactors are presented and compared with code results. In particular, good agreement is obtained in the simulation of MOX experiments, where the cladding failed due to propagation of cracks originated in SCC.

  8. Uncertainty evaluation of nuclear reaction model parameters using integral and microscopic measurements. Covariances evaluation with CONRAD code

    Directory of Open Access Journals (Sweden)

    Tommasi J.

    2010-10-01

    Full Text Available In the [eV;MeV] energy range, modelling of the neutron induced reactions are based on nuclear reaction models having parameters. Estimation of co-variances on cross sections or on nuclear reaction model parameters is a recurrent puzzle in nuclear data evaluation. Major breakthroughs were asked by nuclear reactor physicists to assess proper uncertainties to be used in applications. In this paper, mathematical methods developped in the CONRAD code[2] will be presented to explain the treatment of all type of uncertainties, including experimental ones (statistical and systematic and propagate them to nuclear reaction model parameters or cross sections. Marginalization procedure will thus be exposed using analytical or Monte-Carlo solutions. Furthermore, one major drawback found by reactor physicist is the fact that integral or analytical experiments (reactor mock-up or simple integral experiment, e.g. ICSBEP, … were not taken into account sufficiently soon in the evaluation process to remove discrepancies. In this paper, we will describe a mathematical framework to take into account properly this kind of information.

  9. A code to simulate nuclear reactor inventories and associated gamma-ray spectra.

    Science.gov (United States)

    Cresswell, A J; Allyson, J D; Sanderson, D C

    2001-01-01

    A computer code has been developed to simulate the gamma-ray spectra that would be measured by airborne gamma spectrometry (AGS) systems from sources containing short-lived fission products. The code uses simple numerical methods to simulate the production and decay of fission products and generates spectra for sodium iodide (NaI) detectors using Monte Carlo codes. A new Monte Carlo code using a virtual array of detectors to reduce simulation times for airborne geometries is described. Spectra generated for a short irradiation and laboratory geometry have been compared with an experimental data set. The agreement is good. Spectra have also been generated for airborne geometries and longer irradiation periods. The application of this code to generate AGS spectra for accident scenarios and their uses in the development and evaluation of spectral analysis methods for such situations are discussed.

  10. The 2010 fib Model Code for Structural Concrete: A new approach to structural engineering

    NARCIS (Netherlands)

    Walraven, J.C.; Bigaj-Van Vliet, A.

    2011-01-01

    The fib Model Code is a recommendation for the design of reinforced and prestressed concrete which is intended to be a guiding document for future codes. Model Codes have been published before, in 1978 and 1990. The draft for fib Model Code 2010 was published in May 2010. The most important new elem

  11. Comparison of Codes and Neutronics Data Used in the United States and Russia for the TOPAZ-2 Nuclear Safety Assessment

    Science.gov (United States)

    Glushkov, Y. S.; Ponomarov-Stepnoy, N. N.; Kompaniets, G. V.; Gomin, Y. A.; Mayorov, L. V.; Lobyntsev, V. A.; Polyakov, D. N.; Sapir, Joe; Pelowitz, Denise; Streetman, J. Robert

    1994-07-01

    The TOPAZ-2 reactor system is a heterogeneous epithermal system fueled with highly-enriched fuel based on uranium oxide, cooled by a sodium-potassium liquid metal (NaK), using a zirconium hydride moderator, with 37 thermionic fuel elements (TFEs) built into the core. The core is surrounded by a radial beryllium reflector which contains rotating regulating drums with moderating segments. An important problem is the guaranteeing of nuclear safety upon the accidental falling of the TOPAZ-2 reactor into water, which leads to the growth of the reactivity of the reactor. It has turned out that it is necessary to use the Monte-Carlo method for the conduct of neutronics calculations of such a complex reactor. In the United States (U.S.) and Russia, different codes based on the Monte-Carlo method are used for calculations - the MCNP code in the U.S., and the MCU-2 code in Russia. The goal of this work is the comparison of the codes and neutronics data used in the U.S. and Russia for the basis of the TOPAZ-2 nuclear safety. With this goal, a joint computer model benchmark of the TOPAZ-2 reactor was developed and the calculations of a series of variants, differing by the presence and absence of water in the reactor cavities and behind the radial reflector, in the position of the regulating drums, in the presence of the radial reflector, etc. were done independently by specialists in both the U.S. and Russia. Along with the reactor calculations, calculations were also done of the nuclei of the core using the MCNP code (U.S.) and the MCU-2 code (Russia). The work done allowed one to obtain results comparing the MCNP code to the MCU-2 code which gave somewhat different results both for the absolute values of Keff and for reactivity effects. In the future it remains to conduct a detailed analysis of the reasons for the discrepancies. For this it is necessary to exchange neutronics data used for TOPAZ-2 reactor calculations in the U.S. and Russia.

  12. 78 FR 37848 - ASME Code Cases Not Approved for Use

    Science.gov (United States)

    2013-06-24

    ... COMMISSION ASME Code Cases Not Approved for Use AGENCY: Nuclear Regulatory Commission. ACTION: Draft... public comment draft regulatory guide (DG), DG-1233, ``ASME Code Cases not Approved for Use.'' This regulatory guide lists the American Society of Mechanical Engineers (ASME) Code Cases that the NRC...

  13. Comparative study of Monte Carlo particle transport code PHITS and nuclear data processing code NJOY for recoil cross section spectra under neutron irradiation

    Science.gov (United States)

    Iwamoto, Yosuke; Ogawa, Tatsuhiko

    2017-04-01

    Because primary knock-on atoms (PKAs) create point defects and clusters in materials that are irradiated with neutrons, it is important to validate the calculations of recoil cross section spectra that are used to estimate radiation damage in materials. Here, the recoil cross section spectra of fission- and fusion-relevant materials were calculated using the Event Generator Mode (EGM) of the Particle and Heavy Ion Transport code System (PHITS) and also using the data processing code NJOY2012 with the nuclear data libraries TENDL2015, ENDF/BVII.1, and JEFF3.2. The heating number, which is the integral of the recoil cross section spectra, was also calculated using PHITS-EGM and compared with data extracted from the ACE files of TENDL2015, ENDF/BVII.1, and JENDL4.0. In general, only a small difference was found between the PKA spectra of PHITS + TENDL2015 and NJOY + TENDL2015. From analyzing the recoil cross section spectra extracted from the nuclear data libraries using NJOY2012, we found that the recoil cross section spectra were incorrect for 72Ge, 75As, 89Y, and 109Ag in the ENDF/B-VII.1 library, and for 90Zr and 55Mn in the JEFF3.2 library. From analyzing the heating number, we found that the data extracted from the ACE file of TENDL2015 for all nuclides were problematic in the neutron capture region because of incorrect data regarding the emitted gamma energy. However, PHITS + TENDL2015 can calculate PKA spectra and heating numbers correctly.

  14. Comparative study of Monte Carlo particle transport code PHITS and nuclear data processing code NJOY for recoil cross section spectra under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Iwamoto, Yosuke, E-mail: iwamoto.yosuke@jaea.go.jp; Ogawa, Tatsuhiko

    2017-04-01

    Because primary knock-on atoms (PKAs) create point defects and clusters in materials that are irradiated with neutrons, it is important to validate the calculations of recoil cross section spectra that are used to estimate radiation damage in materials. Here, the recoil cross section spectra of fission- and fusion-relevant materials were calculated using the Event Generator Mode (EGM) of the Particle and Heavy Ion Transport code System (PHITS) and also using the data processing code NJOY2012 with the nuclear data libraries TENDL2015, ENDF/BVII.1, and JEFF3.2. The heating number, which is the integral of the recoil cross section spectra, was also calculated using PHITS-EGM and compared with data extracted from the ACE files of TENDL2015, ENDF/BVII.1, and JENDL4.0. In general, only a small difference was found between the PKA spectra of PHITS + TENDL2015 and NJOY + TENDL2015. From analyzing the recoil cross section spectra extracted from the nuclear data libraries using NJOY2012, we found that the recoil cross section spectra were incorrect for {sup 72}Ge, {sup 75}As, {sup 89}Y, and {sup 109}Ag in the ENDF/B-VII.1 library, and for {sup 90}Zr and {sup 55}Mn in the JEFF3.2 library. From analyzing the heating number, we found that the data extracted from the ACE file of TENDL2015 for all nuclides were problematic in the neutron capture region because of incorrect data regarding the emitted gamma energy. However, PHITS + TENDL2015 can calculate PKA spectra and heating numbers correctly.

  15. Calculation of electron and isotopes dose point kernels with FLUKA Monte Carlo code for dosimetry in nuclear medicine therapy

    CERN Document Server

    Mairani, A; Valente, M; Battistoni, G; Botta, F; Pedroli, G; Ferrari, A; Cremonesi, M; Di Dia, A; Ferrari, M; Fasso, A

    2011-01-01

    Purpose: The calculation of patient-specific dose distribution can be achieved by Monte Carlo simulations or by analytical methods. In this study, FLUKA Monte Carlo code has been considered for use in nuclear medicine dosimetry. Up to now, FLUKA has mainly been dedicated to other fields, namely high energy physics, radiation protection, and hadrontherapy. When first employing a Monte Carlo code for nuclear medicine dosimetry, its results concerning electron transport at energies typical of nuclear medicine applications need to be verified. This is commonly achieved by means of calculation of a representative parameter and comparison with reference data. Dose point kernel (DPK), quantifying the energy deposition all around a point isotropic source, is often the one. Methods: FLUKA DPKS have been calculated in both water and compact bone for monoenergetic electrons (10-3 MeV) and for beta emitting isotopes commonly used for therapy ((89)Sr, (90)Y, (131)I, (153)Sm, (177)Lu, (186)Re, and (188)Re). Point isotropic...

  16. SENDIN and SENTINEL: two computer codes to assess the effects of nuclear data changes

    Energy Technology Data Exchange (ETDEWEB)

    Marable, J. H.; Drischler, J. D.; Weisbin, C. R.

    1977-07-01

    A description is given of the computer code SENTINEL, which provides a simple means for finding the effects on calculated reactor and shielding performance parameters due to proposed changes in the cross section data base. This code uses predetermined detailed sensitivity coefficients in SENPRO format, which is described in Appendix A. Knowledge of details of the particular reactor and/or shielding assemblies is not required of the user. Also described is the computer code SENDIN, which converts unformatted (binary) sensitivity files to card image form and vice versa. This is useful for transferring sensitivity files from one installation to another.

  17. Project-Based Learning in the Masters degree in Nuclear Engineering at BarcelonaTECH. Experience gained in the area of Management of Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Reventos, F.; Vives, E.; Brunet, A.; Sabate, R.; Calvino, F.; Batet, L.

    2014-07-01

    From its first edition, that took place in 2011-2012, the Masters degree in Nuclear Engineering from BarcelonaTECH has been using techniques of Project-Based Learning to fulfill the purpose of training nuclear engineers with a profile suitable for positions in the industry. The Master is sponsored by ENDESA and relies on the collaboration with institutions and companies. The Master is embedded in EMINE, the European Master in Innovation in Nuclear Energy, supported by KIC-InnoEnergy and the European Institute of Technology. (Author)

  18. Resolving arthropod phylogeny: exploring phylogenetic signal within 41 kb of protein-coding nuclear gene sequence.

    Science.gov (United States)

    Regier, Jerome C; Shultz, Jeffrey W; Ganley, Austen R D; Hussey, April; Shi, Diane; Ball, Bernard; Zwick, Andreas; Stajich, Jason E; Cummings, Michael P; Martin, Joel W; Cunningham, Clifford W

    2008-12-01

    This study attempts to resolve relationships among and within the four basal arthropod lineages (Pancrustacea, Myriapoda, Euchelicerata, Pycnogonida) and to assess the widespread expectation that remaining phylogenetic problems will yield to increasing amounts of sequence data. Sixty-eight regions of 62 protein-coding nuclear genes (approximately 41 kilobases (kb)/taxon) were sequenced for 12 taxonomically diverse arthropod taxa and a tardigrade outgroup. Parsimony, likelihood, and Bayesian analyses of total nucleotide data generally strongly supported the monophyly of each of the basal lineages represented by more than one species. Other relationships within the Arthropoda were also supported, with support levels depending on method of analysis and inclusion/exclusion of synonymous changes. Removing third codon positions, where the assumption of base compositional homogeneity was rejected, altered the results. Removing the final class of synonymous mutations--first codon positions encoding leucine and arginine, which were also compositionally heterogeneous--yielded a data set that was consistent with a hypothesis of base compositional homogeneity. Furthermore, under such a data-exclusion regime, all 68 gene regions individually were consistent with base compositional homogeneity. Restricting likelihood analyses to nonsynonymous change recovered trees with strong support for the basal lineages but not for other groups that were variably supported with more inclusive data sets. In a further effort to increase phylogenetic signal, three types of data exploration were undertaken. (1) Individual genes were ranked by their average rate of nonsynonymous change, and three rate categories were assigned--fast, intermediate, and slow. Then, bootstrap analysis of each gene was performed separately to see which taxonomic groups received strong support. Five taxonomic groups were strongly supported independently by two or more genes, and these genes mostly belonged to the slow

  19. Assessing the Predictive Capability of the LIFEIV Nuclear Fuel Performance Code using Sequential Calibration

    Energy Technology Data Exchange (ETDEWEB)

    Stull, Christopher J. [Los Alamos National Laboratory; Williams, Brian J. [Los Alamos National Laboratory; Unal, Cetin [Los Alamos National Laboratory

    2012-07-05

    This report considers the problem of calibrating a numerical model to data from an experimental campaign (or series of experimental tests). The issue is that when an experimental campaign is proposed, only the input parameters associated with each experiment are known (i.e. outputs are not known because the experiments have yet to be conducted). Faced with such a situation, it would be beneficial from the standpoint of resource management to carefully consider the sequence in which the experiments are conducted. In this way, the resources available for experimental tests may be allocated in a way that best 'informs' the calibration of the numerical model. To address this concern, the authors propose decomposing the input design space of the experimental campaign into its principal components. Subsequently, the utility (to be explained) of each experimental test to the principal components of the input design space is used to formulate the sequence in which the experimental tests will be used for model calibration purposes. The results reported herein build on those presented and discussed in [1,2] wherein Verification & Validation and Uncertainty Quantification (VU) capabilities were applied to the nuclear fuel performance code LIFEIV. In addition to the raw results from the sequential calibration studies derived from the above, a description of the data within the context of the Predictive Maturity Index (PMI) will also be provided. The PMI [3,4] is a metric initiated and developed at Los Alamos National Laboratory to quantitatively describe the ability of a numerical model to make predictions in the absence of experimental data, where it is noted that 'predictions in the absence of experimental data' is not synonymous with extrapolation. This simply reflects the fact that resources do not exist such that each and every execution of the numerical model can be compared against experimental data. If such resources existed, the justification for

  20. Development of a severe accident module of a nuclear power plant based in the MELCOR nuclear code and its incorporation to the room simulator; Desarrollo del modulo de accidentes severos de una central nucleoelectrica basado en el codigo nuclear MELCOR y su incorporacion al simulador de aula

    Energy Technology Data Exchange (ETDEWEB)

    Cortes M, F.S.; Ramos P, J.C.; Nelson E, P.; Chavez M, C. [Facultad de Ingenieria, Division de Ingenieria Electrica, Grupo de Ingenieria Nuclear, UNAM, Ciudad Universitaria, Distrito Federal (Mexico)]. E-mail: samuelcortes@correo.unam.mx

    2004-07-01

    This work describes the development of the Severe Accidents Module (MAS) based on the Code MELCOR and its incorporation to the Simulator of Classroom of the Group of Nuclear Engineering of the Engineering Faculty (GrINFI) of the National Autonomous University of Mexico (UNAM). The module of Severe Accidents has the purpose of counting with installed and operational capacity for the simulation of accident sequences with capacitation purposes, training, analysis and design. A shallow description of SimAula is presented, and the philosophy used to obtain the interactive version of MELCOR are discussed, as well as its implementation in the atmosphere of SimAula. Finally, after confirming the correct operation of the development of the tool, some possible topics are discussed for specific applications of the MAS. (Author)

  1. Determination of the costs of the nuclear desalination using the DEEP code from IAEA; Determinacion de los costos de la desalacion nuclear utilizando el codigo DEEP del OIEA

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J.R.; Palacios H, J.C.; Alonso V, G. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: jrrs@nuclear.inin.mx

    2005-07-01

    The desalination of seawater is being an important solution to satisfy the demands of drinking water to population's centers that have hydric resources very limited, like it is the case of some Arab countries and arid regions of the planet, in where they have settled desalination plants that use as energy source to those fossil fuels or nuclear energy plants. Taking into account that the desalination of seawater is a process that consumes a lot of thermal and/or electric energy, it is necessary to quantify the costs of the supply and that of the desalination plant for different options and technologies, looking for this way the but appropriate for the specific conditions of the region where it has planned the desalination of seawater. In this report the three technologies but promising for the desalination are described and by means of the DEEP code the costs of production of water and energy are evaluated, using as thermal source different types of power nuclear reactors. It was obtained according to DEEP that the costs of the electricity generation for the considered reactors are around 40 USD/MWh. With these costs of electric power generation and using the DEEP code is obtained that the costs of production of drinking water are around 1 USD/m{sup 3}. (Author)

  2. Accuracy and convergence of coupled finite-volume/Monte Carlo codes for plasma edge simulations of nuclear fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ghoos, K., E-mail: kristel.ghoos@kuleuven.be [KU Leuven, Department of Mechanical Engineering, Celestijnenlaan 300A, 3001 Leuven (Belgium); Dekeyser, W. [KU Leuven, Department of Mechanical Engineering, Celestijnenlaan 300A, 3001 Leuven (Belgium); Samaey, G. [KU Leuven, Department of Computer Science, Celestijnenlaan 200A, 3001 Leuven (Belgium); Börner, P. [Institute of Energy and Climate Research (IEK-4), FZ Jülich GmbH, D-52425 Jülich (Germany); Baelmans, M. [KU Leuven, Department of Mechanical Engineering, Celestijnenlaan 300A, 3001 Leuven (Belgium)

    2016-10-01

    The plasma and neutral transport in the plasma edge of a nuclear fusion reactor is usually simulated using coupled finite volume (FV)/Monte Carlo (MC) codes. However, under conditions of future reactors like ITER and DEMO, convergence issues become apparent. This paper examines the convergence behaviour and the numerical error contributions with a simplified FV/MC model for three coupling techniques: Correlated Sampling, Random Noise and Robbins Monro. Also, practical procedures to estimate the errors in complex codes are proposed. Moreover, first results with more complex models show that an order of magnitude speedup can be achieved without any loss in accuracy by making use of averaging in the Random Noise coupling technique.

  3. Statement of work for sytem design and engineering of the spent nuclear fuel multi-cansiter overpack

    Energy Technology Data Exchange (ETDEWEB)

    Smith, K.E., Fluor Daniel Hanford

    1997-03-03

    This Statement of Work (SOW) describes the work scope for the preparation of the Phase 2 (final) design for the Multiple Canister Overpack (MCO) equipment. The MCO is to be used as the radiological containment device for the Spent Nuclear Fuel (SNF) assemblies, currently in wet storage in K East and West Basins, to be transported and stored in the Canister Storage Building (CSB) until final disposal facilities are made available. The engineering services contractor will be requested to provide reports, studies, analyses, engineering, drawings, specifications, estimates and schedules. The overall goal of this task order is to do the following: 1. Prepare a fabrication specification, ASME Code exception report, a packaging, shipping and warehouse plan, and detailed fabrication drawings of the MCO in accordance with the MCO Performance Specification (HNF-S-0426, Rev. 3) for procurement activities by the SNF MCO Subproject. 2. Establish and maintain a comment data base on the comments, resolutions, changes to the design of the MCO. 3. Support fabrication activities through the review of vendor fabrication drawings and shop test reports.

  4. A new code for modelling the near field diffusion releases from the final disposal of nuclear waste

    Science.gov (United States)

    Vopálka, D.; Vokál, A.

    2003-01-01

    The canisters with spent nuclear fuel produced during the operation of WWER reactors at the Czech power plants are planned, like in other countries, to be disposed of in an underground repository. Canisters will be surrounded by compacted bentonite that will retard the migration of safety-relevant radionuclides into the host rock. A new code that enables the modelling of the critical radionuclides transport from the canister through the bentonite layer in the cylindrical geometry was developed. The code enables to solve the diffusion equation for various types of initial and boundary conditions by means of the finite difference method and to take into account the non-linear shape of the sorption isotherm. A comparison of the code reported here with code PAGODA, which is based on analytical solution of the transport equation, was made for the actinide chain 4N+3 that includes 239Pu. A simple parametric study of the releases of 239Pu, 129I, and 14C into geosphere is discussed.

  5. Engineering on abolishment measure of nuclear fuel facilities. Application of 3D-CAD to abolishment measure of nuclear fuel facilities

    Energy Technology Data Exchange (ETDEWEB)

    Annen, Sotonori; Sugitsue, Noritake [Japan Nuclear Cycle Development Inst., Ningyo Toge Environmental Engineering Center, Kamisaibara, Okayama (Japan)

    2001-12-01

    The Japan Nuclear Cycle Development Institute (JNC) progresses some advancing R and Ds required for establishment of the nuclear fuel cycle under considering on safety, economical efficiency, environmental compatibility, and so on. An important item among them is a technology on safe abolishment of a nuclear energy facility ended its role, which is called the abolishment measure technique. Here was introduced at a center of viewpoint called on use of three dimensional CAD (3D-CAD), on outlines of engineering system for abolishment measure (subdivision engineering system) under an object of nuclear fuel facilities, constructed through subdivision and removal of refinement conversion facilities, by the Ningyo-toge Environmental Engineering Center of JNC. (G.K.)

  6. An improved heat transfer configuration for a solid-core nuclear thermal rocket engine

    Science.gov (United States)

    Clark, John S.; Walton, James T.; Mcguire, Melissa L.

    1992-01-01

    Interrupted flow, impingement cooling, and axial power distribution are employed to enhance the heat-transfer configuration of a solid-core nuclear thermal rocket engine. Impingement cooling is introduced to increase the local heat-transfer coefficients between the reactor material and the coolants. Increased fuel loading is used at the inlet end of the reactor to enhance heat-transfer capability where the temperature differences are the greatest. A thermal-hydraulics computer program for an unfueled NERVA reactor core is employed to analyze the proposed configuration with attention given to uniform fuel loading, number of channels through the impingement wafers, fuel-element length, mass-flow rate, and wafer gap. The impingement wafer concept (IWC) is shown to have heat-transfer characteristics that are better than those of the NERVA-derived reactor at 2500 K. The IWC concept is argued to be an effective heat-transfer configuration for solid-core nuclear thermal rocket engines.

  7. Utilization Practice of the Concept Mapping Program for Nuclear Engineer Training

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bae Joo; Ko, Byung Moo; Heo, Yuk [Korea Hydro and Nuclear Power, Daejeon (Korea, Republic of)

    2009-05-15

    Knowledge is the most important factor in the safe and reliable operation of the NPP. Many methods are used to enhance the knowledge level of the personnel in the NPP. Generally, classroom lecture method is used for nuclear engineers. But this method has some pitfalls as an adult training method because students have already a lot of knowledge, so they want to participate actively in the learning process. KNPEI undertook a research project from March 2006 to September 2007 to capture the experience knowledge from senior staff and transfer it to junior staff. As part of the research activity KNPEI introduced a Concept Mapping Program and set up a Concept Mapping server to capture the experience knowledge of the senior staff. This Concept Mapping Program has some characteristics that can be used in learning about conceptual knowledge. The purpose of this report is to introduce the utilization method and practice at KNPEI for the nuclear engineer training using the Concept Mapping Program.

  8. Reverse engineering nuclear properties from rare earth abundances in the r process

    Science.gov (United States)

    Mumpower, M. R.; McLaughlin, G. C.; Surman, R.; Steiner, A. W.

    2017-03-01

    The bulk of the rare earth elements are believed to be synthesized in the rapid neutron capture process or r process of nucleosynthesis. The solar r-process residuals show a small peak in the rare earths around A∼ 160, which is proposed to be formed dynamically during the end phase of the r process by a pileup of material. This abundance feature is of particular importance as it is sensitive to both the nuclear physics inputs and the astrophysical conditions of the main r process. We explore the formation of the rare earth peak from the perspective of an inverse problem, using Monte Carlo studies of nuclear masses to investigate the unknown nuclear properties required to best match rare earth abundance sector of the solar isotopic residuals. When nuclear masses are changed, we recalculate the relevant β-decay properties and neutron capture rates in the rare earth region. The feedback provided by this observational constraint allows for the reverse engineering of nuclear properties far from stability where no experimental information exists. We investigate a range of astrophysical conditions with this method and show how these lead to different predictions in the nuclear properties influential to the formation of the rare earth peak. We conclude that targeted experimental campaigns in this region will help to resolve the type of conditions responsible for the production of the rare earth nuclei, and will provide new insights into the longstanding problem of the astrophysical site(s) of the r process.

  9. Pluronic F127 nanomicelles engineered with nuclear localized functionality for targeted drug delivery.

    Science.gov (United States)

    Li, Yong-Yong; Li, Lan; Dong, Hai-Qing; Cai, Xiao-Jun; Ren, Tian-Bin

    2013-07-01

    PKKKRKV (Pro-Lys-Lys-Lys-Arg-Lys-Val, PV7), a seven amino acid peptide, has emerged as one of the primary nuclear localization signals that can be targeted into cell nucleus via the nuclear import machinery. Taking advantage of chemical diversity and biological activities of this short peptide sequence, in this study, Pluronic F127 nanomicelles engineered with nuclear localized functionality were successfully developed for intracellular drug delivery. These nanomicelles with the size ~100 nm were self-assembled from F127 polymer that was flanked with two PV7 sequences at its both terminal ends. Hydrophobic anticancer drug doxorubicin (DOX) with inherent fluorescence was chosen as the model drug, which was found to be efficiently encapsulated into nanomicelles with the encapsulation efficiency at 72.68%. In comparison with the non-functionalized namomicelles, the microscopic observation reveals that PV7 functionalized nanomicelles display a higher cellular uptake, especially into the nucleus of HepG2 cells, due to the nuclear localization signal effects. Both cytotoxicity and apoptosis studies show that the DOX-loaded nanomicelles were more potent than drug nanomicelles without nuclear targeting functionality. It was thus concluded that PV7 functionalized nanomicelles could be a potentially alternative vehicle for nuclear targeting drug delivery.

  10. Engineered barrier development for a nuclear waste repository in basalt: an integration of current knowledge

    Energy Technology Data Exchange (ETDEWEB)

    Smith, M.J.

    1980-05-01

    This document represents a compilation of data and interpretive studies conducted as part of the engineered barriers program of the Basalt Waste Isolation Project. The overall objective of these studies is to provide information on barrier system designs, emplacement and isolation techniques, and chemical reactions expected in a nuclear waste repository located in the basalts underlying the Hanford Site within the state of Washington. Backfills, waste-basalt interactions, sorption, borehole plugging, etc., are among the topics discussed.

  11. "Cloud" functions and templates of engineering calculations for nuclear power plants

    Science.gov (United States)

    Ochkov, V. F.; Orlov, K. A.; Ko, Chzho Ko

    2014-10-01

    The article deals with an important problem of setting up computer-aided design calculations of various circuit configurations and power equipment carried out using the templates and standard computer programs available in the Internet. Information about the developed Internet-based technology for carrying out such calculations using the templates accessible in the Mathcad Prime software package is given. The technology is considered taking as an example the solution of two problems relating to the field of nuclear power engineering.

  12. Nuclear Engineering: Enrollments and Degrees. Enrollments-Fall 1973, Degrees Granted-July 1965-June 1973.

    Science.gov (United States)

    Atomic Energy Commission, Washington, DC. Office of Industrial Relations.

    This document presents statistical data concerning enrollments for fall 1973 and degrees granted 1965-June 1973 in nuclear engineering. Highlights of this survey of educational institutions indicated: (1) Ph.D.'s decreased to 126 from 149 in 1971-72 and from 181 in 1969-70. (2) MS's increased to 442 from 428 in 1971-72. (3) BS's increased to 551…

  13. Space and Terrestrial Power System Integration Optimization Code BRMAPS for Gas Turbine Space Power Plants With Nuclear Reactor Heat Sources

    Science.gov (United States)

    Juhasz, Albert J.

    2007-01-01

    expected within the next 30 to 50 years, as predicted by the Hubbert model and confirmed by other global energy consumption prognoses. Having invested national resources into the development of NGNP, the technology and experience accumulated during the project needs to be documented clearly and in sufficient detail for young engineers coming on-board at both DOE and NASA to acquire it. Hands on training on reactor operation, test rigs of turbomachinery, and heat exchanger components, as well as computational tools will be needed. Senior scientist/engineers involved with the development of NGNP should also be encouraged to participate as lecturers, instructors, or adjunct professors at local universities having engineering (mechanical, electrical, nuclear/chemical, and/or materials) as one of their fields of study.

  14. Enrico Fermi and the Physics and Engineering of a nuclear pile: the retrieval of novel documents

    CERN Document Server

    Esposito, S

    2008-01-01

    We give a detailed account of the recent retrieval of a consistent amount (about 600 pages) of documents written by Enrico Fermi and/or his collaborators, coming from different sources previously unexplored. These documents include articles, patents, reports, notes on scientific and technical meetings and other papers, mainly testifying Fermi's activity in the 1940s about nuclear pile physics and engineering. All of them have been carefully described, pointing out the relevance of the given papers for their scientific or even historical content. From the analysis of these papers, a number of important scientific and technical points comes out, putting a truly new light on the Fermi's (and others') scientific activity about nuclear piles and their applications. Quite unexpectedly intriguing historical remarks, such as those regarding the relationships between U.S. and Britain, just after the end of the war, about nuclear power for pacific and/or military use, or even regarding long term physics research and po...

  15. [Prospects of systemic radioecology in solving innovative tasks of nuclear power engineering].

    Science.gov (United States)

    Spiridonov, S I

    2014-01-01

    A need of systemic radioecological studies in the strategy developed by the atomic industry in Russia in the XXI century has been justified. The priorities in the radioecology of nuclear power engineering of natural safety associated with the development of the radiation-migration equivalence concept, comparative evaluation of innovative nuclear technologies and forecasting methods of various emergencies have been identified. Also described is an algorithm for the integrated solution of these tasks that includes elaboration of methodological approaches, methods and software allowing dose burdens to humans and biota to be estimated. The rationale of using radioecological risks for the analysis of uncertainties in the environmental contamination impacts,at different stages of the existing and innovative nuclear fuel cycles is shown.

  16. Complex composite engineering architectures for nuclear and high-radiation environments

    Energy Technology Data Exchange (ETDEWEB)

    Kornreich, Drew E [Los Alamos National Laboratory; Vaidya, Rajendra U [Los Alamos National Laboratory; Ammerman, Curtt N [Los Alamos National Laboratory

    2010-01-01

    Integrated Computational Materials Engineering (ICME) is a novel overarching approach to bridge length and time scales in computational materials science and engineering. This approach integrates all elements of multi-scale modeling (including various empirical and science-based models) with materials informatics to provide users the opportunity to tailor material selections based on stringent application needs. Typically, materials engineering has focused on structural requirements (stress, strain, modulus, fracture toughness etc.) while multi-scale modeling has been science focused (mechanical threshold strength model, grain-size models, solid-solution strengthening models etc.). Materials informatics (mechanical property inventories) on the other hand, is extensively data focused. All of these elements are combined within the framework of ICME to create architecture for the development, selection and design new composite materials for challenging environments. We propose development of the foundations for applying ICME to composite materials development for nuclear and high-radiation environments (including nuclear-fusion energy reactors, nuclear-fission reactors, and accelerators). We expect to combine all elements of current material models (including thermo-mechanical and finite-element models) into the ICME framework. This will be accomplished through the use of a various mathematical modeling constructs. These constructs will allow the integration of constituent models, which in tum would allow us to use the adaptive strengths of using a combinatorial scheme (fabrication and computational) for creating new composite materials. A sample problem where these concepts are used is provided in this summary.

  17. Pluronic F127 nanomicelles engineered with nuclear localized functionality for targeted drug delivery

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yong-Yong; Li, Lan; Dong, Hai-Qing, E-mail: inano_donghq@tongji.edu.cn; Cai, Xiao-Jun; Ren, Tian-Bin, E-mail: rentianbin@yeah.net

    2013-07-01

    PKKKRKV (Pro-Lys-Lys-Lys-Arg-Lys-Val, PV7), a seven amino acid peptide, has emerged as one of the primary nuclear localization signals that can be targeted into cell nucleus via the nuclear import machinery. Taking advantage of chemical diversity and biological activities of this short peptide sequence, in this study, Pluronic F127 nanomicelles engineered with nuclear localized functionality were successfully developed for intracellular drug delivery. These nanomicelles with the size ∼ 100 nm were self-assembled from F127 polymer that was flanked with two PV7 sequences at its both terminal ends. Hydrophobic anticancer drug doxorubicin (DOX) with inherent fluorescence was chosen as the model drug, which was found to be efficiently encapsulated into nanomicelles with the encapsulation efficiency at 72.68%. In comparison with the non-functionalized namomicelles, the microscopic observation reveals that PV7 functionalized nanomicelles display a higher cellular uptake, especially into the nucleus of HepG2 cells, due to the nuclear localization signal effects. Both cytotoxicity and apoptosis studies show that the DOX-loaded nanomicelles were more potent than drug nanomicelles without nuclear targeting functionality. It was thus concluded that PV7 functionalized nanomicelles could be a potentially alternative vehicle for nuclear targeting drug delivery. - Highlights: ► A new nuclear targeted drug delivery system based on micelles is developed. ► This micellar system features a core-shell structure with the size peaked at 100 nm. ► PV7, a short peptide sequence, is adopted as a nuclear targeting ligand. ► PV7 functionalized drug loaded micelles are more potent in killing tumor cells.

  18. 64 International conference "NUCLEUS-2014" Fundamental problems of nuclear physics, atomic power engineering and nuclear technologies

    OpenAIRE

    Vlasnikov, A. K.

    2014-01-01

    Тезисы 64 международной конференции «ЯДРО-2014» (Фундаментальные проблемы ядерной физики, атомной энергетики и ядерных технологий), БГУ, Минск, 1 – 4 июля 2014 года. The scientific program of the conference covers almost all problems in nuclear physics and its applications such as: neutron-rich nuclei, nuclei far from stability valley, giant resonances, many-phonon and many-quasiparticle states in nuclei, high-spin and super-deformed states in nuclei, synthesis of super-heavy elements, ...

  19. Research on the improvement of nuclear safety -The development of LOCA analysis codes for nuclear power plant-

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Won Pyo; Jung, Yung Jong; Kim, Kyung Doo; Jung, Jae Joon; Kim, Won Suk; Han, Doh Heui; Hah, Kooi Suk; Jung, Bub Dong; Lee, Yung Jin; Hwang, Tae Suk; Lee, Sang Yong; Park, Chan Uk; Choi, Han Rim; Lee, Sang Jong; Choi, Jong Hoh; Ban, Chang Hwan; Bae, Kyoo Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    The present research aims at development of both a best estimate methodology on LOCA analysis and, as an application, performance analyses of safety systems. SBLOCA analyses have been continued to examine the capacity reduction effect of ECCS since the second project year. As a results, core uncovery, which is requirement of URD has not been occurred in 6`` cold leg break. Although core uncovery has been predicted when DVI line has been broken for DVI+4-Train HPIS, the calculated PCT has lied well within the criterion. The effect of safety injection position and SIT characteristics are also analyzed for LBLOCA. The results show that cold leg injection is the most effective way and the adaption of advanced SIT could lead to elimination of LPSI pump from the safety system. On the other hand, the quantified uncertainties obtained from THTF and FLECHT/SEASET which represents blowdown and reflood phenomena, respectively, have been confirmed using IET(LOFT test). The application uncertainty for Kori unit 3 has been analyzed. Finally, application of the best estimate methodology using the uncertainties concerned with the code, the bais, and the application, leads to overall uncertainty of about 200K for Kori unit 3. 244 figs, 22 tabs, 92 refs. (Author).

  20. Nuclear Engine System Simulation (NESS). Version 2.0: Program user's guide

    Science.gov (United States)

    Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman

    1993-01-01

    This Program User's Guide discusses the Nuclear Thermal Propulsion (NTP) engine system design features and capabilities modeled in the Nuclear Engine System Simulation (NESS): Version 2.0 program (referred to as NESS throughout the remainder of this document), as well as its operation. NESS was upgraded to include many new modeling capabilities not available in the original version delivered to NASA LeRC in Dec. 1991, NESS's new features include the following: (1) an improved input format; (2) an advanced solid-core NERVA-type reactor system model (ENABLER 2); (3) a bleed-cycle engine system option; (4) an axial-turbopump design option; (5) an automated pump-out turbopump assembly sizing option; (6) an off-design gas generator engine cycle design option; (7) updated hydrogen properties; (8) an improved output format; and (9) personal computer operation capability. Sample design cases are presented in the user's guide that demonstrate many of the new features associated with this upgraded version of NESS, as well as design modeling features associated with the original version of NESS.

  1. Investigation of Nuclear Data Libraries with TRIPOLI-4 Monte Carlo Code for Sodium-cooled Fast Reactors

    Science.gov (United States)

    Lee, Y.-K.; Brun, E.

    2014-04-01

    The Sodium-cooled fast neutron reactor ASTRID is currently under design and development in France. Traditional ECCO/ERANOS fast reactor code system used for ASTRID core design calculations relies on multi-group JEFF-3.1.1 data library. To gauge the use of ENDF/B-VII.0 and JEFF-3.1.1 nuclear data libraries in the fast reactor applications, two recent OECD/NEA computational benchmarks specified by Argonne National Laboratory were calculated. Using the continuous-energy TRIPOLI-4 Monte Carlo transport code, both ABR-1000 MWth MOX core and metallic (U-Pu) core were investigated. Under two different fast neutron spectra and two data libraries, ENDF/B-VII.0 and JEFF-3.1.1, reactivity impact studies were performed. Using JEFF-3.1.1 library under the BOEC (Beginning of equilibrium cycle) condition, high reactivity effects of 808 ± 17 pcm and 1208 ± 17 pcm were observed for ABR-1000 MOX core and metallic core respectively. To analyze the causes of these differences in reactivity, several TRIPOLI-4 runs using mixed data libraries feature allow us to identify the nuclides and the nuclear data accounting for the major part of the observed reactivity discrepancies.

  2. Research on the improvement of nuclear safety -The development of a severe accident analysis code-

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Heui Dong; Cho, Sung Won; Park, Jong Hwa; Hong, Sung Wan; Yoo, Dong Han; Hwang, Moon Kyoo; Noh, Kee Man; Song, Yong Man [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    For prevention and mitigation of the containment failure during severe accident, the study is focused on the severe accident phenomena, especially, the ones occurring inside the cavity and is intended to improve existing models and develop analytical tools for the assessment of severe accidents. A correlation equation of the flame velocity of pre mixture gas of H{sub 2}/air/steam has been suggested and combustion flame characteristic was analyzed using a developed computer code. For the analysis of the expansion phase of vapor explosion, the mechanical model has been developed. The development of a debris entrainment model in a reactor cavity with captured volume has been continued to review and examine the limitation and deficiencies of the existing models. Pre-test calculation was performed to support the severe accident experiment for molten corium concrete interaction study and the crust formation process and heat transfer characteristics of the crust have been carried out. A stress analysis code was developed using finite element method for the reactor vessel lower head failure analysis. Through international program of PHEBUS-FP and participation in the software development, the research on the core degradation process and fission products release and transportation are undergoing. CONTAIN and MELCOR codes were continuously updated under the cooperation with USNRC and French developed computer codes such as ICARE2, ESCADRE, SOPHAEROS were also installed into the SUN workstation. 204 figs, 61 tabs, 87 refs. (Author).

  3. Selection of a computer code for Hanford low-level waste engineered-system performance assessment. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    McGrail, B.P.; Bacon, D.H.

    1998-02-01

    Planned performance assessments for the proposed disposal of low-activity waste (LAW) glass produced from remediation of wastes stored in underground tanks at Hanford, Washington will require calculations of radionuclide release rates from the subsurface disposal facility. These calculations will be done with the aid of computer codes. The available computer codes with suitable capabilities at the time Revision 0 of this document was prepared were ranked in terms of the feature sets implemented in the code that match a set of physical, chemical, numerical, and functional capabilities needed to assess release rates from the engineered system. The needed capabilities were identified from an analysis of the important physical and chemical processes expected to affect LAW glass corrosion and the mobility of radionuclides. This analysis was repeated in this report but updated to include additional processes that have been found to be important since Revision 0 was issued and to include additional codes that have been released. The highest ranked computer code was found to be the STORM code developed at PNNL for the US Department of Energy for evaluation of arid land disposal sites.

  4. Engineering

    National Research Council Canada - National Science Library

    Includes papers in the following fields: Aerospace Engineering, Agricultural Engineering, Chemical Engineering, Civil Engineering, Electrical Engineering, Environmental Engineering, Industrial Engineering, Materials Engineering, Mechanical...

  5. THATCH: A computer code for modelling thermal networks of high- temperature gas-cooled nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kroeger, P.G.; Kennett, R.J.; Colman, J.; Ginsberg, T. (Brookhaven National Lab., Upton, NY (United States))

    1991-10-01

    This report documents the THATCH code, which can be used to model general thermal and flow networks of solids and coolant channels in two-dimensional r-z geometries. The main application of THATCH is to model reactor thermo-hydraulic transients in High-Temperature Gas-Cooled Reactors (HTGRs). The available modules simulate pressurized or depressurized core heatup transients, heat transfer to general exterior sinks or to specific passive Reactor Cavity Cooling Systems, which can be air or water-cooled. Graphite oxidation during air or water ingress can be modelled, including the effects of added combustion products to the gas flow and the additional chemical energy release. A point kinetics model is available for analyzing reactivity excursions; for instance due to water ingress, and also for hypothetical no-scram scenarios. For most HTGR transients, which generally range over hours, a user-selected nodalization of the core in r-z geometry is used. However, a separate model of heat transfer in the symmetry element of each fuel element is also available for very rapid transients. This model can be applied coupled to the traditional coarser r-z nodalization. This report described the mathematical models used in the code and the method of solution. It describes the code and its various sub-elements. Details of the input data and file usage, with file formats, is given for the code, as well as for several preprocessing and postprocessing options. The THATCH model of the currently applicable 350 MW{sub th} reactor is described. Input data for four sample cases are given with output available in fiche form. Installation requirements and code limitations, as well as the most common error indications are listed. 31 refs., 23 figs., 32 tabs.

  6. An Introduction to Thermodynamic Performance Analysis of Aircraft Gas Turbine Engine Cycles Using the Numerical Propulsion System Simulation Code

    Science.gov (United States)

    Jones, Scott M.

    2007-01-01

    This document is intended as an introduction to the analysis of gas turbine engine cycles using the Numerical Propulsion System Simulation (NPSS) code. It is assumed that the analyst has a firm understanding of fluid flow, gas dynamics, thermodynamics, and turbomachinery theory. The purpose of this paper is to provide for the novice the information necessary to begin cycle analysis using NPSS. This paper and the annotated example serve as a starting point and by no means cover the entire range of information and experience necessary for engine performance simulation. NPSS syntax is presented but for a more detailed explanation of the code the user is referred to the NPSS User Guide and Reference document (ref. 1).

  7. Numerical Zooming Between a NPSS Engine System Simulation and a One-Dimensional High Compressor Analysis Code

    Science.gov (United States)

    Follen, Gregory; auBuchon, M.

    2000-01-01

    Within NASA's High Performance Computing and Communication (HPCC) program, NASA Glenn Research Center is developing an environment for the analysis/design of aircraft engines called the Numerical Propulsion System Simulation (NPSS). NPSS focuses on the integration of multiple disciplines such as aerodynamics, structures, and heat transfer along with the concept of numerical zooming between zero-dimensional to one-, two-, and three-dimensional component engine codes. In addition, the NPSS is refining the computing and communication technologies necessary to capture complex physical processes in a timely and cost-effective manner. The vision for NPSS is to create a "numerical test cell" enabling full engine simulations overnight on cost-effective computing platforms. Of the different technology areas that contribute to the development of the NPSS Environment, the subject of this paper is a discussion on numerical zooming between a NPSS engine simulation and higher fidelity representations of the engine components (fan, compressor, burner, turbines, etc.). What follows is a description of successfully zooming one-dimensional (row-by-row) high-pressure compressor analysis results back to a zero-dimensional NPSS engine simulation and a discussion of the results illustrated using an advanced data visualization tool. This type of high fidelity system-level analysis, made possible by the zooming capability of the NPSS, will greatly improve the capability of the engine system simulation and increase the level of virtual test conducted prior to committing the design to hardware.

  8. Reverse engineering nuclear properties from rare earth abundances in the $r$ process

    CERN Document Server

    Mumpower, M R; Surman, R; Steiner, A W

    2016-01-01

    The bulk of the rare earth elements are believed to be synthesized in the rapid neutron capture process or $r$ process of nucleosynthesis. The solar $r$-process residuals show a small peak in the rare earths around $A\\sim 160$, which is proposed to be formed dynamically during the end phase of the $r$ process by a pileup of material. This abundance feature is of particular importance as it is sensitive to both the nuclear physics inputs and the astrophysical conditions of the main $r$ process. We explore the formation of the rare earth peak from the perspective of an inverse problem, using Monte Carlo studies of nuclear masses to investigate the unknown nuclear properties required to best match rare earth abundance sector of the solar isotopic residuals. When nuclear masses are changed, we recalculate the relevant $\\beta$-decay properties and neutron capture rates in the rare earth region. The feedback provided by this observational constraint allows for the reverse engineering of nuclear properties far from ...

  9. Starting Point, Keys and Milestones of a Computer Code for the Simulation of the Behaviour of a Nuclear Fuel Rod

    Directory of Open Access Journals (Sweden)

    Armando C. Marino

    2011-01-01

    Full Text Available The BaCo code (“Barra Combustible” was developed at the Atomic Energy National Commission of Argentina (CNEA for the simulation of nuclear fuel rod behaviour under irradiation conditions. We present in this paper a brief description of the code and the strategy used for the development, improvement, enhancement, and validation of a BaCo during the last 30 years. “Extreme case analysis”, parametric (or sensitivity, probabilistic (or statistic analysis plus the analysis of the fuel performance (full core analysis are the tools developed in the structure of BaCo in order to improve the understanding of the burnup extension in the Atucha I NPP, and the design of advanced fuel elements as CARA and CAREM. The 3D additional tools of BaCo can enhance the understanding of the fuel rod behaviour, the fuel design, and the safety margins. The modular structure of the BaCo code and its detailed coupling of thermo-mechanical and irradiation-induced phenomena make it a powerful tool for the prediction of the influence of material properties on the fuel rod performance and integrity.

  10. A Review on the Regulatory Strategy of Human Factors Engineering Consideration in Pakistan Nuclear Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Sohail, Sabir [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Choi, Seong Nam [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    In this paper, the legal and regulatory infrastructure available in Pakistan for HFE requirements is assessed, and the methodology for strengthening of legal infrastructure is presented. The regulatory strategy on evaluation of HFE consideration should provide reviewers with guidance on review process. Therefore, the suggested methodology is based on preparation of guidance documents such as checklist, working procedures, S and Gs etc.; incorporation of PRM elements in regulatory system; and finally the development of PRM implementation criteria. Altogether, the scheme provide the enhancement in regulatory infrastructure and also the effective and efficient review process. The Three Mile Island (TMI) accident brought the general consensus among the nuclear community on the integration of human factors engineering (HFE) principles in all phases of nuclear power. This notion has further strengthened after the recent Fukushima nuclear accident. Much effort has been put over to incorporate the lesson learned and continuous technical evolution on HFE to device different standards. The total of 174 ergonomics standards are alone identified by Dul et al. (2004) published by International Organization for Standardization (ISO) and the European Committee for Standardization (CEN) and number of standards and HFE guidelines (S and Gs) are also published by organizations like Institute for Electrical and Electronics Engineering (IEEE), International Electrotechnical Commission (IEC), International Atomic Energy Agency (IAEA), United States Nuclear Regulatory Commission (USNRC), etc. The ambition of effective review on HFE integration in nuclear facility might be accomplished through the development of methodology for systematic implementation of S and Gs. Such kind of methodology would also be beneficial for strengthening the regulatory framework and practices for countries new in the nuclear arena and with small scale nuclear program. The objective of paper is to review the

  11. OECD/NEA International Benchmark exercises: Validation of CFD codes applied nuclear industry; OECD/NEA internatiion Benchmark exercices: La validacion de los codigos CFD aplicados a la industria nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Pena-Monferrer, C.; Miquel veyrat, A.; Munoz-Cobo, J. L.; Chiva Vicent, S.

    2016-08-01

    In the recent years, due, among others, the slowing down of the nuclear industry, investment in the development and validation of CFD codes, applied specifically to the problems of the nuclear industry has been seriously hampered. Thus the International Benchmark Exercise (IBE) sponsored by the OECD/NEA have been fundamental to analyze the use of CFD codes in the nuclear industry, because although these codes are mature in many fields, still exist doubts about them in critical aspects of thermohydraulic calculations, even in single-phase scenarios. The Polytechnic University of Valencia (UPV) and the Universitat Jaume I (UJI), sponsored by the Nuclear Safety Council (CSN), have actively participated in all benchmark's proposed by NEA, as in the expert meetings,. In this paper, a summary of participation in the various IBE will be held, describing the benchmark itself, the CFD model created for it, and the main conclusions. (Author)

  12. Reactivity effects in VVER-1000 of the third unit of the kalinin nuclear power plant at physical start-up. Computations in ShIPR intellectual code system with library of two-group cross sections generated by UNK code

    Science.gov (United States)

    Zizin, M. N.; Zimin, V. G.; Zizina, S. N.; Kryakvin, L. V.; Pitilimov, V. A.; Tereshonok, V. A.

    2010-12-01

    The ShIPR intellectual code system for mathematical simulation of nuclear reactors includes a set of computing modules implementing the preparation of macro cross sections on the basis of the two-group library of neutron-physics cross sections obtained for the SKETCH-N nodal code. This library is created by using the UNK code for 3D diffusion computation of first VVER-1000 fuel loadings. Computation of neutron fields in the ShIPR system is performed using the DP3 code in the two-group diffusion approximation in 3D triangular geometry. The efficiency of all groups of control rods for the first fuel loading of the third unit of the Kalinin Nuclear Power Plant is computed. The temperature, barometric, and density effects of reactivity as well as the reactivity coefficient due to the concentration of boric acid in the reactor were computed additionally. Results of computations are compared with the experiment.

  13. Use of liquid metals in nuclear and thermonuclear engineering, and in other innovative technologies

    Science.gov (United States)

    Rachkov, V. I.; Arnol'dov, M. N.; Efanov, A. D.; Kalyakin, S. G.; Kozlov, F. A.; Loginov, N. I.; Orlov, Yu. I.; Sorokin, A. P.

    2014-05-01

    By now, a good deal of experience has been gained with using liquid metals as coolants in nuclear power installations; extensive knowledge has been gained about the physical, thermophysical, and physicochemical properties of these coolants; and the scientific principles and a set of methods and means for handling liquid metals as coolants for nuclear power installations have been elaborated. Prototype and commercialgrade sodium-cooled NPP power units have been developed, including the BOR-60, BN-350, and BN-600 power units (the Soviet Union); the Rapsodie, Phenix, and Superphenix power units (France), the EBR-II power unit (the United States); and the PFR power unit (the United Kingdom). In Russia, dedicated nuclear power installations have been constructed, including those with a lead-bismuth coolant for nuclear submarines and with sodium-potassium alloy for spacecraft (the Buk and Topol installations), which have no analogs around the world. Liquid metals (primarily lithium and its alloy with lead) hold promise for use in thermonuclear power engineering, where they can serve not only as a coolant, but also as tritium-producing medium. In this article, the physicochemical properties of liquid metal coolants, as well as practical experience gained from using them in nuclear and thermonuclear power engineering and in innovative technologies are considered, and the lines of further research works are formulated. New results obtained from investigations carried out on the Pb-Bi and Pb for the SVBR and BREST fast-neutron reactors (referred to henceforth as fast reactors) and for controlled accelerator systems are described.

  14. Interdisciplinary Team-Teaching Experience for a Computer and Nuclear Energy Course for Electrical and Computer Engineering Students

    Science.gov (United States)

    Kim, Charles; Jackson, Deborah; Keiller, Peter

    2016-01-01

    A new, interdisciplinary, team-taught course has been designed to educate students in Electrical and Computer Engineering (ECE) so that they can respond to global and urgent issues concerning computer control systems in nuclear power plants. This paper discusses our experience and assessment of the interdisciplinary computer and nuclear energy…

  15. Non-Standard Genetic Codes Define New Concepts for Protein Engineering

    OpenAIRE

    Bezerra, Ana R; Guimarães, Ana R.; Santos, Manuel A. S.

    2015-01-01

    The essential feature of the genetic code is the strict one-to-one correspondence between codons and amino acids. The canonical code consists of three stop codons and 61 sense codons that encode 20% of the amino acid repertoire observed in nature. It was originally designated as immutable and universal due to its conservation in most organisms, but sequencing of genes from the human mitochondrial genomes revealed deviations in codon assignments. Since then, alternative codes have been reporte...

  16. Uncertainty propagation in a 3-D thermal code for performance assessment of a nuclear waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Dutfoy, A. [Electricite de France (EDF), Research and Development Div., Safety and Reliability Branch, ESF, 92 - Clamart (France); Ritz, J.B. [Electricite de France (EDF), Research and Development Div., Fluid Mechanics and Heat Transfer, MFTT, 78 - Chatou (France)

    2001-07-01

    Given the very large time scale involved, the performance assessment of a nuclear waste repository requires numerical modelling. Because we are uncertain of the exact value of the input parameters, we have to analyse the impact of these uncertainties on the outcome of the physical models. The EDF Division Research and Development has set a reliability method to propagate these uncertainties or variability through models which requires much less physical simulations than the usual simulation methods. We apply the reliability method MEFISTO to a base case modelling the heat transfers in a virtual disposal in the future site of the French underground research laboratory, in the East of France. This study is led in collaboration with ANDRA which is the French Nuclear Waste Management Agency. With this exercise, we want to evaluate the thermal behaviour of a concept related to the variation of physical parameters and their uncertainty. (author)

  17. Joining the Nuclear Renaissance with the Engineering Business Unit of AREVA

    Energy Technology Data Exchange (ETDEWEB)

    Hubert, Nathalie; Menguy, Stephane [SGN, AREVA Group, 1 rue des Herons, 78182 Saint-Quentin en Yvelines Cedex (France); Valery, Jean-Francois [AREVA NC, AREVA Group, Tour AREVA, 1 place de la Coupole, 92084 Paris La Defense Cedex (France)

    2008-07-01

    The reality of the nuclear renaissance is no longer a question. All over the world, new nuclear plants are going to be deployed; the whole fuel cycle has to be adjusted to fulfil their needs, the front-end to produce the fuel and the back-end to properly manage radioactive waste. AREVA fuel cycle engineering teams have been involved in the design of a variety of industrial plants covering the entire fuel cycle for 50 years. The consistency of the French nuclear policy has been a major factor to acquire and renew the competencies and workforce of AREVA Engineering Business Unit. Our partnership with our customers, French ones but also Japanese, Americans and from other countries, has led us to develop a comprehensive approach of the services that we can deliver, in order to give them the best answer. SGN teams have been involved in the R and D phases in order to take into account the industrialisation aspects as early as possible, and our work does not end with the delivery of the plants; it includes assistance to the operators to optimise and keep their facilities in line with the changing rules and constraints, which ensures the integration of a wide operational experience feedback and the ability to design flexible facilities. This paper will present through our experience how this global approach has been developed and continuously improved and how we are preparing our teams to be ready to answer to the coming needs. (authors)

  18. The Need for Cyber-Informed Engineering Expertise for Nuclear Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, Robert Stephen [Idaho National Laboratory

    2015-12-01

    Engineering disciplines may not currently understand or fully embrace cyber security aspects as they apply towards analysis, design, operation, and maintenance of nuclear research reactors. Research reactors include a wide range of diverse co-located facilities and designs necessary to meet specific operational research objectives. Because of the nature of research reactors (reduced thermal energy and fission product inventory), hazards and risks may not have received the same scrutiny as normally associated with power reactors. Similarly, security may not have been emphasized either. However, the lack of sound cybersecurity defenses may lead to both safety and security impacts. Risk management methodologies may not contain the foundational assumptions required to address the intelligent adversary’s capabilities in malevolent cyber attacks. Although most research reactors are old and may not have the same digital footprint as newer facilities, any digital instrument and control function must be considered as a potential attack platform that can lead to sabotage or theft of nuclear material, especially for some research reactors that store highly enriched uranium. This paper will provide a discussion about the need for cyber-informed engineering practices that include the entire engineering lifecycle. Cyber-informed engineering as referenced in this paper is the inclusion of cybersecurity aspects into the engineering process. A discussion will consider several attributes of this process evaluating the long-term goal of developing additional cyber safety basis analysis and trust principles. With a culture of free information sharing exchanges, and potentially a lack of security expertise, new risk analysis and design methodologies need to be developed to address this rapidly evolving (cyber) threatscape.

  19. A versatile, bar-coded nuclear marker/reporter for live cell fluorescent and multiplexed high content imaging.

    Directory of Open Access Journals (Sweden)

    Irina Krylova

    Full Text Available The screening of large numbers of compounds or siRNAs is a mainstay of both academic and pharmaceutical research. Most screens test those interventions against a single biochemical or cellular output whereas recording multiple complementary outputs may be more biologically relevant. High throughput, multi-channel fluorescence microscopy permits multiple outputs to be quantified in specific cellular subcompartments. However, the number of distinct fluorescent outputs available remains limited. Here, we describe a cellular bar-code technology in which multiple cell-based assays are combined in one well after which each assay is distinguished by fluorescence microscopy. The technology uses the unique fluorescent properties of assay-specific markers comprised of distinct combinations of different 'red' fluorescent proteins sandwiched around a nuclear localization signal. The bar-code markers are excited by a common wavelength of light but distinguished ratiometrically by their differing relative fluorescence in two emission channels. Targeting the bar-code to cell nuclei enables individual cells expressing distinguishable markers to be readily separated by standard image analysis programs. We validated the method by showing that the unique responses of different cell-based assays to specific drugs are retained when three assays are co-plated and separated by the bar-code. Based upon those studies, we discuss a roadmap in which even more assays may be combined in a well. The ability to analyze multiple assays simultaneously will enable screens that better identify, characterize and distinguish hits according to multiple biologically or clinically relevant criteria. These capabilities also enable the re-creation of complex mixtures of cell types that is emerging as a central area of interest in many fields.

  20. NUPEC (Nuclear Power Engineering Corporation) annual report 1998, activities in fiscal 1997

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-12-01

    NUPEC was founded in March, 1976 under the initiative of scholars and private corporations including electric power companies, electric machinery and general construction companies. Ever since, NUPEC has been proceeding with its operations to meet the needs of the times with the support and cooperation of the government and academic circles. The specific activities so far include solving problems in the initial stage of light water reactors(LWRs) operation, engineering tests for improvement and standardization programs to develop domestic LWRs, seismic technology development and testing, assistance with accident analysis during safety examinations by government agency, human factor research, safety analysis of nuclear facilities, research of safety-related information, countermeasures for aging of LWRs and public acceptance activities. For such purposes, NUPEC has engineering laboratories in Tadotsu, Takasago, Isogo and Katsuta as well as a high performance parallel computer system for safety analysis at its headquarters. Among these facilities, the large-scale high-performance shaking table at Tadotsu Engineering Laboratory is attracting international attention for its capability for seismic testing. NUPEC is actively promoting international cooperation with international organizations and partners in the U.S., France, Germany, Russia and Asian countries through joint projects, information exchange, etc. NUPEC`s testing and analysis have contributed to improvement of safety and credibility of Nuclear power generation and to establishment and improvement of Japanese-originated LWR technology. A summary of our achievements in fiscal 1997 is presented in this annual report. (J.P.N.)

  1. DIONISIO 2.0: New version of the code for simulating a whole nuclear fuel rod under extended irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Soba, Alejandro, E-mail: soba@cnea.gov.ar; Denis, Alicia

    2015-10-15

    Highlights: • A new version of the DIONISIO code is developed. • DIONISIO is devoted to simulating the behavior of a nuclear fuel rod in operation. • The formerly two-dimensional simulation of a pellet-cladding segment is now extended to the whole rod length. • An acceptable and more realistic agreement with experimental data is obtained. • The prediction range of our code is extended up to average burnup of 60 MWd/kgU. - Abstract: The version 2.0 of the DIONISIO code, that incorporates diverse new aspects, has been recently developed. One of them is referred to the code architecture that allows taking into account the axial variation of the conditions external to the rod. With this purpose, the rod is divided into a number of axial segments. In each one the program considers the system formed by a pellet and the corresponding cladding portion and solves the numerous phenomena that take place under the local conditions of linear power and coolant temperature, which are given as input parameters. To do this a bi-dimensional domain in the r–z plane is considered where cylindrical symmetry and also symmetry with respect to the pellet mid-plane are assumed. The results obtained for this representative system are assumed valid for the complete segment. The program thus produces in each rod section the values of the temperature, stress, strain, among others as outputs, as functions of the local coordinates r and z. Then, the general rod parameters (internal rod pressure, amount of fission gas released, pellet stack elongation, etc.) are evaluated. Moreover, new calculation tools designed to extend the application range of the code to high burnup, which were reported elsewhere, have also been incorporated to DIONISIO 2.0 in recent times. With these improvements, the code results are compared with some 33 experiments compiled in the IFPE data base, that cover more than 380 fuel rods irradiated up to average burnup levels of 40–60 MWd/kgU. The results of these

  2. ASME Code Efforts Supporting HTGRs

    Energy Technology Data Exchange (ETDEWEB)

    D.K. Morton

    2011-09-01

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

  3. ASME Code Efforts Supporting HTGRs

    Energy Technology Data Exchange (ETDEWEB)

    D.K. Morton

    2010-09-01

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

  4. ASME Code Efforts Supporting HTGRs

    Energy Technology Data Exchange (ETDEWEB)

    D.K. Morton

    2012-09-01

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

  5. Users Guide to SAMINT: A Code for Nuclear Data Adjustment with SAMMY Based on Integral Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Sobes, Vladimir [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Leal, Luiz C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Arbanas, Goran [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-10-01

    The purpose of this project is to couple differential and integral data evaluation in a continuous-energy framework. More specifically, the goal is to use the Generalized Linear Least Squares methodology employed in TSURFER to update the parameters of a resolved resonance region evaluation directly. Recognizing that the GLLS methodology in TSURFER is identical to the mathematical description of the simple Bayesian updating carried out in SAMMY, the computer code SAMINT was created to help use the mathematical machinery of SAMMY to update resolved resonance parameters based on integral data. Minimal modifications of SAMMY are required when used with SAMINT to make resonance parameter updates based on integral experimental data.

  6. TLD environmental monitoring at the Institute of Nuclear Engineering in Brazil.

    Science.gov (United States)

    Taam, I H; da Rosa, L A R; Crispim, V R

    2008-09-01

    Since 2003 the Institute of Nuclear Engineering in Rio de Janeiro city, Brazil, operates a new cyclotron, RDS-111, to produce (18)F-Fluorodeoxyglucose to be used in nuclear medicine. Additionally, the IEN radioactive waste repository has been enlarged during the past last years, receiving a considerable amount of radioactive materials. Therefore, it became necessary to evaluate a possible increase of the environmental gamma exposure rates at the institute site due to the operation of the new accelerator and the enlargement of the institute waste repository as well. LiF:Mg,Cu,P, TLD-100H, and TL detectors were employed for environmental kerma rate evaluation and the results were compared with previous results obtained before the RDS-111 operation initialisation and the enlargement of IEN waste repository. No significant contribution for the enhancement of environmental gamma kerma rates was detected.

  7. Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) verification and validation plan. version 1.

    Energy Technology Data Exchange (ETDEWEB)

    Bartlett, Roscoe Ainsworth; Arguello, Jose Guadalupe, Jr.; Urbina, Angel; Bouchard, Julie F.; Edwards, Harold Carter; Freeze, Geoffrey A.; Knupp, Patrick Michael; Wang, Yifeng; Schultz, Peter Andrew; Howard, Robert (Oak Ridge National Laboratory, Oak Ridge, TN); McCornack, Marjorie Turner

    2011-01-01

    The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. To meet this objective, NEAMS Waste IPSC M&S capabilities will be applied to challenging spatial domains, temporal domains, multiphysics couplings, and multiscale couplings. A strategic verification and validation (V&V) goal is to establish evidence-based metrics for the level of confidence in M&S codes and capabilities. Because it is economically impractical to apply the maximum V&V rigor to each and every M&S capability, M&S capabilities will be ranked for their impact on the performance assessments of various components of the repository systems. Those M&S capabilities with greater impact will require a greater level of confidence and a correspondingly greater investment in V&V. This report includes five major components: (1) a background summary of the NEAMS Waste IPSC to emphasize M&S challenges; (2) the conceptual foundation for verification, validation, and confidence assessment of NEAMS Waste IPSC M&S capabilities; (3) specifications for the planned verification, validation, and confidence-assessment practices; (4) specifications for the planned evidence information management system; and (5) a path forward for the incremental implementation of this V&V plan.

  8. Simulation about Self-absorption of Ni-63 Nuclear Battery Using Monte Carlo Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Ho; Kim, Ji Hyun [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2014-05-15

    The radioisotope batteries have an energy density of 100-10000 times greater than chemical batteries. Also, Li ion battery has the fundamental problems such as short life time and requires recharge system. In addition to these things, the existing batteries are hard to operate at internal human body, national defense arms or space environment. Since the development of semiconductor process and materials technology, the micro device is much more integrated. It is expected that, based on new semiconductor technology, the conversion device efficiency of betavoltaic battery will be highly increased. Furthermore, the radioactivity from the beta particle cannot penetrate a skin of human body, so it is safer than Li battery which has the probability to explosion. In the other words, the interest for radioisotope battery is increased because it can be applicable to an artificial internal organ power source without recharge and replacement, micro sensor applied to arctic and special environment, small size military equipment and space industry. However, there is not enough data for beta particle fluence from radioisotope source using nuclear battery. Beta particle fluence directly influences on battery efficiency and it is seriously affected by radioisotope source thickness because of self-absorption effect. Therefore, in this article, we present a basic design of Ni-63 nuclear battery and simulation data of beta particle fluence with various thickness of radioisotope source and design of battery.

  9. Calculation of electron and isotopes dose point kernels with FLUKA Monte Carlo code for dosimetry in nuclear medicine therapy.

    Science.gov (United States)

    Botta, F; Mairani, A; Battistoni, G; Cremonesi, M; Di Dia, A; Fassò, A; Ferrari, A; Ferrari, M; Paganelli, G; Pedroli, G; Valente, M

    2011-07-01

    The calculation of patient-specific dose distribution can be achieved by Monte Carlo simulations or by analytical methods. In this study, FLUKA Monte Carlo code has been considered for use in nuclear medicine dosimetry. Up to now, FLUKA has mainly been dedicated to other fields, namely high energy physics, radiation protection, and hadrontherapy. When first employing a Monte Carlo code for nuclear medicine dosimetry, its results concerning electron transport at energies typical of nuclear medicine applications need to be verified. This is commonly achieved by means of calculation of a representative parameter and comparison with reference data. Dose point kernel (DPK), quantifying the energy deposition all around a point isotropic source, is often the one. FLUKA DPKS have been calculated in both water and compact bone for monoenergetic electrons (10-3 MeV) and for beta emitting isotopes commonly used for therapy (89Sr, 90Y, 131I 153Sm, 177Lu, 186Re, and 188Re). Point isotropic sources have been simulated at the center of a water (bone) sphere, and deposed energy has been tallied in concentric shells. FLUKA outcomes have been compared to PENELOPE v.2008 results, calculated in this study as well. Moreover, in case of monoenergetic electrons in water, comparison with the data from the literature (ETRAN, GEANT4, MCNPX) has been done. Maximum percentage differences within 0.8.RCSDA and 0.9.RCSDA for monoenergetic electrons (RCSDA being the continuous slowing down approximation range) and within 0.8.X90 and 0.9.X90 for isotopes (X90 being the radius of the sphere in which 90% of the emitted energy is absorbed) have been computed, together with the average percentage difference within 0.9.RCSDA and 0.9.X90 for electrons and isotopes, respectively. Concerning monoenergetic electrons, within 0.8.RCSDA (where 90%-97% of the particle energy is deposed), FLUKA and PENELOPE agree mostly within 7%, except for 10 and 20 keV electrons (12% in water, 8.3% in bone). The

  10. A binary mixed integer coded genetic algorithm for multi-objective optimization of nuclear research reactor fuel reloading

    Energy Technology Data Exchange (ETDEWEB)

    Binh, Do Quang [University of Technical Education Ho Chi Minh City (Viet Nam); Huy, Ngo Quang [University of Industry Ho Chi Minh City (Viet Nam); Hai, Nguyen Hoang [Centre for Research and Development of Radiation Technology, Ho Chi Minh City (Viet Nam)

    2014-12-15

    This paper presents a new approach based on a binary mixed integer coded genetic algorithm in conjunction with the weighted sum method for multi-objective optimization of fuel loading patterns for nuclear research reactors. The proposed genetic algorithm works with two types of chromosomes: binary and integer chromosomes, and consists of two types of genetic operators: one working on binary chromosomes and the other working on integer chromosomes. The algorithm automatically searches for the most suitable weighting factors of the weighting function and the optimal fuel loading patterns in the search process. Illustrative calculations are implemented for a research reactor type TRIGA MARK II loaded with the Russian VVR-M2 fuels. Results show that the proposed genetic algorithm can successfully search for both the best weighting factors and a set of approximate optimal loading patterns that maximize the effective multiplication factor and minimize the power peaking factor while satisfying operational and safety constraints for the research reactor.

  11. Calculation of electron and isotopes dose point kernels with fluka Monte Carlo code for dosimetry in nuclear medicine therapy

    Energy Technology Data Exchange (ETDEWEB)

    Botta, F; Di Dia, A; Pedroli, G; Mairani, A; Battistoni, G; Fasso, A; Ferrari, A; Ferrari, M; Paganelli, G

    2011-06-01

    The calculation of patient-specific dose distribution can be achieved by Monte Carlo simulations or by analytical methods. In this study, fluka Monte Carlo code has been considered for use in nuclear medicine dosimetry. Up to now, fluka has mainly been dedicated to other fields, namely high energy physics, radiation protection, and hadrontherapy. When first employing a Monte Carlo code for nuclear medicine dosimetry, its results concerning electron transport at energies typical of nuclear medicine applications need to be verified. This is commonly achieved by means of calculation of a representative parameter and comparison with reference data. Dose point kernel (DPK), quantifying the energy deposition all around a point isotropic source, is often the one.Methods: fluka DPKs have been calculated in both water and compact bone for monoenergetic electrons (10–3 MeV) and for beta emitting isotopes commonly used for therapy (89Sr, 90Y, 131I, 153Sm, 177Lu, 186Re, and 188Re). Point isotropic sources have been simulated at the center of a water (bone) sphere, and deposed energy has been tallied in concentric shells. fluka outcomes have been compared to penelope v.2008 results, calculated in this study as well. Moreover, in case of monoenergetic electrons in water, comparison with the data from the literature (etran, geant4, mcnpx) has been done. Maximum percentage differences within 0.8·RCSDA and 0.9·RCSDA for monoenergetic electrons (RCSDA being the continuous slowing down approximation range) and within 0.8·X90 and 0.9·X90 for isotopes (X90 being the radius of the sphere in which 90% of the emitted energy is absorbed) have been computed, together with the average percentage difference within 0.9·RCSDA and 0.9·X90 for electrons and isotopes, respectively.Results: Concerning monoenergetic electrons, within 0.8·RCSDA (where 90%–97% of the particle energy is deposed), fluka and penelope agree mostly within 7%, except for 10 and 20 keV electrons (12% in water, 8

  12. Calculation of electron and isotopes dose point kernels with fluka Monte Carlo code for dosimetry in nuclear medicine therapy

    Energy Technology Data Exchange (ETDEWEB)

    Botta, F.; Mairani, A.; Battistoni, G.; Cremonesi, M.; Di Dia, A.; Fasso, A.; Ferrari, A.; Ferrari, M.; Paganelli, G.; Pedroli, G.; Valente, M. [Medical Physics Department, European Institute of Oncology, Via Ripamonti 435, 20141 Milan (Italy); Istituto Nazionale di Fisica Nucleare (I.N.F.N.), Via Celoria 16, 20133 Milan (Italy); Medical Physics Department, European Institute of Oncology, Via Ripamonti 435, 20141 Milan (Italy); Jefferson Lab, 12000 Jefferson Avenue, Newport News, Virginia 23606 (United States); CERN, 1211 Geneva 23 (Switzerland); Medical Physics Department, European Institute of Oncology, Milan (Italy); Nuclear Medicine Department, European Institute of Oncology, Via Ripamonti 435, 2014 Milan (Italy); Medical Physics Department, European Institute of Oncology, Via Ripamonti 435, 20141 Milan (Italy); FaMAF, Universidad Nacional de Cordoba and CONICET, Cordoba, Argentina C.P. 5000 (Argentina)

    2011-07-15

    Purpose: The calculation of patient-specific dose distribution can be achieved by Monte Carlo simulations or by analytical methods. In this study, fluka Monte Carlo code has been considered for use in nuclear medicine dosimetry. Up to now, fluka has mainly been dedicated to other fields, namely high energy physics, radiation protection, and hadrontherapy. When first employing a Monte Carlo code for nuclear medicine dosimetry, its results concerning electron transport at energies typical of nuclear medicine applications need to be verified. This is commonly achieved by means of calculation of a representative parameter and comparison with reference data. Dose point kernel (DPK), quantifying the energy deposition all around a point isotropic source, is often the one. Methods: fluka DPKs have been calculated in both water and compact bone for monoenergetic electrons (10{sup -3} MeV) and for beta emitting isotopes commonly used for therapy ({sup 89}Sr, {sup 90}Y, {sup 131}I, {sup 153}Sm, {sup 177}Lu, {sup 186}Re, and {sup 188}Re). Point isotropic sources have been simulated at the center of a water (bone) sphere, and deposed energy has been tallied in concentric shells. fluka outcomes have been compared to penelope v.2008 results, calculated in this study as well. Moreover, in case of monoenergetic electrons in water, comparison with the data from the literature (etran, geant4, mcnpx) has been done. Maximum percentage differences within 0.8{center_dot}R{sub CSDA} and 0.9{center_dot}R{sub CSDA} for monoenergetic electrons (R{sub CSDA} being the continuous slowing down approximation range) and within 0.8{center_dot}X{sub 90} and 0.9{center_dot}X{sub 90} for isotopes (X{sub 90} being the radius of the sphere in which 90% of the emitted energy is absorbed) have been computed, together with the average percentage difference within 0.9{center_dot}R{sub CSDA} and 0.9{center_dot}X{sub 90} for electrons and isotopes, respectively. Results: Concerning monoenergetic electrons

  13. Concept study of a hydrogen containment process during nuclear thermal engine ground testing

    Science.gov (United States)

    Wang, Ten-See; Stewart, Eric T.; Canabal, Francisco

    A new hydrogen containment process was proposed for ground testing of a nuclear thermal engine. It utilizes two thermophysical steps to contain the hydrogen exhaust. First, the decomposition of hydrogen through oxygen-rich combustion at higher temperature; second, the recombination of remaining hydrogen with radicals at low temperature. This is achieved with two unit operations: an oxygen-rich burner and a tubular heat exchanger. A computational fluid dynamics methodology was used to analyze the entire process on a three-dimensional domain. The computed flammability at the exit of the heat exchanger was less than the lower flammability limit, confirming the hydrogen containment capability of the proposed process.

  14. Idaho Nuclear Technology and Engineering Center Newly Generated Liquid Waste Demonstration Project Feasibility Study

    Energy Technology Data Exchange (ETDEWEB)

    Herbst, A.K.

    2000-02-01

    A research, development, and demonstration project for the grouting of newly generated liquid waste (NGLW) at the Idaho Nuclear Technology and Engineering Center is considered feasible. NGLW is expected from process equipment waste, decontamination waste, analytical laboratory waste, fuel storage basin waste water, and high-level liquid waste evaporator condensate. The potential grouted waste would be classed as mixed low-level waste, stabilized and immobilized to meet RCRA LDR disposal in a grouting process in the CPP-604 facility, and then transported to the state.

  15. The use of low power dual mode nuclear thermal rocket engines to support space exploration missions

    Science.gov (United States)

    Zubrin, Robert M.

    1991-01-01

    The evolution of dual mode concepts is presented, focusing on advantages and problems associated with both low and high temperature dual mode conversion systems. It is concluded that dual mode nuclear thermal rocket (NTR) systems using high temperature Brayton cycle conversion technology offer a high payoff enhancement of conventional NTR, with a comparatively minor increase of technological challenge. It is recommended that NTR engines be designed so that dual mode conversion systems can be attached to them in a modular way, thus enabling the production of electric power on all missions where it is needed.

  16. Calcine Waste Storage at the Idaho Nuclear Technology and Engineering Center

    Energy Technology Data Exchange (ETDEWEB)

    M. D. Staiger

    1999-06-01

    A potential option in the program for long-term management of high-level wastes at the Idaho Nuclear Technology and Engineering Center (INTEC), at the Idaho National Engineering and Environmental Laboratory, calls for retrieving calcine waste and converting it to a more stable and less dispersible form. An inventory of calcine produced during the period December 1963 to May 1999 has been prepared based on calciner run, solids storage facilities operating, and miscellaneous operational information, which gives the range of chemical compositions of calcine waste stored at INTEC. Information researched includes calciner startup data, waste solution analyses and volumes calcined, calciner operating schedules, solids storage bin capacities, calcine storage bin distributor systems, and solids storage bin design and temperature monitoring records. Unique information on calcine solids storage facilities design of potential interest to remote retrieval operators is given.

  17. Engineering of Deinococcus radiodurans R1 for bioprecipitation of uranium from dilute nuclear waste.

    Science.gov (United States)

    Appukuttan, Deepti; Rao, Amara Sambasiva; Apte, Shree Kumar

    2006-12-01

    Genetic engineering of radiation-resistant organisms to recover radionuclides/heavy metals from radioactive wastes is an attractive proposition. We have constructed a Deinococcus radiodurans strain harboring phoN, a gene encoding a nonspecific acid phosphatase, obtained from a local isolate of Salmonella enterica serovar Typhi. The recombinant strain expressed an approximately 27-kDa active PhoN protein and efficiently precipitated over 90% of the uranium from a 0.8 mM uranyl nitrate solution in 6 h. The engineered strain retained uranium bioprecipitation ability even after exposure to 6 kGy of 60Co gamma rays. The PhoN-expressing D. radiodurans offers an effective and eco-friendly in situ approach to biorecovery of uranium from dilute nuclear waste.

  18. Design and analysis of a single stage to orbit nuclear thermal rocket reactor engine

    Energy Technology Data Exchange (ETDEWEB)

    Labib, Satira, E-mail: Satira.Labib@duke-energy.com; King, Jeffrey, E-mail: kingjc@mines.edu

    2015-06-15

    Graphical abstract: - Highlights: • Three NTR reactors are optimized for the single stage launch of 1–15 MT payloads. • The proposed rocket engines have specific impulses in excess of 700 s. • Reactivity and submersion criticality requirements are satisfied for each reactor. - Abstract: Recent advances in the development of high power density fuel materials have renewed interest in nuclear thermal rockets (NTRs) as a viable propulsion technology for future space exploration. This paper describes the design of three NTR reactor engines designed for the single stage to orbit launch of payloads from 1 to 15 metric tons. Thermal hydraulic and rocket engine analyses indicate that the proposed rocket engines are able to reach specific impulses in excess of 800 s. Neutronics analyses performed using MCNP5 demonstrate that the hot excess reactivity, shutdown margin, and submersion criticality requirements are satisfied for each NTR reactor. The reactors each consist of a 40 cm diameter core packed with hexagonal tungsten cermet fuel elements. The core is surrounded by radial and axial beryllium reflectors and eight boron carbide control drums. The 40 cm long reactor meets the submersion criticality requirements (a shutdown margin of at least $1 subcritical in all submersion scenarios) with no further modifications. The 80 and 120 cm long reactors include small amounts of gadolinium nitride as a spectral shift absorber to keep them subcritical upon submersion in seawater or wet sand following a launch abort.

  19. FX2-TH: a two-dimensional nuclear reactor kinetics code with thermal-hydraulic feedback

    Energy Technology Data Exchange (ETDEWEB)

    Shober, R.A.; Daly, T.A.; Ferguson, D.R.

    1978-10-01

    FX2-TH is a two-dimensional, time-dependent nuclear reactor kinetics program with thermal and hydraulic feedback. The neutronics model used is multigroup neutron diffusion theory. The following geometry options are available: x, r, x-y, r-z, theta-r, and triangular. FX2-TH contains two basic thermal and hydraulic models: a simple adiabatic fuel temperature calculation, and a more detailed model consisting of an explicit representation of a fuel pin, gap, clad, and coolant. FX2-TH allows feedback effects from both fuel temperature (Doppler) and coolant temperature (density) changes. FX2-TH will calculate a consistent set of steady state conditions by iterating between the neutronics and thermal-hydraulics until convergence is reached. The time-dependent calculation is performed by the use of the improved quasistatic method. A disk editing capability is available. FX2-TH is operational on IBM system 360 or 370 computers and on the CDC 7600.

  20. Thermal-hydraulic analysis techniques for axisymmetric pebble bed nuclear reactor cores. [PEBBLE code

    Energy Technology Data Exchange (ETDEWEB)

    Stroh, K.R.

    1979-03-01

    The pebble bed reactor's cylindrical core volume contains a random bed of small, spherical fuel-moderator elements. These graphite spheres, containing a central region of dispersed coated-particle fissile and fertile material, are cooled by high pressure helium flowing through the connected interstitial voids. A mathematical model and numerical solution technique have been developed which allow calculation of macroscopic values of thermal-hydraulic variables in an axisymmetric pebble bed nuclear reactor core. The computer program PEBBLE is based on a mathematical model which treats the bed macroscopically as a generating, conducting porous medium. The steady-state model uses a nonlinear Forchheimer-type relation between the coolant pressure gradient and mass flux, with newly derived coefficients for the linear and quadratic resistance terms. The remaining equations in the model make use of mass continuity, and thermal energy balances for the solid and fluid phases.

  1. Comparable frequencies of coding mutations and loss of imprinting in human pluripotent cells derived by nuclear transfer and defined factors.

    Science.gov (United States)

    Johannesson, Bjarki; Sagi, Ido; Gore, Athurva; Paull, Daniel; Yamada, Mitsutoshi; Golan-Lev, Tamar; Li, Zhe; LeDuc, Charles; Shen, Yufeng; Stern, Samantha; Xu, Nanfang; Ma, Hong; Kang, Eunju; Mitalipov, Shoukhrat; Sauer, Mark V; Zhang, Kun; Benvenisty, Nissim; Egli, Dieter

    2014-11-06

    The recent finding that reprogrammed human pluripotent stem cells can be derived by nuclear transfer into human oocytes as well as by induced expression of defined factors has revitalized the debate on whether one approach might be advantageous over the other. Here we compare the genetic and epigenetic integrity of human nuclear-transfer embryonic stem cell (NT-ESC) lines and isogenic induced pluripotent stem cell (iPSC) lines, derived from the same somatic cell cultures of fetal, neonatal, and adult origin. The two cell types showed similar genome-wide gene expression and DNA methylation profiles. Importantly, NT-ESCs and iPSCs had comparable numbers of de novo coding mutations, but significantly more than parthenogenetic ESCs. As iPSCs, NT-ESCs displayed clone- and gene-specific aberrations in DNA methylation and allele-specific expression of imprinted genes. The occurrence of these genetic and epigenetic defects in both NT-ESCs and iPSCs suggests that they are inherent to reprogramming, regardless of derivation approach.

  2. Educating nuclear engineers at German universities; Die Ausbildung von Kerntechnikern an deutschen Hochschulen

    Energy Technology Data Exchange (ETDEWEB)

    Knorr, J. [Fakultaet Maschinenwesen, Inst. fuer Energietechnik, Lehrstuhl fuer Kernenergietechnik, Technische Univ. Dresden (Germany)

    1995-05-01

    Nuclear technology is a relatively young university discipline. Yet, as a consequence of the declining public acceptance of the peaceful use of nuclear power, its very existence is already being threatened at many universities. However, if Germany needs nuclear power, which undoubtedly is the case, highly qualified, committed experts are required above all. Nuclear technology develops internationally. Consequently, also university education must meet international standards. Generally, university education has been found to be the most effective way of increasing the number of scientific and engineering personnel. Nuclear techniques have meanwhile found acceptance in many other scientific disciplines, thus advancing those branches of science. Teaching needs research; like research in nucelar technology at the national research centers, also the universities are suffering massive financial disadvantages. Research is possible only if outside funds are solicited, which increase dependency and decreases basic research. (orig.) [Deutsch] An den Hochschulen ist das Studium der Kerntechnik noch relativ jung, und doch schon - bedingt durch die sinkende Akzeptanz der friedlichen Nutzung der Kernenergie in der Oeffentlichkeit - an vielen Wissenschaftsstandorten in seiner Existenz bedroht. Wenn Deutschland jedoch die Kernkraft braucht - und daran besteht kein Zweifel - dann werden in erster Linie hochqualifizierte, engagierte Fachleute benoetigt. Die Entwicklung der Kerntechnik vollzieht sich international. Darum muss auch die Hochschulausbildung internationalen Massstaeben genuegen. Die universitaere Ausbildung hat sich generell als die effektivste Form zur Reproduktion wissenschaftlichen und ingenieurtechnischen Personals erwiesen. Kerntechnische Methoden haben Eingang in viele andere Wissenschaftsdisziplinen gefunden und dadurch deren Entwicklung vorangetrieben. Lehre lebt von der Forschung; wie die kerntechnische Forschung in den Grossforschungseinrichtungen, so werden

  3. Convergence Nanorobot Analysis for Radiation Therapy-Industrial Innovations in Nuclear Engineering

    Energy Technology Data Exchange (ETDEWEB)

    Woo, Taeho [Yonsei Univ., Wonju (Korea, Republic of)

    2015-10-15

    The important step of the commercialization is the make the prototype nanorobot where lots of applications could be introduced for the industry. For the much more advanced operations of the nanorobot, it is needed to imagine the strategy for the operation in the non-regular shaped organs like the lung which shows the different feature following breaths. The biological stuffs are usually in the irregular shape and could be changed by the external force or the infected viruses. The biological substance could be made by the amorphous material which is used frequently in the industry. The antibody reaction is a particular matter which could be happen in the human body. So, the adaptations of the nanorobot could be increased for the practical purposed. Fig. 7 is the newly imagined convergence nuclear technology with nanorobotics for nuclear engineering fields in which many kinds of applications are imagined. Following the new applications of the nanorobot, it is possible to challenge for the difficult matters in the conventional nuclear industry. Fig. 8 shows the historic mistakes in commercialized nuclear power plants (NPPs) considering the nuclear reactor analysis and safety system induced by the accident. Firstly, the non-matched flux shapes made by the multiplications of Bessel function and cosine function by the cylindrical core shape, which is different from the spherical or rectangular core shape, couldn't describe the exact flux shape. Secondly, the safety system installed to start in the accident is the piping-based injection equipment. However, the safety injection systems have failed in three major sever accidents as Three Mile Island (TMI), Chernobyl, and Fukushima cases due to the significant piping failures.

  4. Thermohydraulic Design Analysis Modeling for Korea Advanced NUclear Thermal Engine Rocket for Space Application

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Seung Hyun; Choi, Jae Young; Venneria, Paolo F.; Jeong, Yong Hoon; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    Space exploration is a realistic and profitable goal for long-term humanity survival, although the harsh space environment imposes lots of severe challenges to space pioneers. To date, almost all space programs have relied upon Chemical Rockets (CRs) rating superior thrust level to transit from the Earth's surface to its orbit. However, CRs inherently have insurmountable barrier to carry out deep space missions beyond Earth's orbit due to its low propellant efficiency, and ensuing enormous propellant requirement and launch costs. Meanwhile, nuclear rockets typically offer at least two times the propellant efficiency of a CR and thus notably reduce the propellant demand. Particularly, a Nuclear Thermal Rocket (NTR) is a leading candidate for near-term manned missions to Mars and beyond because it satisfies a relatively high thrust as well as a high efficiency. The superior efficiency of NTRs is due to both high energy density of nuclear fuel and the low molecular weight propellant of Hydrogen (H{sub 2}) over the chemical reaction by-products. A NTR uses thermal energy released from a nuclear fission reactor to heat the H{sub 2} propellant and then exhausted the highly heated propellant through a propelling nozzle to produce thrust. A propellant efficiency parameter of rocket engines is specific impulse (I{sub s}p) which represents the ratio of the thrust over the propellant consumption rate. If the average exhaust H{sub 2} temperature of a NTR is around 3,000 K, the I{sub s}p can be achieved as high as 1,000 s as compared with only 450 - 500 s of the best CRs. For this reason, NTRs are favored for various space applications such as orbital tugs, lunar transports, and manned missions to Mars and beyond. The best known NTR development effort was conducted from 1955 to1974 under the ROVER and NERVA programs in the USA. These programs had successfully designed and tested many different reactors and engines. After these projects, the researches on NERVA derived

  5. Genome-Scale Analysis of Cell-Specific Regulatory Codes Using Nuclear Enzymes.

    Science.gov (United States)

    Baek, Songjoon; Sung, Myong-Hee

    2016-01-01

    High-throughput sequencing technologies have made it possible for biologists to generate genome-wide profiles of chromatin features at the nucleotide resolution. Enzymes such as nucleases or transposes have been instrumental as a chromatin-probing agent due to their ability to target accessible chromatin for cleavage or insertion. On the scale of a few hundred base pairs, preferential action of the nuclear enzymes on accessible chromatin allows mapping of cell state-specific accessibility in vivo. Such accessible regions contain functionally important regulatory sites, including promoters and enhancers, which undergo active remodeling for cells adapting in a dynamic environment. DNase-seq and the more recent ATAC-seq are two assays that are gaining popularity. Deep sequencing of DNA libraries from these assays, termed genomic footprinting, has been proposed to enable the comprehensive construction of protein occupancy profiles over the genome at the nucleotide level. Recent studies have discovered limitations of genomic footprinting which reduce the scope of detectable proteins. In addition, the identification of putative factors that bind to the observed footprints remains challenging. Despite these caveats, the methodology still presents significant advantages over alternative techniques such as ChIP-seq or FAIRE-seq. Here we describe computational approaches and tools for analysis of chromatin accessibility and genomic footprinting. Proper experimental design and assay-specific data analysis ensure the detection sensitivity and maximize retrievable information. The enzyme-based chromatin profiling approaches represent a powerful and evolving methodology which facilitates our understanding of how the genome is regulated.

  6. Design and Development of the MITEE-B Bi-Modal Nuclear Propulsion Engine

    Science.gov (United States)

    Paniagua, John C.; Powell, James R.; Maise, George

    2003-01-01

    Previous studies of compact, ultra-lightweight high performance nuclear thermal propulsion engines have concentrated on systems that only deliver high thrust. However, many potential missions also require substantial amounts of electric power. Studies of a new, very compact and lightweight bi-modal nuclear engine that provides both high propulsive thrust and high electric power for planetary science missions are described. The design is a modification of the MITEE nuclear thermal engine concept that provided only high propulsive thrust. In the new design, MITEE-B, separate closed cooling circuits are incorporated into the reactor, which transfers useful amounts of thermal energy to a small power conversion system that generates continuous electric power over the full life of the mission, even when the engine is not delivering propulsive thrust. Two versions of the MITEE-B design are described and analyzed. Version 1 generates 1 kW(e) of continuous power for control of the spacecraft, sensors, data transmission, etc. This power level eliminates the need for RTG's on missions to the outer planets, and allowing considerably greater operational capability for the spacecraft. This, plus its high thrust and high specific impulse propulsive capabilities, makes MITEE-B very attractive for such missions. In Version 2, of MITEE-B, a total of 20 kW(e) is generated, enabling the use of electric propulsion. The combination of high open cycle propulsion thrust (20,000 Newtons) with a specific impulse of ~1000 seconds for short impulse burns, and long term (months to years), electric propulsion greatly increases MITEE's ΔV capability. Version 2 of MITEE-B also enables the production and replenishment of H2 propellant using in-situ resources, such as electrolysis of water from the ice sheet on Europa and other Jovian moons. This capability would greatly increase the ΔV available for certain planetary science missions. The modifications to the MITEE multiple pressure tube

  7. Study on the Promotion in the Citation of the Nuclear Engineering and Technology

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Young Choon; Yoo, J. B.; Yi, J. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    The Korean journal published in English, Nuclear Engineering and Technology (here under NET) has been enlisted in the global citation database SCI E(Science Citation Index Expanded) of Thomson Reuters (past ISI), beginning with NET vol.39 No.1 (Feb. 2007). As of July 2009, the citation index of NET as reported by JCR (Journal Citation Report) based on the cumulative data from ISI (Institute for Scientific Information) reached to 0.991. This index ranks on 12{sup th} among the 33 journals in the area of nuclear science and technology in the science and technology covered by JCR, meaning fairly high impact factor. The following year 2010, however, witnessed the JCR figure dropping down to 0.465. The reason behind such drastic fall would be the decreased citation and in a lesser extent self-citation in 2010, in comparison with 2009, despite the increased number of paper publication. This study attempts to give an analysis as of the end of 2011 on the NET citation frequency in SCI Source Journal and the citation frequency by KAERI authors, together with the nationalities of NET authors and SCI journals that refer to NET most. Based on the analysis, the paper suggests some ways to promoting the position of NET as a journal in the international nuclear sector

  8. S-MATE: Secure Coding-based Multipath Adaptive Traffic Engineering

    CERN Document Server

    Aly, Salah A; Walid, Anwar I; Poor, H Vincent

    2010-01-01

    There have been several approaches to provisioning traffic between core network nodes in Internet Service Provider (ISP) networks. Such approaches aim to minimize network delay, increase network capacity, and enhance network security services. MATE (Multipath Adaptive Traffic Engineering) protocol has been proposed for multipath adaptive traffic engineering between an ingress node (source) and an egress node (destination). Its novel idea is to avoid network congestion and attacks that might exist in edge and node disjoint paths between two core network nodes. This paper builds an adaptive, robust, and reliable traffic engineering scheme for better performance of communication network operations. This will also provision quality of service (QoS) and protection of traffic engineering to maximize network efficiency. Specifically, we present a new approach, S-MATE (secure MATE) is developed to protect the network traffic between two core nodes (routers or switches) in a cloud network. S-MATE secures against a sin...

  9. Multiphysics Computational Analysis of a Solid-Core Nuclear Thermal Engine Thrust Chamber

    Science.gov (United States)

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

    2007-01-01

    The objective of this effort is to develop an efficient and accurate computational heat transfer methodology to predict thermal, fluid, and hydrogen environments for a hypothetical solid-core, nuclear thermal engine - the Small Engine. In addition, the effects of power profile and hydrogen conversion on heat transfer efficiency and thrust performance were also investigated. The computational methodology is based on an unstructured-grid, pressure-based, all speeds, chemically reacting, computational fluid dynamics platform, while formulations of conjugate heat transfer were implemented to describe the heat transfer from solid to hydrogen inside the solid-core reactor. The computational domain covers the entire thrust chamber so that the afore-mentioned heat transfer effects impact the thrust performance directly. The result shows that the computed core-exit gas temperature, specific impulse, and core pressure drop agree well with those of design data for the Small Engine. Finite-rate chemistry is very important in predicting the proper energy balance as naturally occurring hydrogen decomposition is endothermic. Locally strong hydrogen conversion associated with centralized power profile gives poor heat transfer efficiency and lower thrust performance. On the other hand, uniform hydrogen conversion associated with a more uniform radial power profile achieves higher heat transfer efficiency, and higher thrust performance.

  10. Structural Analyses of the Support Trusses for the Nuclear Thermal Rocket Engines and Drop Tanks

    Science.gov (United States)

    Myers, David E.; Kosareo, Daniel N.

    2006-01-01

    Finite element structural analyses were performed on the support trusses of the Nuclear Thermal Rocket (NTR) engines and drop tanks to verify that the proper amount of mass was allocated for these components in the vehicle sizing model. The verification included a static stress analysis, a modal analysis, and a buckling analysis using the MSC/NASTRAN™ structural analysis software package. In addition, a crippling stress analysis was performed on the truss beams using a handbook equation. Two truss configurations were examined as possible candidates for the drop tanks truss while a baseline was examined for the engine support thrust structure. For the drop tanks trusses, results showed that both truss configurations produced similar results although one performed slightly better in buckling. In addition, it was shown that the mass allocated in the vehicle sizing model was adequate although the engine thrust structure may need to be modified slightly to increase its lateral natural frequency above the minimum requirement of 8 Hz that is specified in the Delta IV Payload Planners Guide.

  11. Functions of an engineered barrier system for a nuclear waste repository in basalt

    Energy Technology Data Exchange (ETDEWEB)

    Coons, W.E.; Moore, E.L.; Smith, M.J.; Kaser, J.D.

    1980-01-01

    Defined in this document are the functions of components selected for an engineered barrier system for a nuclear waste repository in basalt. The definitions provide a focal point for barrier material research and development by delineating the purpose and operative lifetime of each component of the engineered system. A five-component system (comprised of waste form, canister, buffer, overpack, and tailored backfill) is discussed in terms of effective operation throughout the course of repository history, recognizing that the emplacement environment changes with time. While components of the system are mutually supporting, redundancy is provided by subsystems of physical and chemical barriers which act in concert with the geology to provide a formidable barrier to transport of hazardous materials to the biosphere. The operating philosophy of the conceptual engineered barrier system is clarified by examples pertinent to storage in basalt, and a technical approach to barrier design and material selection is proposed. A method for system validation and qualification is also included which considers performance criteria proposed by external agencies in conjunction with site-specific models and risk assessment to define acceptable levels of system performance.

  12. Highly sensitive detection of protein biomarkers via nuclear magnetic resonance biosensor with magnetically engineered nanoferrite particles.

    Science.gov (United States)

    Jeun, Minhong; Park, Sungwook; Lee, Hakho; Lee, Kwan Hyi

    Magnetic-based biosensors are attractive for on-site detection of biomarkers due to the low magnetic susceptibility of biological samples. Here, we report a highly sensitive magnetic-based biosensing system that is composed of a miniaturized nuclear magnetic resonance (NMR) device and magnetically engineered nanoferrite particles (NFPs). The sensing performance, also identified as the transverse relaxation (R2) rate, of the NMR device is directly related to the magnetic properties of the NFPs. Therefore, we developed magnetically engineered NFPs (MnMg-NFP) and used them as NMR agents to exhibit a significantly improved R2 rate. The magnetization of the MnMg-NFPs was increased by controlling the Mn and Mg cation concentration and distribution during the synthesis process. This modification of the Mn and Mg cation directly contributed to improving the R2 rate. The miniaturized NMR system, combined with the magnetically engineered MnMg-NFPs, successfully detected a small amount of infectious influenza A H1N1 nucleoprotein with high sensitivity and stability.

  13. Using Automatic Code Generation in the Attitude Control Flight Software Engineering Process

    Science.gov (United States)

    McComas, David; O'Donnell, James R., Jr.; Andrews, Stephen F.

    1999-01-01

    This paper presents an overview of the attitude control subsystem flight software development process, identifies how the process has changed due to automatic code generation, analyzes each software development phase in detail, and concludes with a summary of our lessons learned.

  14. Using Automatic Code Generation in the Attitude Control Flight Software Engineering Process

    Science.gov (United States)

    McComas, David; O'Donnell, James R., Jr.; Andrews, Stephen F.

    1999-01-01

    This paper presents an overview of the attitude control subsystem flight software development process, identifies how the process has changed due to automatic code generation, analyzes each software development phase in detail, and concludes with a summary of our lessons learned.

  15. Military Adaptation of Commercial Items: Laboratory Evaluation of the Code E-436 Engine

    Science.gov (United States)

    1984-02-01

    Coolant, engine outlet 120-250 + 2 (13) Combustion air at meter ( Meriam 30-160 ± 2 flow meter) (14) Cooling water, tower inlet * 35-100 (15) Cooling...2) Air, before turbo (in.H 2 0)-S 0 to -25 ± 1 (3) Air, after turbo (in. hg) 0 to +60 +.2 (4) Air across Meriam flow meter 0 to -28 +.1 (in. H 20) (5...meter for measuring engine blowby. (8) Temperature reference bath (Maintain at 2000 F). (9) Meriam air flow meter. e. The folloing monitors vill be

  16. Further two-dimensional code development for Stirling space engine components

    Science.gov (United States)

    Ibrahim, Mounir; Tew, Roy C.; Dudenhoefer, James E.

    1990-01-01

    The development of multidimensional models of Stirling engine components is described. Two-dimensional parallel plate models of an engine regenerator and a cooler were used to study heat transfer under conditions of laminar, incompressible oscillating flow. Substantial differences in the nature of the temperature variations in time over the cycle were observed for the cooler as contrasted with the regenerator. When the two-dimensional cooler model was used to calculate a heat transfer coefficient, it yields a very different result from that calculated using steady-flow correlations. Simulation results for the regenerator and the cooler are presented.

  17. Performance Engineering: Understanding and Improving thePerformance of Large-Scale Codes

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, David H.; Lucas, Robert; Hovland, Paul; Norris, Boyana; Yelick, Kathy; Gunter, Dan; de Supinski, Bronis; Quinlan, Dan; Worley,Pat; Vetter, Jeff; Roth, Phil; Mellor-Crummey, John; Snavely, Allan; Hollingsworth, Jeff; Reed, Dan; Fowler, Rob; Zhang, Ying; Hall, Mary; Chame, Jacque; Dongarra, Jack; Moore, Shirley

    2007-10-01

    Achieving good performance on high-end computing systems is growing ever more challenging due to enormous scale, increasing architectural complexity, and increasing application complexity. To address these challenges in DOE's SciDAC-2 program, the Performance Engineering Research Institute (PERI) has embarked on an ambitious research plan encompassing performance modeling and prediction, automatic performance optimization and performance engineering of high profile applications. The principal new component is a research activity in automatic tuning software, which is spurred by the strong user preference for automatic tools.

  18. Idaho National Engineering Laboratory code assessment of the Rocky Flats transuranic waste

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    This report is an assessment of the content codes associated with transuranic waste shipped from the Rocky Flats Plant in Golden, Colorado, to INEL. The primary objective of this document is to characterize and describe the transuranic wastes shipped to INEL from Rocky Flats by item description code (IDC). This information will aid INEL in determining if the waste meets the waste acceptance criteria (WAC) of the Waste Isolation Pilot Plant (WIPP). The waste covered by this content code assessment was shipped from Rocky Flats between 1985 and 1989. These years coincide with the dates for information available in the Rocky Flats Solid Waste Information Management System (SWIMS). The majority of waste shipped during this time was certified to the existing WIPP WAC. This waste is referred to as precertified waste. Reassessment of these precertified waste containers is necessary because of changes in the WIPP WAC. To accomplish this assessment, the analytical and process knowledge available on the various IDCs used at Rocky Flats were evaluated. Rocky Flats sources for this information include employee interviews, SWIMS, Transuranic Waste Certification Program, Transuranic Waste Inspection Procedure, Backlog Waste Baseline Books, WIPP Experimental Waste Characterization Program (headspace analysis), and other related documents, procedures, and programs. Summaries are provided of: (a) certification information, (b) waste description, (c) generation source, (d) recovery method, (e) waste packaging and handling information, (f) container preparation information, (g) assay information, (h) inspection information, (i) analytical data, and (j) RCRA characterization.

  19. AMS at the National Institute of Nuclear Physics and Engineering in Bucharest

    Science.gov (United States)

    Stan-Sion, C.; Ivascu, M.; Plostinaru, D.; Catana, D.; Marinescu, L.; Radulescu, M.; Nolte, E.

    2000-10-01

    A new beam line and injector deck for AMS measurements have been built at the 8 MV tandem accelerator of the National Institute of Nuclear Physics and Engineering, Bucharest, Romania. The main components on the low-energy side are a high-current cesium sputter source, a 90° injection magnet and a pre-acceleration stage. At the high-energy side the beam line is achromatic, consisting of two 90° analysing magnets with mass energy product 120 MeV amu and a gas-filled ionization chamber. The system will be complete with a Wien filter and a multi-anode gas detector with time-of-flight discrimination. Presently, the AMS facility is undergoing tests and routine measurements are expected to start soon.

  20. AMS at the National Institute of Nuclear Physics and Engineering in Bucharest

    Energy Technology Data Exchange (ETDEWEB)

    Stan-Sion, C. E-mail: stansion@ifin.nipne.ro; Ivascu, M.; Plostinaru, D.; Catana, D.; Marinescu, L.; Radulescu, M.; Nolte, E

    2000-10-01

    A new beam line and injector deck for AMS measurements have been built at the 8 MV tandem accelerator of the National Institute of Nuclear Physics and Engineering, Bucharest, Romania. The main components on the low-energy side are a high-current cesium sputter source, a 90 deg. injection magnet and a pre-acceleration stage. At the high-energy side the beam line is achromatic, consisting of two 90 deg. analysing magnets with mass energy product 120 MeV amu and a gas-filled ionization chamber. The system will be complete with a Wien filter and a multi-anode gas detector with time-of-flight discrimination. Presently, the AMS facility is undergoing tests and routine measurements are expected to start soon.

  1. WE-AB-204-11: Development of a Nuclear Medicine Dosimetry Module for the GPU-Based Monte Carlo Code ARCHER

    Energy Technology Data Exchange (ETDEWEB)

    Liu, T; Lin, H; Xu, X [Rensselaer Polytechnic Institute, Troy, NY (United States); Stabin, M [Vanderbilt Univ Medical Ctr, Nashville, TN (United States)

    2015-06-15

    Purpose: To develop a nuclear medicine dosimetry module for the GPU-based Monte Carlo code ARCHER. Methods: We have developed a nuclear medicine dosimetry module for the fast Monte Carlo code ARCHER. The coupled electron-photon Monte Carlo transport kernel included in ARCHER is built upon the Dose Planning Method code (DPM). The developed module manages the radioactive decay simulation by consecutively tracking several types of radiation on a per disintegration basis using the statistical sampling method. Optimization techniques such as persistent threads and prefetching are studied and implemented. The developed module is verified against the VIDA code, which is based on Geant4 toolkit and has previously been verified against OLINDA/EXM. A voxelized geometry is used in the preliminary test: a sphere made of ICRP soft tissue is surrounded by a box filled with water. Uniform activity distribution of I-131 is assumed in the sphere. Results: The self-absorption dose factors (mGy/MBqs) of the sphere with varying diameters are calculated by ARCHER and VIDA respectively. ARCHER’s result is in agreement with VIDA’s that are obtained from a previous publication. VIDA takes hours of CPU time to finish the computation, while it takes ARCHER 4.31 seconds for the 12.4-cm uniform activity sphere case. For a fairer CPU-GPU comparison, more effort will be made to eliminate the algorithmic differences. Conclusion: The coupled electron-photon Monte Carlo code ARCHER has been extended to radioactive decay simulation for nuclear medicine dosimetry. The developed code exhibits good performance in our preliminary test. The GPU-based Monte Carlo code is developed with grant support from the National Institute of Biomedical Imaging and Bioengineering through an R01 grant (R01EB015478)

  2. Delivering Single-Walled Carbon Nanotubes to the Nucleus Using Engineered Nuclear Protein Domains.

    Science.gov (United States)

    Boyer, Patrick D; Ganesh, Sairaam; Qin, Zhao; Holt, Brian D; Buehler, Markus J; Islam, Mohammad F; Dahl, Kris Noel

    2016-02-10

    Single-walled carbon nanotubes (SWCNTs) have great potential for cell-based therapies due to their unique intrinsic optical and physical characteristics. Consequently, broad classes of dispersants have been identified that individually suspend SWCNTs in water and cell media in addition to reducing nanotube toxicity to cells. Unambiguous control and verification of the localization and distribution of SWCNTs within cells, particularly to the nucleus, is needed to advance subcellular technologies utilizing nanotubes. Here we report delivery of SWCNTs to the nucleus by noncovalently attaching the tail domain of the nuclear protein lamin B1 (LB1), which we engineer from the full-length LMNB1 cDNA. More than half of this low molecular weight globular protein is intrinsically disordered but has an immunoglobulin-fold composed of a central hydrophobic core, which is highly suitable for associating with SWCNTs, stably suspending SWCNTs in water and cell media. In addition, LB1 has an exposed nuclear localization sequence to promote active nuclear import of SWCNTs. These SWCNTs-LB1 dispersions in water and cell media display near-infrared (NIR) absorption spectra with sharp van Hove peaks and an NIR fluorescence spectra, suggesting that LB1 individually disperses nanotubes. The dispersing capability of SWCNTs by LB1 is similar to that by albumin proteins. The SWCNTs-LB1 dispersions with concentrations ≥150 μg/mL (≥30 μg/mL) in water (cell media) remain stable for ≥75 days (≥3 days) at 4 °C (37 °C). Further, molecular dynamics modeling of association of LB1 with SWCNTs reveal that the exposure of the nuclear localization sequence is independent of LB1 binding conformation. Measurements from confocal Raman spectroscopy and microscopy, NIR fluorescence imaging of SWCNTs, and fluorescence lifetime imaging microscopy show that millions of these SWCNTs-LB1 complexes enter HeLa cells, localize to the nucleus of cells, and interact with DNA. We postulate that the

  3. Improvement of level-1 PSA computer code package - Modeling and analysis for dynamic reliability of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chang Hoon; Baek, Sang Yeup; Shin, In Sup; Moon, Shin Myung; Moon, Jae Phil; Koo, Hoon Young; Kim, Ju Shin [Seoul National University, Seoul (Korea, Republic of); Hong, Jung Sik [Seoul National Polytechnology University, Seoul (Korea, Republic of); Lim, Tae Jin [Soongsil University, Seoul (Korea, Republic of)

    1996-08-01

    The objective of this project is to develop a methodology of the dynamic reliability analysis for NPP. The first year`s research was focused on developing a procedure for analyzing failure data of running components and a simulator for estimating the reliability of series-parallel structures. The second year`s research was concentrated on estimating the lifetime distribution and PM effect of a component from its failure data in various cases, and the lifetime distribution of a system with a particular structure. Computer codes for performing these jobs were also developed. The objectives of the third year`s research is to develop models for analyzing special failure types (CCFs, Standby redundant structure) that were nor considered in the first two years, and to complete a methodology of the dynamic reliability analysis for nuclear power plants. The analysis of failure data of components and related researches for supporting the simulator must be preceded for providing proper input to the simulator. Thus this research is divided into three major parts. 1. Analysis of the time dependent life distribution and the PM effect. 2. Development of a simulator for system reliability analysis. 3. Related researches for supporting the simulator : accelerated simulation analytic approach using PH-type distribution, analysis for dynamic repair effects. 154 refs., 5 tabs., 87 figs. (author)

  4. Novel methods for the molecular discrimination of Fasciola spp. on the basis of nuclear protein-coding genes.

    Science.gov (United States)

    Shoriki, Takuya; Ichikawa-Seki, Madoka; Suganuma, Keisuke; Naito, Ikunori; Hayashi, Kei; Nakao, Minoru; Aita, Junya; Mohanta, Uday Kumar; Inoue, Noboru; Murakami, Kenji; Itagaki, Tadashi

    2016-06-01

    Fasciolosis is an economically important disease of livestock caused by Fasciola hepatica, Fasciola gigantica, and aspermic Fasciola flukes. The aspermic Fasciola flukes have been discriminated morphologically from the two other species by the absence of sperm in their seminal vesicles. To date, the molecular discrimination of F. hepatica and F. gigantica has relied on the nucleotide sequences of the internal transcribed spacer 1 (ITS1) region. However, ITS1 genotypes of aspermic Fasciola flukes cannot be clearly differentiated from those of F. hepatica and F. gigantica. Therefore, more precise and robust methods are required to discriminate Fasciola spp. In this study, we developed PCR restriction fragment length polymorphism and multiplex PCR methods to discriminate F. hepatica, F. gigantica, and aspermic Fasciola flukes on the basis of the nuclear protein-coding genes, phosphoenolpyruvate carboxykinase and DNA polymerase delta, which are single locus genes in most eukaryotes. All aspermic Fasciola flukes used in this study had mixed fragment pattern of F. hepatica and F. gigantica for both of these genes, suggesting that the flukes are descended through hybridization between the two species. These molecular methods will facilitate the identification of F. hepatica, F. gigantica, and aspermic Fasciola flukes, and will also prove useful in etiological studies of fasciolosis. Copyright © 2015 Elsevier Ireland Ltd. All rights reserved.

  5. Construction of a bibliographic information database for the Nuclear Science and Engineering (IX)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Whan; Oh, Jeong Hun; Choi, Kwang; Keum, Jong Yong [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-12-01

    The major goal of this project is to construct a database in nuclear science and engineering information materials, support the R and D activities of users in this field, finally to support KRISTAL(Korea Research Information for Science and Technology Access Line)'s DB through the KREONet(Korea Research Environment Open Network), as one of the five national information networks. The contents of this project are as follows: 1) Materials selection and collection, 2) Indexing and abstract preparation, 3) Data input and transmission, and 4) document delivery service. In this seventh year, 40,000 records, as total of inputted data, are added to the existing SATURN DB. These records are covered with the articles of nuclear-related core journals, proceedings, seminars, and research reports, etc. And using the Web, this project was important for users to get their needed information itself and to receive the materials of online requested information. And then it will give the chance users not only to promote the the effectiveness of R and D activities, but also to obviate the duplicated research works. 1 fig., 1 tab. (Author)

  6. Updating Human Factors Engineering Guidelines for Conducting Safety Reviews of Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    O, J.M.; Higgins, J.; Stephen Fleger - NRC

    2011-09-19

    The U.S. Nuclear Regulatory Commission (NRC) reviews the human factors engineering (HFE) programs of applicants for nuclear power plant construction permits, operating licenses, standard design certifications, and combined operating licenses. The purpose of these safety reviews is to help ensure that personnel performance and reliability are appropriately supported. Detailed design review procedures and guidance for the evaluations is provided in three key documents: the Standard Review Plan (NUREG-0800), the HFE Program Review Model (NUREG-0711), and the Human-System Interface Design Review Guidelines (NUREG-0700). These documents were last revised in 2007, 2004 and 2002, respectively. The NRC is committed to the periodic update and improvement of the guidance to ensure that it remains a state-of-the-art design evaluation tool. To this end, the NRC is updating its guidance to stay current with recent research on human performance, advances in HFE methods and tools, and new technology being employed in plant and control room design. This paper describes the role of HFE guidelines in the safety review process and the content of the key HFE guidelines used. Then we will present the methodology used to develop HFE guidance and update these documents, and describe the current status of the update program.

  7. Reliability study: digital engineered safety feature actuation system of Korean Standard Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Sudarno [National Nuclear Energy Agency, Batan (Indonesia); Kang, H. G.; Jang, S. C.; Eom, H. S.; Ha, J. J. [KAERI, Taejon (Korea, Republic of)

    2003-04-01

    The usage of digital Instrumentation and Control (I and C) in a nuclear power plant becomes more extensive, including safety related systems. The PSA application of these new designs are very important in order to evaluate their reliability. In particular, Korean Standard Nuclear Power Plants (KSNPPs), typically Ulchin 5 and 6 (UCN 5 and 6) reactor units, adopted the digital safety-critical systems such as Digital Plant Protection System (DPPS) and Digital Engineered Safety Feature Actuation System (DESFAS). In this research, we developed fault tree models for assessing the unavailability of the DESFAS functions. We also performed an analysis of the quantification results. The unavailability results of different DESFAS functions showed that their values are comprised from 5.461E-5 to 3.14E-4. The system unavailability of DESFAS AFAS-1 is estimated as 5.461E-5, which is about 27% less than that of analog system if we consider the difference of human failure probability estimation between both analyses. The results of this study could be utilized in risk-effect analysis of KSNPP. We expect that the safety analysis result will contribute to design feedback.

  8. Anticipated Degradation Modes of Metallic Engineered Barriers for High-Level Nuclear Waste Repositories

    Science.gov (United States)

    Rodríguez, Martín A.

    2014-03-01

    Metallic engineered barriers must provide a period of absolute containment to high-level radioactive waste in geological repositories. Candidate materials include copper alloys, carbon steels, stainless steels, nickel alloys, and titanium alloys. The national programs of nuclear waste management have to identify and assess the anticipated degradation modes of the selected materials in the corresponding repository environment, which evolves in time. Commonly assessed degradation modes include general corrosion, localized corrosion, stress-corrosion cracking, hydrogen-assisted cracking, and microbiologically influenced corrosion. Laboratory testing and modeling in metallurgical and environmental conditions of similar and higher aggressiveness than those expected in service conditions are used to evaluate the corrosion resistance of the materials. This review focuses on the anticipated degradation modes of the selected or reference materials as corrosion-resistant barriers in nuclear repositories. These degradation modes depend not only on the selected alloy but also on the near-field environment. The evolution of the near-field environment varies for saturated and unsaturated repositories considering backfilled and unbackfilled conditions. In saturated repositories, localized corrosion and stress-corrosion cracking may occur in the initial aerobic stage, while general corrosion and hydrogen-assisted cracking are the main degradation modes in the anaerobic stage. Unsaturated repositories would provide an oxidizing environment during the entire repository lifetime. Microbiologically influenced corrosion may be avoided or minimized by selecting an appropriate backfill material. Radiation effects are negligible provided that a thick-walled container or an inner shielding container is used.

  9. The engineering of a nuclear thermal landing and ascent vehicle utilizing indigenous Martian propellant

    Science.gov (United States)

    Zubrin, Robert M.

    1991-01-01

    The following paper reports on a design study of a novel space transportation concept known as a 'NIMF' (Nuclear rocket using Indigenous Martian Fuel). The NIMF is a ballistic vehicle which obtains its propellant out of the Martian air by compression and liquefaction of atmospheric CO2. This propellant is subsequently used to generate rocket thrust at a specific impulse of 264 s by being heated to high temperature (2800 K) gas in the NIMFs' nuclear thermal rocket engines. The vehicle is designed to provide surface to orbit and surface to surface transportation, as well as housing, for a crew of three astronauts. It is capable of refueling itself for a flight to its maximum orbit in less than 50 days. The ballistic NIMF has a mass of 44.7 tonnes and, with the assumed 2800 K propellant temperature, is capable of attaining highly energetic (250 km by 34,000 km elliptical) orbits. This allows it to rendezvous with interplanetary transfer vehicles which are only very loosely bound into orbit around Mars. If a propellant temperature of 2000 K is assumed, then low Mars orbit can be attained; while if 3100 K is assumed, then the ballistic NIMF is capable of injecting itself onto a minimum energy transfer orbit to Earth in a direct ascent from the Martian surface.

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

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Per

    2012-10-30

    The objective of the 3-year project was to collect integral effects test (IET) data to validate the RELAP5-3D code and other thermal hydraulics codes for use in predicting the transient thermal hydraulics response of liquid salt cooled reactor systems, including integral transient response for forced and natural circulation operation. The reference system for the project is a modular, 900-MWth Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a specific type of Fluoride salt-cooled High temperature Reactor (FHR). Two experimental facilities were developed for thermal-hydraulic integral effects tests (IETs) and separate effects tests (SETs). The facilities use simulant fluids for the liquid fluoride salts, with very little distortion to the heat transfer and fluid dynamics behavior. The CIET Test Bay facility was designed, built, and operated. IET data for steady state and transient natural circulation was collected. SET data for convective heat transfer in pebble beds and straight channel geometries was collected. The facility continues to be operational and will be used for future experiments, and for component development. The CIET 2 facility is larger in scope, and its construction and operation has a longer timeline than the duration of this grant. The design for the CIET 2 facility has drawn heavily on the experience and data collected on the CIET Test Bay, and it was completed in parallel with operation of the CIET Test Bay. CIET 2 will demonstrate start-up and shut-down transients and control logic, in addition to LOFC and LOHS transients, and buoyant shut down rod operation during transients. Design of the CIET 2 Facility is complete, and engineering drawings have been submitted to an external vendor for outsourced quality controlled construction. CIET 2 construction and operation continue under another NEUP grant. IET data from both CIET facilities is to be used for validation of system codes used for FHR modeling, such as RELAP5-3D. A set of

  11. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs draft environmental impact statement. Volume 1, Appendix B: Idaho National Engineering Laboratory Spent Nuclear Fuel Management Program

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    The US Department of Energy (DOE) has prepared this report to assist its management in making two decisions. The first decision, which is programmatic, is to determine the management program for DOE spent nuclear fuel. The second decision is on the future direction of environmental restoration, waste management, and spent nuclear fuel management activities at the Idaho National Engineering Laboratory. Volume 1 of the EIS, which supports the programmatic decision, considers the effects of spent nuclear fuel management on the quality of the human and natural environment for planning years 1995 through 2035. DOE has derived the information and analysis results in Volume 1 from several site-specific appendixes. Volume 2 of the EIS, which supports the INEL-specific decision, describes environmental impacts for various environmental restoration, waste management, and spent nuclear fuel management alternatives for planning years 1995 through 2005. This Appendix B to Volume 1 considers the impacts on the INEL environment of the implementation of various DOE-wide spent nuclear fuel management alternatives. The Naval Nuclear Propulsion Program, which is a joint Navy/DOE program, is responsible for spent naval nuclear fuel examination at the INEL. For this appendix, naval fuel that has been examined at the Naval Reactors Facility and turned over to DOE for storage is termed naval-type fuel. This appendix evaluates the management of DOE spent nuclear fuel including naval-type fuel.

  12. Development of improved thermal hydraulics and fuel performance technology; development of turbulence model and simulation code for flow analysis in nuclear fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Myung, H. K.; Yang, S. Y.; Kim, B. H.; Song, J. H.; Oh, J. Z. [Kookmin University, Seoul (Korea)

    2002-03-01

    The flow through a nuclear rod bundle with mixing vanes is very complex and so required a suitable turbulence model for its accurate prediction. Subchannel flow in a nuclear bundle having vanes to mix flow appears complex turbulent flow. Objective of this study is to investigate performance of prediction about turbulence model contained in STAR-CD code and to develop suitable turbulence model which can predict complex flow in nuclear assembly. For several nonlinear {kappa}-{epsilon} turbulence models, their performance were investigated in the prediction of the flow in nuclear fuel assembly, and also their problems were discussed in detail. The results obtained from the present research would give a help for the development of turbulence model which can accurately predict the flow through the rod bundles with mixing vanes. 19 refs., 32 figs., 3 tabs. (Author)

  13. Investigation of metallic, ceramic, and polymeric materials for engineered barrier applications in nuclear-waste packages

    Energy Technology Data Exchange (ETDEWEB)

    Westerman, R.E.

    1980-10-01

    An effort to develop licensable engineered barrier systems for the long-term (about 1000 yr) containment of nuclear wastes under conditions of deep continental geologic disposal has been underway at Pacific Northwest Laboratory since January 1979, under the auspices of the High-Level Waste Immobilization Program. In the present work, the barrier system comprises the hard or structural elements of the package: the canister, the overpack(s), and the hole sleeve. A number of candidate metallic, ceramic, and polymeric materials were put through mechanical, corrosion, and leaching screening tests to determine their potential usefulness in barrier-system applications. Materials demonstrating adequate properties in the screening tests will be subjected to more detailed property tests, and, eventually, cost/benefit analyses, to determine their ultimate applicability to barrier-system design concepts. The following materials were investigated: two titanium alloys of Grade 2 and Grade 12; 300 and 400 series stainless steels, Inconels, Hastelloy C-276, titanium, Zircoloy, copper-nickel alloys and cast irons; total of 14 ceramic materials, including two grades of alumina, plus graphite and basalt; and polymers such as polyamide-imide, polyarylene, polyimide, polyolefin, polyphenylene sulfide, polysulfone, fluoropolymer, epoxy, furan, silicone, and ethylene-propylene terpolymer (EPDM) rubber. The most promising candidates for further study and potential use in engineered barrier systems were found to be rubber, filled polyphenylene sulfide, fluoropolymer, and furan derivatives.

  14. Idaho Nuclear Technology and Engineering Center (INTEC) (formerly ICPP) ash reutilization study

    Energy Technology Data Exchange (ETDEWEB)

    Langenwalter, T.; Pettet, M.; Ochoa, R.; Jensen, S.

    1998-05-01

    Since 1984, the coal-fired plant at the Idaho Nuclear Technology and Engineering Center (INTEC, formerly Idaho Chemical Processing Plant) has been generating fly ash at a rate of approximately 1,000 tons per year. This ash is hydrated and placed in an ash bury pit near the coal-fired plant. The existing ash bury pit will be full in less than 1 year at its present rate of use. A conceptual design to build a new ash bury pit was completed, and the new pit is estimated to cost $1.7 million. This report evaluates ash reutilization alternatives that propose to eliminate this waste stream and save the $1.7 million required to build a new pit. The alternatives include using ash for landfill day cover, concrete admixture, flowable fill, soil stabilization, waste remediation, and carbon recovery technology. Both physical and chemical testing, under the guidance of the American Society for Testing and Materials, have been performed on ash from the existing pit and from different steps within the facility`s processes. The test results have been evaluated, compared to commercial ash, and are discussed as they relate to reutilization alternatives. This study recommends that the ash be used in flowable fill concrete for Deactivation and Demolition work at the Idaho National Engineering and Environmental Laboratory.

  15. Calcined Waste Storage at the Idaho Nuclear Technology and Engineering Center

    Energy Technology Data Exchange (ETDEWEB)

    Staiger, M. Daniel, Swenson, Michael C.

    2011-09-01

    This comprehensive report provides definitive volume, mass, and composition (chemical and radioactivity) of calcined waste stored at the Idaho Nuclear Technology and Engineering Center. Calcine composition data are required for regulatory compliance (such as permitting and waste disposal), future treatment of the caline, and shipping the calcine to an off-Site-facility (such as a geologic repository). This report also contains a description of the calcine storage bins. The Calcined Solids Storage Facilities (CSSFs) were designed by different architectural engineering firms and built at different times. Each CSSF has a unique design, reflecting varying design criteria and lessons learned from historical CSSF operation. The varying CSSF design will affect future calcine retrieval processes and equipment. Revision 4 of this report presents refinements and enhancements of calculations concerning the composition, volume, mass, chemical content, and radioactivity of calcined waste produced and stored within the CSSFs. The historical calcine samples are insufficient in number and scope of analysis to fully characterize the entire inventory of calcine in the CSSFs. Sample data exist for all the liquid wastes that were calcined. This report provides calcine composition data based on liquid waste sample analyses, volume of liquid waste calcined, calciner operating data, and CSSF operating data using several large Microsoft Excel (Microsoft 2003) databases and spreadsheets that are collectively called the Historical Processing Model. The calcine composition determined by this method compares favorably with historical calcine sample data.

  16. MELCOR computer code manuals

    Energy Technology Data Exchange (ETDEWEB)

    Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.; Stuart, D.S.; Thompson, S.L. [Sandia National Labs., Albuquerque, NM (United States); Hodge, S.A.; Hyman, C.R.; Sanders, R.L. [Oak Ridge National Lab., TN (United States)

    1995-03-01

    MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.

  17. Engineering development of a digital replacement protection system at an operating US PWR nuclear power plant: Installation and operational experiences

    Energy Technology Data Exchange (ETDEWEB)

    Miller, M.H. [Duke Power Co., Seneca, SC (United States)

    1995-04-01

    The existing Reactor Protection Systems (RPSs) at most US PWRs are systems which reflect 25 to 30 year-old designs, components and manufacturing techniques. Technological improvements, especially in relation to modern digital systems, offer improvements in functionality, performance, and reliability, as well as reductions in maintenance and operational burden. The Nuclear power industry and the US nuclear regulators are poised to move forward with the issues that have slowed the transition to modern digital replacements for nuclear power plant safety systems. The electric utility industry is now more than ever being driven by cost versus benefit decisions. Properly designed, engineered, and installed digital systems can provide adequate cost-benefit and allow continued nuclear generated electricity. This paper describes various issues and areas related to an ongoing RPS replacement demonstration project which are pertinant for a typical US nuclear plant to consider cost-effective replacement of an aging analog RPS with a modern digital RPS. The following subject areas relative to the Oconee Nuclear Station ISAT{trademark} Demonstrator project are discussed: Operator Interface Development; Equipment Qualification; Validation and Verification of Software; Factory Testing; Field Changes and Verification Testing; Utility Operational, Engineering and Maintenance; Experiences with Demonstration System; and Ability to operate in parallel with the existing Analog RPS.

  18. Development of 3D models of buildings for containment of the nuclear power plant of Almaraz and of the Trillo Nuclear with the GOTHIC 8.0 code; Desarrollo de modelos 3D de los edificios de conten cion de la Central Nuclear de Almaraz y de la Central Nuclear de Trillo con el codigo GOTHIC 8.0

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, G.; Bocanegra Melian, R.; Fernandez Cosils, K.; Barreira Pereira, P.; Rey Peinado, L.; Posada Barral, J. M.

    2014-07-01

    The objective of the first phase of the research of CNAT and the UPM project is the construction of several three-dimensional models detailed GOTHIC 8.0 code of containment of a buildings plant type PWR-W and KWU, corresponding to the Central Nuclear de Almaraz (CNA) and Trillo (CNT) respectively. (Author)

  19. A preliminary approach to the extension of the Transuranus code to the fuel rod performance analysis of HLM-cooled nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Luzzi, L.; Botazzoli, P.; Devita, M.; Di Marcello, V.; Pastore, G. [Department of Energy, Politecnico di Milano, Enrico Fermi Center for Nuclear Studies - CeSNEF, via Ponzio 34/3, 20133 Milano (Italy)

    2010-07-01

    This paper briefly presents a preliminary modelling approach, aimed at the extension of the TRANSURANUS code to the fuel rod performance analysis of Heavy Liquid Metal (HLM) cooled nuclear reactors, with specific reference to the employment of the T91 steel as cladding material and of the liquid Lead-Bismuth Eutectic (LBE) as coolant. On the basis of literature indications, correlations for heat transfer to LBE, corrosion behaviour and thermo-mechanical properties of T91 are proposed, and some open issues are discussed in prospect of more reliable fuel rod performance analysis of HLM-cooled nuclear reactors. (authors)

  20. Sources of signal in 62 protein-coding nuclear genes for higher-level phylogenetics of arthropods.

    Directory of Open Access Journals (Sweden)

    Jerome C Regier

    Full Text Available BACKGROUND: This study aims to investigate the strength of various sources of phylogenetic information that led to recent seemingly robust conclusions about higher-level arthropod phylogeny and to assess the role of excluding or downweighting synonymous change for arriving at those conclusions. METHODOLOGY/PRINCIPAL FINDINGS: The current study analyzes DNA sequences from 68 gene segments of 62 distinct protein-coding nuclear genes for 80 species. Gene segments analyzed individually support numerous nodes recovered in combined-gene analyses, but few of the higher-level nodes of greatest current interest. However, neither is there support for conflicting alternatives to these higher-level nodes. Gene segments with higher rates of nonsynonymous change tend to be more informative overall, but those with lower rates tend to provide stronger support for deeper nodes. Higher-level nodes with bootstrap values in the 80% - 99% range for the complete data matrix are markedly more sensitive to substantial drops in their bootstrap percentages after character subsampling than those with 100% bootstrap, suggesting that these nodes are likely not to have been strongly supported with many fewer data than in the full matrix. Data set partitioning of total data by (mostly synonymous and (mostly nonsynonymous change improves overall node support, but the result remains much inferior to analysis of (unpartitioned nonsynonymous change alone. Clusters of genes with similar nonsynonymous rate properties (e.g., faster vs. slower show some distinct patterns of node support but few conflicts. Synonymous change is shown to contribute little, if any, phylogenetic signal to the support of higher-level nodes, but it does contribute nonphylogenetic signal, probably through its underlying heterogeneous nucleotide composition. Analysis of seemingly conservative indels does not prove useful. CONCLUSIONS: Generating a robust molecular higher-level phylogeny of Arthropoda is

  1. Phylogenetic relationships within Echinococcus and Taenia tapeworms (Cestoda: Taeniidae): an inference from nuclear protein-coding genes.

    Science.gov (United States)

    Knapp, Jenny; Nakao, Minoru; Yanagida, Tetsuya; Okamoto, Munehiro; Saarma, Urmas; Lavikainen, Antti; Ito, Akira

    2011-12-01

    The family Taeniidae of tapeworms is composed of two genera, Echinococcus and Taenia, which obligately parasitize mammals including humans. Inferring phylogeny via molecular markers is the only way to trace back their evolutionary histories. However, molecular dating approaches are lacking so far. Here we established new markers from nuclear protein-coding genes for RNA polymerase II second largest subunit (rpb2), phosphoenolpyruvate carboxykinase (pepck) and DNA polymerase delta (pold). Bayesian inference and maximum likelihood analyses of the concatenated gene sequences allowed us to reconstruct phylogenetic trees for taeniid parasites. The tree topologies clearly demonstrated that Taenia is paraphyletic and that the clade of Echinococcus oligarthrus and Echinococcusvogeli is sister to all other members of Echinococcus. Both species are endemic in Central and South America, and their definitive hosts originated from carnivores that immigrated from North America after the formation of the Panamanian land bridge about 3 million years ago (Ma). A time-calibrated phylogeny was estimated by a Bayesian relaxed-clock method based on the assumption that the most recent common ancestor of E. oligarthrus and E. vogeli existed during the late Pliocene (3.0 Ma). The results suggest that a clade of Taenia including human-pathogenic species diversified primarily in the late Miocene (11.2 Ma), whereas Echinococcus started to diversify later, in the end of the Miocene (5.8 Ma). Close genetic relationships among the members of Echinococcus imply that the genus is a young group in which speciation and global radiation occurred rapidly. Copyright © 2011 Elsevier Inc. All rights reserved.

  2. Engineering light-inducible nuclear localization signals for precise spatiotemporal control of protein dynamics in living cells.

    Science.gov (United States)

    Niopek, Dominik; Benzinger, Dirk; Roensch, Julia; Draebing, Thomas; Wehler, Pierre; Eils, Roland; Di Ventura, Barbara

    2014-07-14

    The function of many eukaryotic proteins is regulated by highly dynamic changes in their nucleocytoplasmic distribution. The ability to precisely and reversibly control nuclear translocation would, therefore, allow dissecting and engineering cellular networks. Here we develop a genetically encoded, light-inducible nuclear localization signal (LINuS) based on the LOV2 domain of Avena sativa phototropin 1. LINuS is a small, versatile tag, customizable for different proteins and cell types. LINuS-mediated nuclear import is fast and reversible, and can be tuned at different levels, for instance, by introducing mutations that alter AsLOV2 domain photo-caging properties or by selecting nuclear localization signals (NLSs) of various strengths. We demonstrate the utility of LINuS in mammalian cells by controlling gene expression and entry into mitosis with blue light.

  3. Qualification and application of nuclear reactor accident analysis code with the capability of internal assessment of uncertainty; Qualificacao e aplicacao de codigo de acidentes de reatores nucleares com capacidade interna de avaliacao de incerteza

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Ronaldo Celem

    2001-10-15

    This thesis presents an independent qualification of the CIAU code ('Code with the capability of - Internal Assessment of Uncertainty') which is part of the internal uncertainty evaluation process with a thermal hydraulic system code on a realistic basis. This is done by combining the uncertainty methodology UMAE ('Uncertainty Methodology based on Accuracy Extrapolation') with the RELAP5/Mod3.2 code. This allows associating uncertainty band estimates with the results obtained by the realistic calculation of the code, meeting licensing requirements of safety analysis. The independent qualification is supported by simulations with RELAP5/Mod3.2 related to accident condition tests of LOBI experimental facility and to an event which has occurred in Angra 1 nuclear power plant, by comparison with measured results and by establishing uncertainty bands on safety parameter calculated time trends. These bands have indeed enveloped the measured trends. Results from this independent qualification of CIAU have allowed to ascertain the adequate application of a systematic realistic code procedure to analyse accidents with uncertainties incorporated in the results, although there is an evident need of extending the uncertainty data base. It has been verified that use of the code with this internal assessment of uncertainty is feasible in the design and license stages of a NPP. (author)

  4. DEVELOPING OF A NEW COMPREHENSIVE SPARK IGNITION ENGINES CODE FOR HEAT LOSS ANALYSIS WITHIN COMBUSTION CHAMBER WALLS

    Directory of Open Access Journals (Sweden)

    Shahram Khalilarya

    2010-01-01

    Full Text Available The objective of this work is to develop the existing a zero-dimensional model named ODES to provide detailed insights into the internal process of the modern high speed spark ignition engines. Therefore, it has been concentrated on the development of new sub models for incorporation in an extended form of ODES, as follows: - the existing semi-empirical combustion model has been replaced by a new comprehensive model, which is based on the turbulent flame speed in the combustion chamber. - the existing three wall heat transfer model has been replaced by a new one in which, the combustion chamber is divided in to three zones including cylinder head, cylinder wall, and piston head. The steady-state heat transfer equation is solved through finite difference method with replaced boundary and initial conditions. The results gave the temperature distribution of combustion chamber walls. The rate of heat losses from combustion chamber to the coolant is calculated by using the mean temperature of each part. The code has been extensively validated with respect to performance and heat transfer against experimental results obtained on XU7JP spark ignition engine with two kinds of fuel, gasoline and compresed natural gas and gave good agreement with available experimental.

  5. Identification of kakusei, a nuclear non-coding RNA, as an immediate early gene from the honeybee, and its application for neuroethological study

    OpenAIRE

    Taketoshi Kiya; Atsushi Ugajin; Takekazu Kunieda; Takeo Kubo

    2012-01-01

    The honeybee is a social insect that exhibits various social behaviors. To elucidate the neural basis of honeybee behavior, we detected neural activity in freely-moving honeybee workers using an immediate early gene (IEG) that is expressed in a neural activity-dependent manner. In European honeybees (Apis mellifera), we identified a novel nuclear non-coding RNA, termed kakusei, as the first insect IEG, and revealed the neural activity pattern in foragers. In addition, we isolated a homologue ...

  6. Master’s degree in Nuclear Engineering UPC-ENDESA. Creating synergy at industrial and academic levels

    Energy Technology Data Exchange (ETDEWEB)

    Batet, I.; Calviño, F.; Duch, M.A.; Dies, J.; León, P.; Fernández-Olano, P.

    2015-07-01

    The Master’s degree in Nuclear Engineering, born from the alignment of objectives of Academy and Industry, aims to prepare competent engineers to assume managerial positions within the Nuclear Industry. MNE is completely taught in English. Synergies are established at both industrial and academic levels. MNE syllabus has been designed (and is being continuously improved) with the help of industrial partners and the Spanish Regulatory Body (CSN). One half of the lectures are delivered by professionals external to the university. Besides ENDESA, other companies (ANAV, AREVA, ENRESA, ENSA, ENUSA, IDOM, Nuclenor, Tecnatom, Westinghouse) collaborate in the master. Lecturers from CSN and CIEMAT (the major Spanish research centre) participate in the Master as well. A large portion of the master contents is delivered as Project Based Learning, In general, active learning and team work activities are thoroughly used so as to help the students achieve the learning objectives and acquire a number of soft skills required by industry. MNE is embedded in EMINE, the European Master in Nuclear Energy (European Institute of Technology, KIC-InnoEnergy). As well, MNE is part of a double degree in the Barcelona Engineering School (ETSEIB) with the official Master in Industrial Engineering (MUEI). Having in the same classroom EMINE and MNE students creates a good working atmosphere, while allowing the future engineers work in a multicultural and international environment. The double degree MNE-MUEI allows students to acquire the MNE competencies and, at the same time, legal engineering attributions. It has been useful to attract good engineering students to the master. (Author)

  7. Application of simulation codes in the optimization of the design of fire protection at nuclear power plants; Aplicacion de codigos de simulacion en la optimizacion del diseno de la proteccion contra incendios en centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Amezcua, V.

    2011-07-01

    The evolution of fire protection codes and standards has permitted the use of fire simulation tools for design and optimization of fire protection solution, as an alternative to the traditional deterministic approach. This alternative results in a more flexible design, suiting the solution to the real conditions and risks. In this context, Empresarios Agrupados (EEAA) is carrying out a project jointly with the CSN (the Spanish Nuclear Safety Council), in the area of fire modelling and simulation, aimed for developing a method for the reliable application of the fire simulation models to nuclear power plants scenarios. (Author)

  8. Preoperational Subsurface Conditions at the Idaho Nuclear Technology and Engineering Center Service Wastewater Discharge Facility

    Energy Technology Data Exchange (ETDEWEB)

    Ansley, Shannon L.

    2002-02-20

    The Idaho Nuclear Technology and Engineering Center (INTEC) Service Wastewater Discharge Facility replaces the existing percolation ponds as a disposal facility for the INTEC Service Waste Stream. A preferred alternative for helping decrease water content in the subsurface near INTEC, closure of the existing ponds is required by the INTEC Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) Record of Decision (ROD) for Waste Area Group 3 Operable Unit 3-13 (DOE-ID 1999a). By August 2002, the replacement facility was constructed approximately 2 miles southwest of INTEC, near the Big Lost River channel. Because groundwater beneath the Idaho National Engineering and Environmental Laboratory (INEEL) is protected under Federal and State of Idaho regulations from degradation due to INEEL activities, preoperational data required by U.S. Department of Energy (DOE) Order 5400.1 were collected. These data include preexisting physical, chemical, and biological conditions that could be affected by the discharge; background levels of radioactive and chemical components; pertinent environmental and ecological parameters; and potential pathways for human exposure or environmental impact. This document presents specific data collected in support of DOE Order 5400.1, including: four quarters of groundwater sampling and analysis of chemical and radiological parameters; general facility description; site specific geology, stratigraphy, soils, and hydrology; perched water discussions; and general regulatory requirements. However, in order to avoid duplication of previous information, the reader is directed to other referenced publications for more detailed information. Documents that are not readily available are compiled in this publication as appendices. These documents include well and borehole completion reports, a perched water evaluation letter report, the draft INEEL Wellhead Protection Program Plan, and the Environmental Checklist.

  9. Preoperational Subsurface Conditions at the Idaho Nuclear Technology and Engineering Center Service Waste Disposal Facility

    Energy Technology Data Exchange (ETDEWEB)

    Ansley, Shannon Leigh

    2002-02-01

    The Idaho Nuclear Technology and Engineering Center (INTEC) Service Wastewater Discharge Facility replaces the existing percolation ponds as a disposal facility for the INTEC Service Waste Stream. A preferred alternative for helping decrease water content in the subsurface near INTEC, closure of the existing ponds is required by the INTEC Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) Record of Decision (ROD) for Waste Area Group 3 Operable Unit 3-13 (DOE-ID 1999a). By August 2002, the replacement facility was constructed approximately 2 miles southwest of INTEC, near the Big Lost River channel. Because groundwater beneath the Idaho National Engineering and Environmental Laboratory (INEEL) is protected under Federal and State of Idaho regulations from degradation due to INEEL activities, preoperational data required by U.S. Department of Energy (DOE) Order 5400.1 were collected. These data include preexisting physical, chemical, and biological conditions that could be affected by the discharge; background levels of radioactive and chemical components; pertinent environmental and ecological parameters; and potential pathways for human exposure or environmental impact. This document presents specific data collected in support of DOE Order 5400.1, including: four quarters of groundwater sampling and analysis of chemical and radiological parameters; general facility description; site specific geology, stratigraphy, soils, and hydrology; perched water discussions; and general regulatory requirements. However, in order to avoid duplication of previous information, the reader is directed to other referenced publications for more detailed information. Documents that are not readily available are compiled in this publication as appendices. These documents include well and borehole completion reports, a perched water evaluation letter report, the draft INEEL Wellhead Protection Program Plan, and the Environmental Checklist.

  10. Application of the coupled code RELAP5-QUABOX/CUBBOX in the system analysis of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Bencik, V.; Feretic, D.; Debrecin, N. [Faculty of Electrical Engineering and Computing, Zagreb (Croatia)

    2002-11-01

    Best estimate codes and methods for the realistic simulation of operational transients and accidents are being developed in two directions. First, computer codes with models of the interaction between multidimensional neutron kinetic and NPP dynamic behavior enable realistic simulation of transients characterized by strong coupling between neutronics and thermal-hydraulics as well as of transients that result in asymmetrical spatial core power distribution. Coupled codes consisting of a system thermal-hydraulic code and a multidimensional neutronic code are being developed worldwide in order to accomplish that task. Secondly, development of the qualified plant nodalization and of the models of plant protection and control systems is important for the realistic system analysis of operational transients and accidents. Comparison of the coupled code and point kinetic results is important for the validation of the coupled code and to gain more experience in the use of the coupled code in realistic analyses. In this paper the results of two transients for NPP Krsko using the coupled code RELAP5-QUABOX/CUBBOX (R5QC) and RELAP5 stand alone code are discussed. (orig.)

  11. Nuclear receptor engineering based on novel structure activity relationships revealed by farnesyl pyrophosphate.

    Science.gov (United States)

    Goyanka, Ritu; Das, Sharmistha; Samuels, Herbert H; Cardozo, Timothy

    2010-11-01

    Nuclear receptors (NRs) comprise the second largest protein family targeted by currently available drugs, acting via specific ligand interactions within the ligand binding domain (LBD). Recently, farnesyl pyrophosphate (FPP) was shown to be a unique promiscuous NR ligand, activating a subset of NR family members and inhibiting wound healing in skin. The current study aimed at visualizing the unique basis of FPP interaction with multiple receptors in order to identify general structure-activity relationships that operate across the NR family. Docking of FPP to the 3D structures of the LBDs of a diverse set of NRs consistently revealed an electrostatic FPP pyrophosphate contact with an NR arginine conserved in the NR family, a hydrophobic farnesyl contact with NR helix-12 and a ligand binding pocket volume between 300 and 430 Å(3) as the minimal requirements for FPP activation of any NR. Lack of any of these structural features appears to render a given NR resistant to FPP activation. We used these structure-activity relationships to rationally design and successfully engineer several mutant human estrogen receptors that retain responsiveness to estradiol but no longer respond to FPP.

  12. Results and lessons learned of the first edition of the master in nuclear engineering and applications (MINA)

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, Luis E.; Garcia, Juan c.; Falcon, Susana; Marco, Maria l.; Gonzalez Romero, Enrique M. [Centro de Investigaciones Energeticas Medioambientales y Tecnologicas (CIEMAT), Avda. Complutense, 22. 28040 Madrid (Spain); Casas, Jose A. [Universidad Autonoma de Madrid, Seccion Departamental de Ingenieria Quimica, 28049 Cantoblanco, Madrid (Spain)

    2010-07-01

    The Master in Nuclear Engineering and Applications (MINA) was born to build up a bridge between University education and the technical skills demanded by nuclear industry and organizations, particularly in Spain. Motivated by nuclear renaissance, knowledge preservation and the bases of the European Education area, the new approach adopted to accomplish such a challenge has been heavily based on a professional profile defined by the Spanish nuclear community. The first edition success (MINA-2008) has been assessed through a set of indicators, which encompass a broad range of aspects, from the number of registrations to the employment rate. This paper summarizes and discusses such an assessment. Additionally, a critical thorough review has allowed identifying a few aspects that could be improved. All the lessons learned have been translated into specific measures implemented in the MINA-2009 edition. Among the indicators, participation and industrial support were considered of utmost importance. MINA-2008 had 18 students, out of which 60% were financially supported to some extent thanks to the nuclear industry and organizations (during the conduction of the master project, this support was even enhanced). Beyond the economic contribution, nuclear companies and institutions were strongly involved in all the phases of MINA-2008, from the definition of the program up to the supervision of more than 70 % of the master projects. As a result of the lessons learned, the subjects have been grouped in modules and a more practical approach has been pursued in the teaching/learning process. (authors)

  13. What they have in common the engineering from the Spanish nuclear power plants?; Que tienen en comun las ingenierias de las centrales nucleares espanolas

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez Mendez, M.

    2012-11-01

    In recent years, Spain Nuclear Power Plant Engineering have switched their project/task management method to Critical Chain multi-project management, developed by Dr. Goldratt, achieving outstanding results in improving quality and productivity. Multitasking reduction, task and resource synchronizing without the need of exact schedules, implementing a real-time priority information system, relying on the software Concerto, and daily decision making are the basis for the management change that has generated productivity increases of between 20% to 50%, opening new horizons for improvement in other scenarios such as optimizing refueling shutdowns. (Author)

  14. A model of turbocharger radial turbines appropriate to be used in zero- and one-dimensional gas dynamics codes for internal combustion engines modelling

    Energy Technology Data Exchange (ETDEWEB)

    Serrano, J.R.; Arnau, F.J.; Dolz, V.; Tiseira, A. [CMT-Motores Termicos, Universidad Politecnica de Valencia, Camino de Vera s/n, 46022 Valencia (Spain); Cervello, C. [Conselleria de Cultura, Educacion y Deporte, Generalitat Valenciana (Spain)

    2008-12-15

    The paper presents a model of fixed and variable geometry turbines. The aim of this model is to provide an efficient boundary condition to model turbocharged internal combustion engines with zero- and one-dimensional gas dynamic codes. The model is based from its very conception on the measured characteristics of the turbine. Nevertheless, it is capable of extrapolating operating conditions that differ from those included in the turbine maps, since the engines usually work within these zones. The presented model has been implemented in a one-dimensional gas dynamic code and has been used to calculate unsteady operating conditions for several turbines. The results obtained have been compared with success against pressure-time histories measured upstream and downstream of the turbine during on-engine operation. (author)

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

    Science.gov (United States)

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

    2013-01-01

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

  16. An improved ARS2-derived nuclear reporter enhances the efficiency and ease of genetic engineering in Chlamydomonas.

    Science.gov (United States)

    Specht, Elizabeth A; Nour-Eldin, Hussam Hassan; Hoang, Kevin T D; Mayfield, Stephen P

    2015-03-01

    The model alga Chlamydomonas reinhardtii has been used to pioneer genetic engineering techniques for high-value protein and biofuel production from algae. To date, most studies of transgenic Chlamydomonas have utilized the chloroplast genome due to its ease of engineering, with a sizeable suite of reporters and well-characterized expression constructs. The advanced manipulation of algal nuclear genomes has been hampered by limited strong expression cassettes, and a lack of high-throughput reporters. We have improved upon an endogenous reporter gene - the ARS2 gene encoding an arylsulfatase enzyme - that was first cloned and characterized decades ago but has not been used extensively. The new construct, derived from ARS2 cDNA, expresses significantly higher levels of reporter protein and transforms more efficiently, allowing qualitative and quantitative screening using a rapid, inexpensive 96-well assay. The improved arylsulfatase expression cassette was used to screen a new transgene promoter from the ARG7 gene, and found that the ARG7 promoter can express the ARS2 reporter as strongly as the HSP70-RBCS2 chimeric promoter that currently ranks as the best available promoter, thus adding to the list of useful nuclear promoters. This enhanced arylsulfatase reporter construct improves the efficiency and ease of genetic engineering within the Chlamydomonas nuclear genome, with potential application to other algal strains.

  17. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Draft Environmental Impact Statement. Volume 1, Appendix D, Part B: Naval spent nuclear fuel management

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    This volume contains the following attachments: transportation of Naval spent nuclear fuel; description of Naval spent nuclear receipt and handling at the Expended Core Facility at the Idaho National Engineering Laboratory; comparison of storage in new water pools versus dry container storage; description of storage of Naval spent nuclear fuel at servicing locations; description of receipt, handling, and examination of Naval spent nuclear fuel at alternate DOE facilities; analysis of normal operations and accident conditions; and comparison of the Naval spent nuclear fuel storage environmental assessment and this environmental impact statement.

  18. Innovations in the supply chain and construction engineering of nuclear-based heat transfer components

    Energy Technology Data Exchange (ETDEWEB)

    Perales, A. [Equipos Nucleares, S.A. - Ensa, Jose Ortega y Gasset, 28006 Madrid (Spain); Woolf, G. [Tecnicas Reunidas - TR, Arapiles, 28014 Madrid (Spain)

    2010-07-01

    Equipos Nucleares S.A. (Ensa) and Tecnicas Reunidas S.A. (TR), both long-established Spanish companies, have brought together innovative approaches for the supply of heat transfer solution packages by combining their respective experiences in heat exchanger design (TR), manufacturing (Ensa), and nuclear materials procurement (Ensa and TR), thereby founding a new potent European component supplier for nuclear power plants with over 50 years of experience in the global nuclear market. The combined strategy of the Ensa-TR association which addresses the problems currently faced by nuclear component suppliers is described herein. (authors)

  19. Prosthetic Engineering

    Science.gov (United States)

    ... Overview CoE for Limb Loss Prevention and Prosthetic Engineering Menu Menu VA Center of Excellence for Limb ... ZIP code here Enter ZIP code here Prosthetic Engineering - Overview Our aim is to improve prosthetic prescription ...

  20. Application of a moment tensor inversion code developed for mining-induced seismicity to fracture monitoring of civil engineering materials

    Science.gov (United States)

    Linzer, Lindsay; Mhamdi, Lassaad; Schumacher, Thomas

    2015-01-01

    A moment tensor inversion (MTI) code originally developed to compute source mechanisms from mining-induced seismicity data is now being used in the laboratory in a civil engineering research environment. Quantitative seismology methods designed for geological environments are being tested with the aim of developing techniques to assess and monitor fracture processes in structural concrete members such as bridge girders. In this paper, we highlight aspects of the MTI_Toolbox programme that make it applicable to performing inversions on acoustic emission (AE) data recorded by networks of uniaxial sensors. The influence of the configuration of a seismic network on the conditioning of the least-squares system and subsequent moment tensor results for a real, 3-D network are compared to a hypothetical 2-D version of the same network. This comparative analysis is undertaken for different cases: for networks consisting entirely of triaxial or uniaxial sensors; for both P and S-waves, and for P-waves only. The aim is to guide the optimal design of sensor configurations where only uniaxial sensors can be installed. Finally, the findings of recent laboratory experiments where the MTI_Toolbox has been applied to a concrete beam test are presented and discussed.

  1. Development of a dynamical model of a nuclear processes simulator for analysis and training in classroom based in the RELAP/SCDAP codes; Desarrollo del modulo dinamico del simulador de procesos nucleares para analisis y entrenamiento en aula basado en los codigos RELAP/SCDAP

    Energy Technology Data Exchange (ETDEWEB)

    Salazar C, J.H.; Ramos P, J.C.; Salazar S, E.; Chavez M, C. [UNAM, Fac. de Ingenieria, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, DEPFI Campus Morelos, Cuernavaca (Mexico)]. e-mail: cchavez2@cableonline.com.mx

    2003-07-01

    The present work illustrates the application of the concept of a simulator for analysis, design, instruction and training in a classroom environment associated to a nuclear power station. Emphasis is made on the methodology used to incorporate the best estimate codes RELAP/SCDAP to a prototype under development at the Nuclear Reactor Engineering Analysis Laboratory (NREAL). This methodology is based on a modular structure where multiple processes can be executed in an independent way and where the generated information is stored in shared memory segments and distributed by means of communication routines developed in the C programming language. The utility of the system is demonstrated using highly interactive graphics (mimic diagrams, pictorials and tendency graphs) for the simultaneous dynamic visualization of the most significant variables of a typical transient event (feed water controller failure in a BWR). A fundamental part of the system is its advanced graphic interface. This interface, of the type of direct manipulation, reproduces instruments and controls whose functionality is similar to those found in the current replica simulator for the Laguna Verde Nuclear Power Station. Finally the evaluation process is described. The general behavior of the main variables for the selected transitory event is interpreted, corroborating that they follow the same tendency that those reported for a BWR. The obtained results allow to conclude that the developed system works satisfactorily and that the use of al 1 x 1 real time visualization tools offers important advantages regarding other traditional methods of analysis. (Author)

  2. Summary engineering description of underwater fuel storage facility for foreign research reactor spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Dahlke, H.J.; Johnson, D.A.; Rawlins, J.K.; Searle, D.K.; Wachs, G.W.

    1994-10-01

    This document is a summary description for an Underwater Fuel Storage Facility (UFSF) for foreign research reactor (FRR) spent nuclear fuel (SNF). A FRR SNF environmental Impact Statement (EIS) is being prepared and will include both wet and dry storage facilities as storage alternatives. For the UFSF presented in this document, a specific site is not chosen. This facility can be sited at any one of the five locations under consideration in the EIS. These locations are the Idaho National Engineering Laboratory, Savannah River Site, Hanford, Oak Ridge National Laboratory, and Nevada Test Site. Generic facility environmental impacts and emissions are provided in this report. A baseline fuel element is defined in Section 2.2, and the results of a fission product analysis are presented. Requirements for a storage facility have been researched and are summarized in Section 3. Section 4 describes three facility options: (1) the Centralized-UFSF, which would store the entire fuel element quantity in a single facility at a single location, (2) the Regionalized Large-UFSF, which would store 75% of the fuel element quantity in some region of the country, and (3) the Regionalized Small-UFSF, which would store 25% of the fuel element quantity, with the possibility of a number of these facilities in various regions throughout the country. The operational philosophy is presented in Section 5, and Section 6 contains a description of the equipment. Section 7 defines the utilities required for the facility. Cost estimates are discussed in Section 8, and detailed cost estimates are included. Impacts to worker safety, public safety, and the environment are discussed in Section 9. Accidental releases are presented in Section 10. Standard Environmental Impact Forms are included in Section 11.

  3. Potential contribution of materials investigations in reducing the risks of unavailability of nuclear civil engineering infrastructures

    Energy Technology Data Exchange (ETDEWEB)

    Goy, R.; Gaine, C.; Cornish-Bowden, I.; Auge, L.; Thillard, G.; Capra, B. [Oxand (France)

    2011-07-01

    Many consulting assignments undertaken by Oxand within the framework of life-cycle management of existing nuclear civil engineering structures, relied on taking into account the structure's level of deterioration, the quantitative estimate of its residual strength capacity and the future progression of this capacity over time. The developments of structure assessment techniques allow more and more the integration of the real state of the existing structure by enhancing the value of measures and investigations carried out on structures. On one hand, this information allows to take into account the structure's ageing and therefore estimate the residual resistance and reliability. On the other hand, by investigating the real properties of the structure and of its operation, refined calculations can sometimes highlight a resistance margin of the structure 'such as built' higher than the one considered during design. The precision on the 'real' reliability of the structure allows to refine its management and thus to optimize the infrastructure's performance. Developments of assessment methods are increasingly oriented toward probabilistic approaches. Although they go out of the 'classical regulatory' framework, they bring precision on the 'real' reliability of the structure. Using higher assessment levels needs to have more important and precise input data, what can lead to the need to carry out more detailed investigations. However, these methods offer interesting ways, which can provide, in certain cases, useful additional information for the decision-making of actions to be engaged. The methodology is illustrated with the case of a network of reinforced concrete beams, which are subjected to a rapid increase in their mechanical stresses. In particular this example shows that investigations on the real performances of materials can provided a significant contribution in managing the risks of unavailability

  4. Strengthening the fission reactor nuclear science and engineering program at UCLA. Final technical report

    Energy Technology Data Exchange (ETDEWEB)

    Okrent, D.

    1997-06-23

    This is the final report on DOE Award No. DE-FG03-92ER75838 A000, a three year matching grant program with Pacific Gas and Electric Company (PG and E) to support strengthening of the fission reactor nuclear science and engineering program at UCLA. The program began on September 30, 1992. The program has enabled UCLA to use its strong existing background to train students in technological problems which simultaneously are of interest to the industry and of specific interest to PG and E. The program included undergraduate scholarships, graduate traineeships and distinguished lecturers. Four topics were selected for research the first year, with the benefit of active collaboration with personnel from PG and E. These topics remained the same during the second year of this program. During the third year, two topics ended with the departure o the students involved (reflux cooling in a PWR during a shutdown and erosion/corrosion of carbon steel piping). Two new topics (long-term risk and fuel relocation within the reactor vessel) were added; hence, the topics during the third year award were the following: reflux condensation and the effect of non-condensable gases; erosion/corrosion of carbon steel piping; use of artificial intelligence in severe accident diagnosis for PWRs (diagnosis of plant status during a PWR station blackout scenario); the influence on risk of organization and management quality; considerations of long term risk from the disposal of hazardous wastes; and a probabilistic treatment of fuel motion and fuel relocation within the reactor vessel during a severe core damage accident.

  5. Summary engineering description of underwater fuel storage facility for foreign research reactor spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Dahlke, H.J.; Johnson, D.A.; Rawlins, J.K.; Searle, D.K.; Wachs, G.W.

    1994-10-01

    This document is a summary description for an Underwater Fuel Storage Facility (UFSF) for foreign research reactor (FRR) spent nuclear fuel (SNF). A FRR SNF environmental Impact Statement (EIS) is being prepared and will include both wet and dry storage facilities as storage alternatives. For the UFSF presented in this document, a specific site is not chosen. This facility can be sited at any one of the five locations under consideration in the EIS. These locations are the Idaho National Engineering Laboratory, Savannah River Site, Hanford, Oak Ridge National Laboratory, and Nevada Test Site. Generic facility environmental impacts and emissions are provided in this report. A baseline fuel element is defined in Section 2.2, and the results of a fission product analysis are presented. Requirements for a storage facility have been researched and are summarized in Section 3. Section 4 describes three facility options: (1) the Centralized-UFSF, which would store the entire fuel element quantity in a single facility at a single location, (2) the Regionalized Large-UFSF, which would store 75% of the fuel element quantity in some region of the country, and (3) the Regionalized Small-UFSF, which would store 25% of the fuel element quantity, with the possibility of a number of these facilities in various regions throughout the country. The operational philosophy is presented in Section 5, and Section 6 contains a description of the equipment. Section 7 defines the utilities required for the facility. Cost estimates are discussed in Section 8, and detailed cost estimates are included. Impacts to worker safety, public safety, and the environment are discussed in Section 9. Accidental releases are presented in Section 10. Standard Environmental Impact Forms are included in Section 11.

  6. Development of Few Group Cross Section Calculation Model for Core Nuclear Design Code CYCAS%堆芯核设计程序CYCAS少群截面模型开发

    Institute of Scientific and Technical Information of China (English)

    杨伟焱; 汤春桃; 毕光文; 杨波

    2016-01-01

    少群截面模型为堆芯三维扩散计算提供实时的节块均匀少群截面,是堆芯计算程序的关键模型之一.CYCAS程序是上海核工程研究设计院最新开发的堆芯三维核设计程序.本文在详细解析影响节块截面的各种因素的基础上,提出应用于CYCAS程序的少群截面的模型.该模型采用能谱修正方法处理由于能谱变化所引入的二次效应,采用微观燃耗修正方法处理燃耗历史效应.单组件和A P1000核电厂的数值验证计算表明,该模型具有很高的计算精度.%The few group cross section calculation model generates node homogeneous few group cross section for core 3D diffusion calculation ,w hich is one of the key models of core calculation code .CYCAS is the new core 3D nuclear design code developed by Shanghai Nuclear Engineering Research & Design Institute (SNERDI) .A new model based on detail analysis of the factors affecting node cross section was developed for CYCAS .In the model ,the energy spectrum correction method was used to process the second order effect introduced by energy spectrum change , and the micro-depletion correction method was utilized to treat depletion history effect .The numerical results of unit assembly and AP1000 core validate the high accuracy of the new model within CYCAS .

  7. SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Benedetti, R. L.; Lords, L. V.; Kiser, D. M.

    1978-02-01

    The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocity and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage.

  8. Computer code and users' guide for the preliminary analysis of dual-mode space nuclear fission solid core power and propulsion systems, NUROC3A. AMS report No. 1239b

    Energy Technology Data Exchange (ETDEWEB)

    Nichols, R.A.; Smith, W.W.

    1976-06-30

    The three-volume report describes a dual-mode nuclear space power and propulsion system concept that employs an advanced solid-core nuclear fission reactor coupled via heat pipes to one of several electric power conversion systems. The second volume describes the computer code and users' guide for the preliminary analysis of the system.

  9. A dominant nuclear mutation in Chlamydomonas identifies a factor controlling chloroplast mRNA stability by acting on the coding region of the atpA transcript.

    Science.gov (United States)

    Drapier, Dominique; Girard-Bascou, Jacqueline; Stern, David B; Wollman, Francis-André

    2002-09-01

    We have characterized a nuclear mutation, mda1-ncc1, that affects mRNA stability for the atpA gene cluster in the chloroplast of Chlamydomonas. Unlike all nuclear mutations altering chloroplast gene expression described to date, mda1-ncc1 is a dominant mutation that still allows accumulation of detectable amounts of atpA mRNAs. At variance with the subset of these mutations that affect mRNA stability through the 5' UTR of a single chloroplast transcript, the mutated version of MDA1 acts on the coding region of the atpA message. We discuss the action of MDA1 in relation to the unusual pattern of expression of atpA that associates particularly short lived-transcripts with a very high translational efficiency.

  10. SPARC-90 code improvement to simulate vapour generator breakages; Mejora del codigo SPARC90 para simular roturas en el generado de vapor

    Energy Technology Data Exchange (ETDEWEB)

    Escriva Castells, A.

    2010-07-01

    The Thermohydraulic and Nuclear Engineering Group of the UPV; together with the CIEMAT; is developing improvements in the decontamination factor calculation in order to introduce them in SPARC90 code (Suppression Pool Aerosol Removal Code). This code calculates the capture of aerosols in pools, which is especially important in case of accident.

  11. Identification of kakusei, a Nuclear Non-Coding RNA, as an Immediate Early Gene from the Honeybee, and Its Application for Neuroethological Study

    Directory of Open Access Journals (Sweden)

    Taketoshi Kiya

    2012-11-01

    Full Text Available The honeybee is a social insect that exhibits various social behaviors. To elucidate the neural basis of honeybee behavior, we detected neural activity in freely-moving honeybee workers using an immediate early gene (IEG that is expressed in a neural activity-dependent manner. In European honeybees (Apis mellifera, we identified a novel nuclear non-coding RNA, termed kakusei, as the first insect IEG, and revealed the neural activity pattern in foragers. In addition, we isolated a homologue of kakusei, termed Acks, from the Japanese honeybee (Apis cerana, and detected active neurons in workers fighting with the giant hornet.

  12. Identification of kakusei, a nuclear non-coding RNA, as an immediate early gene from the honeybee, and its application for neuroethological study.

    Science.gov (United States)

    Kiya, Taketoshi; Ugajin, Atsushi; Kunieda, Takekazu; Kubo, Takeo

    2012-11-22

    The honeybee is a social insect that exhibits various social behaviors. To elucidate the neural basis of honeybee behavior, we detected neural activity in freely-moving honeybee workers using an immediate early gene (IEG) that is expressed in a neural activity-dependent manner. In European honeybees (Apis mellifera), we identified a novel nuclear non-coding RNA, termed kakusei, as the first insect IEG, and revealed the neural activity pattern in foragers. In addition, we isolated a homologue of kakusei, termed Acks, from the Japanese honeybee (Apis cerana), and detected active neurons in workers fighting with the giant hornet.

  13. Second conference on nuclear science and engineering in Australia, 1997. Conference handbook

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-31

    The conference handbook contains the text of papers presented orally and as posters. Leading experts in various areas of nuclear science and technology discussed the following topics: uranium resources, radioactive waste management, research reactor safety and applications, radiation and related research, applications of accelerators and related facilities and nuclear regulation in Australia. The posters include two from the winners of the David Culley Award in 1995 and 1996, instituted by the Australian Nuclear Association to encourage work in nuclear science and technology in school and colleges.

  14. Proceedings of the 2013 International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering - M and C 2013

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-07-01

    The Mathematics and Computation Division of the American Nuclear (ANS) and the Idaho Section of the ANS hosted the 2013 International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M and C 2013). This proceedings contains over 250 full papers with topics ranging from reactor physics; radiation transport; materials science; nuclear fuels; core performance and optimization; reactor systems and safety; fluid dynamics; medical applications; analytical and numerical methods; algorithms for advanced architectures; and validation verification, and uncertainty quantification.

  15. The Air Force Nuclear Engineering Center Structural Activation and Integrity Evaluation

    Science.gov (United States)

    1990-03-01

    topic involves operation of a major computer pro- gram, as this one, ORIGEN2 , it is always questionable about whether you can complete the required...Material Composition..............................9 ORIGEN2 Computer Code............................10 ORIGEN2 Results...A-1 Appendix B: ORIGEN2 Computer Code Input................ B-i Appendix C: ORIGEN2 Results............................. C-i Appendix D

  16. Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Heuser, Brent [Univ. of Illinois, Urbana-Champaign, IL (United States); Stubbins, James [Univ. of Illinois, Urbana-Champaign, IL (United States); Kozlowski, Tomasz [Univ. of Illinois, Urbana-Champaign, IL (United States); Uddin, Rizwan [Univ. of Illinois, Urbana-Champaign, IL (United States); Trinkle, Dallas [Univ. of Illinois, Urbana-Champaign, IL (United States); Downar, Thoms [Univ. of Michigan, Ann Arbor, MI (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States); ang, Yong [Univ. of Florida, Gainesville, FL (United States); Phillpot, Simon [Univ. of Florida, Gainesville, FL (United States); Sabharwall, piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-07-25

    The DOE NEUP sponsored IRP on accident tolerant fuel (ATF) entitled Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel involved three academic institutions, Idaho National Laboratory (INL), and ATI Materials (ATI). Detailed descriptions of the work at the University of Illinois (UIUC, prime), the University of Florida (UF), the University of Michigan (UMich), and INL are included in this document as separate sections. This summary provides a synopsis of the work performed across the IRP team. Two ATF solution pathways were initially proposed, coatings on monolithic Zr-based LWR cladding material and selfhealing modifications of Zr-based alloys. The coating pathway was extensively investigated, both experimentally and in computations. Experimental activities related to ATF coatings were centered at UIUC, UF, and UMich and involved coating development and testing, and ion irradiation. Neutronic and thermal hydraulic aspects of ATF coatings were the focus of computational work at UIUC and UMich, while materials science aspects were the focus of computational work at UF and INL. ATI provided monolithic Zircaloy 2 and 4 material and a binary Zr-Y alloy material. The selfhealing pathway was investigated with advanced computations only. Beryllium was identified as a valid self-healing additive early in this work. However, all attempts to fabricate a Zr-Be alloy failed. Several avenues of fabrication were explored. ATI ultimately declined our fabrication request over health concerns associated with Be (we note that Be was not part of the original work scope and the ATI SOW). Likewise, Ames Laboratory declined our fabrication request, citing known litigation dating to the 1980s and 1990s involving the U.S. Federal government and U.S. National Laboratory employees involving the use of Be. Materion (formerly, Brush Wellman) also declined our fabrication request, citing the difficulty in working with a highly reactive Zr and Be

  17. Assessing controls on perched saturated zones beneath the Idaho Nuclear Technology and Engineering Center, Idaho

    Science.gov (United States)

    Mirus, Benjamin B.; Perkins, Kim S.; Nimmo, John R.

    2011-01-01

    Waste byproducts associated with operations at the Idaho Nuclear Technology and Engineering Center (INTEC) have the potential to contaminate the eastern Snake River Plain (ESRP) aquifer. Recharge to the ESRP aquifer is controlled largely by the alternating stratigraphy of fractured volcanic rocks and sedimentary interbeds within the overlying vadose zone and by the availability of water at the surface. Beneath the INTEC facilities, localized zones of saturation perched on the sedimentary interbeds are of particular concern because they may facilitate accelerated transport of contaminants. The sources and timing of natural and anthropogenic recharge to the perched zones are poorly understood. Simple approaches for quantitative characterization of this complex, variably saturated flow system are needed to assess potential scenarios for contaminant transport under alternative remediation strategies. During 2009-2011, the U.S. Geological Survey (USGS), in cooperation with the U.S. Department of Energy, employed data analysis and numerical simulations with a recently developed model of preferential flow to evaluate the sources and quantity of recharge to the perched zones. Piezometer, tensiometer, temperature, precipitation, and stream-discharge data were analyzed, with particular focus on the possibility of contributions to the perched zones from snowmelt and flow in the neighboring Big Lost River (BLR). Analysis of the timing and magnitude of subsurface dynamics indicate that streamflow provides local recharge to the shallow, intermediate, and deep perched saturated zones within 150 m of the BLR; at greater distances from the BLR the influence of streamflow on recharge is unclear. Perched water-level dynamics in most wells analyzed are consistent with findings from previous geochemical analyses, which suggest that a combination of annual snowmelt and anthropogenic sources (for example, leaky pipes and drainage ditches) contribute to recharge of shallow and

  18. High-Fidelity Space-Time Adaptive Multiphysics Simulations in Nuclear Engineering

    Energy Technology Data Exchange (ETDEWEB)

    Solin, Pavel [Univ. of Reno, NV (United States); Ragusa, Jean [Texas A & M Univ., College Station, TX (United States)

    2014-03-09

    We delivered a series of fundamentally new computational technologies that have the potential to significantly advance the state-of-the-art of computer simulations of transient multiphysics nuclear reactor processes. These methods were implemented in the form of a C++ library, and applied to a number of multiphysics coupled problems relevant to nuclear reactor simulations.

  19. Updating of ASME Nuclear Code Case N-201 to Accommodate the Needs of Metallic Core Support Structures for High Temperature Gas Cooled Reactors Currently in Development

    Energy Technology Data Exchange (ETDEWEB)

    Mit Basol; John F. Kielb; John F. MuHooly; Kobus Smit

    2007-05-02

    On September 29, 2005, ASME Standards Technology, LLC (ASME ST-LLC) executed a multi-year, cooperative agreement with the United States DOE for the Generation IV Reactor Materials project. The project's objective is to update and expand appropriate materials, construction, and design codes for application in future Generation IV nuclear reactor systems that operate at elevated temperatures. Task 4 was embarked upon in recognition of the large quantity of ongoing reactor designs utilizing high temperature technology. Since Code Case N-201 had not seen a significant revision (except for a minor revision in September, 2006 to change the SA-336 forging reference for 304SS and 316SS to SA-965 in Tables 1.2(a) and 1.2(b), and some minor editorial changes) since December 1994, identifying recommended updates to support the current high temperature Core Support Structure (CSS) designs and potential new designs was important. As anticipated, the Task 4 effort identified a number of Code Case N-201 issues. Items requiring further consideration range from addressing apparent inconsistencies in definitions and certain material properties between CC-N-201 and Subsection NH, to inclusion of additional materials to provide the designer more flexibility of design. Task 4 developed a design parameter survey that requested input from the CSS designers of ongoing high temperature gas cooled reactor metallic core support designs. The responses to the survey provided Task 4 valuable input to identify the design operating parameters and future needs of the CSS designers. Types of materials, metal temperature, time of exposure, design pressure, design life, and fluence levels were included in the Task 4 survey responses. The results of the survey are included in this report. This research proves that additional work must be done to update Code Case N-201. Task 4 activities provide the framework for the Code Case N-201 update and future work to provide input on materials. Candidate

  20. Instrumentation Requirements for the Engineering Evaluation of Nuclear-Electric Spacecraft

    Science.gov (United States)

    Apel, W. C.

    1961-01-01

    Spacecraft employing nuclear-electric propulsion are being proposed for missions to Venus and distances beyond. These spacecraft utilize a nuclear reactor to provide thermal energy to a turboalternator which generates electric power for an ion motor and the other spacecraft systems. This Report discusses the instrumentation and communications system needed to evaluate a nuclear-electric spacecraft in flight, along with the problems expected. A representative spacecraft design is presented, which leads to a discussion of the instrumentation needed to evaluate such a spacecraft. A basic communications system is considered for transmitting the spacecraft data to Earth. The instrumentation and communications system, as well as all electronic systems on a nuclear-electric spacecraft, will be operating in high temperature and nuclear-radiation environments. The problems caused by these environments are discussed, and possible solutions are offered.

  1. Instrumentation Requirements for the Engineering Evaluation of Nuclear-Electric Spacecraft

    Science.gov (United States)

    Apel, W. C.

    1961-01-01

    Spacecraft employing nuclear-electric propulsion are being proposed for missions to Venus and distances beyond. These spacecraft utilize a nuclear reactor to provide thermal energy to a turboalternator which generates electric power for an ion motor and the other spacecraft systems. This Report discusses the instrumentation and communications system needed to evaluate a nuclear-electric spacecraft in flight, along with the problems expected. A representative spacecraft design is presented, which leads to a discussion of the instrumentation needed to evaluate such a spacecraft. A basic communications system is considered for transmitting the spacecraft data to Earth. The instrumentation and communications system, as well as all electronic systems on a nuclear-electric spacecraft, will be operating in high temperature and nuclear-radiation environments. The problems caused by these environments are discussed, and possible solutions are offered.

  2. Dakota Uncertainty Quantification Methods Applied to the CFD code Nek5000

    Energy Technology Data Exchange (ETDEWEB)

    Delchini, Marc-Olivier [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Popov, Emilian L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Pointer, William David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division

    2016-04-29

    This report presents the state of advancement of a Nuclear Energy Advanced Modeling and Simulation (NEAMS) project to characterize the uncertainty of the computational fluid dynamics (CFD) code Nek5000 using the Dakota package for flows encountered in the nuclear engineering industry. Nek5000 is a high-order spectral element CFD code developed at Argonne National Laboratory for high-resolution spectral-filtered large eddy simulations (LESs) and unsteady Reynolds-averaged Navier-Stokes (URANS) simulations.

  3. Dakota Uncertainty Quantification Methods Applied to the CFD code Nek5000

    Energy Technology Data Exchange (ETDEWEB)

    Delchini, Marc-Olivier [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Popov, Emilian L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Pointer, William David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division

    2016-04-29

    This report presents the state of advancement of a Nuclear Energy Advanced Modeling and Simulation (NEAMS) project to characterize the uncertainty of the computational fluid dynamics (CFD) code Nek5000 using the Dakota package for flows encountered in the nuclear engineering industry. Nek5000 is a high order spectral element CFD code developed at Argonne National Laboratory for high resolution spectral-filtered large eddy simulations (LESs) and unsteady Reynolds averaged Navier-Stokes (URANS) simulations.

  4. Investigating the burning characteristics of electric cables used in the nuclear power plant by way of 3-D transient FDS code

    Energy Technology Data Exchange (ETDEWEB)

    Ferng, Y.M., E-mail: ymferng@ess.nthu.edu.t [Department of Engineering and System Science, Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Sec. 2. Kuang-Fu Rd., Hsinchu 30013, Taiwan (China); Liu, C.H. [Department of Engineering and System Science, Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Sec. 2. Kuang-Fu Rd., Hsinchu 30013, Taiwan (China)

    2011-01-15

    Burning characteristics of electrical cables are one of the key parameters for the fire hazard assessment of nuclear power plants (NPPs) since the cables are the essential sources of fire in the plants. A three-dimensional (3-D) transient computational fluid dynamics (CFD) code{sub F}DS is adopted in this paper to simulate these characteristics related to the cable burning. Being one of the NRC licensing fire codes, the FDS includes the thermal-hydraulic equations, the turbulence model and the chemical combustion model, etc. In order to assess the CFD fire models used in this code, a burning test using the control cable with the outer jacket of polyvinylchloride (PVC) and the inner insulation of cross-linked polyethylene (XLPE) is conducted. The measured parameters associated with the burning characteristics include the heat release rate (HRR), O{sub 2} depletion, and CO and CO{sub 2} production, etc. Except the amount of O{sub 2} consumption, the predicted transient behaviors of other parameters can reproduce the measured data. Based on the chemical combustion model in the FDS code, this discrepancy may be essentially resulted from the default value of hydrogen fraction (H{sub frac}) contained in the soot since the soot yield for the burning of PVC material is high enough that the uncertainty in the H{sub frac} value has a prominent effect on the amount of O{sub 2} consumption. This explanation can be confirmed by a benchmark calculation for simulating a burning test with the polymethylmethacrylate (PMMA) fuel of low-soot yield. The present simulation works can provide the useful information for the plant staff or the researcher as they would perform the fire hazard analysis in the NPPs using the FDS code.

  5. Study of advanced professional educational requirements relative to nuclear fuel cycle engineering in industry and government. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Jur, T.A.; Huhns, M.N.; Keating, D.A.; Orloff, D.I.; Rhodes, C.A.; Stanford, T.G.; Stephens, L.M.; Tatterson, G.B.; Van Brunt, V.

    1978-12-01

    An assessment was conducted of educational needs among engineers working in nuclear fuel cycle-related areas, focusing on the nuclear industry in the Southeast. Educational needs addressed were those at the post-baccalaureate professional level. As a result of the study, a list of subject areas has been compiled as best representing the current content of an educational program. In addition to identifying subject areas, a set of course descriptions and reference materials has been developed around each subject. Each course description contains information regarding objectives, anticipated audience, and prerequisites and offers a suggested course outline. An initial modest program of implementation is recommended which would continue to concentrate on the Southeast as a target area.

  6. Radiant Energy Measurements from a Scaled Jet Engine Axisymmetric Exhaust Nozzle for a Baseline Code Validation Case

    Science.gov (United States)

    Baumeister, Joseph F.

    1994-01-01

    A non-flowing, electrically heated test rig was developed to verify computer codes that calculate radiant energy propagation from nozzle geometries that represent aircraft propulsion nozzle systems. Since there are a variety of analysis tools used to evaluate thermal radiation propagation from partially enclosed nozzle surfaces, an experimental benchmark test case was developed for code comparison. This paper briefly describes the nozzle test rig and the developed analytical nozzle geometry used to compare the experimental and predicted thermal radiation results. A major objective of this effort was to make available the experimental results and the analytical model in a format to facilitate conversion to existing computer code formats. For code validation purposes this nozzle geometry represents one validation case for one set of analysis conditions. Since each computer code has advantages and disadvantages based on scope, requirements, and desired accuracy, the usefulness of this single nozzle baseline validation case can be limited for some code comparisons.

  7. The outlook for application of powerful nuclear thermionic reactor - powered space electric jet propulsion engines

    Energy Technology Data Exchange (ETDEWEB)

    Semyonov, Y.P.; Bakanov, Y.A.; Synyavsky, V.V.; Yuditsky, V.D. [Rocket-Space Corp. `Energia`, Moscow (Russian Federation)

    1997-12-31

    This paper summarizes main study results for application of powerful space electric jet propulsion unit (EJPUs) which is powered by Nuclear Thermionic Power Unit (NTPU). They are combined in Nuclear Power/Propulsion Unit (NPPU) which serves as means of spacecraft equipment power supply and spacecraft movement. Problems the paper deals with are the following: information satellites delivery and their on-orbit power supply during 10-15 years, removal of especially hazardous nuclear wastes, mining of asteroid resources and others. Evaluations on power/time/mass relationship for this type of mission are given. EJPU parameters are compatible with Russian existent or being under development launch vehicle. (author)

  8. Affordable Development and Demonstration of a Small Nuclear Thermal Rocket (NTR) Engine and Stage: How Small Is Big Enough?

    Science.gov (United States)

    Borowski, Stanley K.; Sefcik, Robert J.; Fittje, James E.; McCurdy, David R.; Qualls, Arthur L.; Schnitzler, Bruce G.; Werner, James E.; Weitzberg, Abraham; Joyner, Claude R.

    2016-01-01

    The Nuclear Thermal Rocket (NTR) derives its energy from fission of uranium-235 atoms contained within fuel elements that comprise the engine's reactor core. It generates high thrust and has a specific impulse potential of approximately 900 specific impulse - a 100 percent increase over today's best chemical rockets. The Nuclear Thermal Propulsion (NTP) project, funded by NASA's Advanced Exploration Systems (AES) program, includes five key task activities: (1) Recapture, demonstration, and validation of heritage graphite composite (GC) fuel (selected as the Lead Fuel option); (2) Engine Conceptual Design; (3) Operating Requirements Definition; (4) Identification of Affordable Options for Ground Testing; and (5) Formulation of an Affordable Development Strategy. During fiscal year (FY) 2014, a preliminary Design Development Test and Evaluation (DDT&E) plan and schedule for NTP development was outlined by the NASA Glenn Research Center (GRC), Department of Energy (DOE) and industry that involved significant system-level demonstration projects that included Ground Technology Demonstration (GTD) tests at the Nevada National Security Site (NNSS), followed by a Flight Technology Demonstration (FTD) mission. To reduce cost for the GTD tests and FTD mission, small NTR engines, in either the 7.5 or 16.5 kilopound-force thrust class, were considered. Both engine options used GC fuel and a common fuel element (FE) design. The small approximately 7.5 kilopound-force criticality-limited engine produces approximately157 thermal megawatts and its core is configured with parallel rows of hexagonal-shaped FEs and tie tubes (TTs) with a FE to TT ratio of approximately 1:1. The larger approximately 16.5 kilopound-force Small Nuclear Rocket Engine (SNRE), developed by Los Alamos National Laboratory (LANL) at the end of the Rover program, produces approximately 367 thermal megawatts and has a FE to TT ratio of approximately 2:1. Although both engines use a common 35-inch (approximately

  9. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-98 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Herbst, Alan Keith; Mc Cray, John Alan; Rogers, Adam Zachary; Simmons, R. F.; Palethorpe, S. J.

    1999-03-01

    The Low-Activity Waste Process Technology Program at the Idaho Nuclear Technology and Engineering Center (INTEC) anticipates that large volumes of low-level/low-activity wastes will need to be grouted prior to near-surface disposal. During fiscal year 1998, three grout formulations were studied for low-activity wastes derived from INTEC liquid sodium-bearing waste. Compressive strength and leach results are presented for phosphate bonding cement, acidic grout, and alkaline grout formulations. In an additional study, grout formulations are recommended for stabilization of the INTEC underground storage tank residual heels.

  10. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program, FY-98 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Herbst, A.K.; Rogers, A.Z.; McCray, J.A.; Simmons, R.F.; Palethorpe, S.J.

    1999-03-01

    The Low-Activity Waste Process Technology Program at the Idaho Nuclear Technology and Engineering Center (INTEC) anticipates that large volumes of low-level/low-activity wastes will need to be grouted prior to near-surface disposal. During fiscal year 1998, three grout formulations were studied for low-activity wastes derived from INTEC liquid sodium-bearing waste. Compressive strength and leach results are presented for phosphate bonding cement, acidic grout, and alkaline grout formulations. In an additional study, grout formulations are recommended for stabilization of the INTEC underground storage tank residual heels.

  11. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-99 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Herbst, Alan Keith; Mc Cray, John Alan; Kirkham, Robert John; Pao, Jenn Hai; Hinckley, Steve Harold

    1999-10-01

    The Low-Activity Waste Process Technology Program at the Idaho Nuclear Technology and Engineering Center (INTEC) anticipates that large volumes of low-level/low-activity wastes will need to be grouted prior to near-surface disposal. During fiscal year 1999, grout formulations were studied for transuranic waste derived from INTEC liquid sodium-bearing waste and for projected newly generated low-level liquid waste. Additional studies were completed on radionuclide leaching, microbial degradation, waste neutralization, and a small mockup for grouting the INTEC underground storage tank residual heels.

  12. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-99 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    A. K. Herbst; J. A. McCray; R. J. Kirkham; J. Pao; S. H. Hinckley

    1999-09-30

    The Low-Activity Waste Process Technology Program at the Idaho Nuclear Technology and Engineering Center (INTEC) anticipates that large volumes of low-level/low-activity wastes will need to be grouted prior to near-surface disposal. During fiscal year 1999, grout formulations were studied for transuranic waste derived from INTEC liquid sodium-bearing waste and for projected newly generated low-level liquid waste. Additional studies were completed on radionuclide leaching, microbial degradation, waste neutralization, and a small mockup for grouting the INTEC underground storage tank residual heels.

  13. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-2000 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Herbst, Alan Keith; Mc Cray, John Alan; Kirkham, Robert John; Pao, Jenn Hai; Argyle, Mark Don; Lauerhass, Lance; Bendixsen, Carl Lee; Hinckley, Steve Harold

    2000-11-01

    The Low-Activity Waste Process Technology Program anticipated that grouting will be used for disposal of low-level and transuranic wastes generated at the Idaho Nuclear Technology Engineering Center (INTEC). During fiscal year 2000, grout formulations were studied for transuranic waste derived from INTEC liquid sodium-bearing waste and for projected newly generated low-level liquid waste. Additional studies were completed using silica gel and other absorbents to solidify sodium-bearing wastes. A feasibility study and conceptual design were completed for the construction of a grout pilot plant for simulated wastes and demonstration facility for actual wastes.

  14. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-2000 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Herbst, A.K.; McCray, J.A.; Kirkham, R.J.; Pao, J.; Argyle, M.D.; Lauerhass, L.; Bendixsen, C.L.; Hinckley, S.H.

    2000-10-31

    The Low-Activity Waste Process Technology Program anticipated that grouting will be used for disposal of low-level and transuranic wastes generated at the Idaho Nuclear Technology Engineering Center (INTEC). During fiscal year 2000, grout formulations were studied for transuranic waste derived from INTEC liquid sodium-bearing waste and for projected newly generated low-level liquid waste. Additional studies were completed using silica gel and other absorbents to solidify sodium-bearing wastes. A feasibility study and conceptual design were completed for the construction of a grout pilot plant for simulated wastes and demonstration facility for actual wastes.

  15. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program, FY-98 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Herbst, A.K.; Rogers, A.Z.; McCray, J.A.; Simmons, R.F.; Palethorpe, S.J.

    1999-03-01

    The Low-Activity Waste Process Technology Program at the Idaho Nuclear Technology and Engineering Center (INTEC) anticipates that large volumes of low-level/low-activity wastes will need to be grouted prior to near-surface disposal. During fiscal year 1998, three grout formulations were studied for low-activity wastes derived from INTEC liquid sodium-bearing waste. Compressive strength and leach results are presented for phosphate bonding cement, acidic grout, and alkaline grout formulations. In an additional study, grout formulations are recommended for stabilization of the INTEC underground storage tank residual heels.

  16. Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-98 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Herbst, A.K.; McCray, J.A.; Rogers, A.Z.; Simmons, R.F.; Palethrope, S.J.

    1999-03-01

    The Low-Activity Waste Process Technology Program at the Idaho Nuclear Technology and Engineering Center (INTEC) anticipates that large volumes of low-level/low-activity wastes will need to be grouted prior to near-surface disposal. During fiscal year 1998, three grout formulations were studied for low-activity wastes derived from INTEC liquid sodium-bearing waste. Compressive strength and leach results are presented for phosphate bonding cement, acidic grout, and alkaline grout formulations. In an additional study, grout formulations are recommended for stabilization of the INTEC underground storage tank residual heels.

  17. Nuclear Energy -- Knowledge Base for Advanced Modeling and Simulation (NE-KAMS) Code Verification and Validation Data Standards and Requirements: Fluid Dynamics Version 1.0

    Energy Technology Data Exchange (ETDEWEB)

    Greg Weirs; Hyung Lee

    2011-09-01

    V&V and UQ are the primary means to assess the accuracy and reliability of M&S and, hence, to establish confidence in M&S. Though other industries are establishing standards and requirements for the performance of V&V and UQ, at present, the nuclear industry has not established such standards or requirements. However, the nuclear industry is beginning to recognize that such standards are needed and that the resources needed to support V&V and UQ will be very significant. In fact, no single organization has sufficient resources or expertise required to organize, conduct and maintain a comprehensive V&V and UQ program. What is needed is a systematic and standardized approach to establish and provide V&V and UQ resources at a national or even international level, with a consortium of partners from government, academia and industry. Specifically, what is needed is a structured and cost-effective knowledge base that collects, evaluates and stores verification and validation data, and shows how it can be used to perform V&V and UQ, leveraging collaboration and sharing of resources to support existing engineering and licensing procedures as well as science-based V&V and UQ processes. The Nuclear Energy Knowledge base for Advanced Modeling and Simulation (NE-KAMS) is being developed at the Idaho National Laboratory in conjunction with Bettis Laboratory, Sandia National Laboratories, Argonne National Laboratory, Utah State University and others with the objective of establishing a comprehensive and web-accessible knowledge base to provide V&V and UQ resources for M&S for nuclear reactor design, analysis and licensing. The knowledge base will serve as an important resource for technical exchange and collaboration that will enable credible and reliable computational models and simulations for application to nuclear power. NE-KAMS will serve as a valuable resource for the nuclear industry, academia, the national laboratories, the U.S. Nuclear Regulatory Commission (NRC) and

  18. 堆芯核设计程序CYCAS动力学模型开发%Development of Kinetics Model in Core Nuclear Design Code CYCAS

    Institute of Scientific and Technical Information of China (English)

    毕光文; 汤春桃; 杨波

    2016-01-01

    The kinetics model and its numerical verification were studied for core nuclear design code CYCAS .The kinetics model employed by CYCAS code was introduced in detail .In order to verify the effectiveness of the kinetics model , the L M W transient benchmark and the dynamic insertion issue of control rod in AP1000 core were simulated and analyzed .The calculation results show that the kinetics model of CYCAS code could obtain reliable results .%对堆芯核设计程序CYCAS的动力学模型及其数值验证进行了研究.详细介绍了CYCAS程序采用的动力学模型.为验证模型的有效性,对L M W瞬态基准题和基于AP1000堆芯动态插棒问题进行了数值模拟和分析.结果表明,CYCAS程序的动力学模型可获得可靠的计算结果.

  19. Study of a new design of p-N semiconductor detector array for nuclear medicine imaging by monte carlo simulation codes.

    Science.gov (United States)

    Hajizadeh-Safar, M; Ghorbani, M; Khoshkharam, S; Ashrafi, Z

    2014-07-01

    Gamma camera is an important apparatus in nuclear medicine imaging. Its detection part is consists of a scintillation detector with a heavy collimator. Substitution of semiconductor detectors instead of scintillator in these cameras has been effectively studied. In this study, it is aimed to introduce a new design of P-N semiconductor detector array for nuclear medicine imaging. A P-N semiconductor detector composed of N-SnO2 :F, and P-NiO:Li, has been introduced through simulating with MCNPX monte carlo codes. Its sensitivity with different factors such as thickness, dimension, and direction of emission photons were investigated. It is then used to configure a new design of an array in one-dimension and study its spatial resolution for nuclear medicine imaging. One-dimension array with 39 detectors was simulated to measure a predefined linear distribution of Tc(99_m) activity and its spatial resolution. The activity distribution was calculated from detector responses through mathematical linear optimization using LINPROG code on MATLAB software. Three different configurations of one-dimension detector array, horizontal, vertical one sided, and vertical double-sided were simulated. In all of these configurations, the energy windows of the photopeak were ± 1%. The results show that the detector response increases with an increase of dimension and thickness of the detector with the highest sensitivity for emission photons 15-30° above the surface. Horizontal configuration array of detectors is not suitable for imaging of line activity sources. The measured activity distribution with vertical configuration array, double-side detectors, has no similarity with emission sources and hence is not suitable for imaging purposes. Measured activity distribution using vertical configuration array, single side detectors has a good similarity with sources. Therefore, it could be introduced as a suitable configuration for nuclear medicine imaging. It has been shown that using

  20. MMSNF 2005. Materials models and simulations for nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Freyss, M.; Durinck, J.; Carlot, G.; Sabathier, C.; Martin, P.; Garcia, P.; Ripert, M.; Blanpain, P.; Lippens, M.; Schut, H.; Federov, A.V.; Bakker, K.; Osaka, M.; Miwa, S.; Sato, I.; Tanaka, K.; Kurosaki, K.; Uno, M.; Yamanaka, S.; Govers, K.; Verwerft, M.; Hou, M.; Lemehov, S.E.; Terentyev, D.; Govers, K.; Kotomin, E.A.; Ashley, N.J.; Grimes, R.W.; Van Uffelen, P.; Mastrikov, Y.; Zhukovskii, Y.; Rondinella, V.V.; Kurosaki, K.; Uno, M.; Yamanaka, S.; Minato, K.; Phillpot, S.; Watanabe, T.; Shukla, P.; Sinnott, S.; Nino, J.; Grimes, R.; Staicu, D.; Hiernaut, J.P.; Wiss, T.; Rondinella, V.V.; Ronchi, C.; Yakub, E.; Kaye, M.H.; Morrison, C.; Higgs, J.D.; Akbari, F.; Lewis, B.J.; Thompson, W.T.; Gueneau, C.; Gosse, S.; Chatain, S.; Dumas, J.C.; Sundman, B.; Dupin, N.; Konings, R.; Noel, H.; Veshchunov, M.; Dubourg, R.; Ozrin, C.V.; Veshchunov, M.S.; Welland, M.T.; Blanc, V.; Michel, B.; Ricaud, J.M.; Calabrese, R.; Vettraino, F.; Tverberg, T.; Kissane, M.; Tulenko, J.; Stan, M.; Ramirez, J.C.; Cristea, P.; Rachid, J.; Kotomin, E.; Ciriello, A.; Rondinella, V.V.; Staicu, D.; Wiss, T.; Konings, R.; Somers, J.; Killeen, J

    2006-07-01

    The MMSNF Workshop series aims at stimulating research and discussions on models and simulations of nuclear fuels and coupling the results into fuel performance codes.This edition was focused on materials science and engineering for fuel performance codes. The presentations were grouped in three technical sessions: fundamental modelling of fuel properties; integral fuel performance codes and their validation; collaborations and integration of activities. (A.L.B.)

  1. Terpene metabolic engineering via nuclear or chloroplast genomes profoundly and globally impacts off-target pathways through metabolite signalling.

    Science.gov (United States)

    Pasoreck, Elise K; Su, Jin; Silverman, Ian M; Gosai, Sager J; Gregory, Brian D; Yuan, Joshua S; Daniell, Henry

    2016-09-01

    The impact of metabolic engineering on nontarget pathways and outcomes of metabolic engineering from different genomes are poorly understood questions. Therefore, squalene biosynthesis genes FARNESYL DIPHOSPHATE SYNTHASE (FPS) and SQUALENE SYNTHASE (SQS) were engineered via the Nicotiana tabacum chloroplast (C), nuclear (N) or both (CN) genomes to promote squalene biosynthesis. SQS levels were ~4300-fold higher in C and CN lines than in N, but all accumulated ~150-fold higher squalene due to substrate or storage limitations. Abnormal leaf and flower phenotypes, including lower pollen production and reduced fertility, were observed regardless of the compartment or level of transgene expression. Substantial changes in metabolomes of all lines were observed: levels of 65-120 unrelated metabolites, including the toxic alkaloid nicotine, changed by as much as 32-fold. Profound effects of transgenesis on nontarget gene expression included changes in the abundance of 19 076 transcripts by up to 2000-fold in CN; 7784 transcripts by up to 1400-fold in N; and 5224 transcripts by as much as 2200-fold in C. Transporter-related transcripts were induced, and cell cycle-associated transcripts were disproportionally repressed in all three lines. Transcriptome changes were validated by qRT-PCR. The mechanism underlying these large changes likely involves metabolite-mediated anterograde and/or retrograde signalling irrespective of the level of transgene expression or end product, due to imbalance of metabolic pools, offering new insight into both anticipated and unanticipated consequences of metabolic engineering.

  2. Investigations on boron carbide oxidation for nuclear reactors safety-General modelling for ICARE/CATHARE code applications

    Energy Technology Data Exchange (ETDEWEB)

    Seiler, N. [Institut de Radioprotection et de Surete Nucleaire (IRSN), Cadarache, BP 3, 13 115 Saint Paul lez Durance Cedex (France)], E-mail: nathalie.seiler@irsn.fr; Bertrand, F.; Marchand, O.; Repetto, G. [Institut de Radioprotection et de Surete Nucleaire (IRSN), Cadarache, BP 3, 13 115 Saint Paul lez Durance Cedex (France); Ederli, S. [ENEA, Ente per le Nuove Tecnologie l' Energia et l' Ambiente (Italy)

    2008-04-15

    The present paper deals with the problem of boron carbide pellet oxidation which might occur during a severe accident. A basic correlation, involving global variables, has been developed for the simulation of boron carbide oxidation with the ICARE/CATHARE code. This modelling has been based on available experimental data, including the VERDI separate effects experiments performed by IRSN at low pressures and high temperatures. According to the agreement between the measured and the calculated bundle temperatures as well as hydrogen release and oxidized B{sub 4}C, the ICARE/CATHARE code simulates rather well QUENCH experiments involving B{sub 4}C control rod degradation, Zircaloy oxidation under starvation and cooling with steam. Based on simulations results, it has been noticed that the B{sub 4}C degradation has a slight direct effect on global bundle degradation but a non-negligible influence on Zircaloy oxidation through power release, material melting and flowing down.

  3. SCDAP/RELAP5 code development and assessment

    Energy Technology Data Exchange (ETDEWEB)

    Allison, C.M.; Hohorst, J.K. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

    1996-03-01

    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.

  4. Simulation of an operation cycle of nuclear power plant of Laguna Verde with code TACHY and computation package CMS; Simulacion de un ciclo de operacion de la CNLV con el codigo TACHY y el paquete de computo CMS

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, J. A.; Del Valle, E.; Vargas, S.; Xolocostli, J. V. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: joseangel.gonzalez@inin.gob.mx

    2009-10-15

    In this work the code TACHY is used, to simulate an operation cycle of the nuclear power plant of Laguna Verde. The code TACHY was designed originally to analyze recharge patterns of Hindu plants type BWR, that have near 800 assemblies, that is almost double the reactor of nuclear power plant of Laguna Verde. For this reason it was necessary to modify the code to be able to apply it to nuclear power plant of Laguna Verde. The values were modified like: operation power, entrance subcooling, flow through the nucleus, assemblies number in nucleus and dimensions of nucleus. In this work is take like base the cycle 9 of Unit 2 of nuclear power plant of Laguna Verde. This cycle is simulated with code TACHY and with code SIMULATE-3 that is part of computation package Core Management System, with the purpose of comparing the results. The results that are compared with the two codes, for the complete nucleus are: the burnt average of nucleus, the cycle longitude, the effective factor of neutrons multiplication, the pick of radial relative power; and for each assembly: the burnt and the relative power. Of the results obtained with TACHY we can conclude that we have a computation tool that allows to analyze a great number of recharge patterns in a reasonable time. (Author)

  5. Genome-Wide Survey of Nuclear Protein-Coding Markers for Beetle Phylogenetics and Their Application in Resolving both Deep and Shallow-Level Divergences.

    Science.gov (United States)

    Che, Li-Heng; Zhang, Shao-Qian; Li, Yun; Liang, Dan; Pang, Hong; Ślipiński, Adam; Zhang, Peng

    2017-03-03

    Beetles (Coleoptera) are the most diverse and species-rich insect group, representing an impressive explosive radiation in the evolutionary history of insects, and their evolutionary relationships are often difficult to resolve. The amount of "traditional markers" (e.g., mitochondrial genes and nuclear rDNAs) for beetle phylogenetics is small and these markers often lack sufficient signals in resolving relationships for such a rapidly radiating lineage. Here, based on the available genome data of beetles and other related insect species, we performed a genome-wide survey to search nuclear protein-coding (NPC) genes suitable for research on beetle phylogenetics. As a result, we identified 1470 candidate loci, which provided a valuable data resource to the beetle evolutionary research community for NPC marker development. We randomly chose 180 candidate loci from the database to design primers and successfully developed 95 NPC markers which can be PCR amplified from standard genomic DNA extracts. These new nuclear markers are universally applicable across Coleoptera, with an average amplification success rate of 90%. To test the phylogenetic utility, we used them to investigate the backbone phylogeny of Coleoptera (18 families sampled) and the family Coccinellidae (39 species sampled). Both phylogenies are well resolved (average bootstrap support > 95%), showing that our markers can be used to address phylogenetic questions of various evolutionary depth (from species level to family level). In general, the newly developed nuclear markers are much easier to use and more phylogenetically informative than the "traditional markers", and show great potential to expedite resolution of many parts in the Beetle Tree of Life. This article is protected by copyright. All rights reserved.

  6. Development of an interface between MCNP and ORIGEN codes for calculations of fuel evolution in nuclear systems. Initial project; Desenvolvimento de uma interface entre os codigos MCNP e ORIGEN para calculos de evolucao de combustiveis em sistemas nucleares. Projeto inicial

    Energy Technology Data Exchange (ETDEWEB)

    Campolina, Daniel de Almeida Magalhaes

    2009-07-01

    In Many situations of nuclear system study, it is necessary to know the detailed particle flux in a geometry. Deterministic 1-D and 2-D methods aren't suitable to represent some strong 3-D behavior configurations, for example in cores where the neutron flux varies considerably in the space and Monte Carlo analysis are necessary. The majority of Monte Carlo transport calculation codes, performs time static simulations, in terms of fuel isotopic composition. This work is a initial project to incorporate depletion capability to the MCNP code, by means of a connection with ORIGEN2.1 burnup code. The method to develop the program proposed followed the methodology of other programs used to the same purpose. Essentially, MCNP data library are used to generate one group microscopic cross sections that override default ORIGEN libraries. To verify the actual implemented part, comparisons which MCNPX (version 2.6.0) results were made. The neutron flux and criticality value of core agree. The neutron flux and criticality value of the core agree, especially in beginning of burnup when the influence of fission products are not very considerable. The small difference encountered was probably caused by the difference in the number of isotopes considered in the transport models (89 MCNPX x 25 GB). Next step of this work is to adapt MCNP version 4C to work with a memory higher than its standard value (4MB), in order to allow a greater number of isotopes in the transport model. (author)

  7. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs, Draft Environmental Impact Statement. Volume 1, Appendix D: Part A, Naval Spent Nuclear Fuel Management

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    Volume 1 to the Department of Energy`s Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Management Programs Environmental Impact Statement evaluates a range of alternatives for managing naval spent nuclear fuel expected to be removed from US Navy nuclear-powered vessels and prototype reactors through the year 2035. The Environmental Impact Statement (EIS) considers a range of alternatives for examining and storing naval spent nuclear fuel, including alternatives that terminate examination and involve storage close to the refueling or defueling site. The EIS covers the potential environmental impacts of each alternative, as well as cost impacts and impacts to the Naval Nuclear Propulsion Program mission. This Appendix covers aspects of the alternatives that involve managing naval spent nuclear fuel at four naval shipyards and the Naval Nuclear Propulsion Program Kesselring Site in West Milton, New York. This Appendix also covers the impacts of alternatives that involve examining naval spent nuclear fuel at the Expended Core Facility in Idaho and the potential impacts of constructing and operating an inspection facility at any of the Department of Energy (DOE) facilities considered in the EIS. This Appendix also considers the impacts of the alternative involving limited spent nuclear fuel examinations at Puget Sound Naval Shipyard. This Appendix does not address the impacts associated with storing naval spent nuclear fuel after it has been inspected and transferred to DOE facilities. These impacts are addressed in separate appendices for each DOE site.

  8. Site Earthquake Characteristics and Dynamic Parameter Test of Phase Ⅲ Qinshan Nuclear Power Engineering

    Institute of Scientific and Technical Information of China (English)

    ZHOV Nian-qing; ZHAO Zai-li; QIN Min

    2009-01-01

    The earthquake characteristics and geological structure of the site to sitting the Qinshan Nuclear Power Station are closely related. According to site investigation drilling, sampling, seismic sound logging wave test in single-hole and cross-hole, laboratory wave velocity test of intact rock, together with analysis of the site geological conditions, the seismic wave test results of the site between strata lithology and the geologic structure were studied. The relationships of seismic waves with the site lithology and the geologic structure were set up.The dynamic parameters of different grades of weathering profile were deduced. The results assist the seismic design of Phase Ⅲ Qinshan Nuclear Power Plant, China.

  9. TRUMP-BD: A computer code for the analysis of nuclear fuel assemblies under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Lombardo, N.J.; Marseille, T.J.; White, M.D.; Lowery, P.S.

    1990-06-01

    TRUMP-BD (Boil Down) is an extension of the TRUMP (Edwards 1972) computer program for the analysis of nuclear fuel assemblies under severe accident conditions. This extension allows prediction of the heat transfer rates, metal-water oxidation rates, fission product release rates, steam generation and consumption rates, and temperature distributions for nuclear fuel assemblies under core uncovery conditions. The heat transfer processes include conduction in solid structures, convection across fluid-solid boundaries, and radiation between interacting surfaces. Metal-water reaction kinetics are modeled with empirical relationships to predict the oxidation rates of steam-exposed Zircaloy and uranium metal. The metal-water oxidation models are parabolic in form with an Arrhenius temperature dependence. Uranium oxidation begins when fuel cladding failure occurs; Zircaloy oxidation occurs continuously at temperatures above 13000{degree}F when metal and steam are available. From the metal-water reactions, the hydrogen generation rate, total hydrogen release, and temporal and spatial distribution of oxide formations are computed. Consumption of steam from the oxidation reactions and the effect of hydrogen on the coolant properties is modeled for independent coolant flow channels. Fission product release from exposed uranium metal Zircaloy-clad fuel is modeled using empirical time and temperature relationships that consider the release to be subject to oxidation and volitization/diffusion ( bake-out'') release mechanisms. Release of the volatile species of iodine (I), tellurium (Te), cesium (Ce), ruthenium (Ru), strontium (Sr), zirconium (Zr), cerium (Cr), and barium (Ba) from uranium metal fuel may be modeled.

  10. Nuclear Engineering Computer Modules: Reactor Dynamics, RD-1 and RD-2.

    Science.gov (United States)

    Onega, Ronald J.

    The objective of the Reactor Dynamics Module, RD-1, is to obtain the kinetics equation without feedback and solve the kinetics equations numerically for one to six delayed neutron groups for time varying reactivity insertions. The computer code FUMOKI (Fundamental Mode Kinetics) will calculate the power as a function of time for either uranium or…

  11. Educational Programs and Facilities in Nuclear Science and Engineering. Fifth Edition.

    Science.gov (United States)

    Oak Ridge Associated Universities, TN.

    This publication contains detailed descriptions of nuclear programs and facilities of 182 four-year educational institutions. Instead of chapters, the contents are presented in five tables. Table I presents the degrees, graduate appointments, special facilities and programs of the institutions. The institutions are arranged in alphabetical order…

  12. Two Phase Flow Models and Numerical Methods of the Commercial CFD Codes

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Sung Won; Jeong, Jae Jun; Chang, Seok Kyu; Cho, Hyung Kyu

    2007-11-15

    The use of commercial CFD codes extend to various field of engineering. The thermal hydraulic analysis is one of the promising engineering field of application of the CFD codes. Up to now, the main application of the commercial CFD code is focused within the single phase, single composition fluid dynamics. Nuclear thermal hydraulics, however, deals with abrupt pressure changes, high heat fluxes, and phase change heat transfer. In order to overcome the CFD limitation and to extend the capability of the nuclear thermal hydraulics analysis, the research efforts are made to collaborate the CFD and nuclear thermal hydraulics. To achieve the final goal, the current useful model and correlations used in commercial CFD codes should be reviewed and investigated. This report gives the summary information about the constitutive relationships that are used in the FLUENT, STAR-CD, and CFX. The brief information of the solution technologies are also enveloped.

  13. Measurement of Sedimentary Interbed Hydraulic Properties and Their Hydrologic Influence near the Idaho Nuclear Technology and Engineering Center at the Idaho National Engineering and Environmental Laboratory

    Science.gov (United States)

    Perkins, Kim S.

    2003-01-01

    Disposal of wastewater to unlined infiltration ponds near the Idaho Nuclear Technology and Engineering Center (INTEC), formerly known as the Idaho Chemical Processing Plant, at the Idaho National Engineering and Environmental Laboratory (INEEL) has resulted in the formation of perched water bodies in the unsaturated zone (Cecil and others, 1991). The unsaturated zone at INEEL comprises numerous basalt flows interbedded with thinner layers of coarse- to fine-grained sediments and perched ground-water zones exist at various depths associated with massive basalts, basalt-flow contacts, sedimentary interbeds, and sediment-basalt contacts. Perched ground water is believed to result from large infiltration events such as seasonal flow in the Big Lost River and wastewater discharge to infiltration ponds. Evidence from a large-scale tracer experiment conducted in 1999 near the Radioactive Waste Management Complex (RWMC), approximately 13 km from the INTEC, indicates that rapid lateral flow of perched water in the unsaturated zone may be an important factor in contaminant transport at the INEEL (Nimmo and others, 2002b). Because sedimentary interbeds, and possibly baked-zone alterations at sediment-basalt contacts (Cecil and other, 1991) play an important role in the generation of perched water it is important to assess the hydraulic properties of these units.

  14. Nuclear Heating Measurement in Critical Facilities and Experimental Validation of Code and Libraries - An Application to Prompt and Delayed γ Nuclear Data Needs

    Science.gov (United States)

    Blaise, P.; Di Salvo, J.; Vaglio-Gaudard, C.; Bernard, D.; Amharrak, H.; Lemaire, M.; Ravaux, S.

    Energy from prompt and delayed gammas in actual and future nuclear systems are more and more taken into account into design studies as they play an important role in the assessment of performance and safety concerns. Their incomplete knowledge (both prompt and delayed) require to take conservative design margins on local dimensioning parameters, thus reducing the awaited performances or flexibility of these facilities, with costs that are far from being negligible. The local energy photon deposit must be accurately known for Generation-III (Gen-III), Generation-IV (Gen-IV) or the new MTR Jules Horowitz Reactor (JHR). The last 2 decades has seen the realization, in Zero Power Reactors (ZPR), of several programs partially devoted to γ-heating measurements. Experimental programs were and are still conducted in different Cadarache facilities such as MASURCA (for SFR), and later in MINERVE and EOLE (for JHR and Gen-III reactors). The adequacy of the γ-heating calculation was compared to experimental data using thermo-luminescent (TL) detectors and γ-fission chambers. Inconsistencies in C/E and associated uncertainties led to improvement of both libraries and experimental techniques. For these last one, characterization for TL and optically stimulated (OSL) detectors (calibration, individual response), and Monte Carlo calculation of charge repartition in those detectors and their environment were carefully checked and optimized. This step enabled to reduce the associated experimental uncertainty by a factor of 2 (8% at 2σ). Nevertheless, interpretation of integral experiment with updated calculation schemes and improved experimental techniques still tend to prove that there are some nuclei for which there are missing or erroneous data, mainly in structural and absorbing materials. New integral and differential measurements are needed to guide new evaluation efforts, which could benefit from consolidated theoretical and experimental modeling techniques.

  15. Acquired experience on organizing 3D S.UN.COP: international course to support nuclear license by user training in the areas of scaling, uncertainty, and 3D thermal-hydraulics/neutron-kinetics coupled codes

    Energy Technology Data Exchange (ETDEWEB)

    Petruzzi, Alessandro; D' Auria, Francesco [University of Pisa, San Piero a Grado (Italy). Nuclear Research Group San Piero a Grado (GRNSPG); Galetti, Regina, E-mail: regina@cnen.gov.b [National Commission for Nuclear Energy (CNEN), Rio de Janeiro, RJ (Brazil); Bajs, Tomislav [University of Zagreb (Croatia). Fac. of Electrical Engineering and Computing. Dept. of Power Systems; Reventos, Francesc [Technical University of Catalonia, Barcelona (Spain). Dept. of Physics and Nuclear Engineering

    2011-07-01

    Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers, vendors, and research organizations. Computer code user represents a source of uncertainty that may significantly affect the results of system code calculations. Code user training and qualification represent an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes the experience in applying a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In addition, this paper presents the organization and the main features of the 3D S.UN.COP (scaling, uncertainty, and 3D coupled code calculations) seminars during which particular emphasis is given to practical applications in connection with the licensing process of best estimate plus uncertainty methodologies, showing the designer, utility and regulatory approaches. (author)

  16. Conformity of nuclear construction codes with the requirements of the French order dated December 12, 2005 related to nuclear pressure equipment; Conformite des codes de construction nucleaires avec les exigences de l'arrete du 12 decembre 2005 relatif aux equipements sous pression nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Grandemange, J.M.; Renaut, P. [Areva-NP, Tour AREVA, 92084 - Paris La Defense cedex, (France); Paris, D. [EDF-Ceidre 2 rue Ampere - 93206 SAINT-DENIS Cedex (France); Faidy, C. [EDF-Septen 12/14, Avenue Dutrievoz 69628 Villeurbanne Cedex (France)

    2007-07-01

    The French Decree dated December 13, 1999 transposing the Pressure Equipment Directive (PED) has replaced the fundamental texts on which up to now the regulation for pressure equipment important for the safety of nuclear reactors was also founded. By a Ministerial Order - called 'ESPN Order' - dated December 12, 2005, a new regulation has been issued for nuclear pressure equipment. This text makes reference to the Decree transposing the PED while completing these provisions by supplementary requirements having the objective to provide a very high level of integrity guarantee for equipments which are the most important for safety, and to cover the prevention of radioactive release risks. These regulatory evolutions are presented in the Plenary Session of the ESOPE conference. Referencing the Decree and thus the PED, and including specific provisions, the Ministerial Order implies that the Manufacturers update their documents and, if necessary, their prescriptions in the following two domains: - that of the conformity of Codes and Standards used, generally inspired from the ASME Code Section III, with the essential safety requirements of the PED, - that of the respect of the complementary provisions brought by the ESPN Order. This paper presents the more significant conclusions of this work and the resulting amendments of the RCC-M Code, introduced by the 2007 addendum to that Code. The analysis will lead to specify the same type of complementary requirements to Code when a manufacturer wishes to use the German KTA Rules or the ASME Code Section III. (authors) [French] Le decret du 13 decembre 1999 transposant la directive europeenne (DESP) relative aux equipements sous pression a remplace les textes fondamentaux sur lesquels se fondait egalement jusque la la reglementation des appareils a pression importants pour la surete des reacteurs nucleaires. Par arrete - dit 'arrete ESPN' - du 12 decembre 2005, une nouvelle reglementation a ete dictee. Ce

  17. Two-dimensional simulation of hydrogen iodide decomposition reaction using fluent code for hydrogen production using nuclear technology

    Directory of Open Access Journals (Sweden)

    Jung-Sik Choi

    2015-06-01

    Full Text Available The operating characteristics of hydrogen iodide (HI decomposition for hydrogen production were investigated using the commercial computational fluid dynamics code, and various factors, such as hydrogen production, heat of reaction, and temperature distribution, were studied to compare device performance with that expected for device development. Hydrogen production increased with an increase of the surface-to-volume (STV ratio. With an increase of hydrogen production, the reaction heat increased. The internal pressure and velocity of the HI decomposer were estimated through pressure drop and reducing velocity from the preheating zone. The mass of H2O was independent of the STV ratio, whereas that of HI decreased with increasing STV ratio.

  18. Multiple-code benchmark simulation study of coupled THMC processesin the excavation disturbed zone associated with geological nuclear wasterepositories

    Energy Technology Data Exchange (ETDEWEB)

    Rutqvist, J.; Feng, X-T.; Hudson, J.; Jing, L.; Kobayashi, A.; Koyama, T.; Pan, P-Z.; Lee, H-S.; Rinne, M.; Sonnenthal, E.; Yamamoto, Y.

    2006-05-10

    An international, multiple-code benchmark test (BMT) studyis being conducted within the international DECOVALEX project to analysecoupled thermal, hydrological, mechanical and chemical (THMC) processesin the excavation disturbed zone (EDZ) around emplacement drifts of anuclear waste repository. This BMT focuses on mechanical responses andlong-term chemo-mechanical effects that may lead to changes in mechanicaland hydrological properties in the EDZ. This includes time-de-pendentprocesses such as creep, and subcritical crack, or healing of fracturesthat might cause "weakening" or "hardening" of the rock over the longterm. Five research teams are studying this BMT using a wide range ofmodel approaches, including boundary element, finite element, and finitedifference, particle mechanics, and elasto-plastic cellular automatamethods. This paper describes the definition of the problem andpreliminary simulation results for the initial model inception part, inwhich time dependent effects are not yet included.

  19. Experimental differential cross sections, level densities, and spin cutoffs as a testing ground for nuclear reaction codes

    Science.gov (United States)

    Voinov, A. V.; Grimes, S. M.; Brune, C. R.; Bürger, A.; Görgen, A.; Guttormsen, M.; Larsen, A. C.; Massey, T. N.; Siem, S.

    2013-11-01

    Proton double-differential cross sections from 59Co(α,p)62Ni, 57Fe(α,p)60Co, 56Fe(7Li,p)62Ni, and 55Mn(6Li,p)60Co reactions have been measured with 21-MeV α and 15-MeV lithium beams. Cross sections have been compared against calculations with the empire reaction code. Different input level density models have been tested. It was found that the Gilbert and Cameron [A. Gilbert and A. G. W. Cameron, Can. J. Phys.0008-420410.1139/p65-139 43, 1446 (1965)] level density model is best to reproduce experimental data. Level densities and spin cutoff parameters for 62Ni and 60Co above the excitation energy range of discrete levels (in continuum) have been obtained with a Monte Carlo technique. Excitation energy dependencies were found to be inconsistent with the Fermi-gas model.

  20. Determination of the physical parameters of the nuclear subcritical assembly Chicago 9000 of the IPN using the Serpent code; Determinacion de los parametros fisicos del conjunto subcritico nuclear Chicago 9000 del IPN usando el codigo SERPENT

    Energy Technology Data Exchange (ETDEWEB)

    Arriaga R, L.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, U.P. Adolfo Lopez Mateos, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Gomez T, A. M., E-mail: guten_tag_04@hotmail.com [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    For the Serpent code was developed the three-dimensional model corresponding to the nuclear subcritical assembly (S A) Chicago 9000 of the Escuela Superior de Fisica y Matematicas del Instituto Politecnico Nacional (ESFM-IPN). The model includes: a) the core, formed by 312 aluminum pipes that contain 5 nuclear fuel rods (natural uranium in metallic form), b) the multi-perforated plates where they penetrate the inferior part of each pipe to be able to remain in vertical form, c) water, acting as moderator and reflector, and d) the recipient lodging to the core. The pipes arrangement is hexagonal although the transversal section of the recipient that lodges to the core is circular. The entrance file for the Serpent code was generated with the data provided by the manual of the S A use about the composition and density of the fuel rods and others obtained in direct form of the rods, as the interior and external diameter, mass and height. Of the obtained physical parameters, those more approached to that reported in the manual of the subcritical assembly are the effective multiplication factor and the reproduction factor η. The differences can be because the description of the fuel rods provided by the manual of the S A use do not correspond those that are physically in the S A core. This difference consists on the presence of a circular central channel of 1.245 diameter centimeters in each fuel rod. The fuel rods reported in the mentioned manual do not have that channel. Although the obtained results are encouraging, we want to continue improving the model to incorporate in this the detectors, defined this way by the Serpent code, which could determine the existent neutrons flux in diverse points of interest like the axial or radial aligned points and to compare these with those that are obtained in an experimental way when a generating neutrons source (Pu-Be) is introduced. Added to this effort the cross sections for each unitary cell will be determined, so that

  1. Nuclear protein-coding genes support lungfish and not the coelacanth as the closest living relatives of land vertebrates.

    Science.gov (United States)

    Brinkmann, Henner; Venkatesh, Byrappa; Brenner, Sydney; Meyer, Axel

    2004-04-01

    The colonization of land by tetrapod ancestors is one of the major questions in the evolution of vertebrates. Despite intense molecular phylogenetic research on this problem during the last 15 years, there is, until now, no statistically supported answer to the question of whether coelacanths or lungfish are the closest living relatives of tetrapods. We determined DNA sequences of the nuclear-encoded recombination activating genes (Rag1 and Rag2) from all three major lungfish groups, the Australian Neoceratodis forsteri, the South American Lepidosiren paradoxa and the African lungfish Protopterus dolloi, and the Indonesian coelacanth Latimeria menadoensis. Phylogenetic analyses of both the single gene and the concatenated data sets of RAG1 and RAG2 found that the lungfishes are the closest living relatives of the land vertebrates. These results are supported by high bootstrap values, Bayesian posterior probabilities, and likelihood ratio tests.

  2. Multiple-code simulation study of the long-term EDZ evolution of geological nuclear waste repositories

    Energy Technology Data Exchange (ETDEWEB)

    Rutqvist, J.; Backstrom, A.; Chijimatsu, M.; Feng, X.-T.; Pan, P.-Z.; Hudson, J.; Jing, L.; Kobayashi, A.; Koyama, T.; Lee, H.-S.; Huang, X.-H.; Rinne, M.; Shen, B.

    2008-10-23

    This simulation study shows how widely different model approaches can be adapted to model the evolution of the excavation disturbed zone (EDZ) around a heated nuclear waste emplacement drift in fractured rock. The study includes modeling of coupled thermal-hydrological-mechanical (THM) processes, with simplified consideration of chemical coupling in terms of time-dependent strength degradation or subcritical crack growth. The different model approaches applied in this study include boundary element, finite element, finite difference, particle mechanics, and elastoplastic cellular automata methods. The simulation results indicate that thermally induced differential stresses near the top of the emplacement drift may cause progressive failure and permeability changes during the first 100 years (i.e., after emplacement and drift closure). Moreover, the results indicate that time-dependent mechanical changes may play only a small role during the first 100 years of increasing temperature and thermal stress, whereas such time-dependency is insignificant after peak temperature, because decreasing thermal stress.

  3. Investigation on effect of equivalence ratio and engine speed on homogeneous charge compression ignition combustion using chemistry based CFD code

    Directory of Open Access Journals (Sweden)

    Ghafouri Jafar

    2014-01-01

    Full Text Available Combustion in a large-bore natural gas fuelled diesel engine operating under Homogeneous Charge Compression Ignition mode at various operating conditions is investigated in the present paper. Computational Fluid Dynamics model with integrated chemistry solver is utilized and methane is used as surrogate of natural gas fuel. Detailed chemical kinetics mechanism is used for simulation of methane combustion. The model results are validated using experimental data by Aceves, et al. (2000, conducted on the single cylinder Volvo TD100 engine operating at Homogeneous Charge Compression Ignition conditions. After verification of model predictions using in-cylinder pressure histories, the effect of varying equivalence ratio and engine speed on combustion parameters of the engine is studied. Results indicate that increasing engine speed provides shorter time for combustion at the same equivalence ratio such that at higher engine speeds, with constant equivalence ratio, combustion misfires. At lower engine speed, ignition delay is shortened and combustion advances. It was observed that increasing the equivalence ratio retards the combustion due to compressive heating effect in one of the test cases at lower initial pressure. Peak pressure magnitude is increased at higher equivalence ratios due to higher energy input.

  4. 基于RMC程序的keff对核数据的敏感性分析%keff Sensitivity Analysis to Nuclear Data with RMC Code

    Institute of Scientific and Technical Information of China (English)

    丘意书; 余健开; 梁金刚; 王侃

    2015-01-01

    Methods suitable for sensitivity analysis in continuous‐energy Monte Carlo codes become a research hotspot in the field of reactor physics .In this work ,the formu‐las of sensitivity coefficients of five different reaction types were established .Then ,the theoretical basis and the algorithm of the iterated fission probability method which was used widely currently were discussed .Furthermore ,two Monte Carlo codes ,RMC and MCNP6 ,were used to compute eigenvalue sensitivity coefficients to nuclear data .The agreement between RMC and MCNP6 is well .The results indicate that RMC is capable to perform sensitivity analysis preliminarily .%适用于连续能量蒙特卡罗程序的敏感性分析方法是当前的研究热点。本文建立了5种不同反应类型的敏感性系数的计算公式,对当前应用广泛的反复裂变几率法的理论基础及算法进行了分析。分别使用RMC程序和MCNP6程序计算了 kef对核数据的敏感性系数,计算结果吻合良好。本文结果表明RM C程序初步具备了敏感性分析的功能。

  5. WSPEEDI (worldwide version of SPEEDI): A computer code system for the prediction of radiological impacts on Japanese due to a nuclear accident in foreign countries

    Energy Technology Data Exchange (ETDEWEB)

    Chino, Masamichi; Yamazawa, Hiromi; Nagai, Haruyasu; Moriuchi, Shigeru [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ishikawa, Hirohiko

    1995-09-01

    A computer code system has been developed for near real-time dose assessment during radiological emergencies. The system WSPEEDI, the worldwide version of SPEEDI (System for Prediction of Environmental Emergency Dose Information) aims at predicting the radiological impact on Japanese due to a nuclear accident in foreign countries. WSPEEDI consists of a mass-consistent wind model WSYNOP for large-scale wind fields and a particle random walk model GEARN for atmospheric dispersion and dry and wet deposition of radioactivity. The models are integrated into a computer code system together with a system control software, worldwide geographic database, meteorological data processor and graphic software. The performance of the models has been evaluated using the Chernobyl case with reliable source terms, well-established meteorological data and a comprehensive monitoring database. Furthermore, the response of the system has been examined by near real-time simulations of the European Tracer Experiment (ETEX), carried out over about 2,000 km area in Europe. (author).

  6. Environmental tasks of engineering geology in the construction of nuclear power installations

    Energy Technology Data Exchange (ETDEWEB)

    Hrasna, M. (Komenskeho Univ., Bratislava (Czechoslovakia). Prirodovedecka Fakulta)

    1984-01-01

    A nuclear power installation shoUld be sited in an area without large and intensive geodynamic phenomena such as seismicity, gravitation movement, subsidence, etc., an area with mineral and drinking water resources, deposits of mineral raw materials, quality farmland and protected areas. With regard to demands on the quality of foundation soil and the ground water level it will always be necessary to secure the safe foundation of buildings with possible radioactivity escape. Most such buildings are founded on a solid foundation slab. The definitive siting of the nuclear power facility should be decided upon during the project design preparation stage when most geological surveying is done. During the pre-project stage a quantitative determination is made of all characteristics of rocks on the site. Monitoring and surveillance is conducted during construction.

  7. [The characteristics of morbidity of workers of nuclear power engineering enterprise].

    Science.gov (United States)

    Pischugina, A V; Ivanov, A G; Belyakova, N A

    2013-01-01

    The article considers the morbidity endocrine, pathology included, of workers of nuclear power station and body-abled population of the district employed in other areas of professional activities. The statistically reliable exceeding of the level of primarily diagnosed endocrine morbidity in the group of working population of the district as compared with the group of workers of nuclear power station is established. In the compared groups, the structure of pathology of endocrine system is characterized by the prevalence of diseases of thyroid gland and obesity. The official statistics data reflects the level of morbidity of working population depending on appealability to curative preventive institutions, ratio and scope of the periodic medical examinations, availability of shop therapeutic service and possibility to involve physicians-specialists to health posts enterprises. Therefore, the foundation of enhancement of quality of medical care to workers is the improvemnent of organizational activities at the level of primary health care.

  8. The high-temperature sodium coolant technology in nuclear power installations for hydrogen power engineering

    Science.gov (United States)

    Kozlov, F. A.; Sorokin, A. P.; Alekseev, V. V.; Konovalov, M. A.

    2014-05-01

    In the case of using high-temperature sodium-cooled nuclear power installations for obtaining hydrogen and for other innovative applications (gasification and fluidization of coal, deep petroleum refining, conversion of biomass into liquid fuel, in the chemical industry, metallurgy, food industry, etc.), the sources of hydrogen that enters from the reactor plant tertiary coolant circuit into its secondary coolant circuit have intensity two or three orders of magnitude higher than that of hydrogen sources at a nuclear power plant (NPP) equipped with a BN-600 reactor. Fundamentally new process solutions are proposed for such conditions. The main prerequisite for implementing them is that the hydrogen concentration in sodium coolant is a factor of 100-1000 higher than it is in modern NPPs taken in combination with removal of hydrogen from sodium by subjecting it to vacuum through membranes made of vanadium or niobium. Numerical investigations carried out using a diffusion model showed that, by varying such parameters as fuel rod cladding material, its thickness, and time of operation in developing the fuel rods for high-temperature nuclear power installations (HT NPIs) it is possible to exclude ingress of cesium into sodium through the sealed fuel rod cladding. However, if the fuel rod cladding loses its tightness, operation of the HT NPI with cesium in the sodium will be unavoidable. Under such conditions, measures must be taken for deeply purifying sodium from cesium in order to minimize the diffusion of cesium into the structural materials.

  9. Computer code for single-point thermodynamic analysis of hydrogen/oxygen expander-cycle rocket engines

    Science.gov (United States)

    Glassman, Arthur J.; Jones, Scott M.

    1991-01-01

    This analysis and this computer code apply to full, split, and dual expander cycles. Heat regeneration from the turbine exhaust to the pump exhaust is allowed. The combustion process is modeled as one of chemical equilibrium in an infinite-area or a finite-area combustor. Gas composition in the nozzle may be either equilibrium or frozen during expansion. This report, which serves as a users guide for the computer code, describes the system, the analysis methodology, and the program input and output. Sample calculations are included to show effects of key variables such as nozzle area ratio and oxidizer-to-fuel mass ratio.

  10. Differential effects of high-temperature stress on nuclear topology and transcription of repetitive noncoding and coding rye sequences.

    Science.gov (United States)

    Tomás, D; Brazão, J; Viegas, W; Silva, M

    2013-01-01

    The plant stress response has been extensively characterized at the biochemical and physiological levels. However, knowledge concerning repetitive sequence genome fraction modulation during extreme temperature conditions is scarce. We studied high-temperature effects on subtelomeric repetitive sequences (pSc200) and 45S rDNA in rye seedlings submitted to 40°C during 4 h. Chromatin organization patterns were evaluated through fluorescent in situ hybridization and transcription levels were assessed using quantitative real-time PCR. Additionally, the nucleolar dynamics were evaluated through fibrillarin immunodetection in interphase nuclei. The results obtained clearly demonstrated that the pSc200 sequence organization is not affected by high-temperature stress (HTS) and proved for the first time that this noncoding subtelomeric sequence is stably transcribed. Conversely, it was demonstrated that HTS treatment induces marked rDNA chromatin decondensation along with nucleolar enlargement and a significant increase in ribosomal gene transcription. The role of noncoding and coding repetitive rye sequences in the plant stress response that are suggested by their clearly distinct behaviors is discussed. While the heterochromatic conformation of pSc200 sequences seems to be involved in the stabilization of the interphase chromatin architecture under stress conditions, the dynamic modulation of nucleolar and rDNA topology and transcription suggest their role in plant stress response pathways.

  11. Detecting selection in the blue crab, Callinectes sapidus, using DNA sequence data from multiple nuclear protein-coding genes.

    Science.gov (United States)

    Yednock, Bree K; Neigel, Joseph E

    2014-01-01

    The identification of genes involved in the adaptive evolution of non-model organisms with uncharacterized genomes constitutes a major challenge. This study employed a rigorous and targeted candidate gene approach to test for positive selection on protein-coding genes of the blue crab, Callinectes sapidus. Four genes with putative roles in physiological adaptation to environmental stress were chosen as candidates. A fifth gene not expected to play a role in environmental adaptation was used as a control. Large samples (n>800) of DNA sequences from C. sapidus were used in tests of selective neutrality based on sequence polymorphisms. In combination with these, sequences from the congener C. similis were used in neutrality tests based on interspecific divergence. In multiple tests, significant departures from neutral expectations and indicative of positive selection were found for the candidate gene trehalose 6-phosphate synthase (tps). These departures could not be explained by any of the historical population expansion or bottleneck scenarios that were evaluated in coalescent simulations. Evidence was also found for balancing selection at ATP-synthase subunit 9 (atps) using a maximum likelihood version of the Hudson, Kreitmen, and Aguadé test, and positive selection favoring amino acid replacements within ATP/ADP translocase (ant) was detected using the McDonald-Kreitman test. In contrast, test statistics for the control gene, ribosomal protein L12 (rpl), which presumably has experienced the same demographic effects as the candidate loci, were not significantly different from neutral expectations and could readily be explained by demographic effects. Together, these findings demonstrate the utility of the candidate gene approach for investigating adaptation at the molecular level in a marine invertebrate for which extensive genomic resources are not available.

  12. Detecting selection in the blue crab, Callinectes sapidus, using DNA sequence data from multiple nuclear protein-coding genes.

    Directory of Open Access Journals (Sweden)

    Bree K Yednock

    Full Text Available The identification of genes involved in the adaptive evolution of non-model organisms with uncharacterized genomes constitutes a major challenge. This study employed a rigorous and targeted candidate gene approach to test for positive selection on protein-coding genes of the blue crab, Callinectes sapidus. Four genes with putative roles in physiological adaptation to environmental stress were chosen as candidates. A fifth gene not expected to play a role in environmental adaptation was used as a control. Large samples (n>800 of DNA sequences from C. sapidus were used in tests of selective neutrality based on sequence polymorphisms. In combination with these, sequences from the congener C. similis were used in neutrality tests based on interspecific divergence. In multiple tests, significant departures from neutral expectations and indicative of positive selection were found for the candidate gene trehalose 6-phosphate synthase (tps. These departures could not be explained by any of the historical population expansion or bottleneck scenarios that were evaluated in coalescent simulations. Evidence was also found for balancing selection at ATP-synthase subunit 9 (atps using a maximum likelihood version of the Hudson, Kreitmen, and Aguadé test, and positive selection favoring amino acid replacements within ATP/ADP translocase (ant was detected using the McDonald-Kreitman test. In contrast, test statistics for the control gene, ribosomal protein L12 (rpl, which presumably has experienced the same demographic effects as the candidate loci, were not significantly different from neutral expectations and could readily be explained by demographic effects. Together, these findings demonstrate the utility of the candidate gene approach for investigating adaptation at the molecular level in a marine invertebrate for which extensive genomic resources are not available.

  13. Theoretical determination of the strength characteristics of multilayer materials intended for nuclear and thermonuclear engineering

    Science.gov (United States)

    Vitkovskii, I. V.; Leshukov, A. Yu.; Romashin, S. N.; Shorkin, V. S.

    2015-12-01

    A method is developed to estimate the integrity of multilayer structures. This method is based on the version of the theory of adhesion and cohesion interactions of structure elements that only takes into account their thermomechanical properties. The structures to be studied are the material of the multilayer wall of the liquid-metal thermonuclear reactor blanket and a heat-resistant magnet wire with a bimetallic conductor, which is the base of the windings of the magnetohydrodynamic machines and electric motors intended for operation at high temperatures under ionizing radiation in, e.g., the machines and facilities in nuclear and thermonuclear reactors.

  14. Compulsory Checking of Nuclear Power Engineering Materials by Direct and Eddy Current

    Science.gov (United States)

    Larionov, V. V.; Lider, A. M.; Sednev, D. A.; Xu, Shupeng

    2016-08-01

    The testing technology of copper parts designed for dry storage of spent nuclear fuel with application of direct and eddy current has been developed. Measurements results of flaw quantity caused hydrogenation and oxidation processes are presented. Evolution of copper M 001 flaw structure during hydrogenation from gaseous medium is analyzed. It has been demonstrated that the dependence of copper p electrical resistance on number of flaws in its structure has dome shaped character and changes with eddy current frequency change. Number of flaws formed by hydrogen depends on direction (100) or (200) of the crystal structure of copper lattice.

  15. Why Model-Based Engineering and Manufacturing Makes Sense for the Plants and Laboratories of the Nuclear Weapon Complex

    Energy Technology Data Exchange (ETDEWEB)

    Franklin, K W; Howell, L N; Lewis, D G; Neugebauer, C A; O' Brien, D W; Schilling, S A

    2001-05-15

    The purpose of this White Paper is to outline the benefits we expect to receive from Model-Based Engineering and Manufacturing (MBE/M) for the design, analysis, fabrication, and assembly of nuclear weapons for upcoming Life Extension Programs (LEPs). Industry experiences with model-based approaches and the NNSA/DP investments and experiences, discussed in this paper, indicate that model-based methods can achieve reliable refurbished weapons for the stockpile with less cost and time. In this the paper, we list both general and specific benefits of MBE/M for the upcoming LEPs and the metrics for determining the success of model-based approaches. We also present some outstanding issues and challenges to deploying and achieving long-term benefit from the MBE/M. In conclusion, we argue that successful completion of the upcoming LEPs--with very aggressive schedule and funding restrictions--will depend on electronic model-based methods. We ask for a strong commitment from LEP managers throughout the Nuclear Weapons Complex to support deployment and use of MBE/M systems to meet their program needs.

  16. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Draft Environmental Impact Statement. Volume 1, Appendix C, Savannah River Site Spent Nuclear Fuel Mangement Program

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    The US Department of Energy (DOE) is engaged in two related decision making processes concerning: (1) the transportation, receipt, processing, and storage of spent nuclear fuel (SNF) at the DOE Idaho National Engineering Laboratory (INEL) which will focus on the next 10 years; and (2) programmatic decisions on future spent nuclear fuel management which will emphasize the next 40 years. DOE is analyzing the environmental consequences of these spent nuclear fuel management actions in this two-volume Environmental Impact Statement (EIS). Volume 1 supports broad programmatic decisions that will have applicability across the DOE complex and describes in detail the purpose and need for this DOE action. Volume 2 is specific to actions at the INEL. This document, which limits its discussion to the Savannah River Site (SRS) spent nuclear fuel management program, supports Volume 1 of the EIS. Following the introduction, Chapter 2 contains background information related to the SRS and the framework of environmental regulations pertinent to spent nuclear fuel management. Chapter 3 identifies spent nuclear fuel management alternatives that DOE could implement at the SRS, and summarizes their potential environmental consequences. Chapter 4 describes the existing environmental resources of the SRS that spent nuclear fuel activities could affect. Chapter 5 analyzes in detail the environmental consequences of each spent nuclear fuel management alternative and describes cumulative impacts. The chapter also contains information on unavoidable adverse impacts, commitment of resources, short-term use of the environment and mitigation measures.

  17. First steps towards a validation of the new burnup and depletion code TNT

    Energy Technology Data Exchange (ETDEWEB)

    Herber, S.C.; Allelein, H.J. [RWTH Aachen (Germany). Inst. for Reactor Safety and Reactor Technology; Research Center Juelich (Germany). Inst. for Energy and Climate Research - Nuclear Waste Disposal and Reactor Safety (IEK-6); Friege, N. [RWTH Aachen (Germany). Inst. for Reactor Safety and Reactor Technology; Kasselmann, S. [Research Center Juelich (Germany). Inst. for Energy and Climate Research - Nuclear Waste Disposal and Reactor Safety (IEK-6)

    2012-11-01

    In the frame of the fusion of the core design calculation capabilities, represented by V.S.O.P., and the accident calculation capabilities, represented by MGT(-3D), the successor of the TINTE code, difficulties were observed in defining an interface between a program backbone and the ORIGEN code respectively the ORIGENJUEL code. The estimation of the effort of refactoring the ORIGEN code or to write a new burnup code from scratch, led to the decision that it would be more efficient writing a new code, which could benefit from existing programming and software engineering tools from the computer code side and which can use the latest knowledge of nuclear reactions, e.g. consider all documented reaction channels. Therefore a new code with an object-oriented approach was developed at IEK-6. Object-oriented programming is currently state of the art and provides mostly an improved extensibility and maintainability. The new code was named TNT which stands for Topological Nuclide Transformation, since the code makes use of the real topology of the nuclear reactions. Here we want to present some first validation results from code to code benchmarks with the codes ORIGEN V2.2 and FISPACT2005 and whenever possible analytical results also used for the comparison. The 2 reference codes were chosen due to their high reputation in the field of fission reactor analysis (ORIGEN) and fusion facilities (FISPACT). (orig.)

  18. Information model for management and preservation of scientific digital memory of the Institute of Nuclear Engineering, Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Sales, Luana Farias, E-mail: lsales@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Sayao, Luis Fernando, E-mail: isayao@cnen.gov.br [Centro de Informacoes Nucleares (CIN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    In the context of the data-oriented science (eScience), a considerable part of the results of research activities has been created in digital formats. This means that the memory of the scientific institutions involved in this new scientific paradigm may be at risk of being lost by rapid technological obsolescence, the known fragility of digital media and also by the fragmentation of information and knowledge scattered across multiples repositories. Thus, management of research data in a digital networked and distributed environment becomes an increasing challenge for the research world and the whole area of information: information science, librarianship, knowledge management, archival science and information technology; moreover, in the dynamic environment featuring eScience, there is a need for novel concepts of documents establishing a linkage between traditional documents - printed or digital - stored in repositories, with the data sets stored in data repositories. In this new research environment, an important issue is how to preserve these new complex documents so that they maintain their structure, meaning and authenticity and also its ability to be retrieved, accessed and reused through time and space. In this sense, this paper proposes an information model focused on the curation of scientific memory of the Institute of Nuclear Engineering of the Brazilian Commission of Nuclear Energy (CNEN/IEN). The model considers the traditional scientific documents (theses, articles, books, etc.) in digital formats and all other relevant data and information related to them, such as: scientific data, software, simulations, photos, videos, historical facts, news, etc., compounding an enhanced publication type oriented to the nuclear area. (author)

  19. Implementation of the Immersive Virtual Reality Laboratory in Nuclear Engineering Institute

    Energy Technology Data Exchange (ETDEWEB)

    Mol, Antonio Carlos de Abreu; Grecco, Claudio Henrique dos Santos; Carvalho, Paulo Victor R.; Oliveira, Mauro Vitor de; Santos, Isaac J.A. Luquetti; Augusto, Silas Cordeiro; Viana Filho, Alfredo Marques [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil)]. E-mail: mol@ien.gov.br; grecco@ien.gov.br; paulov@ien.gov.br; mvitor@ien.gov.br; luqietti@ien.gov.br; silas@ien.gov.br; marques@ien.gov.br

    2005-07-01

    The Immersive Virtual Reality Laboratory under development in Human System Interface Laboratory constitute a powerful general-purpose facility for experimental and computational work on human perception and perceptually guided action. Virtual reality or virtual environment are computer generated environments with and within people can interact. The advantage of VR is that people can be immersed by the simulated environment, which would sometimes be unavailable due to cost, safety, or perceptual restrictions in the real environment. There are many applications of virtual reality on the nuclear area. Training is one of the most common of them. A significant advantage of a virtual training environment over a real one is it's enormous flexibility. A virtual environment can be used as the basis for training in any number of different scenarios, so that trainees can learn to cope with many different situations, some of which may be impossible to prepare for any other way. Another advantage of using virtual environments for training purposes is that trainees learn by actively performing actions. This has a significant effect on their ability to retain what they learn, and is clearly superior to passive training techniques, such as videos and books, for training where spatial understanding is important. This kind of Laboratory is the first in Brazilian nuclear area. A safe virtual environment can be used to simulate a real environment that is either too dangerous, complex, or expensive to training. Virtual environments can therefore be used to increase safety standards, improve efficiency, and reduce overall training costs. (author)

  20. A preliminary systems-engineering study of an advanced nuclear-electrolytic hydrogen-production facility

    Science.gov (United States)

    Escher, W. J. D.; Donakowski, T. D.; Tison, R. R.

    1975-01-01

    An advanced nuclear-electrolytic hydrogen-production facility concept was synthesized at a conceptual level with the objective of minimizing estimated hydrogen-production costs. The concept is a closely-integrated, fully-dedicated (only hydrogen energy is produced) system whose components and subsystems are predicted on ''1985 technology.'' The principal components are: (1) a high-temperature gas-cooled reactor (HTGR) operating a helium-Brayton/ammonia-Rankine binary cycle with a helium reactor-core exit temperature of 980 C, (2) acyclic d-c generators, (3) high-pressure, high-current-density electrolyzers based on solid-polymer electrolyte technology. Based on an assumed 3,000 MWt HTGR the facility is capable of producing 8.7 million std cu m/day of hydrogen at pipeline conditions, 6,900 kPa. Coproduct oxygen is also available at pipeline conditions at one-half this volume. It has further been shown that the incorporation of advanced technology provides an overall efficiency of about 43 percent, as compared with 25 percent for a contemporary nuclear-electric plant powering close-coupled contemporary industrial electrolyzers.