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. Verification of structural analysis computer codes in nuclear engineering

    International Nuclear Information System (INIS)

    Zebeljan, Dj.; Cizelj, L.

    1990-01-01

    Sources of potential errors, which can take place during use of finite element method based computer programs, are described in the paper. The magnitude of errors was defined as acceptance criteria for those programs. Error sources are described as they are treated by 'National Agency for Finite Element Methods and Standards (NAFEMS)'. Specific verification examples are used from literature of Nuclear Regulatory Commission (NRC). Example of verification is made on PAFEC-FE computer code for seismic response analyses of piping systems by response spectrum method. (author)

  3. Aging plant life management - the requirements defined to date by the KTA nuclear engineering codes

    International Nuclear Information System (INIS)

    Kalinowski, I.

    1996-01-01

    German nuclear engineering codes so far do not enclose a specific aging plant life management programme. However, the existing codes and standards do contain a number of applicable requirements and principles of relevance to objectives and principles of such programmes, as they also cover aging-induced effects on power plants. The major principles relating to preventive safety engineering and quality assurance are laid down in the publications KTA 1401, 1404, 1201, 1202, and KTA 3211. (DG) [de

  4. Kinetic---a system code for analyzing nuclear thermal propulsion rocket engine transients

    International Nuclear Information System (INIS)

    Schmidt, E.; Lazareth, O.; Ludewig, H.

    1993-01-01

    A system code suitable for analyzing Nuclear Thermal Propulsion (NTP) rocket engines is described in this paper. The code consists of a point reactor model and nodes to describe the fluid dynamics and heat transfer mechanism. Feedback from the fuel, coolant, moderator and reflector are allowed for, and the control of the reactor is by motion of controls element (drums or rods). The worth of the control element and feedback coefficients are predetermined. Separate models for the turbo-pump assembly (TPA) and nozzle are also included. The model to be described in this paper is specific for the Particle Bed Reactor (PBR). An illustrative problem is solved. This problem consists of a PBR operating in a blowdown mode

  5. Kinetic—a system code for analyzing nuclear thermal propulsion rocket engine transients

    Science.gov (United States)

    Schmidt, Eldon; Lazareth, Otto; Ludewig, Hans

    1993-01-01

    A system code suitable for analyzing Nuclear Thermal Propulsion (NTP) rocket engines is described in this paper. The code consists of a point reactor model and nodes to describe the fluid dynamics and heat transfer mechanism. Feedback from the fuel, coolant, moderator and reflector are allowed for, and the control of the reactor is by motion of controls element (drums or rods). The worth of the control element and feedback coefficients are predetermined. Separate models for the turbo-pump assembly (TPA) and nozzle are also included. The model to be described in this paper is specific for the Particle Bed Reactor (PBR). An illustrative problem is solved. This problem consists of a PBR operating in a blowdown mode.

  6. KINETIC: A system code for analyzing Nuclear thermal propulsion rocket engine transients

    Science.gov (United States)

    Schmidt, E.; Lazareth, O.; Ludewig, H.

    1993-07-01

    A system code suitable for analyzing Nuclear Thermal Propulsion (NTP) rocket engines is described in this paper. The code consists of a point reactor model and nodes to describe the fluid dynamics and heat transfer mechanism. Feedback from the fuel coolant, moderator and reflector are allowed for, and the control of the reactor is by motion of control elements (drums or rods). The worth of the control clement and feedback coefficients are predetermined. Separate models for the turbo-pump assembly (TPA) and nozzle are also included. The model to be described in this paper is specific for the Particle Bed Reactor (PBR). An illustrative problem is solved. This problem consists of a PBR operating in a blowdown mode.

  7. Uncertainties in calculations of nuclear design code system for the high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Shindo, R.; Yamashita, K.; Murata, I.

    1991-01-01

    The nuclear design code system for the HTTR consists of one dimensional cell burnup computer code, developed in JAERI and the TWOTRAN-2 transport code. In order to satisfy related design criteria, uncertainty of the calculation was investigated by comparing the calculated and experimental results. The experiments were performed with a graphite moderated critical assembly. It was confirmed that discrepancies between calculations and experiments were small enough to be allowed in the nuclear design of HTTR. 8 refs, 6 figs

  8. Accident analysis in the water loop of the nuclear engineering department of IPEN using the RELAP4 code

    International Nuclear Information System (INIS)

    Fernandes Filho, T.L.

    1980-06-01

    A thermal-hydraulic analysis to describe the transient behavior in the water loop of the Nuclear Engineering Department of the Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo, Brazil, was performed. Postulated accidents such as those resulting from (1) loss of coolant, (2) main pump failure and (3) power excursions, were studied. The computer code RELAP4/Mod.3 was employed as the principal tool of analysis. (Author) [pt

  9. A code guidance system for integrated nuclear data evaluation system on the basis of knowledge engineering technology

    International Nuclear Information System (INIS)

    Fukahori, Tokio; Nakagawa, Tsuneo

    1994-01-01

    The integrated nuclear data evaluation system (INDES) is being made in order to support the nuclear data evaluation work. A guidance system in INDES, 'Evaluation Tutor (ET)', is under development in order to support users in selecting the most suitable set of theoretical calculation codes by applying knowledge engineering technology and the experiences of evaluation work for JENDL-3. In this paper, the function of ET is introduced as well as the functions and databases of INDES. An example run of ET for 56 Fe in the 1-20 MeV neutron energy region is also explained. (author)

  10. Nuclear code abstracts (1975 edition)

    International Nuclear Information System (INIS)

    Akanuma, Makoto; Hirakawa, Takashi

    1976-02-01

    Nuclear Code Abstracts is compiled in the Nuclear Code Committee to exchange information of the nuclear code developments among members of the committee. Enlarging the collection, the present one includes nuclear code abstracts obtained in 1975 through liaison officers of the organizations in Japan participating in the Nuclear Energy Agency's Computer Program Library at Ispra, Italy. The classification of nuclear codes and the format of code abstracts are the same as those in the library. (auth.)

  11. Roadmap to an Engineering-Scale Nuclear Fuel Performance and Safety Code

    International Nuclear Information System (INIS)

    Turner, John A.; Clarno, Kevin T.; Hansen, Glen A.

    2009-01-01

    Developing new fuels and qualifying them for large-scale deployment in power reactors is a lengthy and expensive process, typically spanning a period of two decades from concept to licensing. Nuclear fuel designers serve an indispensable role in the process, at the initial exploratory phase as well as in analysis of the testing results. In recent years fuel performance capabilities based on first principles have been playing more of a role in what has traditionally been an empirically dominated process. Nonetheless, nuclear fuel behavior is based on the interaction of multiple complex phenomena, and recent evolutionary approaches are being applied more on a phenomenon-by-phenomenon basis, targeting localized problems, as opposed to a systematic approach based on a fundamental understanding of all interacting parameters. Advanced nuclear fuels are generally more complex, and less understood, than the traditional fuels used in existing reactors (ceramic UO 2 with burnable poisons and other minor additives). The added challenges are primarily caused by a less complete empirical database and, in the case of recycled fuel, the inherent variability in fuel compositions. It is clear that using the traditional approach to develop and qualify fuels over the entire range of variables pertinent to the U.S. Department of Energy (DOE) Office of Nuclear Energy on a timely basis with available funds would be very challenging, if not impossible. As a result the DOE Office of Nuclear Energy has launched the Nuclear Energy Advanced Modeling and Simulation (NEAMS) approach to revolutionize fuel development. This new approach is predicated upon transferring the recent advances in computational sciences and computer technologies into the fuel development program. The effort will couple computational science with recent advances in the fundamental understanding of physical phenomena through ab initio modeling and targeted phenomenological testing to leapfrog many fuel

  12. Engineering opportunities in nuclear engineering

    International Nuclear Information System (INIS)

    Walton, D.G.

    1980-01-01

    The pattern of education and training of Nuclear Engineers in the UK is outlined under the headings; degree courses for professional engineers, postgraduate courses, education of technician engineers. Universities which offer specific courses are stated and useful addresses listed. (UK)

  13. Analytical three-dimensional neutron transport benchmarks for verification of nuclear engineering codes. Final report

    International Nuclear Information System (INIS)

    Ganapol, B.D.; Kornreich, D.E.

    1997-01-01

    Because of the requirement of accountability and quality control in the scientific world, a demand for high-quality analytical benchmark calculations has arisen in the neutron transport community. The intent of these benchmarks is to provide a numerical standard to which production neutron transport codes may be compared in order to verify proper operation. The overall investigation as modified in the second year renewal application includes the following three primary tasks. Task 1 on two dimensional neutron transport is divided into (a) single medium searchlight problem (SLP) and (b) two-adjacent half-space SLP. Task 2 on three-dimensional neutron transport covers (a) point source in arbitrary geometry, (b) single medium SLP, and (c) two-adjacent half-space SLP. Task 3 on code verification, includes deterministic and probabilistic codes. The primary aim of the proposed investigation was to provide a suite of comprehensive two- and three-dimensional analytical benchmarks for neutron transport theory applications. This objective has been achieved. The suite of benchmarks in infinite media and the three-dimensional SLP are a relatively comprehensive set of one-group benchmarks for isotropically scattering media. Because of time and resource limitations, the extensions of the benchmarks to include multi-group and anisotropic scattering are not included here. Presently, however, enormous advances in the solution for the planar Green's function in an anisotropically scattering medium have been made and will eventually be implemented in the two- and three-dimensional solutions considered under this grant. Of particular note in this work are the numerical results for the three-dimensional SLP, which have never before been presented. The results presented were made possible only because of the tremendous advances in computing power that have occurred during the past decade

  14. Catalogue of nuclear fusion codes - 1976

    International Nuclear Information System (INIS)

    1976-10-01

    A catalogue is presented of the computer codes in nuclear fusion research developed by JAERI, Division of Thermonuclear Fusion Research and Division of Large Tokamak Development in particular. It contains a total of about 100 codes under the categories: Atomic Process, Data Handling, Experimental Data Processing, Engineering, Input and Output, Special Languages and Their Application, Mathematical Programming, Miscellaneous, Numerical Analysis, Nuclear Physics, Plasma Physics and Fusion Research, Plasma Simulation and Numerical Technique, Reactor Design, Solid State Physics, Statistics, and System Program. (auth.)

  15. Nuclear chemical engineering

    International Nuclear Information System (INIS)

    Lee, Geon Jae; Shin, Young Jun

    1989-08-01

    The contents of this book are introduction of chemical engineering and related chemistry on an atomic reactor, foundation of the chemistry nuclear chemical engineering, theory on nuclear engineering, the cycle of uranium and nuclear fuel, a product of nuclear division, nuclear reprocessing, management of spent fuel separation of radioisotope, materials of an atomic reactor, technology and chemistry related water in atomic reactors and utilization of radioisotope and radiation. This book has the exercises and reference books for the each chapter.

  16. Bar codes for nuclear safeguards

    International Nuclear Information System (INIS)

    Keswani, A.N.; Bieber, A.M. Jr.

    1983-01-01

    Bar codes similar to those used in supermarkets can be used to reduce the effort and cost of collecting nuclear materials accountability data. A wide range of equipment is now commercially available for printing and reading bar-coded information. Several examples of each of the major types of commercially available equipment are given, and considerations are discussed both for planning systems using bar codes and for choosing suitable bar code equipment

  17. Bar codes for nuclear safeguards

    International Nuclear Information System (INIS)

    Keswani, A.N.; Bieber, A.M.

    1983-01-01

    Bar codes similar to those used in supermarkets can be used to reduce the effort and cost of collecting nuclear materials accountability data. A wide range of equipment is now commercially available for printing and reading bar-coded information. Several examples of each of the major types of commercially-available equipment are given, and considerations are discussed both for planning systems using bar codes and for choosing suitable bar code equipment

  18. KWIC Index of nuclear codes (1975 edition)

    International Nuclear Information System (INIS)

    Akanuma, Makoto; Hirakawa, Takashi

    1976-01-01

    It is a KWIC Index for 254 nuclear codes in the Nuclear Code Abstracts (1975 edition). The classification of nuclear codes and the form of index are the same as those in the Computer Programme Library at Ispra, Italy. (auth.)

  19. The nuclear codes and guidelines

    International Nuclear Information System (INIS)

    Sonter, M.

    1984-01-01

    This paper considers problems faced by the mining industry when implementing the nuclear codes of practice. Errors of interpretation are likely. A major criticism is that the guidelines to the codes must be seen as recommendations only. They are not regulations. Specific clauses in the guidelines are criticised

  20. Development of the next generation code system as an engineering modeling language (6). Development of a cross section adjustment and nuclear design accuracy evaluation solver

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Numata, Kazuyuki

    2008-01-01

    A new cross section adjustment and nuclear design accuracy evaluation solver was developed as a set of modules for MARBLE (multi-purpose advanced reactor physics analysis system based on language of engineering). In order to enhance the system extendibility and flexibility, the object-oriented design and analysis technique was adopted to the development. In the new system, it is easy to add a new design accuracy evaluation method because a new numerical calculation module is independent from other modules. Further, several new functions such as searching and editing calculation data are provided in the new solver. These functions can be easily customised by users because they are designed to work cooperatively with Python scripting language, which is used as a user interface of the MARBLE system. In order to validate the new solver, a test calculation was performed for a realistic calculation case of creating a new unified cross section library. In the test calculation, results calculated by the new solver agreed well with those by the conventional code system. In addition, it is possible to reuse existing input data files prepared for the conventional code system because the new solver utilities support the conventional formats. Because the new solver implements all main functions of the conventional code system, MARBLE can be used as a new calculation code system for cross section adjustment and nuclear design accuracy evaluation

  1. Nuclear safety code study

    Energy Technology Data Exchange (ETDEWEB)

    Hu, H.H.; Ford, D.; Le, H.; Park, S.; Cooke, K.L.; Bleakney, T.; Spanier, J.; Wilburn, N.P.; O' Reilly, B.; Carmichael, B.

    1981-01-01

    The objective is to analyze an overpower accident in an LMFBR. A simplified model of the primary coolant loop was developed in order to understand the instabilities encountered with the MELT III and SAS codes. The computer programs were translated for switching to the IBM 4331. Numerical methods were investigated for solving the neutron kinetics equations; the Adams and Gear methods were compared. (DLC)

  2. Nuclear Reactor Engineering Analysis Laboratory

    International Nuclear Information System (INIS)

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

    1998-01-01

    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

  3. Dictionary of nuclear engineering

    International Nuclear Information System (INIS)

    Sube, R.

    1985-01-01

    This dictionary covers nuclear engineering defined in its general sense as applied nuclear physics: industrial and other applications of nuclear power, isotopes and ionizing radiation, nuclear materials, nuclear facilities and nuclear weapons together with their scientific and technological fundamentals. During the compilation of terms, great attention was only given to generally valid basic expressions and to special terms where these occurred in all four languages. A great number of textbooks and monographs, as well as specialist journals covering many years, have been evaluated. Detailed attention has been paid to standards. Of importance in nuclear engineering are the international standards of the International Atomic Energy Organization (including the terminology employed by the International Nuclear Information System INIS), the International Organization of Standardization, the Council for Mutual Economic Assistance, the World Energy Conference, the International Electrical Engineering Commission, and also a great many national standards which, unfortunately, frequently deviate from one another as regards definition and, in particular, designation. (orig.)

  4. Nuclear ship engineering simulator

    International Nuclear Information System (INIS)

    Itoh, Yasuyoshi; Kusunoki, Tsuyoshi; Hashidate, Koji

    1991-01-01

    The nuclear ship engineering simulator, which analyzes overall system response of nuclear ship numerically, is now being developed by JAERI as an advanced design tool with the latest computer technology in software and hardware. The development of the nuclear ship engineering simulator aims at grasping characteristics of a reactor plant under the situation generated by the combination of ocean, a ship hull and a reactor. The data from various tests with the nuclear ship 'MUTSU' will be used for this simulator to modulate and verify its functions of reproducing realistic response of nuclear ship, and then the simulator will be utilized for the research and development of advanced marine reactors. (author)

  5. Facts in nuclear engineering

    International Nuclear Information System (INIS)

    Buenemann, D.

    1979-07-01

    This compilation of facts has been published on behalf of Kerntechnische Gesellschaft (Society for Nuclear Engineering), as a basis for the discussion between promoters and opponents of nuclear power. It intends to make the nuclear discussion less emotional by providing relevant data material. (orig./HP) [de

  6. Nuclear engineering vocabulary

    International Nuclear Information System (INIS)

    Dumont, X.; Andrieux, C.

    2001-01-01

    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)

  7. Introduction to nuclear engineering

    International Nuclear Information System (INIS)

    Gylys, J.

    1997-01-01

    The textbook, which is the first book in Lithuanian on this subject generalises information on key aspects of nuclear engineering. Specialists in nuclear power for Ignalina NPP and for the infrastructure of nuclear energy sector of Lithuania are prepared at Kaunas University of Technology. The textbooks the students and lecturers have been using to-date were mostly in other languages than Lithuanian and they have not been adapted for teaching in Lithuania's higher educational establishments. This textbook is useful also to anyone who is interested in the issues and future prospects of nuclear power. It contains the chapters on nuclear reactions, theory of nuclear reactors, nuclear reactors kinetics, neutronic analysis, thermalhydraulic calculations of nuclear reactors operation and description of the construction of Ignalina NPP. (author)

  8. Nuclear propulsion systems engineering

    International Nuclear Information System (INIS)

    Madsen, W.W.; Neuman, J.E.: Van Haaften, D.H.

    1992-01-01

    The Nuclear Energy for Rocket Vehicle Application (NERVA) program of the 1960's and early 1970's was dramatically successful, with no major failures during the entire testing program. This success was due in large part to the successful development of a systems engineering process. Systems engineering, properly implemented, involves all aspects of the system design and operation, and leads to optimization of theentire system: cost, schedule, performance, safety, reliability, function, requirements, etc. The process must be incorporated from the very first and continued to project completion. This paper will discuss major aspects of the NERVA systems engineering effort, and consider the implications for current nuclear propulsion efforts

  9. Introduction to nuclear engineering

    International Nuclear Information System (INIS)

    Bouchard, J.; Deffain, J.P.; Gouchet, A.

    2007-01-01

    This book is an introductory course to the technology of nuclear reactors, it is based on lectures given to students engaged in nuclear engineering curricula. A brief historical account of nuclear power is given in which Three Miles Island and Chernobyl accidents appear to be the main milestones because of their big impacts on the way to apprehend the safety in the design of a reactor for the first and on the public controversy of nuclear energy for the second. All the concepts and knowledge required to understand the design of a reactor and how it operates, are described: radioactivity, nuclear safety, irradiation effects on materials, uranium enrichment, reactor kinetics, fission products poisoning,... This book is made up of 4 parts: 1) nuclear power, 2) types of power nuclear reactors (PWR, BWR and fast reactors), 3) the fuel cycle, and 4) neutronics basics. (A.C.)

  10. Application of thermal-hydraulic codes in the nuclear sector

    International Nuclear Information System (INIS)

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

    2011-01-01

    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)

  11. Nuclear engineering vocabulary

    International Nuclear Information System (INIS)

    2006-01-01

    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.)

  12. Design of hoisting device used in nuclear power plants. KTA safety engineering code. Draft amendment, as of 6/98

    International Nuclear Information System (INIS)

    1998-06-01

    The draft amendment specifies, according to the hazards involved, (a) the general provisions, (b) additional provisions supplementing the general provisions, (c) more stringent provisions relating to aggravated risks, applicable to lifting gear, and (d) additional requirements for elevators in reactor containments and refuelling equipment, to be taken into account in the design of hoisting device used in nuclear power plants. The term hoisting device in this context covers elevators, cranes, winches, trolleys, load carrying equipment, and LWR refuelling machines as are used in nuclear power plants. (orig./CB) [de

  13. 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.

  14. Welding in nuclear engineering

    International Nuclear Information System (INIS)

    1982-01-01

    The 3rd international conference 'Welding in nuclear engineering', organized in 1978 by the Deutscher Verband fuer Schweisstechnik e.V., was, like the two foregoing conferences in 1970 and 1974, an absolute success. The noteworthy echo to this meeting in the international technical world - the number of 650 participants from 26 countries is self-evidence - and this fact, was for the Deutscher Verband fuer Schweisstechnik e.V. occasion and at the same time an obligation now to follow in the same way, the meeting that was started 12 years ago, by organizing the international conference 'Welding in nuclear engineering'. The conference this year offers in addition to the two plenary session lectures, 34 short reports and a further 28 single contributions in the form of two poster-sessions. Unfortunately, it was again not possible to accept all the papers submitted because the conference was limited to 2 days only. Nevertheless, the papers will offer a representative cross-section through the total range of welding engineering. In particular, the poster session, which take place for the first time within the scope of a meeting organized by the Working Group 'Welding in Nuclear Engineering', should contribute to the aim that this time again the discussions will form the main point of the conference. (orig./RW) [de

  15. Nuclear rocket engine reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lanin, Anatoly

    2013-07-01

    Covers a new technology of nuclear reactors and the related materials aspects. Integrates physics, materials science and engineering Serves as a basic book for nuclear engineers and nuclear physicists. 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.

  16. Nuclear Medicine Engineering

    International Nuclear Information System (INIS)

    Mateescu, Gheorghe; Craciunescu, Teddy

    2000-01-01

    'An image is more valuable than a thousand words' - this is the thought that underlies the authors' vision about the field of nuclear medicine. The monograph starts with a review of some theoretical and engineering notions that grounds the field of nuclear medicine: nuclear radiation, interaction of radiation with matter, radiation detection and measurement, numerical analysis. Products and methods needed for the implementation of diagnostic and research procedures in nuclear medicine are presented: radioisotopes and radiopharmaceuticals, equipment for in-vitro (radioimmunoassay, liquid scintillation counting) and in-vivo investigations (thyroid uptake, renography, dynamic studies, imaging). A special attention is focused on medical imaging theory and practice as a source of clinical information (morphological and functional). The large variety of parameters, components, biological structures and specific properties of live matter determines the practical use of three-dimensional tomographic techniques based on diverse physical principles: single-photon emission, positron emission, X-rays transmission, nuclear magnetic resonance, ultrasounds transmission and reflection, electrical impedance measurement. The fundamental reconstruction algorithms i.e., algorithms based on the projection theorem and Fourier filtering, algebraic reconstruction techniques and the algorithms based on statistical principles: maximum entropy, maximum likelihood, Monte Carlo algorithms, are depicted in details. A method based on the use of the measured point spread function is suggested. Some classical but often used techniques like linear scintigraphy and Anger gamma camera imaging are also presented together with some image enhancement techniques like Wiener filtering and blind deconvolution. The topic of the book is illustrated with some clinical samples obtained with nuclear medicine devices developed in the Nuclear Medicine Laboratory of the National Institute of Nuclear Physics and

  17. Code of Ethics for Electrical Engineers

    Science.gov (United States)

    Matsuki, Junya

    The Institute of Electrical Engineers of Japan (IEEJ) has established the rules of practice for its members recently, based on its code of ethics enacted in 1998. In this paper, first, the characteristics of the IEEJ 1998 ethical code are explained in detail compared to the other ethical codes for other fields of engineering. Secondly, the contents which shall be included in the modern code of ethics for electrical engineers are discussed. Thirdly, the newly-established rules of practice and the modified code of ethics are presented. Finally, results of questionnaires on the new ethical code and rules which were answered on May 23, 2007, by 51 electrical and electronic students of the University of Fukui are shown.

  18. Nuclear science references coding manual

    International Nuclear Information System (INIS)

    Ramavataram, S.; Dunford, C.L.

    1996-08-01

    This manual is intended as a guide to Nuclear Science References (NSR) compilers. The basic conventions followed at the National Nuclear Data Center (NNDC), which are compatible with the maintenance and updating of and retrieval from the Nuclear Science References (NSR) file, are outlined. In Section H, the structure of the NSR file such as the valid record identifiers, record contents, text fields as well as the major TOPICS for which are prepared are enumerated. Relevant comments regarding a new entry into the NSR file, assignment of , generation of and linkage characteristics are also given in Section II. In Section III, a brief definition of the Keyword abstract is given followed by specific examples; for each TOPIC, the criteria for inclusion of an article as an entry into the NSR file as well as coding procedures are described. Authors preparing Keyword abstracts either to be published in a Journal (e.g., Nucl. Phys. A) or to be sent directly to NNDC (e.g., Phys. Rev. C) should follow the illustrations in Section III. The scope of the literature covered at the NNDC, the categorization into Primary and Secondary sources, etc., is discussed in Section IV. Useful information regarding permitted character sets, recommended abbreviations, etc., is given under Section V as Appendices

  19. Nuclear structure references coding manual

    International Nuclear Information System (INIS)

    Ramavataram, S.; Dunford, C.L.

    1984-02-01

    This manual is intended as a guide to Nuclear Structure References (NSR) compilers. The basic conventions followed at the National Nuclear Data Center (NNDC), which are compatible with the maintenance and updating of and retrieval from the Nuclear Structure References (NSR) file, are outlined. The structure of the NSR file such as the valid record identifiers, record contents, text fields as well as the major topics for which [KEYWORDS] are prepared are ennumerated. Relevant comments regarding a new entry into the NSR file, assignment of [KEYNO ], generation of [SELECTRS] and linkage characteristics are also given. A brief definition of the Keyword abstract is given followed by specific examples; for each TOPIC, the criteria for inclusion of an article as an entry into the NSR file as well as coding procedures are described. Authors submitting articles to Journals which require Keyword abstracts should follow the illustrations. The scope of the literature covered at NNDC, the categorization into Primary and Secondary sources, etc. is discussed. Useful information regarding permitted character sets, recommended abbreviations, etc. is given

  20. Neutrons and Nuclear Engineering

    International Nuclear Information System (INIS)

    Ekkebus, Allen E.

    2007-01-01

    Oak Ridge National Laboratory hosted two workshops in April 2007 relevant to nuclear engineering education. In the Neutron Stress, Texture, and Phase Transformation for Industry workshop (http://neutrons.ornl.gov/workshops/nst2/), several invited speakers gave examples of neutron stress mapping for nuclear engineering applications. These included John Root of National Research Council of Canada, Mike Fitzpatrick of the UK's Open University, and Yan Gao of GE Global Research on their experiences with industrial and academic uses of neutron diffraction. Xun-Li Wang and Camden Hubbard described the new instruments at ORNL that can be used for such studies. This was preceded by the Neutrons for Materials Science and Engineering educational symposium (http://neutrons.ornl.gov/workshops/edsym2007). It was directed to the broad materials science and engineering community based in universities, industry and laboratories who wish to learn what the neutron sources in the US can provide for enhancing the understanding of materials behavior, processing and joining. Of particular interest was the presentation of Donald Brown of Los Alamos about using 'Neutron diffraction measurements of strain and texture to study mechanical behavior of structural materials.' At both workshops, the ORNL neutron scattering instruments relevant to nuclear engineering studies were described. The Neutron Residual Stress Mapping Facility (NRSF2) is currently in operation at the High Flux Isotope Reactor; the VULCAN Engineering Materials Diffractometer will begin commissioning in 2008 at the Spallation Neutron Source. For characteristics of these instruments, as well as details of other workshops, meetings, capabilities, and research proposal submissions, please visit http://neutrons.ornl.gov. To submit user proposals for time on NRSF2 contact Hubbard at hubbardcratornl.gov

  1. Nuclear engineering dictionary. Woerterbuch Kerntechnik

    Energy Technology Data Exchange (ETDEWEB)

    Sube, R

    1985-01-01

    This dictionary treats the subject field of nuclear engineering as a field of applied nuclear physics: Industrial and other applications of nuclear energy, isotopes and ionizing radiation, and their, scientific-technical bases. Emphasis is placed on the terminology of the nuclear fuel cycle. Other applications of nuclear energy include military applications, nuclear fusion technology, and plasma physics, as well as methods and equipment of isotope and radiation technology, without the aspects of biological applications. High-energy physics is also excluded. The terminology presented primarily covers general and basic concepts, special terms have been included as far as available and ascertainable in all four languages. For selection of terms, numerous textbooks and monographies have been searched and compared, as well as various subject-related journals which have been regularly scanned for years. Standards have been a main source of information, as e.g. the international standards of the IAEA (including the INIS terminology), of the ISO, of the COMECON, and of the World Energy Conference and the IEC. Numerous national standards have been evaluated in search for definitions and designations. Users will appreciate the introduction of subject-field codes indicating the main field of usage of a term. Explanations and other hints are numerous and extensive in order to clearly define the terms chosen from other, similar terms, and in order to show homonyms.

  2. Globalization of ASME Nuclear Codes and Standards

    International Nuclear Information System (INIS)

    Swayne, Rick; Erler, Bryan A.

    2006-01-01

    With the globalization of the nuclear industry, it is clear that the reactor suppliers are based in many countries around the world (such as United States, France, Japan, Canada, South Korea, South Africa) and they will be marketing their reactors to many countries around the world (such as US, China, South Korea, France, Canada, Finland, Taiwan). They will also be fabricating their components in many different countries around the world. With this situation, it is clear that the requirements of ASME Nuclear Codes and Standards need to be adjusted to accommodate the regulations, fabricating processes, and technology of various countries around the world. It is also very important for the American Society of Mechanical Engineers (ASME) to be able to assure that products meeting the applicable ASME Code requirements will provide the same level of safety and quality assurance as those products currently fabricated under the ASME accreditation process. To do this, many countries are in the process of establishing or changing their regulations, and it is important for ASME to interface with the appropriate organizations in those countries, in order to ensure there is effective use of ASME Codes and standards around the world. (authors)

  3. Nuclear engine system simulation (NESS) program update

    International Nuclear Information System (INIS)

    Scheil, C.M.; Pelaccio, D.G.; Petrosky, L.J.

    1993-01-01

    The second phase of development of a Nuclear Thermal Propulsion (NTP) engine system design analysis code has been completed. The standalone, versatile Nuclear Engine System Simulation (NESS) code provides an accurate, detailed assessment of engine system operating performance, weight, and sizes. The critical information is required to support ongoing and future engine system and stage design study efforts. This recent development effort included incorporation of an updated solid-core nuclear thermal reactor model that yields a reduced core weight and higher fuel power density when compared to a NERVA type reactor. NESS can now analyze expander, gas generator, and bleed cycles, along with multi-redundant propellant pump feed systems. Performance and weight of efficient multi-stage axial turbopump can now be determined, in addition to the traditional centrifugal pump

  4. Metallurgy for nuclear engineering

    International Nuclear Information System (INIS)

    Kozlov, A.F.

    1986-01-01

    Principal ways of development in metallurgy and metallurgical equipment on nuclear engineering plants are discussed. A great attention is paid to changing welded structures for casted and forged ones. These measures give the possibility to increase reliability of NPP components and decrease labour content. The following processing procedures have been introduced: vacuum carbon reduction providing small amount of nonmetallic inclusions in reactor vessel steel; manufacturing steel large-size castings (360 and 420 t) for WWER vessels; rolling at plate mill 5000 etc

  5. 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...

  6. Computer codes for ventilation in nuclear facilities

    International Nuclear Information System (INIS)

    Mulcey, P.

    1987-01-01

    In this paper the authors present some computer codes, developed in the last years, for ventilation and radioprotection. These codes are used for safety analysis in the conception, exploitation and dismantlement of nuclear facilities. The authors present particularly: DACC1 code used for aerosol deposit in sampling circuit of radiation monitors; PIAF code used for modelization of complex ventilation system; CLIMAT 6 code used for optimization of air conditioning system [fr

  7. History of nuclear engineering curricula

    International Nuclear Information System (INIS)

    Murphy, G.

    1975-01-01

    With the realization that nuclear energy had a vast potential for peacetime development, universities throughout the country began to develop courses in nuclear energy. A pioneering educational effort was necessary because there was an inadequate number of trained faculty, no established curricula, no textbooks, and very little suitable equipment. Nevertheless, by the early 1950's, several programs in nuclear science and engineering were beginning to provide instruction to potential nuclear engineers. At that time, the American Society for Engineering Education (ASEE) established a nuclear committee to cooperate with the U. S. Atomic Energy Commission (AEC) in nuclear education matters. With the financial support of the AEC, textbook material was developed, faculty training programs were instituted, and funds were made available for equipment. Because of the large interest shown in the field, many colleges and universities began to develop nuclear engineering curricula. After a few years, the need arose for general guidelines in curricular development. This led to the development of a Committee on Objective Criteria in Nuclear Engineering Education in which ASEE and the American Nuclear Society cooperated with the support of AEC. The committee report emphasized basic science, nuclear energy concepts, and nuclear technology, which have continued to be the significant components of a nuclear engineering curriculum. The last ten years have brought increased emphasis on BS programs, the introduction of extensive computer-based instruction, and an increasing emphasis on the engineering aspects of nuclear reactor power systems

  8. Nuclear engineering in the linelight

    International Nuclear Information System (INIS)

    Blumentritt, G.; Schwaar, L.

    1979-01-01

    An insight is given into the state of art of nuclear engineering considering only essential problems. The subject is covered under the following headings: (1) the way to nuclear fission, (2) detectors for nuclear radiation, (3) measuring systems for nuclear radiation, (4) radioisotopes in industry, (5) aids in medicine, (6) radiation absorption and its utilization, (7) use of radioisotopes in research, (8) the chain reaction in a nuclear reactor, (9) power from nuclear power plants, (10) pressurized water reactors (PWR), (11) high-temperature reactors (HTGR), (12) fast breeder reactors (FBR), (13) nuclear energetics - a new branch of industry, (14) nuclear explosions, (15) nuclear research at Rossendorf, and (16) the energy of the future. An appendix includes definitions of terms used in nuclear engineering. The book is written for a wide circle of readers who are interested in the peaceful uses of nuclear energy

  9. Choosing nuclear engineering: A survey of nuclear engineering undergraduates

    International Nuclear Information System (INIS)

    Shillenn, J.K.; Klevans, E.H.

    1988-01-01

    Maintaining a reliable pool of qualified nuclear engineering graduates depends on the ability of nuclear engineering undergraduate programs to recruit students. With the prospect of declining enrollments in nuclear engineering it is important for nuclear engineering programs to know what factors influence students to choose nuclear engineering as an undergraduate major and why they choose a particular undergraduate program. This type of information can be very important to nuclear engineering programs that develop recruiting strategies. To provide some insight into this area, a questionnaire was designed and given to undergraduate nuclear engineering students at Pennsylvania State University. The purpose of the survey was to provide information on the reasons that students picked nuclear engineering as a career and chose to attend Penn State. The questionnaire was given to 27 students in their junior year during the spring semester of 1987 and again to 35 junior students during the spring semester of 1988. There was little difference except as noted between the two groups on their responses to the questionnaire. A partial listing of the survey results is provided

  10. 4+ Dimensional nuclear systems engineering

    International Nuclear Information System (INIS)

    Suh, Kune Y.

    2009-01-01

    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 + D) Technology TM , a critical know how for digital management. ESSE (Engineering Super Simulation Emulation) features a 4 + D Technology TM for nuclear energy systems engineering. The technology proposed in the 3D space and time plus cost coordinates, i.e. 4 + 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 + 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 two machines and losses in

  11. 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.

  12. Risks in nuclear engineering

    International Nuclear Information System (INIS)

    Lindackers, K.H.

    1982-01-01

    The German nuclear power plant risk assessment study has not contributed to a higher degree of acceptance of light-water reactors among the general public. One reason is the fact that its predictions are much too inaccurate, and the consequences from severe accidents, regardless of their possibly extremely small likelihood, speak for themselves. The work still to be done in Phase B of the risk analysis will only be useful, if the safety factors in risk assessments can be drastically increased. The results of the risk analysis cannot be used in legislation or the administration of justice, because they are too incomplete and inaccurate. Scientific findings in risk analysis show that new designs and new components complicate objective judgement of changes in reactor safety engineering considerably. Every change in the required precautionary safety measures must be carefully considered, and if need be tested under the appropriate conditions. (orig./HSCH) [de

  13. The nuclear reaction model code MEDICUS

    International Nuclear Information System (INIS)

    Ibishia, A.I.

    2008-01-01

    The new computer code MEDICUS has been used to calculate cross sections of nuclear reactions. The code, implemented in MATLAB 6.5, Mathematica 5, and Fortran 95 programming languages, can be run in graphical and command line mode. Graphical User Interface (GUI) has been built that allows the user to perform calculations and to plot results just by mouse clicking. The MS Windows XP and Red Hat Linux platforms are supported. MEDICUS is a modern nuclear reaction code that can compute charged particle-, photon-, and neutron-induced reactions in the energy range from thresholds to about 200 MeV. The calculation of the cross sections of nuclear reactions are done in the framework of the Exact Many-Body Nuclear Cluster Model (EMBNCM), Direct Nuclear Reactions, Pre-equilibrium Reactions, Optical Model, DWBA, and Exciton Model with Cluster Emission. The code can be used also for the calculation of nuclear cluster structure of nuclei. We have calculated nuclear cluster models for some nuclei such as 177 Lu, 90 Y, and 27 Al. It has been found that nucleus 27 Al can be represented through the two different nuclear cluster models: 25 Mg + d and 24 Na + 3 He. Cross sections in function of energy for the reaction 27 Al( 3 He,x) 22 Na, established as a production method of 22 Na, are calculated by the code MEDICUS. Theoretical calculations of cross sections are in good agreement with experimental results. Reaction mechanisms are taken into account. (author)

  14. Survey of nuclear fuel-cycle codes

    International Nuclear Information System (INIS)

    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

  15. Recent activities on nuclear codes and standards

    International Nuclear Information System (INIS)

    Minematsu, Akiyoshi; Ishimoto, Shozaburo; Honjin, Masao

    2000-01-01

    The technical codes and standards relating to the nuclear power stations in Japan are prepared by shapes of laws (ministerial ordinances and bulletins) issued by the government and obliged to comply with by 'the Law concerning the Regulations of Nuclear Material Substances, Nuclear Fuel Substances and Nuclear Reactors' and 'the Electricity Business Act' and of guides defined by the Nuclear Safety Commission, and further some private standards have been issued at a shape of complement of these laws and guides by receiving national recommendation. On the other hand, in the fields of electricity and heat facilities except atomic energy, simplification and feature stipulation of the national technical codes and standards was recently carried out, by which a system usable for the private standards in and out of Japan were prepared through approval of the private Japan Electrotechnical Standards and Codes Committee (JESC). As the nuclear field was now excepted from simultaneous transfer to the private standard and the standard application system, it is expected in future to realize similar transfer if possible and preparation of the private standards is now being advanced. Here were introduced on present state on technical codes and standards relating to the nuclear power generation facilities and recent trends on their private standardization. (G.K.)

  16. CTCN: Colloid transport code -- nuclear

    International Nuclear Information System (INIS)

    Jain, R.

    1993-01-01

    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

  17. Fluidization in nuclear engineering

    Energy Technology Data Exchange (ETDEWEB)

    Sathiyamoorthy, D; Venugopalan, Ramani; Vijay, P L [Metallurgy Division, Bhabha Atomic Research Centre, Mumbai (India); Varadarajan, T G [Heavy Water Division, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    Fluidization technique has not been fully exploited in nuclear industries mainly due to lack of open literature or unawareness of its applications. Hence in this paper a detailed range of applications of fluidization in uranium extraction, nuclear fuel material preparation, fuel reprocessing and waste disposal is highlighted. A fluidized bed nuclear reactor concept is also presented. The need of fluidization for process improvement and modernization in nuclear programmes is stressed. (author). 40 refs., 3 figs.

  18. Introduction to nuclear facilities engineering

    International Nuclear Information System (INIS)

    Sapy, Georges

    2012-06-01

    Engineering, or 'engineer's art', aims at transforming simple principle schemes into operational facilities often complex especially when they concern the nuclear industry. This transformation requires various knowledge and skills: in nuclear sciences and technologies (nuclear physics, neutronics, thermal-hydraulics, material properties, radiation protection..), as well as in non-nuclear sciences and technologies (civil engineering, mechanics, electricity, computer sciences, instrumentation and control..), and in the regulatory, legal, contractual and financial domains. This book explains how this huge body of knowledge and skills must be organized and coordinated to create a reliable, exploitable, available, profitable and long-lasting facility, together with respecting extremely high safety, quality, and environmental impact requirements. Each aspect of the problem is approached through the commented presentation of nuclear engineering macro-processes: legal procedures and administrative authorizations, nuclear safety/radiation protection/security approach, design and detailed studies, purchase of equipments, on-site construction, bringing into operation, financing, legal, contractual and logistic aspects, all under the global control of a project management. The 'hyper-complexness' of such an approach leads to hard points and unexpected events. The author identifies the most common ones and proposes some possible solutions to avoid, mitigate or deal with them. In a more general way, he proposes some thoughts about the performance factors of a nuclear engineering process

  19. Vectorization, parallelization and porting of nuclear codes. 2001

    International Nuclear Information System (INIS)

    Akiyama, Mitsunaga; Katakura, Fumishige; Kume, Etsuo; Nemoto, Toshiyuki; Tsuruoka, Takuya; Adachi, Masaaki

    2003-07-01

    Several computer codes in the nuclear field have been vectorized, parallelized and transported on the super computer system at Center for Promotion of Computational Science and Engineering in Japan Atomic Energy Research Institute. We dealt with 10 codes in fiscal 2001. In this report, the parallelization of Neutron Radiography for 3 Dimensional CT code NR3DCT, the vectorization of unsteady-state heat conduction code THERMO3D, the porting of initial program of MHD simulation, the tuning of Heat And Mass Balance Analysis Code HAMBAC, the porting and parallelization of Monte Carlo N-Particle transport code MCNP4C3, the porting and parallelization of Monte Carlo N-Particle transport code system MCNPX2.1.5, the porting of induced activity calculation code CINAC-V4, the use of VisLink library in multidimensional two-fluid model code ACD3D and the porting of experiment data processing code from GS8500 to SR8000 are described. (author)

  20. Survey of Codes Employing Nuclear Damage Assessment

    Science.gov (United States)

    1977-10-01

    surveyed codes were com- DO 73Mu 1473 ETN OF 1NOVSSSOLETE UNCLASSIFIED 1 SECURITY CLASSIFICATION OF THIS f AGE (Wh*11 Date Efntered)S<>-~C. I UNCLASSIFIED...level and above) TALLEY/TOTEM not nuclear TARTARUS too highly aggregated (battalion level and above) UNICORN highly aggregated force allocation code...vulnerability data can bq input by the user as he receives them, and there is the abil ’ity to replay any situation using hindsight. The age of target

  1. ENEN - European nuclear engineering network

    International Nuclear Information System (INIS)

    Comsa, Olivia; Paraschiva, M.V.; Banutoiu, Maria

    2002-01-01

    The paper presents the main objectives and expected results of European Project FP5 - ENEN - 'European Nuclear Engineering Network'. The underlying objective of the work is safeguarding the nuclear knowledge and expertise through the preservation of higher nuclear engineering education. Co-operation between universities and universities and research centres, will entail a better use of dwindling teaching capacity, scientific equipment and research infrastructure. 'Today, the priorities of the scientific community regarding basic research lie elsewhere than in nuclear sciences. Taken together, these circumstances create a significantly different situation from three to four decades ago when much of the present competence base was in fact generated. In addition, many of the highly competent engineers and scientists, who helped create the present nuclear industry, and its regulatory structure, are approaching retirement age. These competence issues need to be addressed at Community level and a well designed Community research and training programme should play a role that is more important than ever before. This is an area where the concept of an European research area should be further explored'. The outcome from this project should be a clear road map for the way ahead in nuclear engineering education in Europe. The underlying objective of the concerted action is the preservation of nuclear knowledge and expertise through the preservation of higher nuclear engineering education. 'Many diverse technologies, currently serving nations world-wide, would be affected by an inadequate number of future nuclear scientists and engineers. Nuclear technology is widespread and multidisciplinary: nuclear and reactor physics, thermal hydraulics and mechanics, material science, chemistry, health science, information technology and a variety of other areas. Yet the advancement of this technology, with all its associated benefits, will be threatened if not curtailed unless the

  2. Bar code usage in nuclear materials accountability

    International Nuclear Information System (INIS)

    Mee, W.T.

    1983-01-01

    The age old method of physically taking an inventory of materials by listing each item's identification number has lived beyond its usefulness. In this age of computerization, which offers the local grocery store a quick, sure, and easy means to inventory, it is time for nuclear materials facilities to automate accountability activities. The Oak Ridge Y-12 Plant began investigating the use of automated data collection devices in 1979. At that time, bar code and optical-character-recognition (OCR) systems were reviewed with the purpose of directly entering data into DYMCAS (Dynamic Special Nuclear Materials Control and Accountability System). Both of these systems appeared applicable; however, other automated devices already employed for production control made implementing the bar code and OCR seem improbable. However, the DYMCAS was placed on line for nuclear material accountability, a decision was made to consider the bar code for physical inventory listings. For the past several months a development program has been underway to use a bar code device to collect and input data to the DYMCAS on the uranium recovery operations. Programs have been completed and tested, and are being employed to ensure that data will be compatible and useful. Bar code implementation and expansion of its use for all nuclear material inventory activity in Y-12 is presented

  3. Code of safety for nuclear merchant ships

    International Nuclear Information System (INIS)

    1982-01-01

    The Code is in chapters, entitled: general (including general safety principles and principles of risk acceptance); design criteria and conditions; ship design, construction and equipment; nuclear steam supply system; machinery and electrical installations; radiation safety (including radiological protection design; protection of persons; dosimetry; radioactive waste management); operation (including emergency operation procedures); surveys. Appendices cover: sinking velocity calculations; seaway loads depending on service periods; safety assessment; limiting dose-equivalent rates for different areas and spaces; quality assurance programme; application of single failure criterion. Initial application of the Code is restricted to conventional types of ships propelled by nuclear propulsion plants with pressurized light water type reactors. (U.K.)

  4. Bar code usage in nuclear materials accountability

    International Nuclear Information System (INIS)

    Mee, W.T.

    1983-01-01

    The Oak Ridge Y-12 Plant began investigating the use of automated data collection devices in 1979. At this time, bar code and optical-character-recognition (OCR) systems were reviewed with the purpose of directly entering data into DYMCAS (Dynamic Special Nuclear Materials Control and Accountability System). Both of these systems appeared applicable, however, other automated devices already employed for production control made implementing the bar code and OCR seem improbable. However, the DYMCAS was placed on line for nuclear material accountability, a decision was made to consider the bar code for physical inventory listings. For the past several months a development program has been underway to use a bar code device to collect and input data to the DYMCAS on the uranium recovery operations. Programs have been completed and tested, and are being employed to ensure that data will be compatible and useful. Bar code implementation and expansion of its use for all nuclear material inventory activity in Y-12 is presented

  5. Nuclear industry will be short of engineers

    International Nuclear Information System (INIS)

    Yates, M.

    1990-01-01

    This article discusses the potential shortage of nuclear engineers due to reduction of educational and training facilities and difficulty in attracting minorities into nuclear engineering. The article reports on recommendations from the National Research Council Nuclear Education Study Committee on attracting minorities to nuclear engineering, increasing DOE fellowships, funding for research and development, involvement of utilities and vendors, and support of the American Nuclear Society's advocacy of nuclear engineering education

  6. Inherently safe nuclear-driven internal combustion engines

    International Nuclear Information System (INIS)

    Alesso, P.; Chow, Tze-Show; Condit, R.; Heidrich, J.; Pettibone, J.; Streit, R.

    1991-01-01

    A family of nuclear driven engines is described in which nuclear energy released by fissioning of uranium or plutonium in a prompt critical assembly is used to heat a working gas. Engine performance is modeled using a code that calculates hydrodynamics, fission energy production, and neutron transport self-consistently. Results are given demonstrating a large negative temperature coefficient that produces self-shutoff of energy production. Reduced fission product inventory and the self-shutoff provide inherent nuclear safety. It is expected that nuclear engine reactor units could be scaled from 100 MW on up. 7 refs., 3 figs

  7. Development of Nuclear Energy Security Code

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  8. Accreditation of nuclear engineering programs

    International Nuclear Information System (INIS)

    Williamson, T.G.

    1989-01-01

    The American Nuclear Society (ANS) Professional Development and Accreditation Committee (PDAC) has the responsibility for accreditation of engineering and technology programs for nuclear and similarly named programs. This committee provides society liaison with the Accreditation Board for Engineering and Technology (ABET), is responsible for the appointment and training of accreditation visitors, nomination of members for the ABET Board and Accreditation Commissions, and review of the criteria for accreditation of nuclear-related programs. The committee is composed of 21 members representing academia and industry. The ABET consists of 19 participating bodies, primarily professional societies, and 4 affiliate bodies. Representation on ABET is determined by the size of the professional society and the number of programs accredited. The ANS, as a participating body, has one member on the ABET board, two members on the Engineering Accreditation Commission, and one on the Technology Accreditation Commission. The ABET board sets ABET policy and the commissions are responsible for accreditation visits

  9. On the automated assessment of nuclear reactor systems code accuracy

    International Nuclear Information System (INIS)

    Kunz, Robert F.; Kasmala, Gerald F.; Mahaffy, John H.; Murray, Christopher J.

    2002-01-01

    An automated code assessment program (ACAP) has been developed to provide quantitative comparisons between nuclear reactor systems (NRS) code results and experimental measurements. The tool provides a suite of metrics for quality of fit to specific data sets, and the means to produce one or more figures of merit (FOM) for a code, based on weighted averages of results from the batch execution of a large number of code-experiment and code-code data comparisons. Accordingly, this tool has the potential to significantly streamline the verification and validation (V and V) processes in NRS code development environments which are characterized by rapidly evolving software, many contributing developers and a large and growing body of validation data. In this paper, a survey of data conditioning and analysis techniques is summarized which focuses on their relevance to NRS code accuracy assessment. A number of methods are considered for their applicability to the automated assessment of the accuracy of NRS code simulations. A variety of data types and computational modeling methods are considered from a spectrum of mathematical and engineering disciplines. The goal of the survey was to identify needs, issues and techniques to be considered in the development of an automated code assessment procedure, to be used in United States Nuclear Regulatory Commission (NRC) advanced thermal-hydraulic T/H code consolidation efforts. The ACAP software was designed based in large measure on the findings of this survey. An overview of this tool is summarized and several NRS data applications are provided. The paper is organized as follows: The motivation for this work is first provided by background discussion that summarizes the relevance of this subject matter to the nuclear reactor industry. Next, the spectrum of NRS data types are classified into categories, in order to provide a basis for assessing individual comparison methods. Then, a summary of the survey is provided, where each

  10. The DIT nuclear fuel assembly physics design code

    International Nuclear Information System (INIS)

    Jonsson, A.

    1988-01-01

    The DIT code is the Combustion Engineering, Inc. (C-E) nuclear fuel assembly design code. It belongs to a class of codes, all similar in structure and strategy, that may be characterized by the spectrum and spatial calculations being performed in two dimensions and in a single job step for the entire assembly. The forerunner of this class of codes is the United Kingdom Atomic Energy Authority WIMS code, the first version of which was completed 25 yr ago. The structure and strategy of assembly spectrum codes have remained remarkably similar to the original concept thus proving its usefulness. As other organizations, including C-E, have developed their own versions of the concept, many important variations have been added that significantly influence the accuracy and performance of the resulting computational tool. Those features, which are unique to the DIT code and which might be of interest to the community of fuel assembly physics design code users and developers, are described and discussed

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

    International Nuclear Information System (INIS)

    Hill, Ralph S. III

    2002-01-01

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

  12. Training in nuclear engineering companies

    International Nuclear Information System (INIS)

    Perezagua, R. L.

    2013-01-01

    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)

  13. Matching grant program for university nuclear engineering education

    International Nuclear Information System (INIS)

    Bajorek, Stephen M.

    2002-01-01

    The grant augmented funds from Westinghouse Electric Co. to enhance the Nuclear Engineering program at KSU. The program was designed to provide educational opportunities and to train engineers for careers in the nuclear industry. It provided funding and access to Westinghouse proprietary design codes for graduate and undergraduate studies on topics of current industrial importance. Students had the opportunity to use some of the most advanced nuclear design tools in the industry and to work on actual design problems. The WCOBRA/TRAC code was used to simulate loss of coolant accidents (LOCAs)

  14. The future of nuclear engineering

    International Nuclear Information System (INIS)

    Beeden, Jeffrey

    2003-01-01

    Today, nuclear power refers to the splitting of large uranium atoms into smaller atoms with a net release of energy. Tomorrow, nuclear power will refer to the combining of hydrogen into larger atoms with a net release of energy. Nuclear power's future is fusion. The Mechanical Engineers of tomorrow will need to be familiar with the process of creating and harnessing the energy from a fusion reaction. During the oil shortage in the 1970's, America scrambled to initiate alternative methods of producing power. Nuclear fusion was one of them. As time passes, the solution to the world's energy crisis presses the countries of the world to find alternative forms of energy; nuclear fusion may contain the answer. In the near future, the field of fusion will open up and a new wave of engineers will flood into this field. Mechanical engineers will lead the way with advances in materials, computational fluid dynamics, finite element analysis for thermal and structural systems, and heat transfer designs to optimize nuclear fusion reactors and power plants. All this effort is in anticipation of creating a sustained fusion reaction that can generate enough heat to transfer to steam in order to generate electric power to sustain the fusion reaction and introduce power to the grid. (author)

  15. The mathematics of nuclear engineering

    International Nuclear Information System (INIS)

    Lewins, J.D.

    1982-01-01

    The mathematics of nuclear engineering is considered with especial reference to the problems of; the representation of the transformation of matter at the nuclear level by radioactive decay and neutron transmutation, the problem of the distribution of neutrons and other particles as a transport theory problem including some of the approximation methods used in this problem, particularly diffusion theory with particular emphasis on steady-state problems, time-dependent reactor kinetic and control, and the longer term changes involved with the nuclear fuel cycle both within and without the reactor itself. (U.K.)

  16. Introduction to nuclear test engineering

    International Nuclear Information System (INIS)

    O'Neal, W.C.; Paquette, D.L.

    1982-01-01

    The basic information in this report is from a vu-graph presentation prepared to acquaint new or prospective employees with the Nuclear Test Engineering Division (NTED). Additional information has been added here to enhance a reader's understanding when reviewing the material after hearing the presentation, or in lieu of attending a presentation

  17. Nuclear engineering terms and definitions

    International Nuclear Information System (INIS)

    1981-01-01

    The most important nuclear engineering's terms and definitions are given in this standard. The definitions take into account the Austrian Regulations for Radiation Protection, for and pertinent ISO and DIN-Standards as also the OENORM A7006 and OENORM A6601. (M.T.)

  18. Code on the safety of nuclear power plants: Quality assurance

    International Nuclear Information System (INIS)

    1988-01-01

    This revised Code provides the principles and objectives for the establishment and implementation of quality assurance programmes applied to both the overall and each of the constituent activities associated with a nuclear power plant project. The quality assurance principles enumerated in the present Code can be usefully applied to nuclear facilities other than nuclear power plants. The quality assurance programme encompasses: (1) the activities that are necessary to achieve the appropriate quality of the respective item or service; and (2) the activities that are necessary for verifying that the required quality is achieved and that objective evidence is produced to that effect. Quality assurance is an essential aspect of good management and the quality assurance programme is the main management tool for a disciplined approach to all activities affecting quality, including, where appropriate, verification that each task has been satisfactorily performed and that necessary corrective actions have been implemented. The principles and objectives provided by the Code are applicable by all those responsible for the nuclear power plant, by plant designers, suppliers, architect-engineers, plant constructors, plant operators and other organizations participating in activities affecting quality. The Code is a revision of the previous Code of Practice (1978) on the same subject of interest to regulatory bodies and experts in quality assurance for design, siting and operation of nuclear power plants. Contents: Definitions; 1. Introduction; 2. Quality assurance programmes; 3. Organization; 4. Document control; 5. Design control; 6. Procurement control; 7. Control of items; 8. Process control; 9. Inspection and test control; 10. Non-conformance control; 11. Corrective actions; 12, Records; 13. Audits

  19. Waste management in the nuclear engineering curriculum

    International Nuclear Information System (INIS)

    Tulenko, J.S.

    1989-01-01

    One of the most significant challenges facing the nuclear industry is to successfully close the nuclear fuel cycle and effectively demonstrate to the public that nuclear wastes do not present a health risk. This issue is currently viewed by many as the most important issue affecting public acceptance of nuclear power, and it is imperative that nuclear engineers be able to effectively address the question of nuclear waste from both a generation and disposal standpoint. To address the issue, the area of nuclear waste management has been made one of the fields of specialized study in the Department of Nuclear Engineering Sciences at the University of Florida. The study of radioactive waste management at the University of Florida is designed both for background for the general nuclear engineering student and for those wishing to specialize in it as a multidiscipline study area involving the Departments of Nuclear Engineering Sciences, Environmental Sciences, Material Science and Engineering, Geology, Civil Engineering, and Industrial Engineering

  20. DOZIM - evaluation dose code for nuclear accident

    International Nuclear Information System (INIS)

    Oprea, I.; Musat, D.; Ionita, I.

    2008-01-01

    During a nuclear accident an environmentally significant fission products release can happen. In that case it is not possible to determine precisely the air fission products concentration and, consequently, the estimated doses will be affected by certain errors. The stringent requirement to cope with a nuclear accident, even minor, imposes creation of a computation method for emergency dosimetric evaluations needed to compare the measurement data to certain reference levels, previously established. These comparisons will allow a qualified option regarding the necessary actions to diminish the accident effects. DOZIM code estimates the soil contamination and the irradiation doses produced either by radioactive plume or by soil contamination. Irradiations either on whole body or on certain organs, as well as internal contamination doses produced by isotope inhalation during radioactive plume crossing are taken into account. The calculus does not consider neither the internal contamination produced by contaminated food consumption, or that produced by radioactive deposits resuspension. The code is recommended for dose computation on the wind direction, at distances from 10 2 to 2 x 10 4 m. The DOZIM code was utilized for three different cases: - In air TRIGA-SSR fuel bundle destruction with different input data for fission products fractions released into the environment; - Chernobyl-like accident doses estimation; - Intervention areas determination for a hypothetical severe accident at Cernavoda Nuclear Power Plant. For the first case input data and results (for a 60 m emission height without iodine retention on active coal filters) are presented. To summarize, the DOZIM code conception allows the dose estimation for any nuclear accident. Fission products inventory, released fractions, emission conditions, atmospherical and geographical parameters are the input data. Dosimetric factors are included in the program. The program is in FORTRAN IV language and was run on

  1. The Dit nuclear fuel assembly physics design code

    International Nuclear Information System (INIS)

    Jonsson, A.

    1987-01-01

    DIT is the Combustion Engineering, Inc. (C-E) nuclear fuel assembly design code. It belongs to a class of codes, all similar in structure and strategy, which may be characterized by the spectrum and spatial calculations being performed in 2D and in a single job step for the entire assembly. The forerunner of this class of codes is the U.K.A.E.A. WIMS code, the first version of which was completed 25 years ago. The structure and strategy of assembly spectrum codes have remained remarkably similar to the original concept thus proving its usefulness. As other organizations, including C-E, have developed their own versions of the concept, many important variations have been added which significantly influence the accuracy and performance of the resulting computational tool. This paper describes and discusses those features which are unique to the DIT code and which might be of interest to the community of fuel assembly physics design code users and developers

  2. Career Development in Nuclear Engineering

    International Nuclear Information System (INIS)

    Sibbens, G.

    2015-01-01

    In the eighties it was not common for girls to study engineering. But a few young girls have always been fascinated by science and technical applications and dared to go for a gender untypical education. What are these female engineers doing today? This paper describes the career development of a woman, who completed her Master of Science in Nuclear Engineering, found first a job in an international company as cooperator in the research group of radiation physics and later as head of technical support and quality assurance of medical systems and then succeeded in a competition to be recruited at the European Commission (EC). There she started as an assistant for the primary standardisation of radionuclides and high-resolution alpha-particle spectrometry including the preparation of radioactive sources in the radionuclide metrology sector at the Institute for Reference Materials and Measurements of the European Commission’s Joint Research Centre and consequently published her work in scientific journals. Today, 29 years later, I am the laboratory responsible for the preparation and characterisation of nuclear targets at EC-JRC-IRMM, leading a team that has unique know-how in the preparation of thin film deposits (called targets) tailor-made for nuclear physics measurements at the EC–JRC–IRMM and international accelerator sites. High quality measurements of nuclear data and materials are being asked for in the context of nuclear safety, minimisation of high level nuclear waste and safeguards and security. The different steps of my career development and the repeated process of managing learning, work, family and leisure are presented. The career path across different jobs and responsibilities and the career progress via a certification training programme are also explained to encourage the next generation of female professionals to continue playing a vital role in nuclear science and technology. (author)

  3. 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

  4. Nuclear Engineering Academic Programs Survey, 2004

    International Nuclear Information System (INIS)

    Oak Ridge Institute for Science and Education

    2005-01-01

    This annual report details the number of nuclear engineering bachelor's, master's, and doctoral degrees awarded at a sampling of academic programs from 1998-2004. It also looks at nuclear engineering degrees by curriculum and the number of students enrolled in nuclear engineering degree programs at 31 U.S. universities in 2004

  5. BS degree in nuclear engineering or a nuclear option

    International Nuclear Information System (INIS)

    Williams on, T.G.

    1988-01-01

    Many nuclear engineering educators are concerned about the health of nuclear engineering academic departments. As part of a review of the BS nuclear engineering degree program at the University of Virginia, the authors surveyed several local utilities with operating nuclear plants about their needs for nuclear engineering graduates. The perception of many of the utility executives about a nuclear engineering degree and about a nuclear option in another engineering curriculum does not agree with the way the authors view these two degrees. The responses to two of the survey questions were of particular interest: (1) does your company have a preference between nuclear engineering graduates and graduates in other fields with a nuclear option? (2) what do you consider to be a minimum level of education in nuclear engineering for a nuclear option in mechanical engineering? All of the four utilities that were surveyed stated a preference for mechanical or electrical engineers with a nuclear option, although two indicated that there are certain jobs for which a nuclear engineering graduate is desired

  6. Computational intelligence in nuclear engineering

    International Nuclear Information System (INIS)

    Uhrig, Robert E.; Hines, J. Wesley

    2005-01-01

    Approaches to several recent issues in the operation of nuclear power plants using computational intelligence are discussed. These issues include 1) noise analysis techniques, 2) on-line monitoring and sensor validation, 3) regularization of ill-posed surveillance and diagnostic measurements, 4) transient identification, 5) artificial intelligence-based core monitoring and diagnostic system, 6) continuous efficiency improvement of nuclear power plants, and 7) autonomous anticipatory control and intelligent-agents. Several Changes to the focus of Computational Intelligence in Nuclear Engineering have occurred in the past few years. With earlier activities focusing on the development of condition monitoring and diagnostic techniques for current nuclear power plants, recent activities have focused on the implementation of those methods and the development of methods for next generation plants and space reactors. These advanced techniques are expected to become increasingly important as current generation nuclear power plants have their licenses extended to 60 years and next generation reactors are being designed to operate for extended fuel cycles (up to 25 years), with less operator oversight, and especially for nuclear plants operating in severe environments such as space or ice-bound locations

  7. Reactor physics computations for nuclear engineering undergraduates

    International Nuclear Information System (INIS)

    Huria, H.C.

    1989-01-01

    The undergraduate program in nuclear engineering at the University of Cincinnati provides three-quarters of nuclear reactor theory that concentrate on physical principles, with calculations limited to those that can be conveniently completed on programmable calculators. An additional one-quarter course is designed to introduce the student to realistic core physics calculational methods, which necessarily requires a computer. Such calculations can be conveniently demonstrated and completed with the modern microcomputer. The one-quarter reactor computations course includes a one-group, one-dimensional diffusion code to introduce the concepts of inner and outer iterations, a cell spectrum code based on integral transport theory to generate cell-homogenized few-group cross sections, and a multigroup diffusion code to determine multiplication factors and power distributions in one-dimensional systems. Problem assignments include the determination of multiplication factors and flux distributions for typical pressurized water reactor (PWR) cores under various operating conditions, such as cold clean, hot clean, hot clean at full power, hot full power with xenon and samarium, and a boron concentration search. Moderator and Doppler coefficients can also be evaluated and examined

  8. Regulatory Endorsement Activities for ASME Nuclear Codes and Standards

    International Nuclear Information System (INIS)

    West, Raymond A.

    2006-01-01

    The ASME Board on Nuclear Codes and Standards (BNCS) has formed a Task Group on Regulatory Endorsement (TG-RE) that is currently in discussions with the United States Nuclear Regulatory Commission (NRC) to look at suggestions and recommendations that can be used to help with the endorsement of new and revised ASME Nuclear Codes and Standards (NC and S). With the coming of new reactors in the USA in the very near future we need to look at both the regulations and all the ASME NC and S to determine where we need to make changes to support these new plants. At the same time it is important that we maintain our operating plants while addressing ageing management needs of our existing reactors. This is going to take new thinking, time, resources, and money. For all this to take place the regulations and requirements that we use must be clear concise and necessary for safety and to that end both the NRC and ASME are working together to make this happen. Because of the influence that the USA has in the world in dealing with these issues, this paper is written to inform the international nuclear engineering community about the issues and what actions are being addressed under this effort. (author)

  9. Nuclear engineering enrollments and degrees, 1994: Appendixes

    International Nuclear Information System (INIS)

    1995-05-01

    This survey is designed to include those programs sponsored by the Department of Energy. The survey is designed to include those programs offering a major in nuclear engineering or course work equivalent to a major in other engineering disciplines that prepare the graduates to perform as nuclear engineers. This survey provides data on nuclear engineering enrollments and degrees for use in labor market analyses, information on education programs for students, and information on new graduates to employers, government agencies, academia and professional societies

  10. Engineering and science education for nuclear power

    International Nuclear Information System (INIS)

    1986-01-01

    The Guidebook contains detailed information on curricula which would provide the professional technical education qualifications which have been established for nuclear power programme personnel. The core of the Guidebook consists of model curricula in engineering and science, including relevant practical work. Curricula are provided for specialization, undergraduate, and postgraduate programmes in nuclear-oriented mechanical, chemical, electrical, and electronics engineering, as well as nuclear engineering and radiation health physics. Basic nuclear science and engineering laboratory work is presented together with a list of basic experiments and the nuclear equipment needed to perform them. Useful measures for implementing and improving engineering and science education and training capabilities for nuclear power personnel are presented. Valuable information on the national experiences of IAEA Member States in engineering and science education for nuclear power, as well as examples of such education from various Member States, have been included

  11. Unique nuclear thermal rocket engine

    International Nuclear Information System (INIS)

    Culver, D.W.; Rochow, R.

    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

  12. Reconstruction of nuclear engineering education in universities

    International Nuclear Information System (INIS)

    Kitamura, Masaharu; Tomota, Yo; Tanaka, Shunichi

    2005-01-01

    Nuclear engineering has become the area gradually loosing appeal to the young for these twenty years taking all the circumstances into consideration. However nuclear power is predicted to be primary energy of greatest importance even in the future and this needs highly motivated and excellent personnel in nuclear industry and society so as to develop and maintain nuclear power to a high degree. Under these circumstances discussions on how should be nuclear engineering research and education in the new era were presented from various viewpoints and they led to the direction of reconstruction of nuclear engineering education in universities and relevant organizations to train and ensure personnel. (T. Tanaka)

  13. Code on the safety of nuclear power plants: Siting

    International Nuclear Information System (INIS)

    1988-01-01

    This Code provides criteria and procedures that are recommended for safety in nuclear power plant siting. It forms part of the Agency's programme for establishing Codes and Safety Guides relating to land based stationary thermal neutron power plants

  14. Nuclear reactions in Monte Carlo codes

    CERN Document Server

    Ferrari, Alfredo

    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. (43 refs) .

  15. Code on the safety of civilian nuclear fuel cycle installations

    International Nuclear Information System (INIS)

    1996-01-01

    The 'Code' was promulgated by the National Nuclear Safety Administration (NSSA) on June 17, 1993, which is applicable to civilian nuclear fuel fabrication, processing, storage and reprocessing installations, not including the safety requirements for the use of nuclear fuel in reactors. The contents of the 'Code' involve siting, design, construction, commissioning, operation and decommissioning of fuel cycle installation. The NNSA shall be responsible for the interpretation of this 'Code'

  16. Code development for nuclear reactor simulation

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  17. 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.

  18. US Nuclear Engineering Education: Status and prospects

    International Nuclear Information System (INIS)

    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, as described in this report 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 ageing 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., 24 figs., 49 tabs

  19. The University of Utah Nuclear Engineering Program

    International Nuclear Information System (INIS)

    Jevremovic, T.; McDonald, L. IV; Schow, R.

    2016-01-01

    As of 2014, the University of Utah Nuclear Engineering Program (UNEP) manages and maintains over 7,000 ft 2 (~650 m 2 ) nuclear engineering facilities that includes 100 kW TRIGA Mark I and numerous laboratories such as radiochemistry, microscopy, nuclear forensics, nuclear medicine, radiation detection and instrumentation laboratories. The UNEP offers prestigious educational and training programs in the field of faculty reserach: reactor physics, reactor design and operation, advanced numerical modeling and visualizations in radiation transport, radiochemistry, nuclear forensics, radiation detection and detector designs, signal processing, nuclear medicine, nuclear space and nuclear robotic’s engineering and radiological sciences. With the state-of-the-art nuclear instrumentation and state-of-the-art numerical modeling tools, reserach reactor and modernized educational and training programs, we positioned ourselves in the last five years as the fastest growing national nuclear engineering program attracting the students from many disciplines such as but not limited to: chemical engineering, civil engineering, environmental engineering, chemistry, physics, astronomy, medical sciences, and others. From 2012, we uniquely developed and implemented the nuclear power plants’ safety culture paradigm that we use for day-to-day operation, management and maintenance of our facilities, as well as train all our students at undergraduate and graduate levels of studies. We developed also a new distance-learning approaches in sharing knowledge about experiential learning based on no-cost internet-tools combined with the use of mobile technologies. (author)

  20. A nuclear power plant system engineering workstation

    International Nuclear Information System (INIS)

    Mason, J.H.; Crosby, J.W.

    1989-01-01

    System engineers offer an approach for effective technical support for operation and maintenance of nuclear power plants. System engineer groups are being set up by most utilities in the United States. Institute of Nuclear Power operations (INPO) and U.S. Nuclear Regulatory Commission (NRC) have endorsed the concept. The INPO Good Practice and a survey of system engineer programs in the southeastern United States provide descriptions of system engineering programs. The purpose of this paper is to describe a process for developing a design for a department-level information network of workstations for system engineering groups. The process includes the following: (1) application of a formal information engineering methodology, (2) analysis of system engineer functions and activities; (3) use of Electric Power Research Institute (EPRI) Plant Information Network (PIN) data; (4) application of the Information Engineering Workbench. The resulting design for this system engineer workstation can provide a reference for design of plant-specific systems

  1. Integrated engineering system for nuclear facilities building

    International Nuclear Information System (INIS)

    Tomura, H.; Miyamoto, A.; Futami, F.; Yasuda, S.; Ohtomo, T.

    1995-01-01

    In the construction of buildings for nuclear facilities in Japan, construction companies are generally in charge of the building engineering work, coordinating with plant engineering. An integrated system for buildings (PROMOTE: PROductive MOdeling system for Total nuclear Engineering) described here is a building engineering system including the entire life cycle of buildings for nuclear facilities. A Three-dimensional (3D) building model (PRO-model) is to be in the core of the system (PROMOTE). Data sharing in the PROMOTE is also done with plant engineering systems. By providing these basic technical foundations, PROMOTE is oriented toward offering rational, highquality engineering for the projects. The aim of the system is to provide a technical foundation in building engineering. This paper discusses the characteristics of buildings for nuclear facilities and the outline of the PROMOTE. (author)

  2. European master degree in nuclear engineering

    International Nuclear Information System (INIS)

    Ghitescu, Petre; Prisecaru, Ilie

    2003-01-01

    In order to preserve and to improve the quality of nuclear engineering education and training in Europe, as well to ensure the safe and economic operation of nuclear power plants, the European Nuclear Engineering Network Program (ENEN) started in 2002. It is a program aiming to establish and maintain a set of criteria for specific curricula of nuclear engineering education, in particular, for an European Master Degree in Nuclear Engineering (EMNE). The ENEN program is financed by the FP5 and has the wide support of IAEA, OECD and EU Commission departments dealing with the nuclear engineering knowledge management. The promising results up to now determined the creation of the Asian Nuclear Engineering Network (ANEN) in July 2003 and of the World Nuclear University (WNU) starting in September 2003. The paper presents the future structure of EMNE which will allow the harmonization of the curricula of the universities of Europe until the Bologna Convention will be fully accepted and operational in all European countries. The ENEN program has taken into consideration the curricula of 22 universities and research centres from 15 different European countries and proposed a feasible scheme which allows the undergraduates with a weak to strong nuclear background to continue their graduate education in the nuclear engineering field towards EMNE. As one of the contractors of this program, University 'Politehnica' of Bucharest brings its contribution and actively takes part in all activities establishing the EMNE. (author)

  3. A Semantic Analysis Method for Scientific and Engineering Code

    Science.gov (United States)

    Stewart, Mark E. M.

    1998-01-01

    This paper develops a procedure to statically analyze aspects of the meaning or semantics of scientific and engineering code. The analysis involves adding semantic declarations to a user's code and parsing this semantic knowledge with the original code using multiple expert parsers. These semantic parsers are designed to recognize formulae in different disciplines including physical and mathematical formulae and geometrical position in a numerical scheme. In practice, a user would submit code with semantic declarations of primitive variables to the analysis procedure, and its semantic parsers would automatically recognize and document some static, semantic concepts and locate some program semantic errors. A prototype implementation of this analysis procedure is demonstrated. Further, the relationship between the fundamental algebraic manipulations of equations and the parsing of expressions is explained. This ability to locate some semantic errors and document semantic concepts in scientific and engineering code should reduce the time, risk, and effort of developing and using these codes.

  4. Academic nuclear engineering education - the Dutch way

    International Nuclear Information System (INIS)

    Wallerbos, E.J.M.; Geemert, R. van

    1997-01-01

    The academic nuclear engineering educational program in the Netherlands aims not only to give students a thorough knowledge of reactor physics but also to train them in practical skills and presentation techniques. These three aspects are important to become a successful nuclear engineer. (author)

  5. Compilation of the nuclear codes available in CTA

    International Nuclear Information System (INIS)

    D'Oliveira, A.B.; Moura Neto, C. de; Amorim, E.S. do; Ferreira, W.J.

    1979-07-01

    The present work is a compilation of some nuclear codes available in the Divisao de Estudos Avancados of the Instituto de Atividades Espaciais, (EAV/IAE/CTA). The codes are organized as the classification given by the Argonne National Laboratory. In each code are given: author, institution of origin, abstract, programming language and existent bibliography. (Author) [pt

  6. Establishment of computer code system for nuclear reactor design - analysis

    International Nuclear Information System (INIS)

    Subki, I.R.; Santoso, B.; Syaukat, A.; Lee, S.M.

    1996-01-01

    Establishment of computer code system for nuclear reactor design analysis is given in this paper. This establishment is an effort to provide the capability in running various codes from nuclear data to reactor design and promote the capability for nuclear reactor design analysis particularly from neutronics and safety points. This establishment is also an effort to enhance the coordination of nuclear codes application and development existing in various research centre in Indonesia. Very prospective results have been obtained with the help of IAEA technical assistance. (author). 6 refs, 1 fig., 1 tab

  7. Development of the next generation code system as an engineering modeling language (1)

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Uto, Nariaki; Kasahara, Naoto; Nagura, Fuminori; Ishikawa, Makoto; Ohira, Masanori; Kato, Masayuki

    2002-11-01

    In the fast reactor development, numerical simulation using analytical codes plays an important role for complementing theory and experiment. It is necessary that the engineering models and analysis methods can be flexibly changed, because the phenamine to be investigated become more complicated due to the diversity of the needs for research. And, there are large problems in combining physical properties and engineering models in many different fields. In this study, the goal is to develop a flexible and general-purposive analysis system, in which the physical properties and engineering models are represented as a programming language or a diagrams that are easily understandable for humans and executable for computers. The authors named this concept the Engineering Modeling Language (EML). This report describes the result of the investigation for latest computer technologies and software development techniques which seem to be usable for a realization of the analysis code system for nuclear engineering as an EML. (author)

  8. Lifecycle management for nuclear engineering project documents

    International Nuclear Information System (INIS)

    Zhang Li; Zhang Ming; Zhang Ling

    2010-01-01

    The nuclear engineering project documents with great quantity and various types of data, in which the relationships of each document are complex, the edition of document update frequently, are managed difficultly. While the safety of project even the nuclear safety is threatened seriously by the false documents and mistakes. In order to ensure the integrality, veracity and validity of project documents, the lifecycle theory of document is applied to build documents center, record center, structure and database of document lifecycle management system. And the lifecycle management is used to the documents of nuclear engineering projects from the production to pigeonhole, to satisfy the quality requirement of nuclear engineering projects. (authors)

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

    International Nuclear Information System (INIS)

    Aleshin, Sergey S.; Gorodkov, Sergey S.; Shcherenko, Anna I.

    2015-01-01

    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.

  10. Thermal hydraulics in undergraduate nuclear engineering education

    International Nuclear Information System (INIS)

    Theofanous, T.G.

    1986-01-01

    The intense safety-related research efforts of the seventies in reactor thermal hydraulics have brought about the recognition of the subject as one of the cornerstones of nuclear engineering. Many nuclear engineering departments responded by building up research programs in this area, and mostly as a consequence, educational programs, too. Whether thermal hydraulics has fully permeated the conscience of nuclear engineering, however, remains yet to be seen. The lean years that lie immediately ahead will provide the test. The purpose of this presentation is to discuss the author's own educational activity in undergraduate nuclear engineering education over the past 10 yr or so. All this activity took place at Purdue's School of Nuclear Engineering. He was well satisfied with the results and expects to implement something similar at the University of California in Santa Barbara in the near future

  11. Nuclear industry - challenges in chemical engineering

    International Nuclear Information System (INIS)

    Sen, S.; Sunder Rajan, N.S.; Balu, K.; Garg, R.K.; Murthy, L.G.K.; Ramani, M.P.S.; Rao, M.K.; Sadhukhan, H.K.; Venkat Raj, V.

    1978-01-01

    As chemical engineering processes and operations are closely involved in many areas of nuclear industry, the chemical engineer has a vital role to play in its growth and development. An account of the major achievements of the Indian chemical engineers in this field is given with view of impressing upon the faculty members of the Indian universities the need for taking appropriate steps to prepare chemical engineers suitable for nuclear industry. Some of the major achievements of the Indian chemical engineers in this field are : (1) separation of useful minerals from beach sand, (2) preparation of thorium nitrate of nuclear purity from monazite, (3) processing of zircon sand to obtain nuclear grade zirconium and its separation from hafnium to obtain zirconium metal sponge, (4) recovery of uranium from copper tailings, (5) economic recovery of nuclear grade uranium from low grade uranium ores found in India, (6) fuel reprocessing, (7) chemical processing of both low and high level radioactive wastes. (M.G.B.)

  12. Pressure vessel codes: Their application to nuclear reactor systems

    International Nuclear Information System (INIS)

    1966-01-01

    A survey has been made by the International Atomic Energy Agency of how the problems of applying national pressure vessel codes to nuclear reactor systems have been treated in those Member States that had pressurized reactors in operation or under construction at the beginning of 1963. Fifteen answers received to an official inquiry form the basis of this report, which also takes into account some recently published material. Although the answers to the inquiry in some cases data back to 1963 and also reflect the difficulty of describing local situations in answer to standard questions, it is hoped that the report will be of interest to reactor engineers. 21 refs, 1 fig., 2 tabs

  13. ASME nuclear codes and standards: Scope of coverage and current initiatives

    International Nuclear Information System (INIS)

    Eisenberg, G. M.

    1995-01-01

    The objective of this paper is to address the broad scope of coverage of nuclear codes, standards and guides produced and administered by the American Society of Mechanical Engineers (ASME). Background information is provided regarding the evolution of the present activities. Details are provided on current initiatives intended to permit ASME to meet the needs of a changing nuclear industry on a worldwide scale. During the early years of commercial nuclear power, ASME produced a code for the construction of nuclear vessels used in the reactor coolant pressure boundary, containment and auxiliary systems. In response to industry growth, ASME Code coverage soon broadened to include rules for construction of other nuclear components, and inservice inspection of nuclear reactor coolant systems. In the years following this, the scope of ASME nuclear codes, standards and guides has been broadened significantly to include air cleaning activities for nuclear power reactors, operation and maintenance of nuclear power plants, quality assurance programs, cranes for nuclear facilities, qualification of mechanical equipment, and concrete reactor vessels and containments. ASME focuses on globalization of its codes, standards and guides by encouraging and promoting their use in the international community and by actively seeking participation of international members on its technical and supervisory committees and in accreditation activities. Details are provided on current international representation. Initiatives are underway to separate the technical requirements from administrative and enforcement requirements, to convert to hard metric units, to provide for non-U. S. materials, and to provide for translations into non-English languages. ASME activity as an accredited ISO 9000 registrar for suppliers of mechanical equipment is described. Rules are being developed for construction of containment systems for nuclear spent fuel and high-level waste transport packagings. Intensive

  14. Bar code instrumentation for nuclear safeguards

    International Nuclear Information System (INIS)

    Bieber, A.M. Jr.

    1984-01-01

    This paper presents a brief overview of the basic principles of bar codes and the equipment used to make and to read bar code labels, and a summary of some of the more important factors that need to be considered in integrating bar codes into an information system

  15. Current status of nuclear engineering education

    International Nuclear Information System (INIS)

    Palladino, N.J.

    1975-01-01

    The 65 colleges and universities offering undergraduate degrees in nuclear engineering and the 15 schools offering strong nuclear engineering options are, in general, doing a good job to meet the current spectrum of job opportunities. But, nuclear engineering programs are not producing enough graduates to meet growing demands. They currently receive little aid and support from their customers --industry and government--in the form of scholarships, grants, faculty research support, student thesis and project support, or student summer jobs. There is not enough interaction between industry and universities. Most nuclear engineering programs are geared too closely to the technology of the present family of reactors and too little to the future breeder reactors and controlled thermonuclear reactors. In addition, nuclear engineering programs attract too few women and members of minority ethnic groups. Further study of the reasons for this fact is needed so that effective corrective action can be taken. Faculty in nuclear engineering programs should assume greater initiative to provide attractive and objective nuclear energy electives for technical and nontechnical students in other disciplines to improve their technical understanding of the safety and environmental issues involved. More aggressive and persistent efforts must be made by nuclear engineering schools to obtain industry support and involvement in their programs

  16. Harmonization of nuclear codes and standards, pacific nuclear council working and task group report

    International Nuclear Information System (INIS)

    Dua, S.S.

    2006-01-01

    Full text: The codes and standards, both at the national and international level, have had a major impact on the industry worldwide and served it well in maintaining the performance and safety of the nuclear reactors and facilities. The codes and standards, in general, are consensus documents and do seek public input at various levels before they are finalized and rolled out for use by the nuclear vendors, consultants, utilities and regulatory bodies. However, the extensive development of prescriptive national standards if unchecked against the global environment and trade agreements (NAFTA, WTO, etc.) can also become barriers and cause difficulties to compete in the world market. During the last decade, the national and international writing standards writing bodies have recognized these issues and are moving more towards the rationalization and harmonization of their standards with the more widely accepted generic standards. The Pacific Nuclear Council (PNC) recognized the need for harmonization of the nuclear codes and standards for its member countries and formed a Task Group to achieve its objectives. The Task Group has a number of members from the PNC member countries. In 2005 PNC further raised the importance of this activity and formed a Working Group to cover a broader scope. The Working Group (WG) mandate is to identify and analyze the different codes and standards introduced to the Pacific Basin region, in order to achieve mutual understanding, harmonization and application in each country. This o requires the WG to develop and encourage the use of reasonably consistent criteria for the design and development, engineering, procurement, fabrication, construction, testing, operations, maintenance, waste management, decommissioning and the management of the commercial nuclear power plants in the Pacific Basin so as to: Promote consistent safety, quality, environmental and management standards for nuclear energy and other peaceful applications of nuclear

  17. New Zealand code for nuclear powered shipping

    International Nuclear Information System (INIS)

    1976-06-01

    This report recommends guidelines for the safety precautions and procedures to be implemented when New Zealand ports and approaches are used by nuclear powered merchant ships and nuclear powered naval ships

  18. Performance testing of thermal analysis codes for nuclear fuel casks

    International Nuclear Information System (INIS)

    Sanchez, L.C.

    1987-01-01

    In 1982 Sandia National Laboratories held the First Industry/Government Joint Thermal and Structural Codes Information Exchange and presented the initial stages of an investigation of thermal analysis computer codes for use in the design of nuclear fuel shipping casks. The objective of the investigation was to (1) document publicly available computer codes, (2) assess code capabilities as determined from their user's manuals, and (3) assess code performance on cask-like model problems. Computer codes are required to handle the thermal phenomena of conduction, convection and radiation. Several of the available thermal computer codes were tested on a set of model problems to assess performance on cask-like problems. Solutions obtained with the computer codes for steady-state thermal analysis were in good agreement and the solutions for transient thermal analysis differed slightly among the computer codes due to modeling differences

  19. Gaseous core nuclear-driven engines featuring a self-shutoff mechanism to provide nuclear safety

    International Nuclear Information System (INIS)

    Heidrich, J.; Pettibone, J.; Chow, Tze-Show; Condit, R.; Zimmerman, G.

    1991-11-01

    Nuclear driven engines are described that could be run in either pulsed or steady state modes. In the pulsed mode nuclear energy is released by fissioning of uranium or plutonium in a supercritical assembly of fuel and working gas. In a steady state mode a fuel-gas mixture is injected into a magnetic nozzle where it is compressed into a critical state and produces energy. Engine performance is modeled using a code that calculates hydrodynamics, fission energy production, and neutron transport self-consistently. Results are given demonstrating a large negative temperature coefficient that produces self-shutoff or control of energy production. Reduced fission product inventory and the self-shutoff provide inherent nuclear safety. It is expected that nuclear engine reactor units could be scaled up from about 100 MW e

  20. Software engineers and nuclear engineers: teaming up to do testing

    International Nuclear Information System (INIS)

    Kelly, D.; Cote, N.; Shepard, T.

    2007-01-01

    The software engineering community has traditionally paid little attention to the specific needs of engineers and scientists who develop their own software. Recently there has been increased recognition that specific software engineering techniques need to be found for this group of developers. In this case study, a software engineering group teamed with a nuclear engineering group to develop a software testing strategy. This work examines the types of testing that proved to be useful and examines what each discipline brings to the table to improve the quality of the software product. (author)

  1. Nuclear reactor engineering: Reactor systems engineering. Fourth edition, Volume Two

    International Nuclear Information System (INIS)

    Glasstone, S.; Sesonske, A.

    1994-01-01

    This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in the design and operation of nuclear power plants. Extensively updated, the fourth edition includes new materials on reactor safety and risk analysis, regulation, fuel management, waste management and operational aspects of nuclear power. This volume contains the following: the systems concept, design decisions, and information tools; energy transport; reactor fuel management and energy cost considerations; environmental effects of nuclear power and waste management; nuclear reactor safety and regulation; power reactor systems; plant operations; and advanced plants and the future

  2. Study of nuclear computer code maintenance and management system

    International Nuclear Information System (INIS)

    Ryu, Chang Mo; Kim, Yeon Seung; Eom, Heung Seop; Lee, Jong Bok; Kim, Ho Joon; Choi, Young Gil; Kim, Ko Ryeo

    1989-01-01

    Software maintenance is one of the most important problems since late 1970's.We wish to develop a nuclear computer code system to maintenance and manage KAERI's nuclear software. As a part of this system, we have developed three code management programs for use on CYBER and PC systems. They are used in systematic management of computer code in KAERI. The first program is embodied on the CYBER system to rapidly provide information on nuclear codes to the users. The second and the third programs were embodied on the PC system for the code manager and for the management of data in korean language, respectively. In the requirement analysis, we defined each code, magnetic tape, manual and abstract information data. In the conceptual design, we designed retrieval, update, and output functions. In the implementation design, we described the technical considerations of database programs, utilities, and directions for the use of databases. As a result of this research, we compiled the status of nuclear computer codes which belonged KAERI until September, 1988. Thus, by using these three database programs, we could provide the nuclear computer code information to the users more rapidly. (Author)

  3. The MCEF code for nuclear evaporation and fission calculations

    International Nuclear Information System (INIS)

    Deppman, A.; Pina, S.R. de; Likhachev, V.P.; Mesa, J.; Arruda-Neto, J.D.T.; Rodriguez, O.; Goncalves, M.

    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)

  4. Code on the safety of nuclear power plants: Design

    International Nuclear Information System (INIS)

    1988-01-01

    This Code is a compilation of nuclear safety principles aimed at defining the essential requirements necessary to ensure nuclear safety. These requirements are applicable to structures, systems and components, and procedures important to safety in nuclear power plants embodying thermal neutron reactors, with emphasis on what safety requirements shall be met rather than on specifying how these requirements can be met. It forms part of the Agency's programme for establishing Codes and Safety Guides relating to land based stationary thermal neutron power plants. The document should be used by organizations designing, manufacturing, constructing and operating nuclear power plants as well as by regulatory bodies

  5. Nuclear engineering in the National Polytechnic Institute

    International Nuclear Information System (INIS)

    Del Valle G, E.

    2008-12-01

    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

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

    International Nuclear Information System (INIS)

    Nemoto, Toshiyuki; Kawai, Wataru; Ishizuki, Shigeru; Kawasaki, Nobuo; Kume, Etsuo; Adachi, Masaaki; Ogasawara, Shinobu

    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)

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

    International Nuclear Information System (INIS)

    Ishizuki, Shigeru; Kawai, Wataru; Nemoto, Toshiyuki; Ogasawara, Shinobu; Kume, Etsuo; Adachi, Masaaki; Kawasaki, Nobuo; Yatake, Yo-ichi

    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 vectorization and parallelization on vector processors. In this 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 (n-particle, Temperature and Velocity) 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. In the porting part, the porting of Monte Carlo N-Particle Transport code MCNP4B2 and Reactor Safety Analysis code RELAP5 on the AP3000 are described. (author)

  8. Vectorization of nuclear codes 90-1

    International Nuclear Information System (INIS)

    Nonomiya, Iwao; Nemoto, Toshiyuki; Ishiguro, Misako; Harada, Hiroo; Hori, Takeo.

    1990-09-01

    The vectorization has been made for four codes: SONATINA-2V HTTR version, TRIDOSE, VIENUS, and SCRYU. SONATINA-2V HTTR version is a code for analyzing the dynamic behavior of fuel blocks in the vertical slice of the HTGR (High Temperature Gas-cooled Reactor) core under seismic perturbation, TRIDOSE is a code for calculating environmental tritium concentration and dose, VIENUS is a code for analyzing visco elastic stress of the fuel block of HTTR (High Temperature gas-cooled Test Reactor), and SCRYU is a thermal-hydraulics code with boundary fitted coordinate system. The total speedup ratio of the vectorized versions to the original scalar ones is 5.2 for SONATINA-2V HTTR version. 5.9 ∼ 6.9 for TRIDOSE, 6.7 for VIENUS, 7.6 for SCRYU, respectively. In this report, we describe outline of codes, techniques used for the vectorization, verification of computed results, and speedup effect on the vectorized codes. (author)

  9. Towards the European Nuclear Engineering Education Network

    International Nuclear Information System (INIS)

    Mavko, B.; Giot, M.; Sehgal, B.R.; Goethem, G. Van

    2003-01-01

    Current priorities of the scientific community regarding basic research lie elsewhere than in nuclear sciences. The situation today is significantly different than it was three to four decades ago when much of the present competence base in nuclear sciences was in fact generated. In addition, many of the highly competent engineers and scientists, who helped create the present nuclear industry, and its regulatory structure, are approaching retirement. To preserve nuclear knowledge and expertise through the higher nuclear engineering education in the 5 th framework program of the European Commission the project ENEN (European Nuclear Engineering Education Network) was launched, since the need to keep the university curricula in nuclear sciences and technology alive has been clearly recognized at European level. As the follow up of this project an international nuclear engineering education consortium of universities with partners from the nuclear sector is presently in process of being established This association called ENEN has as founding members: 14 universities and 8 research institutes from 17 European countries. (author)

  10. SIMCRI: a simple computer code for calculating nuclear criticality parameters

    International Nuclear Information System (INIS)

    Nakamaru, Shou-ichi; Sugawara, Nobuhiko; Naito, Yoshitaka; Katakura, Jun-ichi; Okuno, Hiroshi.

    1986-03-01

    This is a user's manual for a simple criticality calculation code SIMCRI. The code has been developed to facilitate criticality calculation on a single unit of nuclear fuel. SIMCRI makes an extensive survey with a little computing time. Cross section library MGCL for SIMCRI is the same one for the Monte Carlo criticality code KENOIV; it is, therefore, easy to compare the results of the two codes. SIMCRI solves eigenvalue problems and fixed source problems based on the one space point B 1 equation. The results include infinite and effective multiplication factor, critical buckling, migration area, diffusion coefficient and so on. SIMCRI is comprised in the criticality safety evaluation code system JACS. (author)

  11. Nuclear industry prepares fore shortage of engineers

    International Nuclear Information System (INIS)

    Gauker, Lynn.

    1991-01-01

    It is predicted that the Canadian nuclear industry will experience a shortage of qualified personnel within the next five to ten years. The reasons for this prediction are as follows: enrollment in engineering courses, particularly five courses in nuclear engineering has been declining; immigration can no longer be expected to fill the gap; the workforce is aging. Solutions may include promotional campaigns, student employment programs, and educating workers to a professional level

  12. Education of nuclear engineering in Japan

    International Nuclear Information System (INIS)

    Ozawa, Yasutomo; Yamamuro, Nobuhiro

    1979-01-01

    The research Committee of Nuclear Engineering Education has two working groups. One group has carried out surveyes on the curriculums of nuclear engineering course of universities in Japan and the activities of graduates in the industrial worlds. The other group conducted an investigation on the present status of energy education in senior high schools. This is an interim report on the activity of the research committee. (author)

  13. 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.

  14. Nuclear model codes available at the Nuclear Energy Agency Computer Program Library (NEA-CPL)

    International Nuclear Information System (INIS)

    Sartori, E.; Garcia Viedma, L. de

    1976-01-01

    This paper briefly outlines the objectives of the NEA-CPL and its activities in the field of Nuclear Model Computer Codes. A short description of the computer codes available from the CPL in this field is also presented. (author)

  15. Applicability of vector processing to large-scale nuclear codes

    International Nuclear Information System (INIS)

    Ishiguro, Misako; Harada, Hiroo; Matsuura, Toshihiko; Okuda, Motoi; Ohta, Fumio; Umeya, Makoto.

    1982-03-01

    To meet the growing trend of computational requirements in JAERI, introduction of a high-speed computer with vector processing faculty (a vector processor) is desirable in the near future. To make effective use of a vector processor, appropriate optimization of nuclear codes to pipelined-vector architecture is vital, which will pose new problems concerning code development and maintenance. In this report, vector processing efficiency is assessed with respect to large-scale nuclear codes by examining the following items: 1) The present feature of computational load in JAERI is analyzed by compiling the computer utilization statistics. 2) Vector processing efficiency is estimated for the ten heavily-used nuclear codes by analyzing their dynamic behaviors run on a scalar machine. 3) Vector processing efficiency is measured for the other five nuclear codes by using the current vector processors, FACOM 230-75 APU and CRAY-1. 4) Effectiveness of applying a high-speed vector processor to nuclear codes is evaluated by taking account of the characteristics in JAERI jobs. Problems of vector processors are also discussed from the view points of code performance and ease of use. (author)

  16. Nuclear engineering education initiative at Ibaraki University

    International Nuclear Information System (INIS)

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

    2015-01-01

    With the help of a grant from the Ministry of Education, Culture, Sports, Science and Technology, Ibaraki University has been engaging for six years in the development and preparation of educational environment on nuclear engineering for each of graduate and undergraduate. Core faculty conducts general services including the design and implementation of curriculum, operational improvement, and implementation of lectures. 'Beginner-friendly introduction for nuclear power education' is provided at the Faculty of Engineering, and 'nuclear engineering education program' at the Graduate School of Science and Engineering. All the students who have interest or concern in the accidents at nuclear power plants or the future of nuclear power engineering have opportunities to learn actively. This university participates in the alliance or association with other universities, builds industry - government - academia cooperation with neighboring institutions such as the Japan Atomic Energy Agency, and makes efforts to promote the learning and development of applied skills related to nuclear engineering through training and study tours at each facility. For example, it established the Frontier Applied Atomic Science Center to analyze the structure and function of materials using the strong neutron source of J-PARC. As the efforts after the earthquake accident, it carried out a radiation survey work in Fukushima Prefecture. In addition, it proposed and practiced the projects such as 'development of methods for the evaluation of transfer/fixation properties and decontamination of radioactive substances,' and 'structure analysis of radioactive substances remaining in soil, litter, and polluted water and its application to the decontamination.' (A.O.)

  17. Nuclear operations summary Engineering organization for Plowshare nuclear operations

    Energy Technology Data Exchange (ETDEWEB)

    Broadman, Gene A [Lawrence Radiation Laboratory, University of California, Livermore, CA (United States)

    1970-05-15

    The availability of nuclear explosives for peaceful projects has given the engineer a new dimension in his thinking. He can now seek methods of adapting Plowshare to a variety of industrial applications. The full potential of the Plowshare Program can only be attained when industry begins to use nuclear explosives on a regular basis, for economically sound projects. It is the purpose of this paper to help the engineer familiarize himself with Plowshare technology to hasten the day when 'Plowsharee goes commercial'. An engineering project utilizing nuclear exposives ordinarily involves three main phases: Phase I (a) The theoretical and empirical analysis of effects. (b) Projected economic and/or scientific evaluation. (c) A safety analysis. Phase II (a) Field construction. (b) Safe detonation of the nuclear explosive. (c) Data acquisition. Phase III The evaluation and/or exploitation of the results. This paper will be restricted to Phase II, referred to collectively as the 'nuclear operation'.

  18. Code on the safety of nuclear power plants: Governmental organization

    International Nuclear Information System (INIS)

    1988-01-01

    This Code recommends requirements for a regulatory body responsible for regulating the siting, design, construction, commissioning, operation and decommissioning of nuclear power plants for safety. It forms part of the Agency's programme for establishing Codes and Safety Guides relating to land based stationary thermal neutron power plants

  19. The applicability of ALPHA/PHOENIX/ANC nuclear design code system on Korean standard PWR's

    International Nuclear Information System (INIS)

    Lee, Kookjong; Choi, Kie-Yong; Lee, Hae-Chan; Roh, Eun-Rae

    1996-01-01

    For the Korean Standard Nuclear Power Plant (KSNPP) designed based on Combustion Engineering (CE) System 80, the Westinghouse nuclear design code system ALPHA/PHOENIX/ANC was applied to the follow-up design of initial and reload core of KSNPP. The follow-up design results of Yonggwang Unit 3 Cycle 1, 2 and Yonggwang Unit 4 Cycle 1 have shown good agreements with the measured data. The assemblywise power distributions have shown less than 2% average differences and critical boron concentrations have shown less than 20 ppm differences. All the low power physics test parameters are in good agreement. Consequently, APA design code system can be applied to KNSPP cores. (author)

  20. ASME nuclear codes and standards: Recent technical initiatives

    International Nuclear Information System (INIS)

    Feigel, R. E.

    1995-01-01

    Although nuclear power construction is currently in a hiatus in the US, ASME and its volunteer committees remain committed to continual improvements in the technical requirements in its nuclear codes. This paper provides an overview of several significant recent revisions to ASME' s nuclear codes. Additionally, other important initiatives currently being addressed by ASME committees will be described. With the largest population of operating light water nuclear plants in the world and worldwide use of its nuclear codes, ASME continues to support technical advancements in its nuclear codes and standards. While revisions of various magnitude are an ongoing process, several recent revisions embody significant changes based on state of the art design philosophy and substantial industry experience. In the design area, a significant revisions has recently been approved which will significantly reduce conservatisms in seismic piping design as well as provide simplified design rules. Major revisions have also been made to the requirements for nuclear material manufacturers and suppliers, which should result in clearer understanding of this difficult administrative area of the code. In the area of Section XI inservice rules, substantial studies are underway to investigate the application of probabilistic, risked based inspection in lieu of the current deterministic inspection philosophy. While much work still is required in this area, it is an important potential application of the emerging field of risk based inspection

  1. Engineering application of in-core fuel management optimization code with CSA algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Zhihong; Hu, Yongming [INET, Tsinghua university, Beijing 100084 (China)

    2009-06-15

    PWR in-core loading (reloading) pattern optimization is a complex combined problem. An excellent fuel management optimization code can greatly improve the efficiency of core reloading design, and bring economic and safety benefits. Today many optimization codes with experiences or searching algorithms (such as SA, GA, ANN, ACO) have been developed, while how to improve their searching efficiency and engineering usability still needs further research. CSA (Characteristic Statistic Algorithm) is a global optimization algorithm with high efficiency developed by our team. The performance of CSA has been proved on many problems (such as Traveling Salesman Problems). The idea of CSA is to induce searching direction by the statistic distribution of characteristic values. This algorithm is quite suitable for fuel management optimization. Optimization code with CSA has been developed and was used on many core models. The research in this paper is to improve the engineering usability of CSA code according to all the actual engineering requirements. Many new improvements have been completed in this code, such as: 1. Considering the asymmetry of burn-up in one assembly, the rotation of each assembly is considered as new optimization variables in this code. 2. Worth of control rods must satisfy the given constraint, so some relative modifications are added into optimization code. 3. To deal with the combination of alternate cycles, multi-cycle optimization is considered in this code. 4. To confirm the accuracy of optimization results, many identifications of the physics calculation module in this code have been done, and the parameters of optimization schemes are checked by SCIENCE code. The improved optimization code with CSA has been used on Qinshan nuclear plant of China. The reloading of cycle 7, 8, 9 (12 months, no burnable poisons) and the 18 months equilibrium cycle (with burnable poisons) reloading are optimized. At last, many optimized schemes are found by CSA code

  2. Nuclear Plant Analyzer development at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Laats, E.T.

    1986-10-01

    The Nuclear Plant Analyzer (NPA) is a state-of-the-art safety analysis and engineering tool being used to address key nuclear power plant safety issues. Under the sponsorship of the US Nuclear Regulatory Commission (NRC), the NPA has been developed to integrate the NRC's computerized reactor behavior simulation codes such as RELAP5, TRAC-BWR and TRAC-PWR, with well-developed computer color graphics programs and large repositories of reactor design and experimental data. An important feature of the NPA is the capability to allow an analyst to redirect a RELAP5 or TRAC calculation as it progresses through its simulated scenario. The analyst can have the same power plant control capabilities as the operator of an actual plant. The NPA resides on the dual Control Data Corporation Cyber 176 mainframe computers at the Idaho National Engineering Laboratory and Cray-1S computers at the Los Alamos National Laboratory (LANL) and Kirtland Air Force Weapons Laboratory (KAFWL)

  3. Engineering development in nuclear power plant construction

    International Nuclear Information System (INIS)

    Guenther, P.

    1979-01-01

    Proceeding from the up-to-now experience in the erection of nuclear power stations, especially of the first and second unit of the Greifswald nuclear power plant, the following essential aspects of the development of constructional engineering are discussed: (1) constructional features and criteria, (2) organizational management, (3) current status and problems in prelimary operations, and (4) possibilities of further expenditure reductions in constructing nuclear power stations

  4. Nuclear engineering education in the United States

    International Nuclear Information System (INIS)

    Williamson, T.G.

    1982-01-01

    In discussing nuclear engineering education in the United States it is shown that the most critical issue facing the nuclear engineering education community today is enrolment in a time of increasing demand for graduate engineers. Related to the issue of enrolment is support for graduate students, whether it be fellowships, traineeships, or research assistantships. Other issues are those of maintaining a vital faculty in the face of a competitive job market, of maintaining research facilities and developing new ones, and of determining the directions of educational efforts in the future. (U.K.)

  5. The engineering function in Scottish Nuclear

    International Nuclear Information System (INIS)

    Grant, J.

    1991-01-01

    The work of the Engineering and Development Division of Scottish Nuclear is described in this article. This company, formed since the privatization of electricity generation in the United Kingdom, owns and operates the two Hunterston Magnox reactors and the Torness Advanced Gass Cooled Reactors. Principle responsibilities such as maintaining safety standards, formulating policy for radioactive waste disposal and decommissioning and optimally controlling the nuclear generation cycle are outlined. Objectives for the next five years are identified and explained separately. The experience, knowledge and expertise of engineering staff is stressed as being of key importance to the future success of Scottish Nuclear. (UK)

  6. Use of the algebraic coding theory in nuclear electronics

    International Nuclear Information System (INIS)

    Nikityuk, N.M.

    1990-01-01

    New results of studies of the development and use of the syndrome coding method in nuclear electronics are described. Two aspects of using the syndrome coding method are considered for sequential coding devices and for the creation of fast parallel data compression devices. Specific examples of the creation of time-to-digital converters based on circular counters are described. Several time intervals can be coded very fast and with a high resolution by means of these converters. The effective coding matrix which can be used for light signal coding. The rule of constructing such coding matrices for arbitrary number of channels and multiplicity n is given. The methods for solving ambiguities in silicon detectors and for creating the special-purpose processors for high-energy spectrometers are given. 21 refs.; 9 figs.; 3 tabs

  7. 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.)

  8. Mobile filters in nuclear engineering

    International Nuclear Information System (INIS)

    Meuter, R.

    1979-01-01

    The need for filters with high efficiencies which may be used at any place originated in nuclear power plants. Filters of this type, called Filtermobil, have been developed by Sulzer. They have been used successfully in nuclear plants for several years. (orig.) [de

  9. Proceedings of the Scientific Meeting in Nuclear Instrumentation Engineering

    International Nuclear Information System (INIS)

    Achmad Suntoro; Rony Djokorayono; Ferry Sujatno; Utaja

    2010-11-01

    The Proceeding of the Scientific Meeting in Nuclear Instrumentation Engineering held on Nov, 30, 2010 by the Centre for Nuclear Instrumentation Engineering - National Nuclear Energy Agency. The Proceedings of the Scientific Contains 40 papers Consist of Nuclear Instrumentation Engineering for Industry, Environment, and Nuclear Facilities. (PPIKSN)

  10. 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...

  11. ASME nuclear codes and standards risk management strategic plan

    International Nuclear Information System (INIS)

    Balkey, Kenneth R.

    2003-01-01

    Over the past 15 years, several risk-informed initiatives have been completed or are under development within the ASME Nuclear Codes and Standards organization. In order to better manage the numerous initiatives in the future, the ASME Board on Nuclear Codes and Standards has recently developed and approved a Risk Management Strategic Plan. This paper presents the latest approved version of the plan beginning with a background of applications completed to date, including the recent issuance of the ASME Standard for Probabilistic Risk Assessment (PRA) for Nuclear Power Plant Applications. The paper discusses potential applications within ASME Nuclear Codes and Standards that may require expansion of the PRA Standard, such as for new generation reactors, or the development of new PRA Standards. A long-term vision for the potential development and evolution to a nuclear systems code that adopts a risk-informed approach across a facility life-cycle (design, construction, operation, maintenance, and closure) is summarized. Finally, near term and long term actions are defined across the ASME Nuclear Codes and Standards organizations related to risk management, and related U.S. regulatory activities are also summarized. (author)

  12. ASME nuclear codes and standards risk management strategic planning

    International Nuclear Information System (INIS)

    Hill, Ralph S. III; Balkey, Kenneth R.; Erler, Bryan A.; Wesley Rowley, C.

    2007-01-01

    This paper is prepared in honor and in memory of the late Professor Emeritus Yasuhide Asada to recognize his contributions to ASME Nuclear Codes and Standards initiatives, particularly those related to risk-informed technology and System Based Code developments. For nearly two decades, numerous risk-informed initiatives have been completed or are under development within the ASME Nuclear Codes and Standards organization. In order to properly manage the numerous initiatives currently underway or planned for the future, the ASME Board on Nuclear Codes and Standards (BNCS) has an established Risk Management Strategic Plan (Plan) that is maintained and updated by the ASME BNCS Risk Management Task Group. This paper presents the latest approved version of the plan beginning with a background of applications completed to date, including the recent probabilistic risk assessment (PRA) standards developments for nuclear power plant applications. The paper discusses planned applications within ASME Nuclear Codes and Standards that will require expansion of the ASME PRA Standard to support new advanced light water reactor and next generation reactor developments, such as for high temperature gas-cooled reactors. Emerging regulatory developments related to risk-informed, performance- based approaches are summarized. A long-term vision for the potential development and evolution to a nuclear systems code that adopts a risk-informed approach across a facility life-cycle (design, construction, operation, maintenance, and closure) is also summarized. Finally, near term and long term actions are defined across the ASME Nuclear Codes and Standards organizations related to risk management, including related U.S. regulatory activities. (author)

  13. Radiation hazards of nuclear engineering

    International Nuclear Information System (INIS)

    Oster, H.

    1981-01-01

    The basic mechanisms and principles of nuclear power plants are discussed, since their knowledge is mandatory for the understanding of the true risk associated with nuclear technology. Differences between predictable and catastrophic accidents are compared, terms which have been frequently confused to the extent that the public has become unjustifiably and irresponsibly alarmed. A description of the jobs and their responsibilities is also given. Known accidents are reported and the role of the physician in the care of accidents and the scheduling of emergency situations is described. Finally, the usefullness, necessity and risk associated with nuclear power are discussed. (orig.) [de

  14. Teaching WWERs at Hacettepe University Nuclear Engineering Department in Turkey

    International Nuclear Information System (INIS)

    Ergun, S.

    2011-01-01

    In this study, the challenges faced in the teaching WWER design for the reactor engineering course, which is taught in the Hcettepe University Nuclear Engineering Department are discussed. Since the course is designated taking a western reactor design into account, the computer programs and class projects prepared for the course include models and correlations suitable for these designs. The attempts for modifying the course and developing codes or programs for the course become a challenge especially in finding proper information sources on design in English. From finding proper material properties to exploring the design ideas, teaching WWER designs and using analysis tools for better teaching are very important to modify the reactor engineering course. With the study presented here, the reactor engineering course taught is described, the teaching tools are listed and attempts of modifying the course to teach and analyze WWER designs are explained

  15. Electronic manual of the nuclear characteristics analysis code-set for FBR

    International Nuclear Information System (INIS)

    Makino, Tohru

    2001-03-01

    Reactor Physics Gr., System Engineering Technology Division, O-arai Engineering Center has consolidated the nuclear design database to improve analytical methods and prediction accuracy for large fast breeder cores such as demonstration or commercial FBRs from the previous research. The up-to-date information about usage of the nuclear characteristics analysis code-set was compiled as a part of the improvement of basic design data base for FBR core. The outlines of the electronic manual are as follows; (1) The electronic manual includes explanations of following codes: JOINT : Code Interface Program. SLAROM, CASUP : Effective Cross Section Calculation Code. CITATION-FBR : Diffusion Analysis Code. PERKY : Perturbative Diffusion Analysis Code. SNPERT, SNPERT-3D : Perturbative Transport Analysis Code. SAGEP, SAGEP-3D : Sensitivity Coefficient Calculation Code. NSHEX : Transport Analysis Code using Nodal Method. ABLE : Cross Section Adjustment Calculation Code. ACCEPT : Predicting Accuracy Evaluation Code. (2) The electronic manual is described using HTML file format and PDF file for easy maintenance, updating and for easy referring through JNC Intranet. User can refer manual pages by usual Web browser software without any special setup. (3) Many of manual pages include link-tags to jump to related pages. String search is available in both HTML and PDF documents. (4) User can download source code, sample input data and shell script files to carry out each analysis from download page of each code (JNC inside only). (5) Usage of the electronic manual and maintenance/updating process are described in this report and it makes possible to enroll new code or new information in the electronic manual. Since the information has been taken into account about modifications and error fixings, added to each code after the last consolidation in 1994, the electronic manual would cover most recent status of the nuclear characteristics analysis code-set. One of other advantages of use

  16. Colloid transport code-nuclear user's manual

    International Nuclear Information System (INIS)

    Jain, R.

    1992-01-01

    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 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

  17. JAERI Nuclear Engineering School and technology transfer

    International Nuclear Information System (INIS)

    Nishimura, Kazuaki; Kawaguchi, Chiyoji

    1978-01-01

    A method is introduced to evaluate the degree of nuclear technology transfer; that is, the output powers of Japanese nuclear reactors constructed in these 20 years are chronologically plotted in a semi-log figure. All reactors plotted are classified into imported and domestic ones according to a value of domestication factor. A space between two historical trajectories of reactor construction may be interpreted as one of the measures indicating the degree of nuclear technology transfer. In connection with this method, historical change of educational and training courses in Nuclear Engineering School of Japan Atomic Energy Research Institute is reviewed in this report. (author)

  18. A nuclear engineer's ethical responsibility to society

    International Nuclear Information System (INIS)

    Kemeny, L.G.

    1989-01-01

    Chernobyl notwithstanding, this paper seeks to illustrate why, on numerous fronts, nuclear technology provides the safest, cleanest and most effective method of base-load power generation. In particular it seeks to demonstrate that, despite the strident rhetoric and media exposure given to the anti-nuclear lobby, the technology is fundamental to the quality of life and the equitable sharing of energy by the year 2000. Therefore, the safety and technological superiority of the nuclear fuel cycle together with its high technology peripheral benefits both societal and fiscal are viewed as an ever increasing challenge and motivation which constitutes a major part of the nuclear engineer's ethical responsibility to society

  19. Civil engineering challenge with nuclear waste

    International Nuclear Information System (INIS)

    Day, D.

    1985-01-01

    The civil engineer can help to solve the problems in disposing of nuclear waste in a deep geologic formation. The site for a nuclear waste repository must be carefully selected so that the geology provides the natural barrier between the waste and the accessible environment specified by the NRC and the EPA. This engineer is familiar with the needed structure and conditions of the host and surrounding rocks, and also the hydraulic mechanisms for limiting the migration of water in the rocks. To dispose of the nuclear waste underground requires stable and long-lasting shafts and tunnels such as civil engineers have designed and constructed for many other uses. The planning, design and construction of the ground surface facilities for a nuclear waste repository involves civil engineering in many ways. The transporation of heavy, metal shielded casks requires special attention to the system of highways and railroads accessing the repository. Structures for handling the shipping casks and transferring the waste onsite and into the deep geologic formation need special considerations. The structures must provide the NRC required containment, including hot cells for remote handling. Therefore, structural design strives for buildings, ventilation structures, shaft headframes, etc., to be earthquake and tornado-proof. These important design bases and considerations for the civil engineer working on a nuclear waste repository are discussed in this paper

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

    International Nuclear Information System (INIS)

    Schoenfelder, Christian

    2014-01-01

    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.

  1. 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.

  2. Engineering and science education for nuclear power

    International Nuclear Information System (INIS)

    Mautner-Markhof, F.

    1988-01-01

    Experience has shown that one of the critical conditions for the successful introduction of a nuclear power programme is the availability of sufficient numbers of personnel having the required education and experience qualifications. For this reason, the introduction of nuclear power should be preceded by a thorough assessment of the relevant capabilities of the industrial and education/training infrastructures of the country involved. The IAEA assists its Member States in a variety of ways in the development of infrastructures and capabilities for engineering and science education for nuclear power. Types of assistance provided by the IAEA to Member States include: Providing information in connection with the establishment or upgrading of academic and non-academic engineering and science education programmes for nuclear power (on the basis of curricula recommended in the Agency's Guidebook on engineering and science education for nuclear power); Expert assistance in setting up or upgrading laboratories and other teaching facilities; Assessing the capabilities and interest of Member States and their institutions/organizations for technical co-operation among countries, especially developing ones, in engineering and science education, as well as its feasibility and usefulness; Preparing and conducting nuclear specialization courses (e.g. on radiation protection) in various Member States

  3. Fire-safety engineering and performance-based codes

    DEFF Research Database (Denmark)

    Sørensen, Lars Schiøtt

    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......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...... thermodynamics • Heat exchange during the fire process • Skin burns • Burning rate, energy release rate and design fires • Proposal to Risk-based design fires • Proposal to a Fire scale • Material ignition and flame spread • Fire dynamics in buildings • Combustion products and toxic gases • Smoke inhalation...

  4. Model-Driven Engineering of Machine Executable Code

    Science.gov (United States)

    Eichberg, Michael; Monperrus, Martin; Kloppenburg, Sven; Mezini, Mira

    Implementing static analyses of machine-level executable code is labor intensive and complex. We show how to leverage model-driven engineering to facilitate the design and implementation of programs doing static analyses. Further, we report on important lessons learned on the benefits and drawbacks while using the following technologies: using the Scala programming language as target of code generation, using XML-Schema to express a metamodel, and using XSLT to implement (a) transformations and (b) a lint like tool. Finally, we report on the use of Prolog for writing model transformations.

  5. Operating nuclear plant feedback to ASME and French codes

    International Nuclear Information System (INIS)

    Journet, J.; O'Donnell, W.J.

    1996-01-01

    The French have an advantage in nuclear plant operating experience feedback due to the highly centralized nature of their nuclear industry. There is only one utility in charge of design as well as operations (EDF) and only one reactor vendor (Framatome). The ASME Code has played a key role in resolving technical issues in the design and operation of nuclear plants since the inception of nuclear power. The committee structure of the Code brings an ideal combination of senior technical people with both broad and specialized experience to bear on complex how safe is safe enough technical issues. The authors now see an even greater role for the ASME Code in a proposed new regulatory era for the US nuclear industry. The current legalistic confrontational regulatory era has been quite destructive. There now appears to be a real opportunity to begin a new era of technical consensus as the primary means for resolving safety issues. This change can quickly be brought about by having the industry take operating plant problems and regulatory technical issues directly to the ASME Code for timely resolution. Surprisingly, there is no institution in the US nuclear industry with such a mandate. In fact, the industry is organized to feedback through the Nuclear Regulatory Commission issues which could be far better resolved through the ASME Code. Major regulatory benefits can be achieved by closing this loop and providing systematic interaction with the ASME Code. The essential elements of a new regulatory era and ideas for organizing US institutional industry responsibilities, taken from the French experience, are described in this paper

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  7. 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.

  8. The computer code system for reactor radiation shielding in design of nuclear power plant

    International Nuclear Information System (INIS)

    Li Chunhuai; Fu Shouxin; Liu Guilian

    1995-01-01

    The computer code system used in reactor radiation shielding design of nuclear power plant includes the source term codes, discrete ordinate transport codes, Monte Carlo and Albedo Monte Carlo codes, kernel integration codes, optimization code, temperature field code, skyshine code, coupling calculation codes and some processing codes for data libraries. This computer code system has more satisfactory variety of codes and complete sets of data library. It is widely used in reactor radiation shielding design and safety analysis of nuclear power plant and other nuclear facilities

  9. The changing face of nuclear engineering education

    International Nuclear Information System (INIS)

    Poston, J.W.

    1991-01-01

    Nuclear engineering education in the US is in a near-crisis situation. Most academic programs are small with limited enrollments and faculty. Some of these programs are being absorbed into larger academic units, while others are being terminated. The number of identifiable academic programs has dropped dramatically over the last several years, and there is genuine concern that this downward trend will continue. The recent report by the National Academy of Sciences highlights the problems, needs, and prospects for nuclear engineering education in this country. At the same time, some programs appear to be relatively healthy and somewhat secure. A closer look at these programs indicates that there has been an evolution in the approach taken by these survivors toward both their academic and research programs. This paper discusses the approaches taken at Texas A and M University over the last 8 to 10 years to strengthen the Department of Nuclear Engineering

  10. Development of REFLA/TRAC code for engineering work station

    International Nuclear Information System (INIS)

    Ohnuki, Akira; Akimoto, Hajime; Murao, Yoshio

    1994-03-01

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

  11. Application of software engineering to development of reactor safety codes

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  12. 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)

  13. Safety in nuclear power plant siting. A code of practice

    International Nuclear Information System (INIS)

    1978-01-01

    This publication is brought out within the framework of establishing Codes of Practice and Safety Guides for nuclear power plants: NUSS programme. The scope of the document encompasses site and site-plant interaction factors related to operational states and accident conditions. The purpose of the Code is to give criteria and procedures to be applied as appropriate to operational states and accident conditions, including those which could lead to emergency situations. This Code is mainly concerned with severe events of low probability which relate to the siting of nuclear power plants and have to be considered in designing a particular nuclear power plant. Annex: Examples of natural and man-made events relevant for design basis evaluation

  14. European Master of Science in Nuclear Engineering

    International Nuclear Information System (INIS)

    Moons, F.; Safieh, J.; Giot, M.; Mavko, B.; Sehgal, B.R.; Schaefer, A.; Goethem, G. van; D'haeseleer, W.

    2004-01-01

    The need to preserve, enhance or strengthen nuclear knowledge is worldwide recognised since a couple of years. It appears that within the European university education and training network, nuclear engineering is presently sufficiently covered, although somewhat fragmented. To take up the challenges of offering top quality, new, attractive and relevant curricula, higher education institutions should cooperate with industry, regulatory bodies and research centres, and more appropriate funding a.o. from public and private is to be re-established. More, European nuclear education and training should benefit from links with international organisations like IAEA, OECD-NEA and others, and should include world-wide cooperation with academic institutions and research centres. The European master in nuclear engineering guarantees a high quality nuclear education in Europe by means of stimulating student and instructor exchange, through mutual checks of the quality of the programmes offered, by close collaboration with renowned nuclear-research groups at universities and laboratories. The concept for a nuclear master programme consists of a solid basket of recommended basic nuclear science and engineering courses, but also contains advanced courses as well as practical training. Some of the advanced courses also serve as part of the curricula for doctoral programmes. A second important issue identified is Continued Professional Development. In order to achieve the objectives and practical goals described above, the ENEN association was formed. This international, non-profit association is be considered as a step towards a virtual European Nuclear University symbolising the active collaboration between various national institutions pursuing nuclear education. (author)

  15. Progress on China nuclear data processing code system

    Science.gov (United States)

    Liu, Ping; Wu, Xiaofei; Ge, Zhigang; Li, Songyang; Wu, Haicheng; Wen, Lili; Wang, Wenming; Zhang, Huanyu

    2017-09-01

    China is developing the nuclear data processing code Ruler, which can be used for producing multi-group cross sections and related quantities from evaluated nuclear data in the ENDF format [1]. The Ruler includes modules for reconstructing cross sections in all energy range, generating Doppler-broadened cross sections for given temperature, producing effective self-shielded cross sections in unresolved energy range, calculating scattering cross sections in thermal energy range, generating group cross sections and matrices, preparing WIMS-D format data files for the reactor physics code WIMS-D [2]. Programming language of the Ruler is Fortran-90. The Ruler is tested for 32-bit computers with Windows-XP and Linux operating systems. The verification of Ruler has been performed by comparison with calculation results obtained by the NJOY99 [3] processing code. The validation of Ruler has been performed by using WIMSD5B code.

  16. Modern Nuclear Data Evaluation with the TALYS Code System

    Science.gov (United States)

    Koning, A. J.; Rochman, D.

    2012-12-01

    This paper presents a general overview of nuclear data evaluation and its applications as developed at NRG, Petten. Based on concepts such as robustness, reproducibility and automation, modern calculation tools are exploited to produce original nuclear data libraries that meet the current demands on quality and completeness. This requires a system which comprises differential measurements, theory development, nuclear model codes, resonance analysis, evaluation, ENDF formatting, data processing and integral validation in one integrated approach. Software, built around the TALYS code, will be presented in which all these essential nuclear data components are seamlessly integrated. Besides the quality of the basic data and its extensive format testing, a second goal lies in the diversity of processing for different type of users. The implications of this scheme are unprecedented. The most important are: 1. Complete ENDF-6 nuclear data files, in the form of the TENDL library, including covariance matrices, for many isotopes, particles, energies, reaction channels and derived quantities. All isotopic data files are mutually consistent and are supposed to rival those of the major world libraries. 2. More exact uncertainty propagation from basic nuclear physics to applied (reactor) calculations based on a Monte Carlo approach: "Total" Monte Carlo (TMC), using random nuclear data libraries. 3. Automatic optimization in the form of systematic feedback from integral measurements back to the basic data. This method of work also opens a new way of approaching the analysis of nuclear applications, with consequences in both applied nuclear physics and safety of nuclear installations, and several examples are given here. This applied experience and feedback is integrated in a final step to improve the quality of the nuclear data, to change the users vision and finally to orchestrate their integration into simulation codes.

  17. Modern Nuclear Data Evaluation with the TALYS Code System

    International Nuclear Information System (INIS)

    Koning, A.J.; Rochman, D.

    2012-01-01

    This paper presents a general overview of nuclear data evaluation and its applications as developed at NRG, Petten. Based on concepts such as robustness, reproducibility and automation, modern calculation tools are exploited to produce original nuclear data libraries that meet the current demands on quality and completeness. This requires a system which comprises differential measurements, theory development, nuclear model codes, resonance analysis, evaluation, ENDF formatting, data processing and integral validation in one integrated approach. Software, built around the TALYS code, will be presented in which all these essential nuclear data components are seamlessly integrated. Besides the quality of the basic data and its extensive format testing, a second goal lies in the diversity of processing for different type of users. The implications of this scheme are unprecedented. The most important are: 1. Complete ENDF-6 nuclear data files, in the form of the TENDL library, including covariance matrices, for many isotopes, particles, energies, reaction channels and derived quantities. All isotopic data files are mutually consistent and are supposed to rival those of the major world libraries. 2. More exact uncertainty propagation from basic nuclear physics to applied (reactor) calculations based on a Monte Carlo approach: “Total” Monte Carlo (TMC), using random nuclear data libraries. 3. Automatic optimization in the form of systematic feedback from integral measurements back to the basic data. This method of work also opens a new way of approaching the analysis of nuclear applications, with consequences in both applied nuclear physics and safety of nuclear installations, and several examples are given here. This applied experience and feedback is integrated in a final step to improve the quality of the nuclear data, to change the users vision and finally to orchestrate their integration into simulation codes.

  18. Exporting nuclear engineering and the industry's viewpoint

    International Nuclear Information System (INIS)

    Barthelt, K.

    1986-01-01

    Nuclear energy offers all possibilities to reduce the energy problems in the world which arise with the world-wide increasing population and the energy demand connected with it. The Federal Republic of Germany lives on the exports of refined technical methods which also include nuclear engineering. The exports of nuclear engineering should lead to a technology transfer with guidance and training on an equal basis between the industrial and developing countries. The preconditions of exporting nuclear-technical systems are a well-functioning domestic market and a certain support by the government, especially with regard to giving guarantees for the special exports risks of these big projects. On the other hand, exports are also needed in order to be able to continue providing high-level technology for the domestic market. (UA) [de

  19. The world nuclear power engineering. 1998 year

    International Nuclear Information System (INIS)

    Preobrazhenskaya, L.B.

    2000-01-01

    The purpose of this article consists in the analysis of the state and prospects of the world nuclear power engineering development. The data on the ratio and value of electrical energy obtained at the NPPs in the world in 1998, the specific capital expenditures on the NPPs construction by 2005, the forecast for the capacity of all NPPs by 2020 are presented. The progress in developing nuclear power engineering conditioned by improvement of the NPPs operation, optimization of their life-cycle and developing of new NPPs projects is noted [ru

  20. Social engineering awareness in Nuclear Malaysia

    International Nuclear Information System (INIS)

    Mohd Dzul Aiman bin Aslan; Mohamad Safuan bin Sulaiman; Abdul Muin bin Abdul Rahman

    2010-01-01

    Social engineering is the best tools to infiltrate an organization weakness. It can go bypass the best fire wall or Intrusion Detection System (IDS) the organization ever had, effectively. Nuclear Malaysia staffs should aware of this technique as information protection it is not only depends on paper and computer. This paper consist a few test cases including e mail, dump ster diving, phishing, malicious web content, and impersonation to acknowledge all Nuclear Malaysia staffs about the method, effect and prevention of social engineering. (author)

  1. Education in nuclear engineering in Slovakia

    International Nuclear Information System (INIS)

    Slugen, V.

    2005-01-01

    Slovak University of Technology is the largest and also the oldest university of technology in Slovakia. Surely more than 50% of high-educated technicians who work nowadays in nuclear industry have graduated from this university. The Department of Nuclear Physics and Technology of the Faculty of Electrical Engineering and Information Technology as a one of seven faculties of this University feels responsibility for proper engineering education and training for Slovak NPP operating staff. The education process is realised via undergraduate (Bc.), graduate (MSc.) and postgraduate (PhD..) study as well as via specialised training courses in a frame of continuous education system. (author)

  2. Nuclear engineering enrollments and degrees, 1981

    International Nuclear Information System (INIS)

    Little, J.R.; Shirley, D.L.

    1982-05-01

    This report presents data on the number of students enrolled and the degrees awarded in academic year 1980-81 from 73 US institutions offering degree programs in nuclear engineering or nuclear options within other engineering fields. Presented here are historical data for the last decade, which provide information such as trends by degree level, foreign national student participation, female and minority student participation, and placement of graduates. Also included is a listing of the universities by type of program and number of students

  3. Hauser*5, a computer code to calculate nuclear cross sections

    International Nuclear Information System (INIS)

    Mann, F.M.

    1979-07-01

    HAUSER*5 is a computer code that uses the statistical (Hauser-Feshbach) model, the pre-equilibrium model, and a statistical model of direct reactions to predict nuclear cross sections. The code is unrestricted as to particle type, includes fission and capture, makes width-fluctuation corrections, and performs three-body calculations - all in minimum computer time. Transmission coefficients can be generated internally or supplied externally. This report describes equations used, necessary input, and resulting output. 2 figures, 4 tables

  4. Code of practice of radiation protection in fixed nuclear gauges

    International Nuclear Information System (INIS)

    Eltayeb, M. A. M.

    2012-09-01

    The present work aims at developing and updating a code of practice of radiation protection in fixed nuclear gauges that comply with current international recommendations. The work also intended to evaluate the current radiation protection situation in two selected companies using nuclear gauges in Sudan. A draft of the code is proposed which includes the basic principle of protection such as source construction and gauges radiation monitoring, storage maintenance and leak testing as well as specific issues related to nuclear gauges. The practical part of this study included investigation of radiation protection in the comparisons using nuclear gauges for level detection, to evaluate the level of radiation protection and the compliance to the regulatory authority regulations. The result revealed that the two companies do not have an effective radiation protection program and that can lead to exposure of workers to unnecessary doses. Some recommendations were stated, if implemented they could improve the status of radiation protection in those companies. (Author)

  5. Application of nuclear air cleaning and treatment codes

    International Nuclear Information System (INIS)

    Kriskovich, J.R.

    1995-01-01

    All modifications to existing ventilation systems, as well as any new ventilation systems used on the Hanford Site are required to meet both American Society of Mechanical Engineers (ASME) codes N509 and N510. Difficulties encountered when applying code N509 at the Hanford Site include the composition of the ventilation air stream and requirements related to ventilation equipment procurement. Also, the existing ventilation systems for the waste tanks at the Hanford Site cannot be tested in accordance with code N510 because of the current configuration of these systems

  6. Application of nuclear air cleaning and treatment codes

    Energy Technology Data Exchange (ETDEWEB)

    Kriskovich, J.R. [Westinghouse Hanford Company, Richland, WA (United States)

    1995-02-01

    All modifications to existing ventilation systems, as well as any new ventilation systems used on the Hanford Site are required to meet both American Society of Mechanical Engineers (ASME) codes N509 and N510. Difficulties encountered when applying code N509 at the Hanford Site include the composition of the ventilation air stream and requirements related to ventilation equipment procurement. Also, the existing ventilation systems for the waste tanks at the Hanford Site cannot be tested in accordance with code N510 because of the current configuration of these systems.

  7. 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.

  8. Chemical engineering side of nuclear fusion power

    International Nuclear Information System (INIS)

    Johnson, E.F.

    1976-10-01

    It is widely recognized that chemical engineering has important roles to play in the development of national and world wide energy resources through optimal utilization of fossil fuel reserves. It is much less appreciated that there are crucial chemical engineering problems in the development of energy production from other sources. In particular the successful development of nuclear fusion power generating systems will require the solution of many problems that are uniquely suited to chemical engineers. This article presents a brief overview of the fusion development program and an identification of the major technological problems remaining to be solved

  9. Nuclear thermal rocket engine operation and control

    International Nuclear Information System (INIS)

    Gunn, S.V.; Savoie, M.T.; Hundal, R.

    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. 9 refs

  10. Independent peer review of nuclear safety computer codes

    International Nuclear Information System (INIS)

    Boyack, B.E.; Jenks, R.P.

    1993-01-01

    A structured, independent computer code peer-review process has been developed to assist the US Nuclear Regulatory Commission (NRC) and the US Department of Energy in their nuclear safety missions. This paper describes a structured process of independent code peer review, benefits associated with a code-independent peer review, as well as the authors' recent peer-review experience. The NRC adheres to the principle that safety of plant design, construction, and operation are the responsibility of the licensee. Nevertheless, NRC staff must have the ability to independently assess plant designs and safety analyses submitted by license applicants. According to Ref. 1, open-quotes this requires that a sound understanding be obtained of the important physical phenomena that may occur during transients in operating power plants.close quotes The NRC concluded that computer codes are the principal products to open-quotes understand and predict plant response to deviations from normal operating conditionsclose quotes and has developed several codes for that purpose. However, codes cannot be used blindly; they must be assessed and found adequate for the purposes they are intended. A key part of the qualification process can be accomplished through code peer reviews; this approach has been adopted by the NRC

  11. Computation of Thermodynamic Equilibria Pertinent to Nuclear Materials in Multi-Physics Codes

    Science.gov (United States)

    Piro, Markus Hans Alexander

    Nuclear energy plays a vital role in supporting electrical needs and fulfilling commitments to reduce greenhouse gas emissions. Research is a continuing necessity to improve the predictive capabilities of fuel behaviour in order to reduce costs and to meet increasingly stringent safety requirements by the regulator. Moreover, a renewed interest in nuclear energy has given rise to a "nuclear renaissance" and the necessity to design the next generation of reactors. In support of this goal, significant research efforts have been dedicated to the advancement of numerical modelling and computational tools in simulating various physical and chemical phenomena associated with nuclear fuel behaviour. This undertaking in effect is collecting the experience and observations of a past generation of nuclear engineers and scientists in a meaningful way for future design purposes. There is an increasing desire to integrate thermodynamic computations directly into multi-physics nuclear fuel performance and safety codes. A new equilibrium thermodynamic solver is being developed with this matter as a primary objective. This solver is intended to provide thermodynamic material properties and boundary conditions for continuum transport calculations. There are several concerns with the use of existing commercial thermodynamic codes: computational performance; limited capabilities in handling large multi-component systems of interest to the nuclear industry; convenient incorporation into other codes with quality assurance considerations; and, licensing entanglements associated with code distribution. The development of this software in this research is aimed at addressing all of these concerns. The approach taken in this work exploits fundamental principles of equilibrium thermodynamics to simplify the numerical optimization equations. In brief, the chemical potentials of all species and phases in the system are constrained by estimates of the chemical potentials of the system

  12. 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.

  13. Nuclear data to support computer code validation

    International Nuclear Information System (INIS)

    Fisher, S.E.; Broadhead, B.L.; DeHart, M.D.; Primm, R.T. III

    1997-04-01

    The rate of plutonium disposition will be a key parameter in determining the degree of success of the Fissile Materials Disposition Program. Estimates of the disposition rate are dependent on neutronics calculations. To ensure that these calculations are accurate, the codes and data should be validated against applicable experimental measurements. Further, before mixed-oxide (MOX) fuel can be fabricated and loaded into a reactor, the fuel vendors, fabricators, fuel transporters, reactor owners and operators, regulatory authorities, and the Department of Energy (DOE) must accept the validity of design calculations. This report presents sources of neutronics measurements that have potential application for validating reactor physics (predicting the power distribution in the reactor core), predicting the spent fuel isotopic content, predicting the decay heat generation rate, certifying criticality safety of fuel cycle facilities, and ensuring adequate radiation protection at the fuel cycle facilities and the reactor. The U.S. in-reactor experience with MOX fuel is first presented, followed by information related to other aspects of the MOX fuel performance information that is valuable to this program, but the data base remains largely proprietary. Thus, this information is not reported here. It is expected that the selected consortium will make the necessary arrangements to procure or have access to the requisite information

  14. Educating nuclear engineers at German universities

    International Nuclear Information System (INIS)

    Knorr, J.

    1995-01-01

    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.) [de

  15. Exporting nuclear engineering and the government's viewpoint

    International Nuclear Information System (INIS)

    Schill, H.

    1986-01-01

    The reasons for the government's positive attitude to nuclear engineering exports are explained, especially with regard to them being a compensation of the decreasing domestic demand. The federal government considers such exports to be necessary and correct for economical and energy-political reasons. Their contribution reaches from accompanying measures to the provision of state guarantees of export financing activities. (UA) [de

  16. Current situation of nuclear engineering education

    International Nuclear Information System (INIS)

    Queral, C.; Minguez, E.

    2001-01-01

    The last few years have seen a growing concern with the decreasing number of suitably qualified engineers and university graduates in the field of Nuclear Technology. The gap between supply and demand is now a fact in several countries, and for the reason the international community has prepared several reports on the issue that are summarized here. (Author) 4 refs

  17. Abbreviations of nuclear power plant engineering

    International Nuclear Information System (INIS)

    Freyberger, G.H.

    1979-01-01

    The edition of this English and German list of abbreviations comprises about 5200 entries in English and about 1400 entries in German as well as the most important American, English, German and other foreign Utilities and component manufacturers frequently quoted in nuclear engineering literature and documentation. (orig./HP) [de

  18. Environment Protection (Nuclear Codes) Act 1978, No. 32

    International Nuclear Information System (INIS)

    1978-01-01

    The purpose of this Act is to make provision for protecting the health and safety of the people of Australia, and the environment from possible harmful effects associated with nuclear activities. The competent Federal Minister, after consultation with the appropriate Minister of each State, may from time to time arrange for the formulation of codes of practice for regulating and controlling nuclear activities. Also, the Governor-General may authorize by order a Minister to take such action as is necessary to control and eliminate hazards to the health and safety of persons, or the environment, resulting from a nuclear activity. (NEA) [fr

  19. Development of simplified decommissioning cost estimation code for nuclear facilities

    International Nuclear Information System (INIS)

    Tachibana, Mitsuo; Shiraishi, Kunio; Ishigami, Tsutomu

    2010-01-01

    The simplified decommissioning cost estimation code for nuclear facilities (DECOST code) was developed in consideration of features and structures of nuclear facilities and similarity of dismantling methods. The DECOST code could calculate 8 evaluation items of decommissioning cost. Actual dismantling in the Japan Atomic Energy Agency (JAEA) was evaluated; unit conversion factors used to calculate the manpower of dismantling activities were evaluated. Consequently, unit conversion factors of general components could be classified into three kinds. Weights of components and structures of the facility were necessary for calculation of manpower. Methods for evaluating weights of components and structures of the facility were studied. Consequently, the weight of components in the facility was proportional to the weight of structures of the facility. The weight of structures of the facility was proportional to the total area of floors in the facility. Decommissioning costs of 7 nuclear facilities in the JAEA were calculated by using the DECOST code. To verify the calculated results, the calculated manpower was compared with the manpower gained from actual dismantling. Consequently, the calculated manpower and actual manpower were almost equal. The outline of the DECOST code, evaluation results of unit conversion factors, the evaluation method of the weights of components and structures of the facility are described in this report. (author)

  20. The status of Korean nuclear codes and standards

    International Nuclear Information System (INIS)

    Namha Kim; Jong-Hae Kim

    2005-01-01

    Korea Electric Power Industry Code (KEPIC), a set of integrated standards applicable to the design, construction and operation of electric power facilities including nuclear power plants, has been developed on the basis of referring to the prevailing U.S. codes and standards which had been applied to the electric power facilities in Korea. Being the developing and managing organization of KEPIC, Korea Electric Association (KEA) published its first edition in 1995, the second in 200,0 and is expected to publish the 2005 edition. KEPIC was applied to the construction of Ulchin Nuclear Units 5 and 6 in 1997, and will be applicable to the construction of forthcoming nuclear power plants in Korea. Along with the effectuation of the Agreement on Technical Barriers to Trade (TBT) in 1995, the international trend related to codes and standards is changing rapidly. The KEA is, therefore, making its utmost efforts so as for KEPIC to keep abreast with the changing environment in international arena. KEA notified ISO/IEC Information Centre of its acceptance of the Code of Good Practice in the Agreement on TBT. The 2005 KEPIC edition will be retrofitted according to the ISO/IEC Guide 21- Adoption of International Standards as regional or national standards. KEA's efforts will help KEPIC correspond with international standards such as ISO/IEC standards, and internationally recognized standards such as ASME codes and standards. (authors)

  1. Midwest Nuclear Science and Engineering Consortium

    International Nuclear Information System (INIS)

    Volkert, Wynn; Kumar, Arvind; Becker, Bryan; Schwinke, Victor; Gonzalez, Angel; McGregor, Douglas

    2010-01-01

    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.

  2. 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.

  3. 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.

  4. Method for quantitative assessment of nuclear safety computer codes

    International Nuclear Information System (INIS)

    Dearien, J.A.; Davis, C.B.; Matthews, L.J.

    1979-01-01

    A procedure has been developed for the quantitative assessment of nuclear safety computer codes and tested by comparison of RELAP4/MOD6 predictions with results from two Semiscale tests. This paper describes the developed procedure, the application of the procedure to the Semiscale tests, and the results obtained from the comparison

  5. EMPIRE-II statistical model code for nuclear reaction calculations

    Energy Technology Data Exchange (ETDEWEB)

    Herman, M [International Atomic Energy Agency, Vienna (Austria)

    2001-12-15

    EMPIRE II is a nuclear reaction code, comprising various nuclear models, and designed for calculations in the broad range of energies and incident particles. A projectile can be any nucleon or Heavy Ion. The energy range starts just above the resonance region, in the case of neutron projectile, and extends up to few hundreds of MeV for Heavy Ion induced reactions. The code accounts for the major nuclear reaction mechanisms, such as optical model (SCATB), Multistep Direct (ORION + TRISTAN), NVWY Multistep Compound, and the full featured Hauser-Feshbach model. Heavy Ion fusion cross section can be calculated within the simplified coupled channels approach (CCFUS). A comprehensive library of input parameters covers nuclear masses, optical model parameters, ground state deformations, discrete levels and decay schemes, level densities, fission barriers (BARFIT), moments of inertia (MOMFIT), and {gamma}-ray strength functions. Effects of the dynamic deformation of a fast rotating nucleus can be taken into account in the calculations. The results can be converted into the ENDF-VI format using the accompanying code EMPEND. The package contains the full EXFOR library of experimental data. Relevant EXFOR entries are automatically retrieved during the calculations. Plots comparing experimental results with the calculated ones can be produced using X4TOC4 and PLOTC4 codes linked to the rest of the system through bash-shell (UNIX) scripts. The graphic user interface written in Tcl/Tk is provided. (author)

  6. Nuclear engineering. Stable industry for bright minds

    International Nuclear Information System (INIS)

    Geisler, Maja

    2009-01-01

    The Deutsches Atomforum (DAtF) invited 35 students and graduate students for 'colloquies for professional orientation' to Luenen on March 8-11, 2009. Another 39 students were guests in Speyer between March 15 and 18 this year. Participants included graduates in physics, chemistry, radiation protection, and mechanical engineering as well as students of process engineering, electrical engineering and environmental technology. The colloquies for professional orientation are a service provided by the Informationskreis Kernenergie (IK) to member firms of DAtF. At the same time, the IK in this way fulfils its duty to promote young scientists and engineers within the framework of the DAtF's basic public relations activities. After all, nuclear technology in Germany is not about to end its life. Firms with international activities are in urgent need of highly qualified young staff members. Personnel is needed for a variety of activities ranging from nuclear power plant construction to fuel fabrication to waste management and the demolition and disposal of nuclear power plants. All these areas are in need of new qualified staff. Some 750 students so far have attended the DAtF colloquies for professional orientation since 2002. Many participants were hired by industries straight away or were given opportunities as trainees or students preparing their diploma theses in the nuclear industry. These contacts with the nuclear industry should not remain a one-off experience for the students. For this reason, the IK invites the participants in colloquies again this year to attend the Annual Meeting on Nuclear Technology in Dresden on May 12-14, 2009. (orig.)

  7. Development of a nuclear power plant system analysis code

    International Nuclear Information System (INIS)

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

    1997-07-01

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

  8. Human modeling in nuclear engineering

    International Nuclear Information System (INIS)

    Yoshikawa, Hidekazu; Furuta, Kazuo.

    1994-01-01

    Review on progress of research and development on human modeling methods is made from the viewpoint of its importance on total man-machine system reliability surrounding nuclear power plant operation. Basic notions on three different approaches of human modeling (behavioristics, cognitives and sociologistics) are firstly introduced, followed by the explanation of fundamental scheme to understand human cognitives at man-machine interface and the mechanisms of human error and its classification. Then, general methodologies on human cognitive model by AI are explained with the brief summary of various R and D activities now prevailing in the human modeling communities around the world. A new method of dealing with group human reliability is also introduced which is based on sociologistic mathematical model. Lastly, problems on human model validation are discussed, followed by the introduction of new experimental method to estimate human cognitive state by psycho-physiological measurement, which is a new methodology plausible for human model validation. (author)

  9. 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)

  10. Transport code and nuclear data in intermediate energy region

    International Nuclear Information System (INIS)

    Hasegawa, Akira; Odama, Naomitsu; Maekawa, F.; Ueki, K.; Kosaka, K.; Oyama, Y.

    1998-01-01

    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)

  11. Merits and difficulties in adopting codes, standards and nuclear regulations

    International Nuclear Information System (INIS)

    El-Saiedi, A.F.; Morsy, S.; Mariy, A.

    1978-01-01

    Developing countries planning for introducing nuclear power plants as a source of energy have to develop or adopt sound regulatory practices. These are necessary to help governmental authorities to assess the safety of nuclear power plants and to perform inspections needed to confirm the established safe and sound limits. The first requirement is to form an independent regulatory body capable of setting up and enforcing proper safety regulations. The formation of this body is governed by several considerations related to local conditions in the developing countries, which may not always be favourable. It is quite impractical for countries with limited experience in the nuclear power field to develop their own codes, standards and regulations required for the nuclear regulatory body to perform its tasks. A practical way is to adopt codes, standards and regulations of a well-developed country. This has merits as well as drawbacks. The latter are related to problems of personnel, software, equipment and facilities. The difficulties involved in forming a nuclear regulatory body, and the merits and difficulties in adopting foreign codes, standards and regulations required for such body to perform its tasks, are discussed in this paper. Discussions are applicable to many developing countries and particular emphasis is given to the conditions and practices in Egypt. (author)

  12. Comparison of Engine Cycle Codes for Rocket-Based Combined Cycle Engines

    Science.gov (United States)

    Waltrup, Paul J.; Auslender, Aaron H.; Bradford, John E.; Carreiro, Louis R.; Gettinger, Christopher; Komar, D. R.; McDonald, J.; Snyder, Christopher A.

    2002-01-01

    This paper summarizes the results from a one day workshop on Rocket-Based Combined Cycle (RBCC) Engine Cycle Codes held in Monterey CA in November of 2000 at the 2000 JANNAF JPM with the authors as primary participants. The objectives of the workshop were to discuss and compare the merits of existing Rocket-Based Combined Cycle (RBCC) engine cycle codes being used by government and industry to predict RBCC engine performance and interpret experimental results. These merits included physical and chemical modeling, accuracy and user friendliness. The ultimate purpose of the workshop was to identify the best codes for analyzing RBCC engines and to document any potential shortcomings, not to demonstrate the merits or deficiencies of any particular engine design. Five cases representative of the operating regimes of typical RBCC engines were used as the basis of these comparisons. These included Mach 0 sea level static and Mach 1.0 and Mach 2.5 Air-Augmented-Rocket (AAR), Mach 4 subsonic combustion ramjet or dual-mode scramjet, and Mach 8 scramjet operating modes. Specification of a generic RBCC engine geometry and concomitant component operating efficiencies, bypass ratios, fuel/oxidizer/air equivalence ratios and flight dynamic pressures were provided. The engine included an air inlet, isolator duct, axial rocket motor/injector, axial wall fuel injectors, diverging combustor, and exit nozzle. Gaseous hydrogen was used as the fuel with the rocket portion of the system using a gaseous H2/O2 propellant system to avoid cryogenic issues. The results of the workshop, even after post-workshop adjudication of differences, were surprising. They showed that the codes predicted essentially the same performance at the Mach 0 and I conditions, but progressively diverged from a common value (for example, for fuel specific impulse, Isp) as the flight Mach number increased, with the largest differences at Mach 8. The example cases and results are compared and discussed in this paper.

  13. 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.

  14. Quality assurance system in nuclear engineering

    International Nuclear Information System (INIS)

    Adams, H.W.; Hoensch, V.

    1985-01-01

    Due to the close connection between the German Atomic Energy Law and the nuclear control regulations, quality systems in nuclear engineering have taken on a special form. Quality assurance systems as a stipulated organisation of structure and procedure to assure quality have implications for the organisation of the electric supply company at the planning, erection and commissioning stage and for the organisation of the nuclear power station facility. To supervise the application and effectiveness of the stipulated organisation of structure and procedure internally and externally among contractors, special organisation units have been set up at the plant suppliers, manufactures, electric supply companies and nuclear power station facilities, which in the electric supply field go by the name of Quality Assurance Supervision. (orig.) [de

  15. Interactive computer codes for education and training on nuclear safety and radioprotection

    International Nuclear Information System (INIS)

    Leszczynski, Francisco

    2008-01-01

    Two interactive computer codes for education and training on nuclear safety and radioprotection developed at RA6 Reactor Division-Bariloche Atomic Center-CNEA are presented on this paper. The first code named SIMREACT has been developed in order to simulate the control of a research nuclear reactor in real time with a simple but accurate approach. The code solves the equations of neutron punctual kinetics with time variable reactivity. Utilizing the timer of the computer and the controls of a PC keyboard, with an adequate graphic interface, a simulation in real time of the temporal behavior of a research reactor is obtained. The reactivity can be changed by means of the extraction or insertion of control rods. It was implemented also the simulation of automatic pilot and scram. The use of this code is focalized on practices of nuclear reactor control like start-up from the subcritical state with external source up to power to a desired level, change of power level, calibration of a control rod with different methods, and approach to critical condition by interpolation of the answer in function of reactivity. The second code named LICEN has been developed in order to help the studies of all the topics included in examination programs for obtaining licenses for research reactor operators and radioprotection officials. Using the PC mouse, with an adequate graphic interface, the student can gradually learn the topics related with general and special licenses. The general option includes nuclear reactor engineering, radioprotection, nuclear safety, documentation and normative. The specific option includes mandatory documentation, description of the installation and task on normal and emergency situations. For each of these topics there are sub-items with all the relevant information. The objective of this code is to joint in one electronic place a large amount of information which usually it is disseminated on difficult to find separated papers. (author)

  16. NUKEM. Innovative solutions for nuclear engineering

    International Nuclear Information System (INIS)

    Scheffler, Beate

    2011-01-01

    Management of radioactive waste, handling spent fuel elements, decommissioning of nuclear facilities, and engineering and consulting activities are services associated with the name of NUKEM all over the world. The company's scientists and engineers develop solution concepts combining the latest technologies with proven techniques and many years of experience. The company;s history and the services offered to the nuclear industry began more than 5 decades ago. The predecessor, NUKEM Nuklear-Chemie-Metallurgie, was founded in 1960 as one of the earliest nuclear companies in Germany. Originally, the firm produced fuel elements for a variety of reactor lines. As early as in the 1970s, logical extensions of these business activities were nuclear engineering and plant construction. In the meantime, NUKEM Technologies GmbH has developed a worldwide reputation for its activities. Numerous reference projects bear witness to optimum project management and customer satisfaction. Since 2009, NUKEM Technologies has been a wholly owned subsidiary of the Russian Atomstroyexport. NUKEM Technologies operates sales and project offices outside Germany, e.g. in Russia, China, Lithuania, France, and Bulgaria. In this way, the company is present in its target markets of Russia, Western and Eastern Europe as well as Asia, offering customers and partners fast and direct contacts. (orig.)

  17. Nuclear industry - challenges in chemical engineering

    International Nuclear Information System (INIS)

    Sen, S.; Sunder Rajan, N.S.; Balu, K.; Garg, R.K.; Murthy, L.G.K.; Ramani, M.P.S.; Rao, M.K.; Sadhukhan, H.K.; Venkat Raj, V.

    1978-01-01

    Chemical engineering processes and operations are closely involved in every step of the nuclear fuel cycle. Starting from mining and milling of the ore through the production of fuel and other materials and their use in nuclear reactors, fuel reprocessing, fissile material recycle and treatment and disposal of fission product wastes, each step presents a challenge to the chemical engineer to evolve and innovate processes and techniques for more efficient utilization of the energy in the atom. The requirement of high recovery of the desired components at high purity levels is in itself a challenge. ''Nuclear Grade'' specifications for materials put a requirement which very few industries can satisfy. Recovery of uranium and thorium from low grade ores, of heavy water from raw water, etc. are examples. Economical and large scale separation of isotopes particularly those of heavy elements is a task for which processess are under various stages of development. Further design of chemical plants such as fuel reprocessing plants and high level waste treatment plants, which are to be operated and maintained remotely due to the high levels of radio-activity call for engineering skills which are being continually evolved. In the reactor, analysis of the fluid mechanics and optimum design of heat removal system are other examples where a chemical engineer can play a useful role. In addition to the above, the activities in the nuclear industry cover a very wide range of chemical engineering applications, such as desalination and other energy intensive processes, radioisotope and radiation applications in industry, medicine and agriculture. (auth.)

  18. Final Technical Report and management: NUCLEAR ENGINEERING RECRUITMENT EFFORT

    International Nuclear Information System (INIS)

    Kerrick, Sharon S.; Vincent, Charles D.

    2007-01-01

    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; and Objective 6--Nuclear engineering enrollment survey at universities

  19. Validation and verification of the ORNL Monte Carlo codes for nuclear safety analysis

    International Nuclear Information System (INIS)

    Emmett, M.B.

    1993-01-01

    The process of ensuring the quality of computer codes can be very time consuming and expensive. The Oak Ridge National Laboratory (ORNL) Monte Carlo codes all predate the existence of quality assurance (QA) standards and configuration control. The number of person-years and the amount of money spent on code development make it impossible to adhere strictly to all the current requirements. At ORNL, the Nuclear Engineering Applications Section of the Computing Applications Division is responsible for the development, maintenance, and application of the Monte Carlo codes MORSE and KENO. The KENO code is used for doing criticality analyses; the MORSE code, which has two official versions, CGA and SGC, is used for radiation transport analyses. Because KENO and MORSE were very thoroughly checked out over the many years of extensive use both in the United States and in the international community, the existing codes were open-quotes baselined.close quotes This means that the versions existing at the time the original configuration plan is written are considered to be validated and verified code systems based on the established experience with them

  20. 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)

  1. Monte Carlo simulation on nuclear energy study. Annual report of Nuclear Code Evaluation Committee

    International Nuclear Information System (INIS)

    Sakurai, Kiyoshi; Yamamoto, Toshihiro

    1999-03-01

    In this report, research results discussed in 1998 fiscal year at Nuclear Code Evaluation Special Committee of Nuclear Code Committee were summarised. Present status of Monte Carlo calculation in high energy region investigated / discussed at Monte Carlo simulation working-group and automatic compilation system for MCNP cross sections developed at MCNP high temperature library compilation working-group were described. The 6 papers are indexed individually. (J.P.N.)

  2. Application of nuclear photon engines for deep-space exploration

    International Nuclear Information System (INIS)

    Gulevich, Andrey V.; Ivanov, Eugeny A.; Kukharchuk, Oleg F.; Poupko, Victor Ya.; Zrodnikov, Anatoly V.

    2001-01-01

    Conception of using the nuclear photon rocket engines for deep space exploration is proposed. Some analytical estimations have been made to illustrate the possibility to travel to 100-10000 AU using a small thrust photon engine. Concepts of high temperature nuclear reactors for the nuclear photon engines are also discussed

  3. Nuclear Engineering Enrollments and Degrees Survey, 2005 Data

    International Nuclear Information System (INIS)

    Oak Ridge Institute for Science and Education

    2006-01-01

    This annual report details the number of nuclear engineering bachelor's, master's, and doctoral degrees awarded at a sampling of academic programs from 1998-2005. it also looks at nuclear engineering degrees by curriculum and the number of students enrolled in nuclear engineering degree programs at 30 U.S. universities in 2005

  4. European Master of Science in Nuclear Engineering

    International Nuclear Information System (INIS)

    Moons, Frans; Safieh, Joseph; Giot, Michel; Mavko, Borut; Sehgal, Bal Raj; Schaefer, Anselm; Goethem, Georges van; D'Haeseleer, William

    2005-01-01

    The need to preserve, enhance or strengthen nuclear knowledge is worldwide recognised since a couple of years. Among others, 'networking to maintain nuclear competence through education and training', was recommended in 2001 by an expert panel to the European Commission [EUR, 19150 EN, Strategic issues related to a 6th Euratom Framework Programme (2002-2006). Scientific and Technical Committee Euratom, pp. 14]. It appears that within the European University education and training framework, nuclear engineering is presently still sufficiently covered, although somewhat fragmented. However, it has been observed that several areas are at risk in the very near future including safety relevant fields such as reactor physics and nuclear thermal-hydraulics. Furthermore, in some countries deficiencies have been identified in areas such as the back-end of the nuclear fuel cycle, waste management and decommissioning. To overcome these risks and deficiencies, it is of very high importance that European countries work more closely together. Harmonisation and improvement of the nuclear education and training have to take place at an international level in order to maintain the knowledge properly and to transfer it throughout Europe for the safe and economic design, operation and dismantling of present and future nuclear systems. To take up the challenges of offering top quality, new, attractive and relevant curricula, higher education institutions should cooperate with industry, regulatory bodies and research centres, and more appropriate funding from public and private sources. In addition, European nuclear education and training should benefit from links with international organisations like IAEA, OECD-NEA and others, and should include worldwide cooperation with academic institutions and research centres. The first and central issue is to establish a European Master of Science in Nuclear Engineering. The concept envisaged is compatible with the projected harmonised European

  5. A study on the nuclear computer codes installation and management system

    International Nuclear Information System (INIS)

    Kim, Yeon Seung; Huh, Young Hwan; Kim, Hee Kyung; Kang, Byung Heon; Kim, Ko Ryeo; Suh, Soong Hyok; Choi, Young Gil; Lee, Jong Bok

    1990-12-01

    From 1987 a number of technical transfer related to nuclear power plant had been performed from C-E for YGN 3 and 4 construction. Among them, installation and management of the computer codes for YGN 3 and 4 fuel and nuclear steam supply system was one of the most important project. Main objectives of this project are to establish the nuclear computer code management system, to develop QA procedure for nuclear codes, to secure the nuclear code reliability and to extend techanical applicabilities including the user-oriented utility programs for nuclear codes. Contents of performing the project in this year was to produce 215 transmittal packages of nuclear codes installation including making backup magnetic tape and microfiche for software quality assurance. Lastly, for easy reference about the nuclear codes information we presented list of code names and information on the codes which were introduced from C-E. (Author)

  6. Nuclear engineering education in the United States

    International Nuclear Information System (INIS)

    Williamson, T.G.

    1982-01-01

    The critical issue facing the nuclear engineering education community today is first and foremost enrollment in a time of increasing demand for graduate engineers. Related to the issue of enrollment is support for graduate students, whether it be fellowships, traineeships, or research assistantships. Other issues are those of maintaining a vital faculty in the face of competitive job market, of maintaining research facilities and developing new ones, and last and certainly not least that of determining the directions of our educational efforts in the future. These issues are examined in the paper. (author)

  7. Civil engineering firms and the nuclear programme

    International Nuclear Information System (INIS)

    Giral, J.L.

    1988-01-01

    Pointing out that the realization of the electronuclear programme accounted for just under 5 % of the annual turnover of civil engineering firms from 1975 to 1987, the author lists the main types of work entrusted to these firms for the bulding of the power stations: electrical work, structural work (civil engineering, metal structures. He then describes the two main problems which the profession has to face in the nuclear field: the management of final contract stages and adaptation to the slowing down in the rate of commitment of power plants [fr

  8. Piping engineering for nuclear power plant

    International Nuclear Information System (INIS)

    Curto, N.; Schmidt, H.; Muller, R.

    1988-01-01

    In order to develop piping engineering, an adequate dimensioning and correct selection of materials must be secured. A correct selection of materials together with calculations and stress analysis must be carried out with a view to minimizing or avoiding possible failures or damages in piping assembling, which could be caused by internal pressure, weight, temperature, oscillation, etc. The piping project for a nuclear power plant is divided into the following three phases. Phase I: Basic piping design. Phase II: Final piping design. Phase III: Detail engineering. (Author)

  9. ASME Section XI trends in developing nuclear codes and standards

    International Nuclear Information System (INIS)

    Hedden, O.F.

    1995-01-01

    When the author began working on nuclear power many years ago, he knew that perfection was the only acceptable technical standard. Unfortunately, this became an obsession with perfection that has had unfavorable consequences in some of the non-technical areas of work in ASME nuclear power Codes and Standards. However, the economic problems of the nuclear power industry now demand a more pragmatic approach if the industry is to continue. Not only does each item considered for action need to be evaluated to criteria that may in some cases be less than perfection, but one needs to consider whether it contributes tangibly to either safety or to reduction in technical or administrative burden. These should be the governing, criteria. The introduction of risk-based inspection methodologies will certainly be an important element in doing this successfully. One needs to consider these criteria collectively, as one discusses each item at the committee level, and individually, as one votes on each item. In the past, the author has been concerned that the industry was not acting quickly enough in taking advantage of opportunities offered by the Code to increase safety or to reduce cost. While he still has some concern, he thinks communication channels have been greatly improved. Now he is becoming more concerned with both the collective and individual actions that delay beneficial changes. The second part of the author's talk has to do with the relevance of the code committees in the nuclear power industry regulatory process

  10. Nuclear-thermal-coupled optimization code for the fusion breeding blanket conceptual design

    International Nuclear Information System (INIS)

    Li, Jia; Jiang, Kecheng; Zhang, Xiaokang; Nie, Xingchen; Zhu, Qinjun; Liu, Songlin

    2016-01-01

    Highlights: • A nuclear-thermal-coupled predesign code has been developed for optimizing the radial build arrangement of fusion breeding blanket. • Coupling module aims at speeding up the efficiency of design progress by coupling the neutronics calculation code with the thermal-hydraulic analysis code. • Radial build optimization algorithm aims at optimal arrangement of breeding blanket considering one or multiple specified objectives subject to the design criteria such as material temperature limit and available TBR. - Abstract: Fusion breeding blanket as one of the key in-vessel components performs the functions of breeding the tritium, removing the nuclear heat and heat flux from plasma chamber as well as acting as part of shielding system. The radial build design which determines the arrangement of function zones and material properties on the radial direction is the basis of the detailed design of fusion breeding blanket. For facilitating the radial build design, this study aims for developing a pre-design code to optimize the radial build of blanket with considering the performance of nuclear and thermal-hydraulic simultaneously. Two main features of this code are: (1) Coupling of the neutronics analysis with the thermal-hydraulic analysis to speed up the analysis progress; (2) preliminary optimization algorithm using one or multiple specified objectives subject to the design criteria in the form of constrains imposed on design variables and performance parameters within the possible engineering ranges. This pre-design code has been applied to the conceptual design of water-cooled ceramic breeding blanket in project of China fusion engineering testing reactor (CFETR).

  11. Nuclear-thermal-coupled optimization code for the fusion breeding blanket conceptual design

    Energy Technology Data Exchange (ETDEWEB)

    Li, Jia, E-mail: lijia@ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027, Anhui (China); Jiang, Kecheng; Zhang, Xiaokang [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, Anhui (China); Nie, Xingchen [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027, Anhui (China); Zhu, Qinjun; Liu, Songlin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, Anhui (China)

    2016-12-15

    Highlights: • A nuclear-thermal-coupled predesign code has been developed for optimizing the radial build arrangement of fusion breeding blanket. • Coupling module aims at speeding up the efficiency of design progress by coupling the neutronics calculation code with the thermal-hydraulic analysis code. • Radial build optimization algorithm aims at optimal arrangement of breeding blanket considering one or multiple specified objectives subject to the design criteria such as material temperature limit and available TBR. - Abstract: Fusion breeding blanket as one of the key in-vessel components performs the functions of breeding the tritium, removing the nuclear heat and heat flux from plasma chamber as well as acting as part of shielding system. The radial build design which determines the arrangement of function zones and material properties on the radial direction is the basis of the detailed design of fusion breeding blanket. For facilitating the radial build design, this study aims for developing a pre-design code to optimize the radial build of blanket with considering the performance of nuclear and thermal-hydraulic simultaneously. Two main features of this code are: (1) Coupling of the neutronics analysis with the thermal-hydraulic analysis to speed up the analysis progress; (2) preliminary optimization algorithm using one or multiple specified objectives subject to the design criteria in the form of constrains imposed on design variables and performance parameters within the possible engineering ranges. This pre-design code has been applied to the conceptual design of water-cooled ceramic breeding blanket in project of China fusion engineering testing reactor (CFETR).

  12. Terminology standardisation in the nuclear engineering field

    International Nuclear Information System (INIS)

    Kraut, A.

    1987-01-01

    Terminological standardisation is made for the purpose of unambiguous understanding, at least among experts in a given field of knowledge. The author explains a number of criteria and aspects to be taken into account in the process of standardisation by referring to the work of the Terminology Committee on Nuclear Engineering. He discusses the word formation in a technical language and the features of standardised terminology. Accepted terminology is a main factor in all procedures concerning design, testing, and approval and licensing of nuclear facilities, and also is of importance in terms of economics. (HP) [de

  13. Reactor physics for non-nuclear engineers

    International Nuclear Information System (INIS)

    Lewis, E.E.

    2011-01-01

    A one-term undergraduate course in reactor physics is described. The instructional format is strongly influenced by its intended audience of non-nuclear engineering students. In contrast to legacy treatments of the subject, the course focuses on the physics of nuclear power reactors with no attempt to include instruction in numerical methods. The multi-physics of power reactors is emphasized highlighting the close interactions between neutronic and thermal phenomena in design and analysis. Consequently, the material's sequencing also differs from traditional treatments, for example treating kinetics before the neutron diffusion is introduced. (author)

  14. Some radiation chemical aspects of nuclear engineering

    International Nuclear Information System (INIS)

    Pikaev, A.K.; Kabakchi, S.A.; Egorov, G.F.

    1988-01-01

    Some radiation chemical aspects of nuclear engineering are discussed (predominantly on the base of the works performed in the Soviet Union). The data on the influence of temperature within the range of 0-300 0 C on the yields of water radiolysis products are considered. The results obtained from the study of reactivity of actinide ions towards inorganic free radicals in acid aqueous solutions are summarized. The information on composition and properties of the products of radiolytic transformations of different extragents and diluents and on their influence on the behaviour of extraction systems during processing of irradiated nuclear fuel is presented. (author)

  15. Approaches to nontraditional delivery of nuclear engineering education

    International Nuclear Information System (INIS)

    Malaviya, B.K.

    1991-01-01

    At Rensselaer Polytechnic Institute, the faculty of the nuclear engineering and engineering physics department have, over the years, been involved in a variety of such approaches in response to the changing needs of nuclear industry personnel. A number of different types of short course and workshop programs have been developed and implemented both on and off campus in such areas as basic nuclear technology, reactor design computer codes and applications, nuclear power plant design and maintenance, reactor operations, health physics, modern developments in boiling heat transfer and two-phase flow, and probabilistic risk assessment. Customized coursed tailored to meet the particular needs of personnel in specialized areas can also be offered on specific industrial site locations, generally resulting in substantial savings of time as well as costs associated with tuition, travel, lodging. The Rensselaer Satellite Video Program (RSVP) brings the latest technological aids to the nontraditional delivery of courses and provides the facilities and opportunities for off-campus students and professional personnel to participate in regular academic programs and courses without leaving their industrial sites

  16. Development of nuclear rocket engine technology

    International Nuclear Information System (INIS)

    Gunn, S.V.

    1989-01-01

    Research sponsored by the Atomic Energy Commission, the USAF, and NASA (later on) in the area of nuclear rocket propulsion is discussed. It was found that a graphite reactor, loaded with highly concentrated Uranium 235, can be used to heat high pressure liquid hydrogen to temperatures of about 4500 R, and to expand the hydrogen through a high expansion ratio rocket nozzle assembly. The results of 20 reactor tests conducted at the Nevada Test Site between July 1959 and June 1969 are analyzed. On the basis of these results, the feasibility of solid graphite reactor/nuclear rocket engines is revealed. It is maintained that this technology will support future space propulsion requirements, using liquid hydrogen as the propellant, for thrust requirements ranging from 25,000 lbs to 250,000 lbs, with vacuum specific impulses of at least 850 sec and with full engine throttle capability. 12 refs

  17. Nuclear engineering education in italian universities

    International Nuclear Information System (INIS)

    Dulla, S.; Panella, B.; Ravetto, P.

    2011-01-01

    The paper illustrates the evolution and the present situation of the university-level nuclear engineering education in Italy. The problems connected with the need of qualified faculty in view of a dramatic increase of students is pointed out. A short description of the programs at present available at Italian universities is also presented, together with some statistics referred to Politecnico di Torino. The mathematical and computation content of each programs is also analyzed. (author)

  18. Nuclear Engineering Academic Programs Survey, 2003

    International Nuclear Information System (INIS)

    Science and Engineering Education, Oak Ridge Institute for Science and Education

    2004-01-01

    The survey includes degrees granted between September 1, 2002 and August 31, 2003. Thirty-three academic programs reported having nuclear engineering programs during the survey time period and all responded (100% response rate). Three of the programs included in last year's report were discontinued or out-of-scope in 2003. One new program has been added to the list. This year the survey data include U.S. citizenship, gender, and race/ethnicity by degree level

  19. Welding problems in nuclear power engineering

    International Nuclear Information System (INIS)

    Zubchenko, A.S.

    1986-01-01

    The problems of welding industry in nuclear power plant engineering, mainly related to the improvement of molten bath protection, are considered. Development of new materials for welding electrodes, for cladding and welding fluxes, is pointed out. Production of the following equipment is brought to a commercial level: welding heads and welding machines for branch pipe welding, anticorrosion cladding, zonal thermal treatment, electron beam welding facilities for the welding and maintenance of turbineblades, equipment for nondestructive testing of welded joints

  20. Safety analysis code input automation using the Nuclear Plant Data Bank

    International Nuclear Information System (INIS)

    Kopp, H.; Leung, J.; Tajbakhsh, A.; Viles, F.

    1985-01-01

    The Nuclear Plant Data Bank (NPDB) is a computer-based system that organizes a nuclear power plant's technical data, providing mechanisms for data storage, retrieval, and computer-aided engineering analysis. It has the specific objective to describe thermohydraulic systems in order to support: rapid information retrieval and display, and thermohydraulic analysis modeling. The Nuclear Plant Data Bank (NPBD) system fully automates the storage and analysis based on this data. The system combines the benefits of a structured data base system and computer-aided modeling with links to large scale codes for engineering analysis. Emphasis on a friendly and very graphically oriented user interface facilitates both initial use and longer term efficiency. Specific features are: organization and storage of thermohydraulic data items, ease in locating specific data items, graphical and tabular display capabilities, interactive model construction, organization and display of model input parameters, input deck construction for TRAC and RELAP analysis programs, and traceability of plant data, user model assumptions, and codes used in the input deck construction process. The major accomplishments of this past year were the development of a RELAP model generation capability and the development of a CRAY version of the code

  1. New trends in nuclear power engineering development

    International Nuclear Information System (INIS)

    Krasin, A.K.

    1974-01-01

    The specific features are considered of three designs of nuclear power plants with fast reactors: three-circuit nuclear power plant with liquid sodium as primary and secondary coolant, in the third circuit water vapor being used as turbine working medium, dual cycle nuclear power plant with pressurized helium as primary coolant and water vapor as turbine working medium, direct cycle nuclear power plant with a dissociating gas (nitrogen tetroxide N 2 O 4 ) as reactor coolant and turbine working medium. The version of the direct cycle nuclear power plant with dissociating N 2 O 4 was proposed and being developed by the Institute of Nuclear Engineering of the Academy of Sciencies of the BSSR. The thermal and physical properties of the dissociating gas allow a high-power-density reactor core to be used with a hard neutron spectra resulting in a high breeding ratio and a short doubling time. The pressure range from 150 to 170 bar was proven for this coolant under laboratory conditions and structural materials were chosen that ensure all the components of the direct cycle nuclear power plant to be workable. At present it is difficult to say which of the three versions is the most advantageous. The further development of a full-scale prototypes of a commercial nuclear power plant with a fast reactor and investigation of their technical and economic parameters remain the problems of utmost importance. A possible use of nuclear reactors is shortly considered for process heat production, in ferrous metallurgy, for hydrogen and new isotope production, and for radiation chemistry as well

  2. Educating nuclear engineers by nuclear science and technology master at UPM

    Energy Technology Data Exchange (ETDEWEB)

    Ahnert, C.; Minguez, E.; Perlado, M. [Universidad Politecnica de Madrid (Spain). Dept. de Ingenieria Nuclear; and others

    2014-05-15

    One of the main objectives of the Master on Nuclear Science and Technology implemented in the Universidad Politecnica de Madrid, is the training for the development of methodologies of simulation and advanced analysis necessary in research and in professional work in the nuclear field, for Fission Reactors and Nuclear Fusion, including fuel cycle and safety aspects. The students are able to use the current computational methodologies/codes for nuclear engineering that covers a difficult gap between nuclear reactor theory and simulations. Also they are able to use some facilities, as the Interactive Graphical Simulator of PWR power plant that is an optimal tool to transfer the knowledge of the physical phenomena that are involved in the nuclear power plants, from the nuclear reactor to the whole set of systems and equipment on a nuclear power plant. The new Internet reactor laboratory to be implemented will help to understand the Reactor Physics concepts. The experimental set-ups for neutron research and for coating fabrication offer new opportunities for training and research activities. All of them are relevant tools for motivation of the students, and to complete the theoretical lessons. They also follow the tendency recommended for the European Space for higher Education (Bologna) adapted studies. (orig.)

  3. Educating nuclear engineers by nuclear science and technology master at UPM

    International Nuclear Information System (INIS)

    Ahnert, C.; Minguez, E.; Perlado, M.

    2014-01-01

    One of the main objectives of the Master on Nuclear Science and Technology implemented in the Universidad Politecnica de Madrid, is the training for the development of methodologies of simulation and advanced analysis necessary in research and in professional work in the nuclear field, for Fission Reactors and Nuclear Fusion, including fuel cycle and safety aspects. The students are able to use the current computational methodologies/codes for nuclear engineering that covers a difficult gap between nuclear reactor theory and simulations. Also they are able to use some facilities, as the Interactive Graphical Simulator of PWR power plant that is an optimal tool to transfer the knowledge of the physical phenomena that are involved in the nuclear power plants, from the nuclear reactor to the whole set of systems and equipment on a nuclear power plant. The new Internet reactor laboratory to be implemented will help to understand the Reactor Physics concepts. The experimental set-ups for neutron research and for coating fabrication offer new opportunities for training and research activities. All of them are relevant tools for motivation of the students, and to complete the theoretical lessons. They also follow the tendency recommended for the European Space for higher Education (Bologna) adapted studies. (orig.)

  4. Nuclear thermal propulsion engine cost trade studies

    International Nuclear Information System (INIS)

    Paschall, R.K.

    1993-01-01

    The NASA transportation strategy for the Mars Exploration architecture includes the use of nuclear thermal propulsion as the primary propulsion system for Mars transits. It is anticipated that the outgrowth of the NERVA/ROVER programs will be a nuclear thermal propulsion (NTP) system capable of providing the propulsion for missions to Mars. The specific impulse (Isp) for such a system is expected to be in the 870 s range. Trade studies were conducted to investigate whether or not it may be cost effective to invest in a higher performance (Isp>870 s) engine for nuclear thermal propulsion for missions to Mars. The basic cost trades revolved around the amount of mass that must be transported to low-earth orbit prior to each Mars flight and the cost to launch that mass. The mass required depended on the assumptions made for Mars missions scenarios including piloted/cargo flights, number of Mars missions, and transit time to Mars. Cost parameters included launch cost, program schedule for development and operations, and net discount rate. The results were very dependent on the assumptions that were made. Under some assumptions, higher performance engines showed cost savings in the billions of dollars; under other assumptions, the additional cost to develop higher performance engines was not justified

  5. VACOSS - variable coding seal system for nuclear material control

    International Nuclear Information System (INIS)

    Kennepohl, K.; Stein, G.

    1977-12-01

    VACOSS - Variable Coding Seal System - is intended to seal: rooms and containers with nuclear material, nuclear instrumentation and equipment of the operator, instrumentation and equipment at the supervisory authority. It is easy to handle, reusable, transportable and consists of three components: 1. Seal. The light guide in fibre optics with infrared light emitter and receiver serves as lead. The statistical treatment of coded data given in the seal via adapter box guarantees an extremely high degree of access reliability. It is possible to store the data of two undue seal openings together with data concerning time and duration of the opening. 2. The adapter box can be used for input or input and output of data indicating the seal integrity. 3. The simulation programme is located in the computing center of the supervisory authority and permits to determine date and time of opening by decoding the seal memory data. (orig./WB) [de

  6. Nuclear Plant Analyzer development at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Laats, E.T.; Beelman, R.J.; Charlton, T.R.; Hampton, N.L.; Burtt, J.D.

    1985-01-01

    The Nuclear Plant Analyzer (NPA) is a state-of-the-art safety analysis and engineering tool being used to address key nuclear power plant safety issues. The NPA has been developed to integrate the NRC's computerized reactor behavior simulation codes such as RELAP5, TRAC-BWR, and TRAC-PWR, with well-developed computer graphics programs and large repositories of reactor design and experimental data. An important feature of the NAP is the capability to allow an analyst to redirect a RELAP5 or TRAC calculation as it progresses through its simulated scenario. The analyst can have the same power plant control capabilities as the operator of an actual plant. The NPA resides on the dual CDS Cyber-176 mainframe computers at the INEL and is being converted to operate on a Cray-1S computer at the LANL. The subject of this paper is the program conducted at the INEL

  7. Development of nuclear decay data library JDDL, and nuclear generation and decay calculation code COMRAD

    International Nuclear Information System (INIS)

    Naito, Yoshitaka; Ihara, Hitoshi; Katakura, Jun-ichi; Hara, Toshiharu.

    1986-08-01

    For safety evaluation of nuclear fuel facilities, a nuclear decay data library named JDDL and a computer code COMRAD have been developed to calculate isotopic composition of each nuclide, radiation source intensity, energy spectrum of γ-ray and neutron, and decay heat of spent fuel. JDDL has been produced mainly from the evaluated nuclear data file ENSDF to use new nuclear data. To supplement the data file for short life nuclides, the JNDC data set were also used which had been evaluated by Japan Nuclear Data Committee. Using these data, calculations became possible from short period to long period after irradiation. (author)

  8. Infiltration of quality concepts in nuclear engineering education

    International Nuclear Information System (INIS)

    Woodall, D.M.

    1993-01-01

    The principles of total quality management (TQM) have been applied increasingly in the nuclear power industry over the last decade. The involvement of industrial professionals on the advisory boards of engineering colleges and departments has increasingly led in recent years to the recommendation that TQM be applied as appropriate to engineering education. This paper describes the concepts of TQM in their application to engineering education, specifically in the nuclear engineering area. A summary of the concerns expressed by nuclear engineering academics, as well as the record of successful implementation of TQM in the nuclear engineering education environment is provided in this paper

  9. Thermohydraulic modeling of nuclear thermal rockets: The KLAXON code

    International Nuclear Information System (INIS)

    Hall, M.L.; Rider, W.J.; Cappiello, M.W.

    1992-01-01

    The hydrogen flow from the storage tanks, through the reactor core, and out the nozzle of a Nuclear Thermal Rocket is an integral design consideration. To provide an analysis and design tool for this phenomenon, the KLAXON code is being developed. A shock-capturing numerical methodology is used to model the gas flow (the Harten, Lax, and van Leer method, as implemented by Einfeldt). Preliminary results of modeling the flow through the reactor core and nozzle are given in this paper

  10. Nuclear densimeter of soil simulated in MCNP-4C code

    International Nuclear Information System (INIS)

    Braga, Mario R.M.S.S.; Penna, Rodrigo; Vasconcelos, Danilo C.; Pereira, Claubia; Guerra, Bruno T.; Silva, Clemente J.G.C.

    2009-01-01

    The Monte Carlo code (MCNPX) was used to simulate a nuclear densimeter for measuring soil density. An Americium source (E = 60 keV) and a NaI (Tl) detector were placed on soil surface. Results from MCNP shown that scattered photon fluxes may be used to determining soil density. Linear regressions between scattered photons fluxes and soil density were calculated and shown correlation coefficients near unity. (author)

  11. Use of computer codes to improve nuclear power plant operation

    International Nuclear Information System (INIS)

    Misak, J.; Polak, V.; Filo, J.; Gatas, J.

    1985-01-01

    For safety and economic reasons, the scope for carrying out experiments on operational nuclear power plants (NPPs) is very limited and any changes in technical equipment and operating parameters or conditions have to be supported by theoretical calculations. In the Nuclear Power Plant Scientific Research Institute (NIIAEhS), computer codes are systematically used to analyse actual operating events, assess safety aspects of changes in equipment and operating conditions, optimize the conditions, preparation and analysis of NPP startup trials and review and amend operating instructions. In addition, calculation codes are gradually being introduced into power plant computer systems to perform real time processing of the parameters being measured. The paper describes a number of specific examples of the use of calculation codes for the thermohydraulic analysis of operating and accident conditions aimed at improving the operation of WWER-440 units at the Jaslovske Bohunice V-1 and V-2 nuclear power plants. These examples confirm that computer calculations are an effective way of solving operating problems and of further increasing the level of safety and economic efficiency of NPP operation. (author)

  12. The impact of changing computing technology on EPRI [Electric Power Research Institute] nuclear analysis codes

    International Nuclear Information System (INIS)

    Breen, R.J.

    1988-01-01

    The Nuclear Reload Management Program of the Nuclear Power Division (NPD) of the Electric Power Research Institute (EPRI) has the responsibility for initiating and managing applied research in selected nuclear engineering analysis functions for nuclear utilities. The computer systems that result from the research projects consist of large FORTRAN programs containing elaborate computational algorithms used to access such areas as core physics, fuel performance, thermal hydraulics, and transient analysis. This paper summarizes a study of computing technology trends sponsored by the NPD. The approach taken was to interview hardware and software vendors, industry observers, and utility personnel focusing on expected changes that will occur in the computing industry over the next 3 to 5 yr. Particular emphasis was placed on how these changes will impact engineering/scientific computer code development, maintenance, and use. In addition to the interviews, a workshop was held with attendees from EPRI, Power Computing Company, industry, and utilities. The workshop provided a forum for discussing issues and providing input into EPRI's long-term computer code planning process

  13. Radioactive releases of nuclear power plants: the code ASTEC

    International Nuclear Information System (INIS)

    Sdouz, G.; Pachole, M.

    1999-11-01

    In order to adopt potential countermeasures to protect the population during the course of an accident in a nuclear power plant a fast prediction of the radiation exposure is necessary. The basic input value for such a dispersion calculation is the source term, which is the description of the physical and chemical behavior of the released radioactive nuclides. Based on a source term data base a pilot system has been developed to determine a relevant source term and to generate the input file for the dispersion code TAMOS of the Zentralanstalt fuer Meteorologie und Geodynamik (ZAMG). This file can be sent directly as an attachment of e-mail to the TAMOS user for further processing. The source terms for 56 European nuclear power plant units are included in the pilot version of the code ASTEC (Austrian Source Term Estimation Code). The use of the system is demonstrated in an example based on an accident in the unit TEMELIN-1. In order to calculate typical core inventories for the data bank the international computer code OBIGEN 2.1 was installed and applied. The report has been completed with a discussion on the optimal data transfer. (author)

  14. MODLIB, library of Fortran modules for nuclear reaction codes

    International Nuclear Information System (INIS)

    Talou, Patrick

    2006-01-01

    1 - Description of program or function: ModLib is a library of Fortran (90-compatible) modules to be used in existing and future nuclear reaction codes. The development of the library is an international effort being undertaken under the auspices of the long-term Subgroup A of the OECD/NEA Working Party on Evaluation and Cooperation. The aim is to constitute a library of well-tested and well-documented pieces of codes that can be used with confidence in all our coding efforts. This effort will undoubtedly help avoid the duplication of work, and most certainly facilitate the very important inter-comparisons between existing codes. 2 - Methods: - Width f luctuations [Talou, Chadwick]: calculates width fluctuation correction factors (output) for a set of transmission coefficients (input). Three methods are available: HRTW, Moldauer, and Verbaarschot (also called GOE approach). So far, no distinction is made according to the type of the coefficients channels (particle emission, gamma-ray emission, fission). - Gamma s trength [Herman]: calculates gamma-ray transmission coefficients using a Giant Resonance formalism. - Level d ensity [Koning]: computes the Gilbert-Cameron-Ignatyuk formalism for the continuum nuclear level density. - CHECKR, FIZCON, INTER, PSYCHE, STANEF [Dunford]: these modules are used in the MODLIB project but are not included in this package. They are available from the NEA Data Bank Computer Program Service under Package Ids: CHECKR (USCD1208), FIZCON (USCD1209), INTER (USCD1212), PSYCHE (USCD1216), STANEF (USCD1218)

  15. A study on the nuclear computer code maintenance and management system

    International Nuclear Information System (INIS)

    Kim, Yeon Seung; Huh, Young Hwan; Lee, Jong Bok; Choi, Young Gil; Suh, Soong Hyok; Kang, Byong Heon; Kim, Hee Kyung; Kim, Ko Ryeo; Park, Soo Jin

    1990-12-01

    According to current software development and quality assurance trends. It is necessary to develop computer code management system for nuclear programs. For this reason, the project started in 1987. Main objectives of the project are to establish a nuclear computer code management system, to secure software reliability, and to develop nuclear computer code packages. Contents of performing the project in this year were to operate and maintain computer code information system of KAERI computer codes, to develop application tool, AUTO-i, for solving the 1st and 2nd moments of inertia on polygon or circle, and to research nuclear computer code conversion between different machines. For better supporting the nuclear code availability and reliability, assistance from users who are using codes is required. Lastly, for easy reference about the codes information, we presented list of code names and information on the codes which were introduced or developed during this year. (Author)

  16. Summary of ground water and surface water flow and contaminant transport computer codes used at the Idaho National Engineering Laboratory (INEL)

    International Nuclear Information System (INIS)

    Bandy, P.J.; Hall, L.F.

    1993-03-01

    This report presents information on computer codes for numerical and analytical models that have been used at the Idaho National Engineering Laboratory (INEL) to model ground water and surface water flow and contaminant transport. Organizations conducting modeling at the INEL include: EG ampersand G Idaho, Inc., US Geological Survey, and Westinghouse Idaho Nuclear Company. Information concerning computer codes included in this report are: agency responsible for the modeling effort, name of the computer code, proprietor of the code (copyright holder or original author), validation and verification studies, applications of the model at INEL, the prime user of the model, computer code description, computing environment requirements, and documentation and references for the computer code

  17. Targeted initiatives. Support for nuclear engineering education in the USA

    International Nuclear Information System (INIS)

    Gutteridge, John

    2001-01-01

    Recruitment and education of a new generation of nuclear engineers stands to benefit in the USA from a range of programmes involving governmental bodies, universities, and industry groups. They are part of efforts to attract more students to consider and prepare for careers in the nuclear industry, and to provide financial support for nuclear research and education. Career prospects in the nuclear field are brightening. The demand for nuclear engineers and nuclear trained personnel is on the rise as the new century opens. During the past year several studies were completed in an attempt to ascertain the problems in nuclear engineering education and define initiatives to address these problems

  18. Code of practice for radiation protection in nuclear medicine

    International Nuclear Information System (INIS)

    Hamed, M. I.

    2010-05-01

    In aim of this study was to develop a draft for a new code practice for radiation protection in nuclear medicine that meets the current relevant international recommendation. The draft includes the following main fields: methods of radiation protection for workers, patients and public. Also, the principles of safe design of nuclear medicine departments, quality assurance program, proper manipulation of radiation sources including radioactive waste and emergency preparedness and response. The practical part of this study includes inspections of three nuclear medicine departments available in Sudan so as to assess the degree of compliance of those departments with what is stated in this code. The inspection missions have been conducted using a checklist that addresses all items that may affect radiation raincoat issues in addition to per formin area radiation monitoring around the installation of the radioactive sources. The results of this revealed that most of the departments do not have effective radiation protection program which in turn could lead to unnecessary exposure to patients, public and workers. Finally, some recommendations are given that - if implemented - could improve the status of radiation protection in nuclear medicine department. (Author)

  19. Nuclear component design ontology building based on ASME codes

    International Nuclear Information System (INIS)

    Bao Shiyi; Zhou Yu; He Shuyan

    2005-01-01

    The adoption of ontology analysis in the study of concept knowledge acquisition and representation for the nuclear component design process based on computer-supported cooperative work (CSCW) makes it possible to share and reuse numerous concept knowledge of multi-disciplinary domains. A practical ontology building method is accordingly proposed based on Protege knowledge model in combination with both top-down and bottom-up approaches together with Formal Concept Analysis (FCA). FCA exhibits its advantages in the way it helps establish and improve taxonomic hierarchy of concepts and resolve concept conflict occurred in modeling multi-disciplinary domains. With Protege-3.0 as the ontology building tool, a nuclear component design ontology based ASME codes is developed by utilizing the ontology building method. The ontology serves as the basis to realize concept knowledge sharing and reusing of nuclear component design. (authors)

  20. French administrative practice and design codes for nuclear vessels

    International Nuclear Information System (INIS)

    Roche, R.L.

    1987-07-01

    French regulations on boilers and pressure vessels have prevailed for a very long time, the first measure having been promulgated on 29 October 1823. Restraining the attention to nuclear pressure vessels it must be pointed out regulations and enforcement by public authorities are more stringent than they are for conventional pressure vessels. The first part of this paper will be devoted to regulations with a special attention to the decree of 26 February 1974 and to the practice of public authorities in this field with special attention given to the Bureau de Controle de la Construction Nucleaire (BCCN = Bureau of Inspection of Nuclear Design and Manufacturing). The second part of this paper will deal with the French construction codes for nuclear components RCC-M (water reactors) and RCC-MR (elevated temperature design)

  1. Future direction of ASME nuclear codes and standards

    International Nuclear Information System (INIS)

    Ennis, Kevin; Sheehan, Mark E.

    2003-01-01

    While the nuclear power industry in the US is in a period of stasis, there continues to be a great deal of activity in the ASME nuclear standards development arena. As plants age, the need for new approaches in standardization changes with the changing needs of the industry. New tools are becoming available in the form of risk analysis, and this is finding its way into more and more of ASME's standards activities. This paper will take a look at the direction that ASME nuclear Codes and Standards are heading in this and other areas, as well as taking a look at some advance reactor concepts and plans for standards to address new technologies

  2. Comparison of GLIMPS and HFAST Stirling engine code predictions with experimental data

    Science.gov (United States)

    Geng, Steven M.; Tew, Roy C.

    1992-01-01

    Predictions from GLIMPS and HFAST design codes are compared with experimental data for the RE-1000 and SPRE free piston Stirling engines. Engine performance and available power loss predictions are compared. Differences exist between GLIMPS and HFAST loss predictions. Both codes require engine specific calibration to bring predictions and experimental data into agreement.

  3. Comparison of GLIMPS and HFAST Stirling engine code predictions with experimental data

    International Nuclear Information System (INIS)

    Geng, S.M.; Tew, R.C.

    1994-01-01

    Predictions from GLIMPS and HFAST design codes are compared with experimental data for the RE-1000 and SPRE free-piston Stirling engines. Engine performance and available power loss predictions are compared. Differences exist between GLIMPS and HFAST loss predictions. Both codes require engine-specific calibration to bring predictions and experimental data into agreement

  4. 4+D digital engineering for advanced nuclear energy systems

    International Nuclear Information System (INIS)

    Jeong, S. G.; Suh, K. Y.; Nam, S. K.

    2007-01-01

    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 automated way of managing the information flow spanning their life cycle. In line with practice in disciplines of naval architecture, aerospace engineering, and automotive manufacturing, the paper proposes total digital systems engineering based on three-dimensional (3D) computer-aided design (CAD) models. The signature in the proposal lies with the four-plus-dimensional (4 + D) Technology T M, a critical know-how for digital management. The so-called OPIUM (Optimized Plant Integrated Ubiquitous Management) features a 4 + D Technology T M for nuclear energy systems engineering. The technology proposed in the 3D space and time plus cost coordinates, i.e. 4 + D, is the backbone of digital engineering in the nuclear systems design and management. Based on an integrated 3D configuration management system, OPIUM consists of solutions NOTUS (Nuclear Optimization Technique Ubiquitous System), VENUS (Virtual Engineering Nuclear Ubiquitous System), INUUS (Informatics Nuclear Utilities Ubiquitous System), JANUS (Junctional Analysis Numerical Ubiquitous System) and EURUS (Electronic Unit Research Ubiquitous System). These solutions will help initial simulation capability for NPPs to supply the crucial information. 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

  5. The Code Aster: a product for mechanical engineers

    International Nuclear Information System (INIS)

    Levesque, J.R.

    1998-01-01

    The Code Aster is a 2D or 3D structural finite element software: analysis of structures and thermo-mechanics for evaluation and research with linear for non linear modelling. Since 1989, it has been the host structure that capitalizes on developments made by the Research and Development Division in the field of numerical modelling in structural mechanics, and user experience feedback. It is an industrial design tool, particularly for engineering of facilities in operation and for the evaluation of new projects. This software was developed using a quality Assurance methodology with independent validation. Upgrades to this product are guided by the objective of satisfying the needs of expertise studies, attempting to make functions coherent and complete. (author)

  6. 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.

  7. 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.

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

    International Nuclear Information System (INIS)

    Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E.; Tills, J.

    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

  9. Developing safety culture in nuclear power engineering

    International Nuclear Information System (INIS)

    Tevlin, S.A.

    2000-01-01

    The new issue (no. 11) of the IAEA publications series Safety Reports, devoted to the safety culture in nuclear engineering Safety culture development in the nuclear activities. Practical recommendations to achieve success, is analyzed. A number of recommendations of international experts is presented and basic general indicators of satisfactory and insufficient safety culture in the nuclear engineering are indicated. It is shown that the safety culture has two foundations: human behavior and high quality of the control system. The necessity of creating the confidence by the management at all levels of the enterprise, development of individual initiative and responsibility of the workers, which make it possible to realize the structural hierarchic system, including technical, human and organizational constituents, is noted. Three stages are traced in the process of introducing the safety culture. At the first stage the require,emts of scientific-technical documentation and provisions of the governmental, regional and control organs are fulfilled. At the second stage the management of the organization accepts the safety as an important direction in its activities. At the third stage the organization accomplishes its work, proceeding from the position of constant safety improvement. The general model of the safety culture development is considered [ru

  10. Impact of quality concepts on nuclear engineering accreditation

    International Nuclear Information System (INIS)

    Woodall, D.M.

    1993-01-01

    This paper is an update of the accreditation process for nuclear engineering education at the undergraduate and graduate level in U.S. universities and colleges. The Engineering Accreditation Commission (EAC) of the Accreditation Board for Engineering and Technology (ABET) has made a number of major changes in the process for engineering accreditation in recent years. This paper identifies those changes that have taken place, discusses the rationale for those changes, and encourages U.S. universities with nuclear engineering programs to respond

  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. Engineering thermal engine rocket adventurer for space nuclear application

    International Nuclear Information System (INIS)

    Nam, Seung H.; Suh, Kune Y.; Kang, Seong G.

    2008-01-01

    The conceptual design for the first-of-a-kind engineering of Thermal Engine Rocket Adventure (TERA) is described. TERA comprising the Battery Omnibus Reactor Integral System (BORIS) as the heat resource and the Space Propulsion Reactor Integral System (SPRIS) as the propulsion system, is one of the advanced Nuclear Thermal Rocket (NTR) engine utilizing hydrogen (H 2 ) propellant being developed at present time. BORIS in this application is an open cycle high temperature gas cooled reactor that has eighteen fuel elements for propulsion and one fuel element for electricity generation and propellant pumping. Each fuel element for propulsion has its own small nozzle. The nineteen fuel elements are arranged into hexagonal prism shape in the core and surrounded by outer Be reflector. The TERA maximum power is 1,000 MW th , specific impulse 1,000 s, thrust 250,000 N, and the total mass is 550 kg including the reactor, turbo pump and auxiliaries. Each fuel element comprises the fuel assembly, moderators, pressure tube and small nozzle. The TERA fuel assembly is fabricated of 93% enriched 1.5 mm (U, Zr, Nb)C wafers in 25.3% voided Square Lattice Honeycomb (SLHC). The H 2 propellant passes through these flow channels. This study is concerned with thermohydrodynamic analysis of the fuel element for propulsion with hypothetical axial power distribution because nuclear analysis of TERA has not been performed yet. As a result, when the power distribution of INSPI's M-SLHC is applied to the fuel assembly, the local heat concentration of fuel is more serious and the pressure of the initial inlet H 2 is higher than those of constant average power distribution applied. This means the fuel assembly geometry of 1.5 mm fuel wafers and 25.3% voided SLHC needs to be changed in order to reduce thermal and mechanical shocks. (author)

  13. Nuclear science and engineering education at a university research reactor

    International Nuclear Information System (INIS)

    Loveland, W.

    1993-01-01

    The role of an on-site irradiation facility in nuclear science and engineering education is examined. Using the example of a university research reactor, the use of such devices in laboratory instruction, public outreach programs, special instructional programs, research, etc. is discussed. Examples from the Oregon State University curriculum in nuclear chemistry, nuclear engineering and radiation health are given. (author) 1 tab

  14. Civil engineering in the nuclear industry

    International Nuclear Information System (INIS)

    Dexter-Smith, R.

    1991-01-01

    Civil Engineering has an important contribution to make at every stage of the nuclear fuel cycle, from the choice of site and conception of the design of a major power station or fuel plan, through modifications during modifications, during operation, to the final stages of designing and building waste management stores and repositories and the decommissioning of stations and plants. The conference papers published here -twenty four in total - cover many of these stages. All the papers are indexed separately. Two international papers are presented, one on French PWRs, the other on repository design. Four papers look at site investigations, four are concerned with earthquake engineering, four with structural analysis, three with quality assurance, three with design and four with in-service performance and decommissioning. (UK)

  15. Engineering design guidelines for nuclear criticality safety

    International Nuclear Information System (INIS)

    Waltz, W.R.

    1988-08-01

    This document provides general engineering design guidelines specific to nuclear criticality safety for a facility where the potential for a criticality accident exists. The guide is applicable to the design of new SRP/SRL facilities and to major modifications Of existing facilities. The document is intended an: A guide for persons actively engaged in the design process. A resource document for persons charged with design review for adequacy relative to criticality safety. A resource document for facility operating personnel. The guide defines six basic criticality safety design objectives and provides information to assist in accomplishing each objective. The guide in intended to supplement the design requirements relating to criticality safety contained in applicable Department of Energy (DOE) documents. The scope of the guide is limited to engineering design guidelines associated with criticality safety and does not include other areas of the design process, such as: criticality safety analytical methods and modeling, nor requirements for control of the design process

  16. Malignant pleural mesothelioma in a nuclear engineer

    International Nuclear Information System (INIS)

    Huncharek, M.

    1988-01-01

    Malignant pleural mesothelioma accounts for a large proportion of deaths among occupational cohorts exposed to asbestos. Of particular interest are recent reports of a high risk of mesothelioma among occupational groups previously thought to be at low risk for developing this neoplasm. In the present report we present a case of pleural mesothelioma associated with bystander exposure to asbestos in a nuclear engineer. To our knowledge, this is the first report of the disease occurring in a member of this occupational group after work related exposure to asbestos. (author)

  17. Nuclear reactor safety: physics and engineering aspects

    International Nuclear Information System (INIS)

    Kinchin, G.H.

    1982-01-01

    In order to carry out the sort of probabilistic analysis referred to by Farmer (Contemp. Phys.; 22:349(1981)), it is necessary to have a good understanding of the processes involved in both normal and accident conditions in a nuclear reactor. Some of these processes, for a variety of different reactor systems, are considered in sections dealing with the neutron chain reaction, the removal of heat from the reactor, material problems, reliability of protective systems and a number of specific topics of particular interest from the point of view of physics or engineering. (author)

  18. Passport of global nuclear business. ASME code certificate acquirement and inspection practices

    International Nuclear Information System (INIS)

    Kawabata, Hiroyuki; Terajima, Makoto; Anami, Kazuhiro

    2010-01-01

    There are possibilities of Japanese nuclear industries to participate in global business such as new and additional construction of nuclear power plants in US and also Asian and other developing countries in the world. It is requisite to acquire ASME code certificate for global business participation, just as passport. This article consists of five papers on present status of ASME code certificate acquirement and inspection practices of nuclear components vendors in the area of Japanese nuclear business. Activities of JSME Committee on Power Generation Facility Codes to make JSME codes corresponded to ASME nuclear codes and standards for their international deployment are also described. (T. Tanaka)

  19. Coded aperture imaging system for nuclear fuel motion detection

    International Nuclear Information System (INIS)

    Stalker, K.T.; Kelly, J.G.

    1980-01-01

    A Coded Aperature Imaging System (CAIS) has been developed at Sandia National Laboratories to image the motion of nuclear fuel rods undergoing tests simulating accident conditions within a liquid metal fast breeder reactor. The tests require that the motion of the test fuel be monitored while it is immersed in a liquid sodium coolant precluding the use of normal optical means of imaging. However, using the fission gamma rays emitted by the fuel itself and coded aperture techniques, images with 1.5 mm radial and 5 mm axial resolution have been attained. Using an electro-optical detection system coupled to a high speed motion picture camera a time resolution of one millisecond can be achieved. This paper will discuss the application of coded aperture imaging to the problem, including the design of the one-dimensional Fresnel zone plate apertures used and the special problems arising from the reactor environment and use of high energy gamma ray photons to form the coded image. Also to be discussed will be the reconstruction techniques employed and the effect of various noise sources on system performance. Finally, some experimental results obtained using the system will be presented

  20. MASTER- an indigenous nuclear design code of KAERI

    International Nuclear Information System (INIS)

    Cho, Byung Oh; Lee, Chang Ho; Park, Chan Oh; Lee, Chong Chul

    1996-01-01

    KAERI has recently developed the nuclear design code MASTER for the application to reactor physics analyses for pressurized water reactors. Its neutronics model solves the space-time dependent neutron diffusion equations with the advanced nodal methods. The major calculation categories of MASTER consist of microscopic depletion, steady-state and transient solution, xenon dynamics, adjoint solution and pin power and burnup reconstruction. The MASTER validation analyses, which are in progress aiming to submit the Uncertainty Topical Report to KINS in the first half of 1996, include global reactivity calculations and detailed pin-by-pin power distributions as well as in-core detector reaction rate calculations. The objective of this paper is to give an overall description of the CASMO/MASTER code system whose verification results are in details presented in the separate papers

  1. Results of the Nonelastic Reaction Code Brieff for Nuclear Data

    International Nuclear Information System (INIS)

    Duarte, H.

    2009-01-01

    We present recent changes in our nonelastic reaction code BRIEFF and especially in the fast stage of reaction described by the intranuclear cascade (INC) code BRIC. Distributions and excitation functions of residual nuclei production cross sections are shown for proton-induced reaction on target nuclei. Slight improvements are seen in the proton-induced reaction on light nuclei with a closed shell when the energy levels are taken into account in the INC stage. On the other hand, fission gives poor results in the current version. To compare to other nuclear models and LA150 libraries, BRIEFF has been incorporated into MCNPX 2.5.0. Examples of neutron production from thick target irradiation by proton beams between 30 and 350 MeV are presented. Except for some discrepancies, a good agreement with data is obtained on average. (authors)

  2. Do nuclear engineering educators have a special responsibility

    International Nuclear Information System (INIS)

    Weinberg, A.M.

    1977-01-01

    Each 1000 MW(e) reactor in equilibrium contains 15 x 10 9 Ci of radioactivity. To handle this material safety requires an extremely high level of expertise and commitment - in many respects, an expertise that goes beyond what is demanded of any other technology. If one grants that nuclear engineering is more demanding than other engineering because the price of failure is greater, one must ask how can we inculcate into the coming generations of nuclear engineers a full sense of the responsibility they bear in practising their profession. Clearly a first requirement is that all elements of the nuclear community -utility executives, equipment engineers, operating engineers, nuclear engineers, administrators - must recognize and accept the idea that nuclear energy is something special, and that therefore its practitioners must be special. This sense must be instilled into young nuclear engineers during their education. A special responsibility therefore devolves upon nuclear engineering educators: first, to recognize the special character of their profession, and second, to convey this sense to their students. The possibility of institutionalizing this sense of responsibility by establishing a nuclear Hippocratic Oath or special canon of ethics for nuclear engineers ought to be discussed within the nuclear community. (author)

  3. MITEE: A new nuclear engine concept for ultra fast, lightweight solar system exploration missions

    Science.gov (United States)

    Powell, James; Paniagua, John; Ludewig, Hans; Maise, George; Todosow, Michael

    1998-01-01

    A new ultra compact nuclear engine concept, MITEE (MIniature R_eactor E_nginE_), is described, and its performance evaluated for various solar system exploration missions. The MITEE concept is based on the Particle Bed Reactor (PBR), with modifications that enable a smaller, lighter nuclear engine. A range of MITEE Engine designs is described. Representative design parameters for the baseline MITEE reactor are: 75MW(th) power level, 1000 second Isp, 100 kilogram mass, 10 MW/Liter fuel element power density, 39 cm core diameter/height. Total engine mass, including turbo pump assembly, nozzles, controls, and contingency, is estimated to be 200 kilograms. Using the MITEE engine, ultra fast, lightweight solar system exploration missions are enabled. A range of such missions has been analyzed using the MULIMP code, and are described.

  4. Environmental and waste disposal options in nuclear engineering curricula

    International Nuclear Information System (INIS)

    Elleman, T.S.; Gilligan, J.G.

    1991-01-01

    The strong national emphasis on waste and environmental issues has prompted increasing interest among nuclear engineering students in study options that will prepare them for careers in these areas. Student interest appears to focus principally on health physics, radioactive waste disposal, and environmental interactions with radionuclides. One motivation for this interest appears to be the growing national programs in environmental restoration and waste remediation that have produced fellowship support for nuclear engineering students as well as employment opportunities. Also, the recent National Academy of sciences study on nuclear engineering education specifically emphasized the importance of expanding nuclear engineering curricula and research programs to include a greater emphasis on radioactive waste and environmental issues. The North Carolina State University (NCSU) Department of Nuclear Engineering is attempting to respond to these needs through the development of course options that will allow students to acquire background in environmental subjects as a complement to the traditional nuclear engineering education

  5. IAEA codes and guides for safety of nuclear power plants

    International Nuclear Information System (INIS)

    Raisic, N.

    1980-01-01

    The objectives and scope of the Agency's programme of nuclear safety standards are described and the role of these documents in regulation of nuclear power im Member States is discussed. For each of the five areas of safety standards development, i.e. siting, design, operation, quality assurance and governmental organization, a set of principles underlying requirements and recommendations contained in the Code of Practice and Safety Guides will be presented. Safety Guides in each of the five areas will be reviewed in respect of the scope and content. A consideration will be given to the future development of the safety standards and to the revision and updating of the published documents. (orig./RW)

  6. Investigation of knowledge structure of nuclear data evaluation code

    International Nuclear Information System (INIS)

    Uenaka, Junji; Kambayashi, Shaw

    1988-08-01

    In this report, investigation results of knowledge structure in a nuclear data evaluation code are described. This investigation is related to the natural language processing and the knowledge base in the research theme of Human Acts Simulation Program (HASP) begun at the Computing Center of JAERI in 1987. By using a machine translation system, an attempt has been made to extract a deep knowledge from Japanese sentences which are equivalent to a FORTRAN program CASTHY for nuclear data evaluation. With the knowledge extraction method used by the authors, the verification of knowledge is more difficult than that of the prototyping method in an ordinary AI technique. In the early stage of building up a knowledge base system, it seems effective to extract and examine knowledge fragments of limited objects. (author)

  7. Engineering Ethics Education Having Reflected Various Values and a Global Code of Ethics

    Science.gov (United States)

    Kanemitsu, Hidekazu

    At the present day, a movement trying to establish a global code of ethics for science and engineering is in activity. The author overviews the context of this movement, and examines the possibility of engineering ethics education which uses global code of ethics. In this paper, the engineering ethics education which uses code of ethics in general will be considered, and an expected function of global code of ethics will be also. Engineering ethics education in the new century should be aimed to share the values among different countries and cultures. To use global code of ethics as a tool for such education, the code should include various values, especially Asian values which engineering ethics has paid little attention to.

  8. Nuclear engineering questions: power, reprocessing, waste, decontamination, fusion

    International Nuclear Information System (INIS)

    Walton, R.D. Jr.

    1979-01-01

    This volume contains papers presented at the chemical engineering symposium on nuclear questions. Specific questions addressed by the speakers included: nuclear power - why and how; commercial reprocessing - permanent death or resurrection; long-term management of commercial high-level wastes; long-term management of defense high-level waste; decontamination and decommissioning of nuclear facilities, engineering aspects of laser fusion I; and engineering aspects of laser fusion II. Individual papers have been input to the Energy Data Base previously

  9. The Nuclear Review: the Institution of Nuclear Engineers' response to the Review of Nuclear Power

    International Nuclear Information System (INIS)

    Anon.

    1994-01-01

    The United Kingdom Government's Nuclear Review currently underway, addresses whether and in what form nuclear power should continue to be part of the country's power generation capability. This article sets out the response of the Institution of Nuclear Engineers to the Nuclear Review. This pro-nuclear group emphasises the benefits to be gained from diversity of generation in the energy supply industry. The environmentally benign nature of nuclear power is emphasised, in terms of gaseous emissions. The industry's excellent safety record also argues in favour of nuclear power. Finally, as power demand increases globally, a health U.K. nuclear industry could generate British wealth through power exports and via the construction industry. The Institution's view on radioactive waste management is also set out. (UK)

  10. Centre for nuclear engineering University of Toronto annual report 1984

    International Nuclear Information System (INIS)

    1984-12-01

    The annual report of the Centre for Nuclear Engineering, University of Toronto covers the following subjects: message from the Dean; Chairman's message; origins of the centre; formation of the centre; new nuclear appointments; and activities of the centre, 1984

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

    International Nuclear Information System (INIS)

    1993-03-01

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

  12. Development of human factors engineering guide for nuclear power project

    International Nuclear Information System (INIS)

    Wu Dangshi; Sheng Jufang

    1997-01-01

    'THE PRACTICAL GUIDE FOR APPLICATION OF HUMAN FACTORS ENGINEERING TO NUCLEAR POWER PROJECT (First Draft, in Chinese)', which was developed under a research program sponsored by National Nuclear Safety Administration (NNSA) is described briefly. It is hoped that more conscious, more systematical and more comprehensive application of Human Factors Engineering to the nuclear power projects from the preliminary feasibility studies up to the commercial operation will benefit the safe, efficient and economical operations of nuclear power plants in China

  13. Study on the evolution of nuclear engineering professions

    International Nuclear Information System (INIS)

    2008-01-01

    Based on interviews of experts belonging to different companies and institutions (EDF, AREVA, CEA, ASN, IRSN, INSTN), subcontractors, engineers and technicians of the nuclear sector, persons in charge of education, pupils and students, this study gives a synthetic vision of the general context of the needs for nuclear engineering professionals, at the world scale, in the French context, the perceived difficulties faced by this sector, the use of subcontracting, the recruitment needs, the educational profile of engineers and technicians, their revenues, their opinion about their work, the adequacy between education and employment in this sector. It gives estimated figures for engineer and technician recruitment needs for different abilities in the French nuclear engineering

  14. A comprehensive program of nuclear engineering and science education

    International Nuclear Information System (INIS)

    Bereznai, G.; Lewis, B.

    2014-01-01

    The University of Ontario Institute of Technology offers undergraduate degrees in nuclear engineering, nuclear power, health physics and radiation science, graduate degrees (masters as well as doctorate) in nuclear engineering, and graduate diplomas that encompass a wide range of nuclear engineering and technology topics. Professional development programs tailored to specific utility needs are also offered, and the sharing of course material between the professional development and university education courses has strengthened both approaches to ensuring the high qualification levels required of professionals in the nuclear industry. (author)

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

    International Nuclear Information System (INIS)

    Philip, Bobby; Clarno, Kevin T.; Cochran, Bill

    2010-01-01

    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.

  16. 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)

  17. 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)

  18. 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)

  19. Nuclear science and engineering education at a university research reactor

    International Nuclear Information System (INIS)

    Loveland, W.

    1990-01-01

    The research and teaching operations of the Nuclear Chemistry Division of the Dept. of Chemistry and the Dept. of Nuclear Engineering are housed at the Oregon State University Radiation Center. This facility which includes a 1.1 MW TRIGA reactor was used for 53 classes from a number of different academic departments last year. About one-half of these classes used the reactor and ∼25% of the reactor's 45 hour week was devoted to teaching. Descriptions will be given of reactor-oriented instructional programs in nuclear engineering, radiation health and nuclear chemistry. In nuclear chemistry, classes in (a) nuclear chemistry for nuclear engineers, (b) radiotracer methods, (c) elementary and advanced activation analysis, and (d) advanced nuclear instrumentation will be described in detail. The use of the facility to promote general nuclear literacy among college students, high school and grade school students and the general population will also be covered

  20. THYDE-NEU: Nuclear reactor system analysis code

    International Nuclear Information System (INIS)

    Asahi, Yoshiro

    2002-03-01

    THYDE-NEU is applicable not only to transient analyses, but also to steady state analyses of nuclear reactor systems (NRSs). In a steady state analysis, the code generates a solution satisfying the transient equations without external disturbances. In a transient analysis, the code calculates temporal NRS behaviors in response to various external disturbances in such a way that mass and energy of the coolant as well as the number of neutrons conserve. The first half of the report is the description of the methods and models for use in the THYDE-NEU code, i.e., (1) the thermal-hydraulic network model, (2) the spatial kinetics model, (3) the heat sources in fuel, (4) the heat transfer correlations, (5) the mechanical behavior of clad and fuel, and (6) the steady state adjustment. The second half of the report is the users' mannual containing the items; (1) the program control, (2) the input requirements, (3) the execution of THYDE-NEU jobs, (4) the output specifications and (5) the sample calculation. (author)

  1. Nuclear data libraries for Tripoli-3.5 code

    International Nuclear Information System (INIS)

    Vergnaud, Th.

    2001-01-01

    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 -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)

  2. ANENT reference curricula for Master Degree in Nuclear Engineering (Draft no. 1 = version 19 Feb 2006)

    International Nuclear Information System (INIS)

    Raghunathan, V.S.; Chung, B.J.; Duan, P.V.

    2007-01-01

    Subject contents in the following areas are presented in detail: Advanced Mathematics for Nuclear Engineering; Advanced Numerical Analysis; Advanced Computer Applications; Engineering Physics; Introduction to Nuclear Engineering (Core); Radiation Detection and Measurements (Core); Radiation Safety and Shielding; Power Plant Instrumentation; Nuclear Safety; Nuclear and Reactor Physics; Health Physics; Nuclear Heat Transfer; Nuclear Power Plants Engineering; Materials Science in Nuclear Engineering; Neutron Transport Theory; Reactor Kinetics; Advanced Nuclear Heat Transfer; Nuclear Reactor Numerical Analysis; Nuclear Fuel Cycle and Non-Proliferation; Power Reactor Design (System Engineering); Advanced Nuclear Safety; Probabilistic Safety Analysis; Strategy and Infrastructure for Nuclear Power; NPP Control and Instrumentation; Nuclear Regulation; Nuclear Material Engineering; Radiation Protection and Shielding; Application of Radioisotope and Radiation Sources; Non-Destructive Testing; Nuclear Imaging; Radioactive Waste Management; Advanced Health Physics; Applied Radiation Measurements; Advanced Laser Application Engineering; Advanced Quantum Engineering; Plasma Diagnostics; Plasma Processing Analysis; Advanced Plasma Engineering; Nuclear Spectroscopy; Thermonuclear Fusion Engineering

  3. Vectorization, parallelization and porting of nuclear codes on the VPP500 system (vectorization). Progress report fiscal 1997

    International Nuclear Information System (INIS)

    Kawasaki, Nobuo; Ogasawara, Shinobu; Adachi, Masaaki; Kume, Etsuo; Ishizuki, Shigeru; Tanabe, Hidenobu; Nemoto, Toshiyuki; Kawai, Wataru; Watanabe, Hideo

    1999-05-01

    Several computer codes in the nuclear field have been vectorized, parallelized and transported on the FUJITSU VPP500 system and/or the AP3000 system at Center for Promotion of Computational Science and Engineering in Japan Atomic Energy Research Institute. We dealt with 14 codes in fiscal 1997. 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 multidimensional two-fluid model code ACE-3D for evaluation of constitutive equations, statistical decay code SD and three-dimensional thermal analysis code for in-core test section (T2) of HENDEL SSPHEAT are described. In the parallelization part, the parallelization of cylindrical direct numerical simulation code CYLDNS44N, worldwide version of system for prediction of environmental emergency dose information code WSPEEDI, extension of quantum molecular dynamics code EQMD and three-dimensional non-steady compressible fluid dynamics code STREAM are described. In the porting part, the porting of transient reactor analysis code TRAC-BF1 and Monte Carlo radiation transport code MCNP4A on the AP3000 are described. In addition, a modification of program libraries for command-driven interactive data analysis plotting program IPLOT is described. (author)

  4. Vectorization, parallelization and porting of nuclear codes on the VPP500 system (porting). Progress report fiscal 1997

    International Nuclear Information System (INIS)

    Ishizuki, Shigeru; Nemoto, Toshiyuki; Kawai, Wataru; Watanabe, Hideo; Tanabe, Hidenobu; Kawasaki, Nobuo; Adachi, Masaaki; Ogasawara, Shinobu; Kume, Etsuo

    1999-05-01

    Several computer codes in the nuclear field have been vectorized, parallelized and transported on the FUJITSU VPP500 system and/or the AP3000 system at Center for Promotion of Computational Science and Engineering in Japan Atomic Energy Research Institute. We dealt with 14 codes in fiscal 1997. 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 transient reactor analysis code TRAC-BF1 and Monte Carlo radiation transport code MCNP4A on the AP3000 are described. In addition, a modification of program libraries for command-driven interactive data analysis plotting program IPLOT is described. In the vectorization part, the vectorization of multidimensional two-fluid model code ACE-3D for evaluation of constitutive equations, statistical decay code SD and three-dimensional thermal analysis code for in-core test section (T2) of HENDEL SSPHEAT are described. In the parallelization part, the parallelization of cylindrical direct numerical simulation code CYLDNS44N, worldwide version of system for prediction of environmental emergency dose information code WSPEEDI, extension of quantum molecular dynamics code EQMD and three-dimensional non-steady compressible fluid dynamics code STREAM are described. (author)

  5. Reactivity Impact of Difference of Nuclear Data Library for PWR Fuel Assembly Calculation by Using AEGIS Code

    International Nuclear Information System (INIS)

    Ohoka, Yasunori; Tatsumi, Masahiro; Sugimura, Naoki; Tabuchi, Masato

    2011-01-01

    In 2010, the latest version of the Japanese Evaluated Nuclear Data Library (JENDL-4.0) has been released by JAEA. JENDL-4.0 is major update from JENDL- 3.3, and confirmed to give good accuracy by integral test for fission reactor systems such as fast neutron system and thermal neutron system. In this study, we evaluated the reactivity impact due to difference between ENDF/B-VII.0 and JENDL-4.0 for PWR fuel assembly burnup calculation using AEGIS code which has been developed by Nuclear Engineering, Ltd. in cooperation with Nuclear Fuel Industries, Ltd. and Nagoya University

  6. Quality assurance for safety in nuclear power plants. A code of practice

    International Nuclear Information System (INIS)

    1978-01-01

    The Code of Practice is a part of the International Atomic Energy Agency's programme, referred to as the NUSS programme (Nuclear Safety Standards), for establishing Codes of Practice and Safety Guides relating to land-based stationary thermal neutron power plants. The documents are based on documentation and experience from various national systems and practices. The present document provides the recommended principles and objectives for the establishment and implementation of a quality assurance programme during design, manufacture, construction, commissioning and operation of structures, system and components important to safety. They are applicable by all those responsible for the power plant, by plant designers, suppliers, architect-engineers, plant constructors, plant operators and other organizations participating in activities affecting quality. The Lists of relevant definition and the Provisional List of NUSS Programme Titles are given

  7. Establishing a code of ethics for nuclear operating organizations

    International Nuclear Information System (INIS)

    2007-01-01

    The IAEA Technical Working Group on Training and Qualification of Nuclear Power Plant Personnel (TWG-T and Q) recommended that the IAEA develop a publication on improving the performance of nuclear facility operating organizations through focusing on the ethics and professionalism of personnel at all levels of such organizations. This publication has been prepared in response to that recommendation. The TWG-T and Q made its recommendation based upon an understanding that an organization's code of ethics should apply to behaviours at all levels of the organization; from the Board Room to the working level. The TWG-T and Q also recognized that having the technical competencies related to nuclear technology is not enough to ensure that an operating organization's performance is at the high standards needed for a sustainable nuclear industry. The values and ethics of individuals and organizational units play an equally important role. This publication is addressed primarily to senior managers of operating organizations, as experience has shown that, in order to succeed, such initiatives need to come from and be continually supported by the highest levels of the organization. This publication was developed under an IAEA project in its 2006-7 programme entitled Achieving Excellence in the Performance of Nuclear Power Plant Personnel. The principal objectives of this project were: - To enhance the capability of Member States to utilize proven practices accumulated, developed and transferred by the Agency for improving personnel performance and maintaining high standards, and - To demonstrate how positive attitudes and professionalism, appropriate performance management, adherence to a systematic approach to training, quality management and the use of effective information and knowledge management technologies contribute to the success in achieving organization objectives in a challenging business environment

  8. Undergraduate education in nuclear engineering in the USA

    International Nuclear Information System (INIS)

    Martin, W.R.

    1993-01-01

    The discipline of nuclear engineering is described, giving some historical background to explain the structure of the curricula commonly found in nuclear engineering programs in the U.S. Typical curricula are described, along with a specific example given by the University of Michigan undergraduate program in nuclear engineering. The National Academy of Sciences report on U.S. nuclear engineering education is summarized, and the major findings are presented, including data on the number of programs, number of degrees, and enrollment trends. Some discussion is made of manpower trends and the degree to which nuclear programs can supply nuclear engineers to meet the anticipated demands of the current decade and into the next century. (author) 12 refs.; 2 figs.; 4 tabs

  9. Experience with performance based training of nuclear criticality safety engineers

    International Nuclear Information System (INIS)

    Taylor, R.G.

    1993-01-01

    For non-reactor nuclear facilities, the U.S. Department of Energy (DOE) does not require that nuclear criticality safety engineers demonstrate qualification for their job. It is likely, however, that more formalism will be required in the future. Current DOE requirements for those positions which do have to demonstrate qualification indicate that qualification should be achieved by using a systematic approach such as performance based training (PBT). Assuming that PBT would be an acceptable mechanism for nuclear criticality safety engineer training in a more formal environment, a site-specific analysis of the nuclear criticality safety engineer job was performed. Based on this analysis, classes are being developed and delivered to a target audience of newer nuclear criticality safety engineers. Because current interest is in developing training for selected aspects of the nuclear criticality safety engineer job, the analysis is incompletely developed in some areas

  10. 75 FR 24323 - American Society of Mechanical Engineers (ASME) Codes and New and Revised ASME Code Cases

    Science.gov (United States)

    2010-05-04

    ...The NRC proposes to amend its regulations to incorporate by reference the 2005 Addenda through 2008 Addenda of Section III, Division 1, and the 2005 Addenda through 2008 Addenda of Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code (ASME B&PV Code); and the 2005 Addenda and 2006 Addenda of the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code). The NRC also proposes to incorporate by reference ASME Code Case N-722-1, ``Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated With Alloy 600/82/182 Materials Section XI, Division 1,'' and Code Case N-770, ``Alternative Examination Requirements and Acceptance Standards for Class 1 PWR [Pressurized- Water Reactor] Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material with or without Application of Listed Mitigation Activities.''

  11. Virginia power nuclear power station engineer training program

    International Nuclear Information System (INIS)

    Williams, T.M.; Haberstroh-Timpano, S.

    1987-01-01

    In response to the Institute of Nuclear Power Operations (INPO) accreditation requirements for technical staff and manager, Virginia Power developed the Nuclear Power Station Engineer Training Programs (NPSETP). The NPSETP is directed toward enhancing the specific knowledge and skills of company engineers, especially newly hired engineers. The specific goals of the program are to promote safe and reliable plant operation by providing engineers and appropriate engineering technicians with (1) station-specific basic skills; (2) station-specific specialized skills in the areas of surveillance and test, plant engineering, nuclear safety, and in-service inspection. The training is designed to develop, maintain, and document through demonstration the required knowledge and skills of the engineers in the identified groups at North Anna and Surry Power Stations. The program responds to American National Standards Institute, INPO, and US Nuclear Regulatory Commission standards

  12. 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...

  13. 78 FR 37885 - Approval of American Society of Mechanical Engineers' Code Cases

    Science.gov (United States)

    2013-06-24

    ... listed in these RGs as alternatives to engineering standards for the construction, inservice inspection... to engineering standards for the construction, ISI, and IST of nuclear power plant components. This...

  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. HIGH SERVE '90 - nuclear engineering services

    International Nuclear Information System (INIS)

    Bauer, K.G.

    1991-01-01

    Nuclear engineering services do not start only with maintenance or repair, but already with the early detection of imminent problems long before they become problems. Services concerning the decommissioning of plants also belong to it. A selection of the extraordinary services rendered nowadays is presented in more than 20 papers in this booklet. These papers may roughly be divided into three groups of subjects: monitoring and operational management; maintenance, repair and improvements; radioactive waste treatment and management. The first group of subjects, in particular, covers papers dealing with early detection, monitoring and diagnosing systems, using highly advanced hard- and software technologies. Modernization of instrumentation and control systems and exchange of process computer systems is another task this service has to accomplish. Process computers of the past have developed into high performance process information systems. (orig./DG) [de

  16. 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

  17. Girls Who Code Club | College of Engineering & Applied Science

    Science.gov (United States)

    Olympiad Girls Who Code Club FIRST Tech Challenge NSF I-Corps Site of Southeastern Wisconsin UW-Milwaukee elizabeth_andrews Join UWM's 2017-18 Girls Who Code Club Click above to let us remind you of registration on August 1, 2016! Our Girls Who Code Club will resume in Spring 2018. The Fall 2017 Level 1A and 2A students

  18. 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

  19. 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

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

    International Nuclear Information System (INIS)

    O'Kula, K.R.; Paik, I.K.; Chung, D.Y.

    1996-01-01

    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

  1. 76 FR 36231 - American Society of Mechanical Engineers (ASME) Codes and New and Revised ASME Code Cases

    Science.gov (United States)

    2011-06-21

    ...The NRC is amending its regulations to incorporate by reference the 2005 Addenda (July 1, 2005) and 2006 Addenda (July 1, 2006) to the 2004 ASME Boiler and Pressure Vessel Code, Section III, Division 1; 2007 ASME Boiler and Pressure Vessel Code, Section III, Division 1, 2007 Edition (July 1, 2007), with 2008a Addenda (July 1, 2008); 2005 Addenda (July 1, 2005) and 2006 Addenda (July 1, 2006) to the 2004 ASME Boiler and Pressure Vessel Code, Section XI, Division 1; 2007 ASME Boiler and Pressure Vessel Code, Section XI, Division 1, 2007 Edition (July 1, 2007), with 2008a Addenda (July 1, 2008); and 2005 Addenda, ASME OMa Code-2005 (approved July 8, 2005) and 2006 Addenda, ASME OMb Code-2006 (approved July 6, 2006) to the 2004 ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code). The NRC is also incorporating by reference (with conditions on their use) ASME Boiler and Pressure Vessel Code Case N-722-1, ``Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated with Alloy 600/82/182 Materials, Section XI, Division 1,'' Supplement 8, ASME approval date: January 26, 2009, and ASME Boiler and Pressure Vessel Code Case N-770-1, ``Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities, Section XI, Division 1,'' ASME approval date: December 25, 2009.

  2. Analysis of the Current Technical Issues on ASME Code and Standard for Nuclear Mechanical Design(2009)

    International Nuclear Information System (INIS)

    Koo, Gyeong Hoi; Lee, B. S.; Yoo, S. H.

    2009-11-01

    This report describes the analysis on the current revision movement related to the mechanical design issues of the U.S ASME nuclear code and standard. ASME nuclear mechanical design in this report is composed of the nuclear material, primary system, secondary system and high temperature reactor. This report includes the countermeasures based on the ASME Code meeting for current issues of each major field. KAMC(ASME Mirror Committee) of this project is willing to reflect a standpoint of the domestic nuclear industry on ASME nuclear mechanical design and play a technical bridge role for the domestic nuclear industry in ASME Codes application

  3. Report on the Current Technical Issues on ASME Nuclear Code and Standard

    International Nuclear Information System (INIS)

    Koo, Gyeong Hoi; Lee, B. S.; Yoo, S. H.

    2008-11-01

    This report describes the analysis on the current revision movement related to the mechanical design issues of the U.S ASME nuclear code and standard. ASME nuclear mechanical design in this report is composed of the nuclear material, primary system, secondary system and high temperature reactor. This report includes the countermeasures based on the ASME Code meeting for current issues of each major field. KAMC(ASME Mirror Committee) of this project is willing to reflect a standpoint of the domestic nuclear industry on ASME nuclear mechanical design and play a technical bridge role for the domestic nuclear industry in ASME Codes application

  4. Analysis of the Current Technical Issues on ASME Code and Standard for Nuclear Mechanical Design(2009)

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Gyeong Hoi; Lee, B. S.; Yoo, S. H.

    2009-11-15

    This report describes the analysis on the current revision movement related to the mechanical design issues of the U.S ASME nuclear code and standard. ASME nuclear mechanical design in this report is composed of the nuclear material, primary system, secondary system and high temperature reactor. This report includes the countermeasures based on the ASME Code meeting for current issues of each major field. KAMC(ASME Mirror Committee) of this project is willing to reflect a standpoint of the domestic nuclear industry on ASME nuclear mechanical design and play a technical bridge role for the domestic nuclear industry in ASME Codes application

  5. Harmonization of nuclear codes and standards. Pacific nuclear council working and task group report

    International Nuclear Information System (INIS)

    Dua, Shami

    2008-01-01

    Nuclear codes and standards have been an integral part of the nuclear industry since its inception. As the industry came into the main stream over the 2nd half of the 20th century, a number of national and international standards were developed to support a specific nuclear reactor concept. These codes and standards have been a key component of the industry to maintain its focus on nuclear safety, reliability and quality. Both national and international standards have served the industry well in obtaining public, shareholder and regulatory acceptance. The existing suite of national and international standards is required to support the emerging nuclear renaissance. However as noted above under Pacific Nuclear Council (PNC), Manufacturing Design Evaluation Program (MDEP) and SMiRT discussions, the time has come now for the codes and standards writing bodies and the industry to take the next step to examine the relevance of existing suite in view of current needs and challenges. This review must account for the changing global environment including global supply chain and regulatory framework, resources, deregulation, free trade, and industry need for competitiveness and performance excellence. The Task Group (TG) has made limited progress in this review period as no additional information on the listing of codes and standards have been received from the members. However, TG Chair has been successful in obtaining considerable interest from some additional individuals from the member countries. It is important that PNC management seek additional participation from the member countries and asks for their active engagement in the Working Group (WG) TG activities to achieve its mandate and deliverables. The harmonization of codes and standards is a key area for the emerging nuclear renaissance and as noted by a number of international organizations (refer to MDEP action noted above) that these tasks can not be completed unless we have the right level of resources and

  6. A new undergraduate course: Problems in nuclear engineering

    International Nuclear Information System (INIS)

    Larsen, Edward W.

    2011-01-01

    During the past five years, a new third-year undergraduate nuclear engineering course has been developed and taught at the University of Michigan. The course was created to correct certain deficiencies in the undergraduate nuclear engineering curriculum. Here we discuss the origins of the new course and our experience with it. (author)

  7. Standardization of computer programs - basis of the Czechoslovak library of nuclear codes

    International Nuclear Information System (INIS)

    Gregor, M.

    1987-01-01

    A standardized form of computer code documentation has been established in the CSSR in the field of reactor safety. Structure and content of the documentation are described and codes already subject to this process are mentioned. The formation of a Czechoslovak nuclear code library and facilitated discussion of safety reports containing results of standardized codes are aimed at

  8. Congressional perspective on the prospects for tomorrow's nuclear engineers

    International Nuclear Information System (INIS)

    Lloyd, M.

    1986-01-01

    This paper reviews in some detail the nature of the directions in the federally supported nuclear energy research program and discusses the potential opportunities in nuclear engineering education to make contributions to the nation's nuclear power research efforts. The potential impacts of deficit reduction measures on the budgets for nuclear fission programs are also described and the subcommittee priorities for the DOE nuclear fission program within the budget framework are discussed

  9. Several aspects of the effect of nuclear power engineering and thermal power engineering on the environment

    Energy Technology Data Exchange (ETDEWEB)

    Malenchenko, A F

    1979-01-01

    A survey is made of the comparative effect of nuclear power engineering and thermal power engineering on environment and man. The most significant approaches to solution of radio-ecological problems of APS are found.

  10. Compilation of the abstracts of nuclear computer codes available at CPD/IPEN

    International Nuclear Information System (INIS)

    Granzotto, A.; Gouveia, A.S. de; Lourencao, E.M.

    1981-06-01

    A compilation of all computer codes available at IPEN in S.Paulo are presented. These computer codes are classified according to Argonne National Laboratory - and Energy Nuclear Agency schedule. (E.G.) [pt

  11. Establishment of professional nuclear power architectural engineering company

    International Nuclear Information System (INIS)

    Guo Dongli; Chen Hua

    2006-01-01

    The rapid development of nuclear power industry in China requires specialized management for the nuclear power engineering projects. It is necessary to establish the nuclear power architectural engineering company to meet the increasing market needs by providing the owner with specialized nuclear engineering project management and overall contracting services. It is imperative that the purpose of establishing the corporation and enterprise core competitiveness should be clearly identified when it is established. Its organizational structure should be geared to the enterprise operation management and development to facilitate the intensified project management and control, and improve its risk-proof ability. (authors)

  12. Design specifications for ASME B and PV Code Section III nuclear class 1 piping

    International Nuclear Information System (INIS)

    Richardson, J.A.

    1978-01-01

    ASME B and PV Code Section III code regulations for nuclear piping requires that a comprehensive Design Specification be developed for ensuring that the design and installation of the piping meets all code requirements. The intent of this paper is to describe the code requirements, discuss the implementation of these requirements in a typical Class 1 piping design specification, and to report on recent piping failures in operating light water nuclear power plants in the US. (author)

  13. Numerical analysis and nuclear standard code application to thermal fatigue

    International Nuclear Information System (INIS)

    Merola, M.

    1992-01-01

    The present work describes the Joint Research Centre Ispra contribution to the IAEA benchmark exercise 'Lifetime Behaviour of the First Wall of Fusion Machines'. The results of the numerical analysis of the reference thermal fatigue experiment are presented. Then a discussion on the numerical analysis of thermal stress is tackled, pointing out its particular aspects in view of their influence on the stress field evaluation. As far as the design-allowable number of cycles are concerned the American nuclear code ASME and the French code RCC-MR are applied and the reasons for the different results obtained are investigated. As regards a realistic fatigue lifetime evaluation, the main problems to be solved are brought out. This work, is intended as a preliminary basis for a discussion focusing on the main characteristics of the thermal fatigue problem from both a numerical and a lifetime assessment point of view. In fact the present margin of discretion left to the analyst may cause undue discrepancies in the results obtained. A sensitivity analysis of the main parameters involved is desirable and more precise design procedures should be stated

  14. Quantitative emission tomography by coded aperture imaging in nuclear medicine

    International Nuclear Information System (INIS)

    Guilhem, J.B.

    1982-06-01

    The coded aperture imaging is applied to nuclear medicine, since ten years. However no satisfactory clinical results have been obtained thus for. The reason is that digital reconstruction methods which have been implemented, in particular the method which use deconvolution filtering are not appropriate for quantification. Indeed these methods which all based on the assumption of shift invariance of the coding procedure, which is contradictory to the geometrical recording conditions giving the best depth resolution, do not take into account gamma rays attenuation by tissues and in most cases give tomograms with artefacts from blurred structures. A method is proposed which has not these limitations and considers the reconstruction problem as the ill-conditioned problem of solving a Fredholm integral equation. The main advantage of this method lies in fact that the transmission kernel of the integral equation is obtained experimentally, and the approximate solution of this equation, close enough to the original 3-D radioactive object, can be obtained in spite of the ill-conditioned nature of the problem, by use of singular values decomposition (S. V. D.) of the kernel [fr

  15. French ESPN order, codes and nuclear industry requirements

    International Nuclear Information System (INIS)

    Laugier, C.; Grandemange, J.M.; Cleurennec, M.

    2010-01-01

    Work on coding safety regulations applicable to large equipment was undertaken in France as of 1978 to accompany the construction of a French nuclear plant. The needs of manufacturers were threefold: translate the design rules from the American licensor, meet the safety objectives expressed in French regulations published at that time through coding of industrial practices (order of February 26, 1974) and stabilize the work reference system between the operator - consultant - and the manufacturer responsible for applying technical recommendations. Significant work was carried out by AFCEN (the French Association for the Design, Construction and Operating Supervision of the equipment for Electronuclear boilers), an association created for this purpose, leading to the publication of a collection of rules related to mechanical equipment for pressurised water reactors, RCC-M and RSE-M, which will be discussed later, and also in several other technical fields: particularly mechanical equipment in fast neutron reactors, RCC-MR, electricity (RCC-E), and fuel (RCC-C). (authors)

  16. The r-Java 2.0 code: nuclear physics

    Science.gov (United States)

    Kostka, M.; Koning, N.; Shand, Z.; Ouyed, R.; Jaikumar, P.

    2014-08-01

    Aims: We present r-Java 2.0, a nucleosynthesis code for open use that performs r-process calculations, along with a suite of other analysis tools. Methods: Equipped with a straightforward graphical user interface, r-Java 2.0 is capable of simulating nuclear statistical equilibrium (NSE), calculating r-process abundances for a wide range of input parameters and astrophysical environments, computing the mass fragmentation from neutron-induced fission and studying individual nucleosynthesis processes. Results: In this paper we discuss enhancements to this version of r-Java, especially the ability to solve the full reaction network. The sophisticated fission methodology incorporated in r-Java 2.0 that includes three fission channels (beta-delayed, neutron-induced, and spontaneous fission), along with computation of the mass fragmentation, is compared to the upper limit on mass fission approximation. The effects of including beta-delayed neutron emission on r-process yield is studied. The role of Coulomb interactions in NSE abundances is shown to be significant, supporting previous findings. A comparative analysis was undertaken during the development of r-Java 2.0 whereby we reproduced the results found in the literature from three other r-process codes. This code is capable of simulating the physical environment of the high-entropy wind around a proto-neutron star, the ejecta from a neutron star merger, or the relativistic ejecta from a quark nova. Likewise the users of r-Java 2.0 are given the freedom to define a custom environment. This software provides a platform for comparing proposed r-process sites.

  17. Nuclear engineering education in the United States: a status report

    International Nuclear Information System (INIS)

    Miller, D.W.; Spinrad, B.I.

    1986-01-01

    The executive summary of the White Paper entitled The Revitalization of Nuclear Energy Education in the United States is the major component of this paper. The White Paper was completed under the auspices of the Nuclear Engineering Department Heads Organization (NEDHO). The presentation highlights events and program changes that have occurred in 1985-1986 following publication of the NEDHO White Paper. Many of these events provide optimism for the revitalization of nuclear engineering education

  18. Design of compact nuclear power marine engineering simulator

    International Nuclear Information System (INIS)

    Gao Jinghui; Xing Hongchuan; Zhang Ronghua; Yang Yanhua; Xu Jijun

    2004-01-01

    The essentiality of compact nuclear power marine engineering simulator (NPMES) is discussed. The technology of nuclear power plant engineering simulator (NPPES) for NPMES development is introduced, and the function design, general design and model design are given in details. A compact NPMES based on the nuclear power marine of 'Mutsu' is developed. The design can help the development of NPMES, which will improve operation safety and management efficiency of marine. (authors)

  19. Introduction to digital instrumentation and control techniques used in nuclear engineering

    International Nuclear Information System (INIS)

    Kurilla, R.G.; Kenney, E.S.

    1988-01-01

    For the past 8 yr, the nuclear engineering department at Pennsylvania State University has been teaching a digital interfacing class at the undergraduate (senior) level. With the ever-increasing use of computers in the nuclear engineering area (such as in the use of automated data acquisition systems) and the complexity of control instrumentation, more than a cursory introduction into electronics and computer controls is needed. Because of the ever-increasing popularity, and hence importance, of IBM-PC compatible microcomputers in the engineering fields, the program has been adapted to the Intel 8086 microprocessor. Courses such as this one are helpful in ensuring the students have an adequate design and practice base as required by accrediting groups. The course, is composed of three parts: (1) machine code/assembly language, (2) interfacing, and (3) final project. Experience demonstrates that a course of this inherent complexity can successfully be taught within a nuclear engineering curriculum without extensive prerequisites. The important ingredient is to treat nuclear engineering students for exactly what they are, engineers. By having them use their creativity and adaptability, they can successfully integrate the digital interfacing techniques now routinely used in the nuclear industry

  20. Nuclear engineering education: A competence based approach to curricula development

    International Nuclear Information System (INIS)

    2014-01-01

    Maintaining nuclear competencies in the nuclear industry is a one of the most critical challenges in the near future. With the development of a number of nuclear engineering educational programmes in several States, this publication provides guidance to decision makers in Member States on a competence based approach to curricula development, presenting the established practices and associated requirements for educational programmes in this field. It is a consolidation of best practices that will ensure sustainable, effective nuclear engineering programmes, contributing to the safe, efficient and economic operation of nuclear power plants. The information presented is drawn from a variety of recognized nuclear engineering programmes around the world and contributes to the main areas that are needed to ensure a viable and robust nuclear industry

  1. Nuclear reactor engineering: Reactor design basics. Fourth edition, Volume One

    International Nuclear Information System (INIS)

    Glasstone, S.; Sesonske, A.

    1994-01-01

    This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in design and operation of nuclear power plants. Extensively updated, the fourth edition includes new material on reactor safety and risk analysis, regulation, fuel management, waste management, and operational aspects of nuclear power. This volume contains the following: energy from nuclear fission; nuclear reactions and radiations; neutron transport; nuclear design basics; nuclear reactor kinetics and control; radiation protection and shielding; and reactor materials

  2. Licensing procedure, nuclear codes and standards in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Schultheiss, G.F.

    1980-01-01

    The present paper deals with legal background of licensing in nuclear technology and atomic energy use, licensing procedures for nuclear power plants and with codes, standards and guidelines in the Federal Republic of Germany. (orig./RW)

  3. Full scale seismic simulation of a nuclear reactor with parallel finite element analysis code for assembled structure

    International Nuclear Information System (INIS)

    Yamada, Tomonori

    2010-01-01

    The safety requirement of nuclear power plant attracts much attention nowadays. With the growing computing power, numerical simulation is one of key technologies to meet this safety requirement. Center for Computational Science and e-Systems of Japan Atomic Energy Agency has been developing a finite element analysis code for assembled structure to accurately evaluate the structural integrity of nuclear power plant in its entirety under seismic events. Because nuclear power plant is very huge assembled structure with tens of millions of mechanical components, the finite element model of each component is assembled into one structure and non-conforming meshes of mechanical components are bonded together inside the code. The main technique to bond these mechanical components is triple sparse matrix multiplication with multiple point constrains and global stiffness matrix. In our code, this procedure is conducted in a component by component manner, so that the working memory size and computing time for this multiplication are available on the current computing environment. As an illustrative example, seismic simulation of a real nuclear reactor of High Temperature engineering Test Reactor, which is located at the O-arai research and development center of JAEA, with 80 major mechanical components was conducted. Consequently, our code successfully simulated detailed elasto-plastic deformation of nuclear reactor and its computational performance was investigated. (author)

  4. 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.

  5. Elaboration of data and documents intended to complement and expand the German series of nuclear engineering codes. 4. Technical report. Current knowledge of the mechanisms of corrosion fatigue in ferritic materials

    International Nuclear Information System (INIS)

    Frank, J.

    1997-01-01

    Concerning the processes of crack initiation under cyclic loading in a high-temperature water environment, examinations are performed in the USA for amending the design-basis curves published in ASME section III, which are comparable to those contained in the KTA code 3201.2. A revision of ASME section III has not yet been published. The same applies to the processes of crack propagation under cyclic loading in a high-temperature water environment, US experts examining for the purpose of amendment the limiting curves of cyclic crack propagation rates of carbon and low-alloyed steels in a water environment published in ASME section XI. (Orig./CB) [de

  6. 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...... structures (1941). The codes were based on the principle of allowable stresses. However, already in 1948 a Danish consulting engineer used a partial safety factor concept for a power station design in order to secure satisfactory safety. The concept was in fact old as it was used by Gerber in his design...

  7. Parallel multiphysics algorithms and software for computational nuclear engineering

    International Nuclear Information System (INIS)

    Gaston, D; Hansen, G; Kadioglu, S; Knoll, D A; Newman, C; Park, H; Permann, C; Taitano, W

    2009-01-01

    There is a growing trend in nuclear reactor simulation to consider multiphysics problems. This can be seen in reactor analysis where analysts are interested in coupled flow, heat transfer and neutronics, and in fuel performance simulation where analysts are interested in thermomechanics with contact coupled to species transport and chemistry. These more ambitious simulations usually motivate some level of parallel computing. Many of the coupling efforts to date utilize simple code coupling or first-order operator splitting, often referred to as loose coupling. While these approaches can produce answers, they usually leave questions of accuracy and stability unanswered. Additionally, the different physics often reside on separate grids which are coupled via simple interpolation, again leaving open questions of stability and accuracy. Utilizing state of the art mathematics and software development techniques we are deploying next generation tools for nuclear engineering applications. The Jacobian-free Newton-Krylov (JFNK) method combined with physics-based preconditioning provide the underlying mathematical structure for our tools. JFNK is understood to be a modern multiphysics algorithm, but we are also utilizing its unique properties as a scale bridging algorithm. To facilitate rapid development of multiphysics applications we have developed the Multiphysics Object-Oriented Simulation Environment (MOOSE). Examples from two MOOSE-based applications: PRONGHORN, our multiphysics gas cooled reactor simulation tool and BISON, our multiphysics, multiscale fuel performance simulation tool will be presented.

  8. Introducing Knowledge Management in Study Program of Nuclear Engineering

    International Nuclear Information System (INIS)

    Pleslic, S.

    2012-01-01

    Nuclear engineering is the branch of engineering concerning application of the fission as well as the fusion of atomic nuclei, and the application of other sub-atomic physics, based on the principles of nuclear physics. In the sub-field of nuclear fission there are many investigations of interactions and maintaining of systems and components like nuclear reactors and nuclear power plants. The field also includes the study of different applications of ionizing radiation (medicine, agriculture...), nuclear safety, the problems of thermodynamics transport, nuclear materials and nuclear fuels, and other related technologies like radioactive waste management. In the area of nuclear science and engineering a big amount of knowledge has been accumulated over the last decades. Different levels of nuclear knowledge were considered in different ways and they were taught to different parts of population as a general human culture and as a general scientific-technical-technological culture (high schools, nuclear information centres, training centres, universities...). An advanced level of nuclear knowledge has been accumulated by many experienced workers, specialists and experts in all nuclear and nuclear-related fields and applications. In the last 20 years knowledge management has established itself as a discipline of enabling individuals, teams and whole organizations to create, share and apply knowledge collectively and systematically, with goal to better achieve their objectives. Also, knowledge management became key strategic approach for management of intellectual assets and knowledge that can improve safety, efficiency and innovation, and lead to preserve and enhance current knowledge. Knowledge management could be applied in education, training, networking, human resource development and capacity building, sharing, pooling and transferring knowledge form centres of knowledge to centres of growth. Considering the critical importance of nuclear knowledge it is important

  9. NET IBK Computer code package for the needs of planning, construction and operation of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Matausek, M V; Kocic, A; Marinkovic, N; Milosevic, M; Stancic, V [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1978-07-01

    Within the Nuclear Engineering Laboratory of the Boris Kidric Institute of Nuclear Sciences (NET IBK) a systematic work has been performed on collecting nuclear data for reactor calculation needs, on developing own methods and computing programs for reactor calculations, as well as on adapting and applying the foreign methods and codes. In this way a complete library of computer programs was formed for precise prediction of nuclear fuel burnup and depletion, for evaluation of the Power distribution variations with irradiation, for computing the amount of produced plutonium and its number densities etc. Programs for evaluation of location of different types of safety and economic analysis have been developed as well. The aim of this paper is to present our abilities to perform complex computations needed for planning, constructing and operating the nuclear power plants, by describing the NET IBK computer programs package. (author)

  10. Expertise preservation in nuclear technology - the new master course ''nuclear safety engineering'' at the RWTH Aachen

    International Nuclear Information System (INIS)

    Backus, Sabine; Heuters, Michael

    2011-01-01

    The energy concept of the German federal Government in 2010 emphasizes the importance of nuclear energy within the energy policy. The lifetime extension of German nuclear power plants and the long-term safety of radioactive waste storage is the new challenge with respect to the expertise preservation in Germany. The owners of nuclear utilities have started to assist new research programs in the field of nuclear engineering at the German universities. RWE Power and ThyssenKrupp have signed a cooperation contract in 2007 with the RWTH Aachen. The companies bear the expenses for professorships ''nuclear fuel cycle'', ''simulation in nuclear engineering'' and ''reactor safety and engineering''. An elongation of the contract is planned. A master course ''nuclear safety engineering'' over 4 semesters covers the complete fuel cycle. The authors discuss issues concerning the information of students, experiences with the expectations of students concerning their future employment, acceptance of nuclear energy and related topics.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

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

    Directory of Open Access Journals (Sweden)

    S.K. KIM

    2013-10-01

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

  13. Appraisal of the PREP, KITT, and SAMPLE computer codes for the evaluation of the reliability characteristics of engineered systems

    Energy Technology Data Exchange (ETDEWEB)

    Shaw, P; White, R F

    1976-01-01

    For the probabilistic approach to reactor safety assessment by the use of event tree and fault tree techniques it is essential to be able to estimate the probabilities of failure of the various engineered safety features provided to mitigate the effects of postulated accident sequences. The PREP, KITT and SAMPLE computer codes, which incorporate Kinetic Tree Theory, perform these calculations and have been used extensively to evaluate the reliability characteristics of engineered safety features of American nuclear reactors. Working versions of these computer codes are now available in SRD, and this report explains the merits, capabilities and ease of application of the PREP, KITT, and SAMPLE programs for the solution of system reliability problems.

  14. Labor market trends for nuclear engineers through 2005

    International Nuclear Information System (INIS)

    Seltzer, N.; Blair, L.M.

    1996-10-01

    Between 1983 and 1989, employment of nuclear engineers in the nuclear energy field increased almost 40 percent while the annual number of nuclear engineering degrees awarded decreased by almost one-fourth. There were, on average, more job openings for new graduates than there were new graduates available to fill the jobs during the 1980s. This trend reversed in the l990s as nuclear engineering employment in the nuclear energy field decreased from 11,500 in 1991 to 9,400 in 1995. During roughly the same period, the annual number of nuclear engineering degrees increased by 11 percent. As a result, from 1990 through 1995, the number of new graduate nuclear engineers available in the labor supply far exceeded the number of job openings for new graduates in the nuclear energy field. This oversupply of new graduates was particularly acute for 1993 through 1995. During 1996--1997, a relative improvement is expected in job opportunities in the nuclear energy field for new graduates; however, a large oversupply is still expected (almost twice as many graduates available for employment as there are job openings). For 1998 through 2000, some improvement is expected in the relative number of job opportunities for new graduates in the nuclear energy field. Nuclear engineering jobs in the nuclear energy field are expected to decrease only slightly (by less than 150) during this period. Also a 10--15% decrease in the annual number of degrees and available supply of new graduates is expected. Overall, an oversupply is expected (140 graduates available per 100 job openings for new graduates in the nuclear energy field), but this is still a substantial improvement over the current period. For 2001 through 2005, if enrollments and degrees continue to decline, the labor market for new graduates is expected to be approximately balanced. This assumes, however, that the number of degrees and the available supply of new graduates will decrease by 25% from 1995 levels

  15. 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.

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

    International Nuclear Information System (INIS)

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

    2015-01-01

    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 B T and I p can result in a high gain Q fus  ∼ 12, P fus  ∼ 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. (paper)

  17. Finally, nuclear engineering textbooks with a Canadian flavour!

    International Nuclear Information System (INIS)

    Bonin, H.W.

    2002-01-01

    The need for nuclear engineering textbooks more appropriate to the Canadian nuclear industry context and the CANDU nuclear reactor program has long been felt not only among the universities offering nuclear engineering programs at the graduate level, but also within the Canadian nuclear industry itself. Coverage of the CANDU reactor system in the textbooks presently supporting teaching is limited to a brief description of the concept. Course instructors usually complement these textbooks with course notes written from their personal experience from past employment within the nuclear industry and from their research interests In the last ten years, the Canadian nuclear industry has been involved on an increasing basis with the issue of the technology transfer to foreign countries which have purchased CANDU reactors or have been in the process of purchasing one or several CANDUs. For some of these countries, the 'turn key' approach is required, in which the Canadian nuclear industry looks after everything up to the commissioning of the nuclear power plant, including the education and training of local nuclear engineers and plant personnel. Atomic Energy of Canada Limited (AECL) in particular has dispatched some personnel tasked to prepare and give short courses on some specific aspects of CANDU design and operation, but a lack of consistency was observed as different persons prepared and gave the courses rather independently. To address the many problems tied with nuclear engineering education, the CANTEACH program was set up involving major partners of the Canadian nuclear industry. Parts of the activities foreseen by CANTEACH consist in the writing of nuclear engineering textbooks and associated computer-based pedagogical material. The present paper discusses the main parts of two textbooks being produced, one in reactor physics at steady state and the other on nuclear fuel management. (author)

  18. 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.

  19. Physical model of the nuclear fuel cycle simulation code SITON

    International Nuclear Information System (INIS)

    Brolly, Á.; Halász, M.; Szieberth, M.; Nagy, L.; Fehér, S.

    2017-01-01

    Finding answers to main challenges of nuclear energy, like resource utilisation or waste minimisation, calls for transient fuel cycle modelling. This motivation led to the development of SITON v2.0 a dynamic, discrete facilities/discrete materials and also discrete events fuel cycle simulation code. The physical model of the code includes the most important fuel cycle facilities. Facilities can be connected flexibly; their number is not limited. Material transfer between facilities is tracked by taking into account 52 nuclides. Composition of discharged fuel is determined using burnup tables except for the 2400 MW thermal power design of the Gas-Cooled Fast Reactor (GFR2400). For the GFR2400 the FITXS method is used, which fits one-group microscopic cross-sections as polynomial functions of the fuel composition. This method is accurate and fast enough to be used in fuel cycle simulations. Operation of the fuel cycle, i.e. material requests and transfers, is described by discrete events. In advance of the simulation reactors and plants formulate their requests as events; triggered requests are tracked. After that, the events are simulated, i.e. the requests are fulfilled and composition of the material flow between facilities is calculated. To demonstrate capabilities of SITON v2.0, a hypothetical transient fuel cycle is presented in which a 4-unit VVER-440 reactor park was replaced by one GFR2400 that recycled its own spent fuel. It is found that the GFR2400 can be started if the cooling time of its spent fuel is 2 years. However, if the cooling time is 5 years it needs an additional plutonium feed, which can be covered from the spent fuel of a Generation III light water reactor.

  20. Estimates of the radiation environment for a nuclear rocket engine

    International Nuclear Information System (INIS)

    Courtney, J.C.; Manohara, H.M.; Williams, M.L.

    1992-01-01

    Ambitious missions in deep space, such as manned expeditions to Mars, require nuclear propulsion if they are to be accomplished in a reasonable length of time. Current technology is adequate to support the use of nuclear fission as a source of energy for propulsion; however, problems associated with neutrons and gammas leaking from the rocket engine must be addressed. Before manned or unmanned space flights are attempted, an extensive ground test program on the rocket engine must be completed. This paper compares estimated radiation levels and nuclear heating rates in and around the rocket engine for both a ground test and space environments

  1. 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.

  2. Integrating design and purchasing [in nuclear engineering] with Ingecad

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    Ingecad was developed by the Ingevision division of Framatome to overcome deficiencies in traditional computer-aided design. It was developed for nuclear power project engineering around the principle of the shared management of a common database, thus making it possible to integrate several engineering disciplines. The multiuser database is managed and accessed by the different application softwares, corresponding to particular aspects of the engineering task: electrical and process control schematics; plant piping design; pressurized equipment design etc. The use of a common database ensures coherence between the different engineering disciplines, particularly between the process engineering, the plant layout design, the piping, and the instrumentation and control engineering. (author)

  3. Nuclear energy and professional engineers. Possibility of utilization of professional engineer system

    International Nuclear Information System (INIS)

    Tanaka, Shunichi; Nariai, Hideki; Madarame, Haruki; Hattori, Takuya; Kitamura, Masaharu; Fujie, Takao

    2008-01-01

    Nuclear and radiation professional engineer system started in 2004 and more than 250 persons have passed the second-step professional engineer examination, while more than 1,000 persons for the first-step examination. This special issue on possibility of utilization of professional engineer system consists of six relevant articles from experts of nuclear organizations and academia. They expect the role of professional engineer in the area of nuclear energy to enhance technology advancement and awareness of professional ethics from their respective standpoints. (T. Tanaka)

  4. Development of System Engineering Technology for Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    Kim, Hodong; Choi, Iljae

    2013-04-01

    The development of efficient process for spent fuel and establishment of system engineering technology to demonstrate the process are required to develop nuclear energy continuously. The demonstration of pyroprocess technology which is proliferation resistance nuclear fuel cycle technology can reduce spent fuel and recycle effectively. Through this, people's trust and support on nuclear power would be obtained. Deriving the optimum nuclear fuel cycle alternative would contribute to establish a policy on back-end nuclear fuel cycle in the future, and developing the nuclear transparency-related technology would contribute to establish amendments of the ROK-U. S. Atomic Energy Agreement scheduled in 2014

  5. INDXENDF: A PC code for indexing nuclear data files in ENDF-6 format

    International Nuclear Information System (INIS)

    Silva, O.O. de; Corcuera, R.P.; Ferreira, P.A.; Moraes Cunha, M. de.

    1992-01-01

    The PC code INDXENDF which creates visual or printed indexes of nuclear data files in ENDF-6 format, is available from the IAEA Nuclear Data Section on a PC diskette, free of charge upon request. The present document describes the features of this code. (author). 11 refs, 9 figs

  6. Engine cycle design considerations for nuclear thermal propulsion systems

    International Nuclear Information System (INIS)

    Pelaccio, D.G.; Scheil, C.M.; Collins, J.T.

    1993-01-01

    A top-level study was performed which addresses nuclear thermal propulsion system engine cycle options and their applicability to support future Space Exploration Initiative manned lunar and Mars missions. Technical and development issues associated with expander, gas generator, and bleed cycle near-term, solid core nuclear thermal propulsion engines are identified and examined. In addition to performance and weight the influence of the engine cycle type on key design selection parameters such as design complexity, reliability, development time, and cost are discussed. Representative engine designs are presented and compared. Their applicability and performance impact on typical near-term lunar and Mars missions are shown

  7. Development of a compact nuclear power station engineering simulator

    International Nuclear Information System (INIS)

    Jian Jianfeng; Yang Yanhua; Lin Meng; Hu Rui

    2003-01-01

    The compact nuclear power plant project simulator is developed based on the Chashma nuclear power plant. This simulator consists of simulation computation code, data communication module and human-machine interface. This paper discusses the design and implementation of the simulator from such aspect as computer system, hydrothermal model, programming language, human-machine interface and data communication in details

  8. 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.

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

    International Nuclear Information System (INIS)

    Schultz, Peter Andrew

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  11. 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

  12. Brief 66 Nuclear Engineering Enrollments and Degrees Survey, 2009 Data

    International Nuclear Information System (INIS)

    Blair, Larry M.

    2010-01-01

    The survey includes degrees granted between September 1, 2008 and August 31, 2009, and fall 2009 enrollments. Thirty-two academic programs reported having nuclear engineering programs during 2009, and data was obtained from all thirty-two.

  13. Nuclear Engineering Enrollments and Degrees Survey, 2007 Data

    International Nuclear Information System (INIS)

    Analysis and Evaluation, Science Education Programs

    2008-01-01

    The survey includes degrees granted between September 1, 2006, and August 1, 2007, and fall 2007 enrollments. Thirty-one academic programs reported having nuclear engineering programs during 2007, and data was obtained for all thirty-one

  14. Nuclear Engineering Enrollments and Degrees Survey, 2008 Data

    International Nuclear Information System (INIS)

    2009-01-01

    The survey includes degrees granted between September 1, 2007, and August 31, 2008, and fall 2008 enrollments. Thirty-one academic programs reported having nuclear engineering programs during 2008, and data was provided by all thirty-one programs

  15. Fuzzy systems and soft computing in nuclear engineering

    International Nuclear Information System (INIS)

    Ruan, D.

    2000-01-01

    This book is an organized edited collection of twenty-one contributed chapters covering nuclear engineering applications of fuzzy systems, neural networks, genetic algorithms and other soft computing techniques. All chapters are either updated review or original contributions by leading researchers written exclusively for this volume. The volume highlights the advantages of applying fuzzy systems and soft computing in nuclear engineering, which can be viewed as complementary to traditional methods. As a result, fuzzy sets and soft computing provide a powerful tool for solving intricate problems pertaining in nuclear engineering. Each chapter of the book is self-contained and also indicates the future research direction on this topic of applications of fuzzy systems and soft computing in nuclear engineering. (orig.)

  16. Human factor engineering applied to nuclear power plant design

    International Nuclear Information System (INIS)

    Manrique, A.; Valdivia, J.C.; Jimenez, A.

    2001-01-01

    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)

  17. 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.

  18. Russian center of nuclear science and education is the way of nuclear engineering skilled personnel training

    International Nuclear Information System (INIS)

    Murogov, V.M.; Sal'nikov, N.L.

    2006-01-01

    Nuclear power engineering as the key of nuclear technologies is not only the element of the power market but also the basis of the country's social-economic progress. Obninsk as the first science town in Russia is the ideal place for the creation of integrated Science-Research Center of Nuclear Science and Technologies - The Russian Center of Nuclear Science and Education (Center for conservation and development of nuclear knowledge) [ru

  19. Development of new Micro-Physics Nuclear Reactor Simulator™ and its possibility for introductory education of nuclear engineering

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Tsujita, Kosuke; Tamari, Yohei

    2015-01-01

    This paper describes recent activity on development of the Micro-Physics Nuclear Reactor Simulator™ and its application to introductory educations of nuclear engineering at high schools and university. The simulator has been continuously improved with active feedbacks from existing and potential users through its applications to exercises in classes/seminars. A newly developed reactor core transient analysis code, RAMBO-T has been adopted in the simulator along with SIMULATE-3K by Studsvik Scandpower Inc. (Borkowski, 1994) The internal data structure has been revised so that any combinations of the target reactor type, the core transient analysis code and the display language can be established. A new graphical user interface was implemented to realize the intuitive and easy-to-understand operations by novice users. The improved version of the Micro-Physics Nuclear Reactor Simulator has been practically used at educational institutions. In order to contribute to the activities on human resource development in the field of nuclear engineering, it is planned to donate the Micro-Physics Simulator™ Lite, a variation of the simulator that supports the only transient core analysis with RAMBO-T, to IAEA, the International Atomic Energy Agency. It will be included into the “NPP Simulators suite for Education” where complimentary copies are distributed to the member states countries. (author)

  20. Basic data, computer codes and integral experiments: The tools for modelling in nuclear technology

    International Nuclear Information System (INIS)

    Sartori, E.

    2001-01-01

    When studying applications in nuclear technology we need to understand and be able to predict the behavior of systems manufactured by human enterprise. First, the underlying basic physical and chemical phenomena need to be understood. We have then to predict the results from the interplay of the large number of the different basic events: i.e. the macroscopic effects. In order to be able to build confidence in our modelling capability, we need then to compare these results against measurements carried out on such systems. The different levels of modelling require the solution of different types of equations using different type of parameters. The tools required for carrying out a complete validated analysis are: - The basic nuclear or chemical data; - The computer codes, and; - The integral experiments. This article describes the role each component plays in a computational scheme designed for modelling purposes. It describes also which tools have been developed and are internationally available. The role of the OECD/NEA Data Bank, the Radiation Shielding Information Computational Center (RSICC), and the IAEA Nuclear Data Section are playing in making these elements available to the community of scientists and engineers is described. (author)

  1. Qualification of code-Saturne for thermal-hydraulics single phase nuclear applications

    International Nuclear Information System (INIS)

    Archambeau, F.; Bechaud, C.; Gest, B.; Martin, A.; Sakiz, M.

    2003-01-01

    Code-Saturne is a general finite volume CFD (computational fluid dynamics) code developed by Electricite de France (EDF) under quality assurance for 2- and 3-dimensional simulations, laminar and turbulent flows, conjugate heat transfer (coupling with thermal code SYRTHES), including combustion modelling and a Lagrangian module. A very large range of meshes can be used. The solver relies on a finite volume method on arbitrary meshes (hybrid, with hanging nodes, any type of element). All variables are located at the cell centres. The solver is time marching, with a predictor-corrector scheme for Navier-Stokes equations. Standard Reynolds Average Navier-Stokes modelling (RANS) is included (k-epsilon, RSM). Code-Saturne is used by EDF in various industrial fields such as process engineering, aeraulics, combustion and nuclear applications. The present paper describes the qualification phase carried out during 2001 for single-phase nuclear applications. Indeed, once an industrial product has been released and validated, it is of major importance, especially in this particular field related to safety matters, to demonstrate the ability of the code to help engineers produce satisfactory conclusions to industrial problems. In coherence with analyses and best practice guidelines such as those published by the ERCOFTAC Special Interest Group, it seemed important to base the qualification phase on well defined and documented experimental facilities, sufficiently complex to be representative of industrial studies. Much attention has been devoted to evaluating sensitivity to numerical parameters such as grid refinement, time step... Moreover, the qualification studies have been carried out in real-life conditions, that is in limited time, with industrial limitations on the number of grid cells, and by the teams usually producing such studies, so as to integrate a real industrial process in the qualification phase. Two test cases chosen to assess certain types of flows in PWR

  2. Code of Practice containing definitions for Safety Codes of Practice for nuclear power plants

    International Nuclear Information System (INIS)

    1979-01-01

    This Code provides definitions of the technical terms used in the licensing applications to be submitted to the Turkish Atomic Energy Commission (TAEC), in accordance with national licensing regulations. The Code is based mainly on the International Atomic Energy Agency's Code of Practice on the subject. (NEA) [fr

  3. Status of nuclear engineering education in the United States

    International Nuclear Information System (INIS)

    Brown, G.J.

    2000-01-01

    Nuclear engineering education in the United States is reflective of the perceived health of the nuclear electric power industry within the country. Just as new commercial reactor orders have vanished and some power plants have shut down, so too have university enrollments shrunk and research reactors closed. This decline in nuclear trained specialists and the disappearance of the nuclear infrastructure is a trend that must be arrested and reversed if the United States is to have a workforce capable of caring for a nuclear power industry to not only meet future electric demand but to ensure that the over 100 existing plants, their supporting facilities and their legacy in the form of high level waste and facility clean-up are addressed. Additionally, the United States has an obligation to support and maintain its nuclear navy and other defence needs. And, lastly, if the United States is to have a meaningful role in the international use of nuclear power with regard to safety, non-proliferation and the environment, then it is imperative that the country continues to produce world-class nuclear engineers and scientists by supporting nuclear engineering education at its universities. The continued support of the federal government. and industry for university nuclear engineering and nuclear energy research and development is essential to sustain the nuclear infrastructure in the United States. Even with this support, and the continued excellent operation of the existing fleet of nuclear electric power plants, it is conceivable that nuclear engineering as an academic discipline may fall victim to poor communications and a tarnished public image. What is needed is a combination of federal and industrial support along with the creativity of the universities to expand their offerings to include more than power production. The objective is a positive message on careers in nuclear related fields, and recognition of the important role of nuclear energy in meeting the country

  4. A Comparison of Nuclear Power Plant Simulator with RELAP5/MOD3 code about Steam Generator Tube Rupture

    International Nuclear Information System (INIS)

    Kim, Sung Hyun; Moon, Chan Ki; Park, Sung Baek; Na, Man Gyun

    2013-01-01

    The RELAP5/MOD3 code introduced in cooperation with U. S. NRC has been utilized mainly for validation calculation of accident analysis submitted by licensee in Korea. The Korea Institute of Nuclear Safety has built a verification system of LWR accident analysis with RELAP5/MOD3 code engine. Therefore, the simulator replicates the design basis accident and its results are compared with RELAP5/MOD3 code results that will have important implications in the verification of the simulator in the future. The SGTR simulations were performed by the simulator and its results were compared with ones by RELAP5/MOD3 code in this study. Thus, the results of this study can be used as materials to build the verification system of the nuclear power plant simulator. We tried to compare with RELAP5/MOD3 verification code by replicating major parameters of steam generator tube rupture using the simulator for OPR-1000 in Yonggwang training center. By comparing the changes in temperature, pressure and inventory of the reactor coolant system and main steam system during the SGTR, it was confirmed that the main behaviors of SGTR which the simulator and RELAP5/MOD3 code showed are similar. However, the behavior of SG pressure and level that are important parameters to diagnose the accident were a little different. We estimated that RELAP5/MOD3 code was not reflected the major control systems in detail, such as FWCS, SBCS and PPCS. The different behaviors of SG level and pressure in this study should be needed an additional review. As a result of the comparison, the major simulation parameters behavior by RELAP5/MOD3 code agreed well with the one by the simulator. Therefore, it is thought that RELAP5/MOD3 code is used as a tool for validation of NPP simulator in the near future through this study

  5. Institute of Nuclear Engineering: report 1974-1976

    International Nuclear Information System (INIS)

    Amyot, L.

    1976-01-01

    The Institute of Nuclear Engineering is described in terms of its objectives, resources, instructional duties, and research. Basically the Institute is involved in the study of technical, economic and ecological aspects of nuclear installations, basic radioisotopic methods, and general energy problems. (E.C.B.)

  6. Annual technical report - 1987 - Nuclear Engineering Institute - Dept. of Physics

    International Nuclear Information System (INIS)

    Silva, A.G. da; Cabral, S.C.; Bastos, M.A.V.

    1987-01-01

    The research reports carried out in the Physics Department of Nuclear Engineering Institute/Brazilian CNEN, in nuclear physics, isotope production and hazards by irradiation using the CV-28 cyclotron capable to accelerate protons, deuterons, helium and alpha particles with maximum energies of 24, 14, 36 and 28 MeV, respectively, are presented. (M.C.K.) [pt

  7. Education and training in nuclear science/engineering in Taiwan

    International Nuclear Information System (INIS)

    Chung, C.

    1994-01-01

    The present status of nuclear education and training in Taiwan is reviewed. The nuclear science/engineering program has been established in Taiwan under the College of Nuclear Science at the National Tsing Hua University since 1956; it remains the only program among 123 universities and colleges in Taiwan where education and training in nuclear fields are offered. The program, with 52 faculty members, offers advanced studies leading to BSc, MSc, and PhD degrees. Lectures and lab classes are given to 600 students currently registered in the program. Career placement program geared for the 200 graduate and 400 undergraduate students is to orientate them into the local nuclear power utilities as well as agricultural, medical, industrial, academic and governmental sectors where nuclear scientists and engineers at all levels are needed. 8 refs., 1 fig

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

    International Nuclear Information System (INIS)

    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

  9. Fundamentals of nuclear science and engineering

    CERN Document Server

    Shultis, J Kenneth

    2007-01-01

    FUNDAMENTAL CONCEPTS Modern Units The Atom Chart of Nuclides MODERN PHYSICS CONCEPTS The Special Theory of Relativity Radiation as Waves and Particles Quantum Mechanics Derivation of Some Special Relativity Results Solutions to Schrodinger's Wave Equation ATOMIC/NUCLEAR MODELS Development of the Modern Atom Model Models of the Nucleus NUCLEAR ENERGETICS Binding Energy Nucleon Separation Energy Nuclear Reactions Examples of Binary Nuclear Reactions Q-Value for a Reaction Conservation of Charge and the Calculation of Q-values Q-Value for reactions Producing Excited Nuclei RADIOACTIVITY Overview Types of Radioactive Decay Radioactive Decay Diagrams Energetics of Radioactive Decay Characteristics of Radioactive Decay Decay Dynamics Naturally Occurring Radionuclides Radiodating Radioactive Decay Data BINARY NUCLEAR REACTIONS Types of Binary Reactions Kinematics of Binary Two-Product Nuclear Reactions Reaction Threshold Energy Applications of Binary Kinematics Reactions...

  10. Development of Czechoslovak nuclear power engineering

    International Nuclear Information System (INIS)

    Keher, J.

    1985-01-01

    The output of Czechoslovak nuclear power plants is envisaged at 2200 MW by 1985, 4400 MW by 1990 and 10,280 MW by the year 2000. The operation so far is assessed of Bohunice V-1 and Bohunice V-2 power plants as is the construction of the Dukovany nuclear power plant. International cooperation in the fulfilment of the nuclear power programme is based on the General Agreement on Cooperation in the Prospective Development and Interlinkage of CMEA Power Systems to the year 1990, the Agreement on Multilateral International Specialization and Cooperation of Production and on Mutual Deliveries of Nuclear Power Plant Equipment. The most important factor in international cooperation is the Programme of Cooperation between the CSSR and the USSR. The primary target in the coming period is the Temelin nuclear power plant project and the establishment of unified control of the nuclear power complex. (M.D.)

  11. Establishing Requirements for Nuclear Engineering Educational Programs

    International Nuclear Information System (INIS)

    Geraskin, N.I.; Kosilov, A.N.; Sbaffoni, M.M.

    2014-01-01

    Conclusions: » There is no single approach in curricula development. » New programmes must fit into national requirements. » Because of the strong international interdependency of all nations using nuclear energy, it is critically important that a competent staff is engaged at all nuclear power plants in every country. » International approach for benchmarking university programs is to be in place with a direct benefit to the countries with new nuclear power projects

  12. Safety margin evaluation of pre-stressed concrete nuclear containment vessel model with BARC code ULCA

    International Nuclear Information System (INIS)

    Basha, S.M.; Patnaik, R.; Ramanujam, S.; Singh, R.K.; Kushwaha, H.S.; Venkat Raj, V.

    2002-01-01

    Full text: Ultimate load capacity assessment of nuclear containments has been a thrust research area for Indian pressurised heavy water reactor (PHWR) power programme. For containment safety assessment of Indian PHWRs a finite element code ULCA was developed at BARC, Trombay. This code has been extensively benchmarked with experimental results and for prediction of safety margins of Indian PHWRs. The present paper highlights the analysis results for prestressed concrete containment vessel (PCCV) tested at Sandia National Labs, USA in a round robin analysis activity co-sponsored by Nuclear Power Engineering Corporation (NUPEC), Japan and the U.S Nuclear Regulatory Commission (NRC). Three levels of failure pressure predictions namely the upper bound, the most probable and the lower bound (all with 90% confidence) were made as per the requirements of the round robin analysis activity. The most likely failure pressure is predicted to be in the range of 2.95 Pd to 3.15 Pd (Pd = design pressure of 0.39 MPa for the PCCV model) depending on the type of liners used in the construction of the PCCV model. The lower bound value of the ultimate pressure of 2.80 Pd and the upper bound of the ultimate pressure of 3.45 Pd are also predicted from the analysis. These limiting values depend on the assumptions of the analysis for simulating the concrete tendon interaction and the strain hardening characteristics of the steel members. The experimental test has been recently concluded at Sandia Laboratory and the peak pressure reached during the test is 3.3 Pd that is enveloped by our upper bound prediction of 3.45 Pd and is close to the predicted most likely pressure of 3.15 Pd

  13. Nuclear engineering career - Phase 2 Argentina. Final report

    International Nuclear Information System (INIS)

    1993-01-01

    The objective of the project was to consolidate and extend the conditions necessary for the development of nuclear technology, and to observe the problems posed by the application of the nuclear energy through the increase and improvement of the scientific and technical infrastructure. The immediate objective of the project was to complete the advancement of research and development activities in nuclear engineering at the Centro Atomico Bariloche and Instituto Balseiro

  14. Master on Nuclear Engineering and Applications (MINA): instrument of knowledge management in the nuclear sector

    International Nuclear Information System (INIS)

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

    2013-01-01

    Knowledge Management in nuclear industry is indespensable 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 occured 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)

  15. 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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  17. 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.

  18. The Belgian Nuclear Higher Education Network: Your way to the European Master in Nuclear Engineering

    International Nuclear Information System (INIS)

    Moons, F.; D'haeseleer, W.; Giot, M.

    2004-01-01

    BNEN, the Belgian Nuclear Higher Education Network has been created in 2001 by five Belgian universities and the Belgian Nuclear Research Centre (SCK CEN) as a joint effort to maintain and further develop a high quality programme in nuclear engineering in Belgium. More information: http://www.sckcen.be/BNEN. (author)

  19. Annual report of Nuclear Engineering Research Laboratory, Faculty of Engineering, University of Tokyo, fiscal year 1995

    International Nuclear Information System (INIS)

    1996-08-01

    This is an annual report prepared on research education action, operation state of research instruments and others in FY 1995 at Nuclear Engineering Research Laboratory, Faculty of Engineering, University of Tokyo. The laboratory has four large instruments such as high speed neutron source reactor, 'Yayoi', electron linac, fundamentally experimental equipment for blanket design of nuclear fusion reactor, and heavy radiation research equipment (HIT), of which former two are used for cooperative research with universities in Japan, and the next blanket and the last HIT are also presented for cooperative researches in Faculty of Engineering and in University of Tokyo, respectively. FY 1995 was the beginning year of earnest discussion on future planning of this facility with concentrated effort. These four large research instruments are all in their active use. And, their further improvement is under preparation. In this report, the progress in FY 1995 on operation and management of the four large instruments are described at first, and on next, research actions, contents of theses for degree and graduation of students as well as research results of laboratory stuffs are summarized. These researches are constituted mainly using these large instruments in the facility, aiming at development of advanced and new field of atomic energy engineering and relates to nuclear reactor first wall engineering, nuclear reactor fuel cycle engineering, electromagnetic structure engineering, thermal-liquid engineering, mathematical information engineering, quantum beam engineering, new type reactor design and so on. (G.K.)

  20. Annual report of Nuclear Engineering Research Laboratory, Faculty of Engineering, University of Tokyo, fiscal year 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-08-01

    This is an annual report prepared on research education action, operation state of research instruments and others in FY 1995 at Nuclear Engineering Research Laboratory, Faculty of Engineering, University of Tokyo. The laboratory has four large instruments such as high speed neutron source reactor, `Yayoi`, electron linac, fundamentally experimental equipment for blanket design of nuclear fusion reactor, and heavy radiation research equipment (HIT), of which former two are used for cooperative research with universities in Japan, and the next blanket and the last HIT are also presented for cooperative researches in Faculty of Engineering and in University of Tokyo, respectively. FY 1995 was the beginning year of earnest discussion on future planning of this facility with concentrated effort. These four large research instruments are all in their active use. And, their further improvement is under preparation. In this report, the progress in FY 1995 on operation and management of the four large instruments are described at first, and on next, research actions, contents of theses for degree and graduation of students as well as research results of laboratory stuffs are summarized. These researches are constituted mainly using these large instruments in the facility, aiming at development of advanced and new field of atomic energy engineering and relates to nuclear reactor first wall engineering, nuclear reactor fuel cycle engineering, electromagnetic structure engineering, thermal-liquid engineering, mathematical information engineering, quantum beam engineering, new type reactor design and so on. (G.K.)

  1. Development and analysis of startup strategies for particle bed nuclear rocket engine

    Science.gov (United States)

    Suzuki, David E.

    1993-06-01

    The particle bed reactor (PBR) nuclear thermal propulsion rocket engine concept is the focus of the Air Force's Space Nuclear Thermal Propulsion program. While much progress has been made in developing the concept, several technical issues remain. Perhaps foremost among these concerns is the issue of flow stability through the porous, heated bed of fuel particles. There are two complementary technical issues associated with this concern: the identification of the flow stability boundary and the design of the engine controller to maintain stable operation. This thesis examines a portion of the latter issue which has yet to be addressed in detail. Specifically, it develops and analyzes general engine system startup strategies which maintain stable flow through the PBR fuel elements while reaching the design conditions as quickly as possible. The PBR engine studies are conducted using a computer model of a representative particle bed reactor and engine system. The computer program utilized is an augmented version of SAFSIM, an existing nuclear thermal propulsion modeling code; the augmentation, dubbed SAFSIM+, was developed by the author and provides a more complete engine system modeling tool.

  2. 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.

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

    International Nuclear Information System (INIS)

    Shin, Ickhyun; Kwon, Kookheui

    2013-01-01

    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

  4. Reconstruction of nuclear science and engineering harmonized with human society

    International Nuclear Information System (INIS)

    2003-03-01

    At the beginning of the 21th century, the use of nuclear power has assumed very serious dimensions, because there are many problems not only safety technologies but also action of technical expert. The situation and problems of nuclear power are explained. It consists of six chapter as followings; introduction, history and R and D of nuclear power, paradigm change of nuclear science and engineering, energy science, investigation of micro world, how to research and development and education and training of special talent. The improvement plans and five proposals are stated as followings; 1) a scholar and engineer related to nuclear power have to understand ethics and build up closer connection with person in the various fields. 2) Nuclear power generation and nuclear fuel cycle are important in future, so that they have to be accepted by the society by means of opening to the public. Safety science, anti-pollution measurements, treatment and disposal of radioactive waste and development of new reactor and fusion reactor should be carried out. 3) It is necessary that the original researches of quantum beam and isotope have to step up. 4) The education of nuclear science and technology and upbringing special talent has to be reconstructed. New educational system such as 'nuclear engineering course crossing with many universities' is established. 5) Cooperation among industry, academic world and government. (S.Y.)

  5. Advances in chemical engineering in nuclear and process industries

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-06-01

    Symposium on Advances in Chemical Engineering in Nuclear and Process Industries dealt with a wide spectrum of areas encompassing various industries such as nuclear, fertilizer, petrochemical, refinery and cement. The topics covered in the symposium dealt with the advancements in the existing fields of science and technologies as well as in some of the emerging technologies such as membrane technology, bio-chemical and photo-chemical engineering etc. with a special emphasis on nuclear related aspects. Papers relevant to INIS are indexed separately.

  6. Advances in chemical engineering in nuclear and process industries

    International Nuclear Information System (INIS)

    1994-06-01

    Symposium on Advances in Chemical Engineering in Nuclear and Process Industries dealt with a wide spectrum of areas encompassing various industries such as nuclear, fertilizer, petrochemical, refinery and cement. The topics covered in the symposium dealt with the advancements in the existing fields of science and technologies as well as in some of the emerging technologies such as membrane technology, bio-chemical and photo-chemical engineering etc. with a special emphasis on nuclear related aspects. Papers relevant to INIS are indexed separately

  7. Educational experiments of radiochemistry in the nuclear engineering school

    International Nuclear Information System (INIS)

    Akatsu, Eiko

    1995-06-01

    Educational experiments of radiochemistry are described. They were an improvement of educational experiment of burn-up measurement as well as experiments on a solvent extraction, a cation exchange behavior of 60 Co, liquid scintillation spectrometry and half-life determination of 87 Rb, and determination of 137 Cs in sea water. Two or one of the experiments were ordinarily studied, depending the occasional situations, by the students of the general course or of the nuclear engineering course in the Nuclear Engineering School, Nuclear Education Center, JAERI from 1976 to 1994. (author)

  8. Experience with performance based training of nuclear criticality safety engineers

    International Nuclear Information System (INIS)

    Taylor, R.G.

    1993-01-01

    Historically, new entrants to the practice of nuclear criticality safety have learned their job primarily by on-the-job training (OJT) often by association with an experienced nuclear criticality safety engineer who probably also learned their job by OJT. Typically, the new entrant learned what he/she needed to know to solve a particular problem and accumulated experience as more problems were solved. It is likely that more formalism will be required in the future. Current US Department of Energy requirements for those positions which have to demonstrate qualification indicate that it should be achieved by using a systematic approach such as performance based training (PBT). Assuming that PBT would be an acceptable mechanism for nuclear criticality safety engineer training in a more formal environment, a site-specific analysis of the nuclear criticality safety engineer job was performed. Based on this analysis, classes are being developed and delivered to a target audience of newer nuclear criticality safety engineers. Because current interest is in developing training for selected aspects of the nuclear criticality safety engineer job, the analysis i's incompletely developed in some areas. Details of this analysis are provided in this report

  9. Chaos and fractals. Applications to nuclear engineering

    International Nuclear Information System (INIS)

    Clausse, A.; Delmastro, D.F.

    1990-01-01

    This work presents a description of the research lines carried out by the authors on chaos and fractal theories, oriented to the nuclear field. The possibilities that appear in the nuclear security branch where the information deriving from chaos and fractal techniques may help to the development of better criteria and more reliable designs, are of special importance. (Author) [es

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

    International Nuclear Information System (INIS)

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

    2011-01-01

    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. 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.

  12. A nuclear engineering curriculum for Asia-Pacific

    International Nuclear Information System (INIS)

    Bereznai, G.; Sumitra, T.; Chankow, N.; Chanyotha, S.

    1996-01-01

    This paper describes the nuclear engineering education and professional development curricula that are being developed at Chulalongkorn University in Bangkok, Thailand. The program was initiated in response to the Thai Government's policy to keep the option of nuclear electric generation available as the country responds to the rapid growth of industrialization and increased standard of living, and the accompanying increase in electricity consumption. The program has three main thrusts: university education, professional development, and public education. Although this paper concentrates on the university curriculum, it is shown how the university program is integrated with the development of industry professionals. The Nuclear Engineering Curricula being developed and implemented at Chulalongkorn University will offer programs at the Bachelor, Master and Doctorate levels. The curricula are designed to provide comprehensive education and training for engineers and scientists planning careers in the peaceful use of nuclear energy, with emphasis on the applications to industry and for nuclear electric generation. The Project of Human Resource Development in the Nuclear Engineering field is the result of a cooperative effort between agencies of the Thai and Canadian Governments, including the Electricity Generating Authority of Thailand, the Office of Atomic Energy for Peace, Chulalongkorn University and several other Thai Universities; Atomic Energy of Canada Limited, the Canadian International Development Agency, several Canadian Universities as well as members of the Canadian Nuclear Industry. (author)

  13. Nuclear Engineering Education in Support of Thailand’s Nuclear Power Programme

    International Nuclear Information System (INIS)

    Chanyotha, S.; Pengvanich, P.; Nilsuwankosit, S.

    2015-01-01

    This paper aims to introduce the nuclear engineering education at the Department of Nuclear Engineering, Chulalongkon University, Bangkok Thailand. The department has been offering curriculum in nuclear engineering to support the national nuclear power programme since 1970s. It is the oldest established nuclear engineering educational programme in the South East Asia region. Nevertheless, since the nuclear power programme has been postponed several times due to various reasons, the educational programme at the department has been continuously adapted to meet the nation’s needs. Several areas of study have been introduced, including nuclear power engineering, industrial applications of radioisotope, nuclear instrumentation, radioisotope production, radiation processing, environment and safety, nuclear materials, as well as the newly created nuclear security and non-proliferation. With the renewed interest in using nuclear power in Thailand in 2007, the department has been actively assisting both the government and the electric utility in preparing human resources to support the nuclear power programme through various educational and training modules. Realizing the importance of establishing and balancing all 3 aspects of the nuclear 3S (safety, security and safeguard) in Thailand and in the Southeast Asian region. The new curriculum of nuclear security and safeguard programme has been offered since 2013. Since the establishment, the department has produced hundreds of graduates (Diploma, Master’s, and Ph.D. levels) to feed the continuously expanding Thai nuclear industry. The full paper will provide detailed information of the curriculum, the challenges and obstacles that the department has encountered, as well as the national and international linkages which have been established over the years. (author)

  14. Nuclear model codes and related software distributed by the OECD/NEA Data Bank

    International Nuclear Information System (INIS)

    Sartori, E.

    1993-01-01

    Software and data for nuclear energy applications is acquired, tested and distributed by several information centres; in particular, relevant computer codes are distributed internationally by the OECD/NEA Data Bank (France) and by ESTSC and EPIC/RSIC (United States). This activity is coordinated among the centres and is extended outside the OECD area through an arrangement with the IAEA. This article covers more specifically the availability of nuclear model codes and also those codes which further process their results into data sets needed for specific nuclear application projects. (author). 2 figs

  15. FIRAC: a computer code to predict fire-accident effects in nuclear facilities

    International Nuclear Information System (INIS)

    Bolstad, J.W.; Krause, F.R.; Tang, P.K.; Andrae, R.W.; Martin, R.A.; Gregory, W.S.

    1983-01-01

    FIRAC is a medium-sized computer code designed to predict fire-induced flows, temperatures, and material transport within the ventilating systems and other airflow pathways in nuclear-related facilities. The code is designed to analyze the behavior of interconnected networks of rooms and typical ventilation system components. This code is one in a family of computer codes that is designed to provide improved methods of safety analysis for the nuclear industry. The structure of this code closely follows that of the previously developed TVENT and EVENT codes. Because a lumped-parameter formulation is used, this code is particularly suitable for calculating the effects of fires in the far field (that is, in regions removed from the fire compartment), where the fire may be represented parametrically. However, a fire compartment model to simulate conditions in the enclosure is included. This model provides transport source terms to the ventilation system that can affect its operation and in turn affect the fire

  16. Two codes used in analysis of rod ejection accident for Qinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Zhu Xinguan

    1987-12-01

    Two codes were developed to analyse rod ejection accident for Qinshan Nuclear Power Plant. One was based on point model with temperature reactivity feedback. In this code, the worth of ejected rod was obtained under'adiabatic' approximation. In the other code, the Nodal Green's Function Method was used to solve space-time dependent neutron diffusion equation. Using these codes, the transient core-power have been calculated for two rod ejection cases at beginning of core-life in Qinshan Nuclear Power Plant

  17. Survey of materials and other problems of relevance in safety engineering, and an assessment of their reflection in regulatory guides for conventional and nuclear engineering (1. technical report)

    International Nuclear Information System (INIS)

    Trunk, M.; Herter, K.H.

    1984-01-01

    Survey and assessment of nuclear engineering specifications and regulatory guides (ASME-BPVC Section III, division 1,2 and KTA, BS 5500) with regard to materials, dimensioning and testing for the purpose of showing to what extent available technical codes, regulatory guides and safety codes are useful in preventing failures and defining the safe limit. The other question examined is that of how these codes ought to be brought up to date in order to reflect the latest state of the art in science and technology. (orig./HP) [de

  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 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.

  20. Nuclear power engineering: Public understanding and public opinion

    International Nuclear Information System (INIS)

    Kryshev, A.I.; Sazykina, T.G.

    1998-01-01

    Subjective and objective reasons for the formation of public opinion about nuclear power engineering of Russia were analyzed. Some methodological errors in work with the Russian public on the problems of nuclear energy and possible methods of their correction were discussed. The social groups of the general public, which are of greatest importance in forming the attitude towards nuclear power engineering were indicated. The conclusion was reached that opinion of the ordinary population is often indicative of real drawbacks in the work of specialists in the nuclear fuel cycle. Consequently, careful surveys of public opinion about the problems of the nuclear industry should be very useful in organizing research work properly and improving the radiation safety. (author)

  1. The midwest workshop on preparing nuclear engineering professionals

    International Nuclear Information System (INIS)

    Danofsky, R.A.; Rohach, A.F.; Spinrad, B.I.; Nodean, W.C.

    1988-01-01

    Personnel training and education are activities of major importance for nuclear utilities and represent fruitful areas for possible cooperation between utilities and educational institutions. Utility personnel have a need for continuing education through advanced and undergraduate degree programs and special courses. Nuclear engineering departments are in a position to meet at least some of these needs. The purpose of the workshop described in this paper was to explore ways to increase the dialogue between utilities and universities and to bring faculty and utility personnel together to discuss the educational needs of nuclear utilities. The workshop was held May 25-27, 1988, at Iowa State University. Planning for the workshop was coordinated by a steering committee with representation from the Department of Nuclear Engineering at Iowa State University, Iowa Electric Light and Power Company (IEL ampersand P), and Kirkwood Community College at Cedar Rapids, Iowa. Participants represented nuclear utilities, nuclear engineering departments, 2- and 4-yr colleges, a nuclear training organization, and the Institute of Nuclear Power Operations

  2. Comparison of Neutron Cross-Sections Using IAEA Nuclear Codes ''ABAREX'' and ''SCAT2''

    International Nuclear Information System (INIS)

    Myint Myint Moe; Win Sin; Sein Htoon

    2004-05-01

    Moel calculations can be used to provide nuclear data for applications in science and technology. The energy averaged neutron induced nuclear reaction cross-sections particular for Al-27, Mg-24, Cr-52, Mn-55, Zn-64 and U-238 with neutrons of energy (0.005 to 10 MeV) are calculated using IAEA nuclear codes ''ABAREX'' and ''SCAT2''. The results are compared with those given in ENDF- 3 nuclear data

  3. The GNASH preequilibrium-statistical nuclear model code

    International Nuclear Information System (INIS)

    Arthur, E. D.

    1988-01-01

    The following report is based on materials presented in a series of lectures at the International Center for Theoretical Physics, Trieste, which were designed to describe the GNASH preequilibrium statistical model code and its use. An overview is provided of the code with emphasis upon code's calculational capabilities and the theoretical models that have been implemented in it. Two sample problems are discussed, the first dealing with neutron reactions on 58 Ni. the second illustrates the fission model capabilities implemented in the code and involves n + 235 U reactions. Finally a description is provided of current theoretical model and code development underway. Examples of calculated results using these new capabilities are also given. 19 refs., 17 figs., 3 tabs

  4. Thermohydraulic analysis of nuclear power plant accidents by computer codes

    International Nuclear Information System (INIS)

    Petelin, S.; Stritar, A.; Istenic, R.; Gregoric, M.; Jerele, A.; Mavko, B.

    1982-01-01

    RELAP4/MOD6, BRUCH-D-06, CONTEMPT-LT-28, RELAP5/MOD1 and COBRA-4-1 codes were successful y implemented at the CYBER 172 computer in Ljubljana. Input models of NPP Krsko for the first three codes were prepared. Because of the high computer cost only one analysis of double ended guillotine break of the cold leg of NPP Krsko by RELAP4 code has been done. BRUCH code is easier and cheaper for use. Several analysis have been done. Sensitivity study was performed with CONTEMPT-LT-28 for double ended pump suction break. These codes are intended to be used as a basis for independent safety analyses. (author)

  5. How the engineers are sinking nuclear power

    International Nuclear Information System (INIS)

    Mintz, J.

    1983-01-01

    Poor concrete work, improper welds, and construction and installation errors at nuclear power plants are blamed on budget and schedule pressures and the nuclear industry's lack of quality assurance. Nuclear Regulatory Commission Chairman Nunzio Palladino, who trained under the exacting Admiral Rickover, has ordered the industry to upgrade its quality assurance and to take safety regulations and training more seriously. The industry's response is a program that will send a team of Institute of Nuclear Power Operators (INPO) investigators to each plant under construction every 18 months to make spot checks of worker training and performance. The Electric Power Research Institute is also developing equipment to test construction quality. Both industry officials and critics remain skeptical that quality assurance will improve with more regulation

  6. Ergonomics in nuclear and human factors engineering

    International Nuclear Information System (INIS)

    Muench, E.; Schultheiss, G.F.

    1988-01-01

    The work situation including man-machine-relationships in nuclear power plants is described. The overview gives only a compact summary of some important ergonomic parameters, i.e. human body dimension, human load, human characteristics and human knowledge. (DG)

  7. Evolutionary computing in Nuclear Engineering Institute/CNEN-Brazil

    International Nuclear Information System (INIS)

    Pereira, Claudio M.N.A.; Lapa, Celso M.F.; Lapa, Nelbia da Silva; Mol, Antonio C.

    2000-01-01

    This paper aims to discuss the importance of evolutionary computation (CE) for nuclear engineering and the development of this area in the Instituto de Engenharia Nuclear (IEN) at the last years. Are describe, briefly, the applications realized in this institute by the technical group of CE. For example: nuclear reactor core design optimization, preventive maintenance scheduling optimizing and nuclear reactor transient identifications. It is also shown a novel computational tool to implementation of genetic algorithm that was development in this institute and applied in those works. Some results were presents and the gains obtained with the evolutionary computation were discussing. (author)

  8. Secure Coding for Safety I and C Systems on Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kim, Y. M.; Park, H. S.; Kim, T. H.

    2015-01-01

    This paper addresses secure coding technologies which can reduce the software vulnerabilities and provides secure coding application guidelines for nuclear safety I and C systems. The use of digital equipment may improve their reliability and reduce maintenance costs. But, the design characteristics of nuclear I and C systems are becoming more complex and the possibility of cyber-attacks using software vulnerabilities has been increased. Software defects, bugs and logic flaws have been consistently the primary causes of software vulnerabilities which can introduce security vulnerabilities. In this study, we described a applying methods for secure coding which can reduce the software vulnerabilities. Software defects lists, countermeasures for each defect and coding rules can be applied properly depending on target system's condition. We expect that the results of this study can help developing the secure coding guidelines and significantly reducing or eliminating vulnerabilities in nuclear safety I and C software

  9. Secure Coding for Safety I and C Systems on Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. M.; Park, H. S. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Kim, T. H. [Formal Works Inc., Seoul (Korea, Republic of)

    2015-10-15

    This paper addresses secure coding technologies which can reduce the software vulnerabilities and provides secure coding application guidelines for nuclear safety I and C systems. The use of digital equipment may improve their reliability and reduce maintenance costs. But, the design characteristics of nuclear I and C systems are becoming more complex and the possibility of cyber-attacks using software vulnerabilities has been increased. Software defects, bugs and logic flaws have been consistently the primary causes of software vulnerabilities which can introduce security vulnerabilities. In this study, we described a applying methods for secure coding which can reduce the software vulnerabilities. Software defects lists, countermeasures for each defect and coding rules can be applied properly depending on target system's condition. We expect that the results of this study can help developing the secure coding guidelines and significantly reducing or eliminating vulnerabilities in nuclear safety I and C software.

  10. Investigation and proposal of the system to affect nuclear fuel type authorization and analysis code certification

    International Nuclear Information System (INIS)

    2006-03-01

    In order to develop the system to affect more advanced and rational regulations of nuclear fuels and earlier introduction of new technologies in nuclear power plants, domestic and overseas safety regulation systems and state of their implementation for water cooled reactor fuel and safety analysis code had been investigated and new regulation system to affect nuclear fuel type authorization and analysis code certification was proposed. Topical report system for common parts related with nuclear fuel type authorization and analysis code certification was firstly proposed for knowledge base. Maintaining consistent safety examination supported by experts, introduction of domestic efficient system for lead irradiation test fuel, and analysis code certification and quality assurance were also proposed. (T. Tanaka)

  11. Evaluation of crack-like flaw in Japanese fitness-for-service code for nuclear power plant components

    International Nuclear Information System (INIS)

    Kashima, Koichi

    2003-01-01

    For evaluation of faults detected at nuclear appliances, establishment of fitness-for-service code in Japan is focused by most of peoples. The code is a management rule to keep features of the appliances under supplying operation to their constant safe level and is a rule composing a pair with design rule. The codes for nuclear power generation facilities-rules of fitness-for-service for nuclear power plants were issued on May, 2002, by the Japan Society of Mechanical Engineering (JSME), which was added on October, 2002, by its inspection code, for its amendment. Under such states, Japan Government is proceeding on establishment of the fitness-for-service code in Japan on a base of the private rule. Here were introduced present state and tasks on content of crack-like flaw evaluation on the code under an example of the private rule of JSME, which is composed of three items of inspection, evaluation, and recovery and exchange. The evaluation of defects consists of 1) the first step of evaluation of defects and 2) the second step of evaluation of defects. The first step determines the size of defect by modeling form. When the size of defect is smaller than the evaluation criterion, the appliances can be used unconditionally. However, its size is larger than the evaluation criterion, the appliances have to be evaluated by the second step. When the estimated defects size at end of evaluation period is smaller than the permissible value, the appliances can be used within the evaluation period. But, if its size is larger than the permissible value, the appliances have to be recovered and exchanged. Modeling, evaluation criterion, evaluation of destruction, safety standards and future problems are described. (S.Y.)

  12. Benefit using reasonable regulations in USA, how to skill up on professional engineers, apply international code, standard, and regulation

    International Nuclear Information System (INIS)

    Turner, S.L.; Morokuzu, Muneo; Amano, Osamu

    2005-01-01

    The reasonable regulations in USA consist of a graduated approach and a risk informed approach (RIA). RIA rationalizes the regulations on the basis of data of operations etc. PSA (Probabilistic Safety Assessment), a general method of RIA, is explained in detail. The benefits of nuclear power plant using RIA are increase of the rate of operation, visualization of risk, application of design standard and design, cost down of nuclear fuel cycle, waste, production and operation, and safety. RIA is supported by the field data, code, standard, regulation and professional engineers. The effects of introduction of RIA are explained. In order to introduce RIA in Japan, all the parties concerned such as the regulation authorities, the electric power industries, makers, universities, have to understand it and work together. A part of scientific society is stated. (S.Y.)

  13. Computer codes to assess risks from nuclear power plants with LWR's

    International Nuclear Information System (INIS)

    Alonso, A.; Blanco, J.; Francia, L.; Gallego, E.; Morales, L.; Ortega, P.; Torres, C.

    1986-01-01

    The codes used to quantify risks from nuclear power plants are described. For QRA level 1 (quantitative risk assessment) qualitative and quantitative codes are described. Codes to estimate uncertainties, importance and dependent failures are also included. For QRA-level 2, the most important codes dealing with thermohydraulics, molten core and aerosols behaviour are described. For QRA-level 3 the list includes integrated as well as separate models. Only light water reactors are considered. The presentation is general but the authors describe with more detail those codes they are more familiar with or the ones they have created through their research effort. (author)

  14. Nuclear engineering career path - Past, present and future in Europe

    International Nuclear Information System (INIS)

    Dragusin, O.; Goicea, A.

    2005-01-01

    Full text: Today's nuclear workforce is growing older. The nuclear industry is at a turning point, mainly when we try to evaluate future needs. The paper tries to explore what is the present status of nuclear engineering career in Europe and which are the future trends. Today's students are either unaware of, or indifferent to, nuclear power. But it is not their fault - after all, an entire generation has grown to adulthood since the Three Miles Island and Chernobyl accidents. If their parents' views about nuclear power have been changed after these unhappy accidents, today's teens appear to be almost ambivalent toward it. What could be done to attract younger to the nuclear field in Europe? Probably the main key is in education and even more in a positive image-making of nuclear industry. Creation of more possibilities for gaining work experience and an improvement of economical situation and successful completion of reforms in the industry could attract young people. Political viewpoint is very important, if we consider that people in general thinks that nuclear will be phased out within a short future. A good advertising about job opportunities to come is another factor that plays a role in information of young generation. The paper provides information about university programmes in nuclear engineering in European countries. (authors)

  15. French codes and standards for design, construction and in-service inspection of nuclear power plants

    International Nuclear Information System (INIS)

    Hugot, G.; Grandemange, J. M.

    1995-01-01

    In 1970, France decided that its future power plants would be of the Pressurized Water Reactor type. This choice proved to be successful since it resulted in more than 60 PWR units in operation or under construction in France and abroad. At the beginning of such a program, the French engineering and manufacturing industry, the national electrical utility and the Safety Authorities had to face the many challenges imposed by the implementation of an imported technology. The government reorganised the licensing process. FRAMATOME, the NSSS vendor, and EDF (Electricite de France), the national utility, decided to create 'AFCEN', the French Association for Design and Construction Rules for Nuclear Island Components. These rules, the RCC's (Regles de Construction et de conception), which are approved by French Safety Authorities deal with mechanical and electrical equipment as well as with nuclear fuel and civil works. They are now being supplemented by in service inspection rules, the RSE's (Regles d'inspection en Service). The paper presents these Codes and their main updating following experience of application, technical progress and evolution of standards. Status of discussion concerning reference to European standardisation and developments of rules applicable to the EPR project will also be discussed

  16. Analytical benchmarks for nuclear engineering applications. Case studies in neutron transport theory

    International Nuclear Information System (INIS)

    2008-01-01

    The developers of computer codes involving neutron transport theory for nuclear engineering applications seldom apply analytical benchmarking strategies to ensure the quality of their programs. A major reason for this is the lack of analytical benchmarks and their documentation in the literature. The few such benchmarks that do exist are difficult to locate, as they are scattered throughout the neutron transport and radiative transfer literature. The motivation for this benchmark compendium, therefore, is to gather several analytical benchmarks appropriate for nuclear engineering applications under one cover. We consider the following three subject areas: neutron slowing down and thermalization without spatial dependence, one-dimensional neutron transport in infinite and finite media, and multidimensional neutron transport in a half-space and an infinite medium. Each benchmark is briefly described, followed by a detailed derivation of the analytical solution representation. Finally, a demonstration of the evaluation of the solution representation includes qualified numerical benchmark results. All accompanying computer codes are suitable for the PC computational environment and can serve as educational tools for courses in nuclear engineering. While this benchmark compilation does not contain all possible benchmarks, by any means, it does include some of the most prominent ones and should serve as a valuable reference. (author)

  17. Labor market trends for nuclear engineers through 2000

    International Nuclear Information System (INIS)

    Seltzer, N.; Blair, L.M.; Baker, J.G.

    1995-01-01

    Throughout most of the 1980s, both private organizations and government agencies were concerned about the availability of an adequate supply of qualified nuclear engineers. This concern was primarily the result of a number of nuclear engineering academic programs being eliminated coupled with a continuous decline in graduate and undergraduate enrollments and degrees. By the early 1990s, the number of degrees and available supply had declined to new lows, but cutbacks in funding for the nuclear weapons program and nuclear energy R ampersand D, and in hiring by the electric utility industry, offset in large measure the declining supply. Recently, concerns about environment and waste management and about nuclear safety have again generated questions about the adequacy of supply of qualified personnel for nuclear energy activities. This report briefly examines the nuclear engineering labor market. Trends in employment, new graduates, job openings, and salaries are reviewed as a basis for understanding the current labor market. This review is then used as a basis for assessing future employment needs and new graduate supply to provide an outlook for future labor market conditions through 2000

  18. MCNP and other nuclear codes output graphical representation using python scripts

    International Nuclear Information System (INIS)

    Cadenas Mendicoa, A. M.

    2016-01-01

    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)

  19. Corrosion engineering in nuclear power industry

    International Nuclear Information System (INIS)

    Prazak, M.; Tlamsa, J.; Jirousova, D.; Silber, K.

    1990-01-01

    Corrosion problems in nuclear power industry are discussed from the point of view of anticorrosion measures, whose aim is not only increasing the lifetime of the equipment but, first of all, securing ecological safety. A brief description is given of causes of corrosion damage that occurred at Czechoslovak nuclear power plants and which could have been prevented. These involve the corrosion of large-volume radioactive waste tanks made of the CSN 17247 steel and of waste piping of an ion exchange station made of the same material, a crack in a steam generator collector, contamination of primary circuit water with iron, and corrosion of CrNi corrosion-resistant steel in a spent fuel store. It is concluded that if a sufficient insight into the corrosion relationships exists and a reasonable volume of data is available concerning the corrosion state during the nuclear facility performance, the required safety can be achieved without adopting extremely costly anticorrosion measures. (Z.M.)

  20. Nuclear science and engineering in China

    Energy Technology Data Exchange (ETDEWEB)

    von Becker, K

    1979-01-01

    A brief review of the development of nuclear science and technology in China is given. It is stated that the change of leadership in China has brought about a radical revision of the attitude towards the science and technology. In the plan of the development of nuclear science and technology adopted in 1973 a great emphasis is laid on investigations in the field of high energy physics. For instance, it is planned to construct, before 1983, a 30-50 GeV proton accelerator. A brief description is given of main nuclear research institutes in Phangshan, Peking and Shanghai which are shown to Western visitors. It is indicated that at these institutes there are the only two research reactors in China, a 3.5-MW LWR and 10 MW HWR, two cyclotrons and a 90-cm tokamak. These institutes also conduct investigations on solid-state physics, low-temperature physics, high-pressure physics, lasers, radiation biology, radiation chemistry, etc.

  1. Review of ASME nuclear codes and standards- subcommittee on repairs, replacements, and modifications

    International Nuclear Information System (INIS)

    Mawson, T.J.

    1990-01-01

    As requested by the ASME board on Nuclear Codes and Standards, the Pressure Vessel Research Committee initiated a project to review Sections III and XI of the ASME Boiler and Pressure Vessel Code for the purposes of improving, clarifying, providing transition, consistency, compatibility, and simplifying code requirements. The project was organized with six subcommittees to address various Code activities: design; tests and examinations; documentation; quality assurance; repair, replacement and modification; and general requirements. This paper discusses how the subcommittee on repair, replacement and modification was organized to review the repair, replacement and modification requirements of the ASME boiler and pressure vessel code, Section III and Section XI for Class 1, 2, and 3 and MC components and their supports, and other documents of the nuclear industry related to the repair, replacement and modification requirements of the ASME code

  2. Master's degree in nuclear engineering by videotaped courses

    International Nuclear Information System (INIS)

    Corradini, M.L.; Vogelsang, W.F.

    1991-01-01

    In 1986, a group of northern midwest utilities met with faculty from the nuclear engineering department at the University of Wisconsin (UW) to discuss the possibility of offering graduate courses by videotape for academic credit and earning a master's degree. Four years later, two utility employees from Northern States Power (NSP) and Wisconsin Electric Power Companies (WEPCO) graduated from the University of Wisconsin with master's degrees earned entirely by taking videotape graduate courses at their individual nuclear power plant sites. Within these 4 years, more than a dozen videotaped graduate courses were developed by the faculty of the department in a formalized master's degree program in nuclear engineering and engineering physics. This paper outlines the program's development and its current features

  3. IAEA provisional code of practice on management of radioactive waste from nuclear power plants

    International Nuclear Information System (INIS)

    1982-10-01

    This Code of Practice defines the minimum requirements for operations and design of structures, systems and components important for management of wastes from thermal nuclear power plants. It emphasizes what safety requirements shall be met rather than specifies how these requirements can be met; the latter aspect is covered in Safety Guides. The Code defines the need for a Government to assume responsibility for regulating waste management practices in conjunction with the regulation of a nuclear power plant. The Code does not prejudge the organization of the regulatory authority, which may differ from one Member State to another, and may involve more than one body. Similarly, the Code does not deal specifically with the functions of a regulatory authority responsible for such matters, although it may be of value to Member States in providing a basis for consideration of such functions. The Code deals with the entire management system for all wastes from nuclear power plants embodying thermal reactors including PWR, BWR, HWR and HTGR technologies. Topics included are: design, normal and abnormal operation, and regulation of management systems for gaseous, liquid and solid wastes, including decommissioning wastes. The Code includes measures to be taken with regard to the wastes arising from spent fuel management at nuclear power plants. However, the options for further management of spent fuel are only outlined since it is the subject of decisions by individual Member States. The Code does not require that an option(s) be decided upon prior to construction or operation of a nuclear power plant

  4. Reactor use in nuclear engineering programs

    International Nuclear Information System (INIS)

    Murray, R.L.

    1975-01-01

    Nuclear reactors for dual use in training and research were established at about 50 universities in the period since 1950, with assistance by the U. S. Atomic Energy Commission and the National Science Foundation. Most of the reactors are in active use for a variety of educational functions--laboratory teaching of undergraduates and graduate students, graduate research, orientation of visitors, and nuclear power plant reactor operator training, along with service to the technical community. As expected, the higher power reactors enjoy a larger average weekly use. Among special programs are reactor sharing and high-school teachers' workshops

  5. Impact of ACI-ASME code on design and construction of nuclear containment structures

    International Nuclear Information System (INIS)

    Reedy, R.F.

    1978-01-01

    The effect of the ACI-ASME code for design and construction of concrete containment structures on the nuclear and concrete industries is examined. Topics covered include purpose of the code, general requirements, responsibilities and duties, design and construction specifications, quality assurance, inspection, the liner, and stamping

  6. Sensibility analysis of fuel depletion using different nuclear fuel depletion codes

    International Nuclear Information System (INIS)

    Martins, F.; Velasquez, C.E.; Castro, V.F.; Pereira, C.; Silva, C. A. Mello da

    2017-01-01

    Nowadays, the utilization of different nuclear codes to perform the depletion and criticality calculations has been used to simulated nuclear reactors problems. Therefore, the goal is to analyze the sensibility of the fuel depletion of a PWR assembly using three different nuclear fuel depletion codes. The burnup calculations are performed using the codes MCNP5/ORIGEN2.1 (MONTEBURNS), KENO-VI/ORIGEN-S (TRITONSCALE6.0) and MCNPX (MCNPX/CINDER90). Each nuclear code performs the burnup using different depletion codes. Each depletion code works with collapsed energies from a master library in 1, 3 and 63 groups, respectively. Besides, each code uses different ways to obtain neutron flux that influences the depletions calculation. The results present a comparison of the neutronic parameters and isotopes composition such as criticality and nuclides build-up, the deviation in results are going to be assigned to features of the depletion code in use, such as the different radioactive decay internal libraries and the numerical method involved in solving the coupled differential depletion equations. It is also seen that the longer the period is and the more time steps are chosen, the larger the deviation become. (author)

  7. Sensibility analysis of fuel depletion using different nuclear fuel depletion codes

    Energy Technology Data Exchange (ETDEWEB)

    Martins, F.; Velasquez, C.E.; Castro, V.F.; Pereira, C.; Silva, C. A. Mello da, E-mail: felipmartins94@gmail.com, E-mail: carlosvelcab@hotmail.com, E-mail: victorfariascastro@gmail.com, E-mail: claubia@nuclear.ufmg.br, E-mail: clarysson@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-07-01

    Nowadays, the utilization of different nuclear codes to perform the depletion and criticality calculations has been used to simulated nuclear reactors problems. Therefore, the goal is to analyze the sensibility of the fuel depletion of a PWR assembly using three different nuclear fuel depletion codes. The burnup calculations are performed using the codes MCNP5/ORIGEN2.1 (MONTEBURNS), KENO-VI/ORIGEN-S (TRITONSCALE6.0) and MCNPX (MCNPX/CINDER90). Each nuclear code performs the burnup using different depletion codes. Each depletion code works with collapsed energies from a master library in 1, 3 and 63 groups, respectively. Besides, each code uses different ways to obtain neutron flux that influences the depletions calculation. The results present a comparison of the neutronic parameters and isotopes composition such as criticality and nuclides build-up, the deviation in results are going to be assigned to features of the depletion code in use, such as the different radioactive decay internal libraries and the numerical method involved in solving the coupled differential depletion equations. It is also seen that the longer the period is and the more time steps are chosen, the larger the deviation become. (author)

  8. Graduate nuclear engineering programmes motivate educational and research activities

    International Nuclear Information System (INIS)

    Mavko, B.

    2000-01-01

    Some fifteen years ago the University of Ljubljana, Faculty for Mathematics and Physics together with the national research organisation the J. Stefan jointly established a Graduate programme of Nuclear Engineering. From the onset, the programme focused on nuclear technology, nuclear safety, and reactor physics and environment protection. Over the years this graduate programme has became the focal point of nuclear related, research and educational activities in Slovenia. It has grown into a meeting ground for recognised national and distinguished foreign educators and experienced professionals from the industry. In conjunction with an important national project, supported by the Slovenian government, entitled 'Jung Researcher' it also enhances the knowledge transfer to the next generation. Since the programme was introduced, the interest for this programme has been steadily growing. Accordingly, a number of PhD and MS degrees in NE have been awarded. The graduates of this programme have encountered very good job opportunities in nuclear as well as in non-nuclear sector. (author)

  9. Digital solutions of some problems bordering on nuclear engineering

    International Nuclear Information System (INIS)

    Goldstein, S.

    1974-01-01

    The essential features of the PASTEL code, developed at the Saclay nuclear Research Centre and using the method of finite elements in bidimensional geometry, are outlined. Some examples of application are given development of stresses in a graphite block of an HTR reactor; creep in fast reactor vessel; shock between two blocks of graphite; thermoplastic cycle [fr

  10. About the application of MCNP4 code in nuclear reactor core design calculations

    International Nuclear Information System (INIS)

    Svarny, J.

    2000-01-01

    This paper provides short review about application of MCNP code for reactor physics calculations performed in SKODA JS. Problems of criticality safety analysis of spent fuel systems for storage and transport of spent fuel are discussed and relevant applications are presented. Application of standard Monte Carlo code for accelerator driven system for LWR waste destruction is shown and conclusions are reviewed. Specific heterogeneous effects in neutron balance of WWER nuclear cores are solved for adjusting standard design codes. (Authors)

  11. Engineered barrier development for a nuclear waste repository in basalt

    International Nuclear Information System (INIS)

    Smith, M.J.

    1980-05-01

    The BWIP Engineered Barrier Program has been developed to provide an integrated approach to the development of site-specific Engineered Barrier assemblages for a repository located in basalt. The goal of this program is to specify engineered and natural barriers which will ensure that nuclear and non-radioactive hazardous materials emplaced in a repository in basalt do not exceed acceptable rates of release to the biosphere. A wide range of analytical and experimental activities related to the basalt repository environment, waste package environment, waste/barrier/rock interactions, and barrier performance assessment provide the basis for selection of systems capable of meeting licensing requirements. Work has concentrated on specifying and testing natural and man-made materials which can be used to plug boreholes in basalt and which can be used as multiple barriers to surround nuclear waste forms and containers. The Engineered Barriers Program is divided into two major activities: multiple barrier studies and borehole plugging. 8 figures, 4 tables

  12. Nuclear plant engineering work and integrated management system

    International Nuclear Information System (INIS)

    Ohkubo, Y.; Obata, T.; Tanaka, K.

    1992-01-01

    The Application of computers to the design, engineering, manufacturing and construction works of nuclear power plants has greatly contributed to improvement of productivity and reliability in the nuclear power plants constructed by Mitsubishi Nuclear Group for more than ten years. However, in most cases, those systems have been developed separately and utilized independently in different computer software and hardware environments and have not been fully utilized to achieve high efficiency and reliability. In order to drastically increase the productivity and efficiency, development of NUclear power plant engineering Work and INtegrated manaGement System (NUWINGS) started in 1987 to unify and integrate various conventional and developing systems using the state-of-the-art computer technology. The NUWINGS is almost completed and is now applied to actual plant construction. (author)

  13. Nuclear engineering in Belgium and abroad

    International Nuclear Information System (INIS)

    Gaube, M.

    1992-01-01

    The new facilities for the treatment and storage of radioactive waste in Mol/Dessel are described. In 1985 - 1990 the Doel-1 and Doel-2 nuclear power stations were fitted with new safety systems. These are presented briefly as is Belgian work for Poland (assessment of the Zarnowiec Project) and Taiwan (data bank system for radioactive waste). 2 figs

  14. Human factor engineering applied to nuclear power plant design

    International Nuclear Information System (INIS)

    Manrique, A.; Valdivia, J.C.

    2007-01-01

    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. 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. Such a plan should state: -) Activities to be performed, and -) Creation of a Human Factor Engineering team adequately qualified. The Human Factor Engineering team is an integral part of the design team and is strongly linked to the engineering organizations but simultaneously has independence to act and is free to evaluate designs and propose changes in order to enhance human behavior. TECNATOM S.A. (a Spanish company) has been a part of the Design and Human Factor Engineering Team and has collaborated in the design of an advanced Nuclear Power Plant, developing methodologies and further implementing those methodologies in the design of the plant systems through the development of the plant systems operational analysis and of the man-machine interface design. The methodologies developed are made up of the following plans: -) Human Factor Engineering implementation in the Man-Machine Interface design; -) Plant System Functional Requirement Analysis; -) Allocation of Functions to man/machine; -) Task Analysis; -) Human-System Interface design; -) Control Room Verification and -) Validation

  15. Certain aspects of the environmental impact of nuclear power engineering and thermal power engineering

    Energy Technology Data Exchange (ETDEWEB)

    Malenchenko, A F [AN Belorusskoj SSR, Minsk. Inst. Yadernoj Ehnergetiki

    1979-01-01

    A review is made of the both environmental impact and hazard to man resulting from nuclear power engineering as compared with those of thermal power engineering. At present, in addition to such criteria, as physical-chemical characteristic of energy sources, their efficiency and accessibility for exploitation, new requirements were substantiated in relation to safety of their utilization for environment. So, one of essential problems of nuclear power engineering development consists in assessment and prediction of radioecological consequence. The analysis and operating experience of more than 1000 reactor/years with no accidents and harm for pupulation show, that in respect to impact on environment and man nuclear power engineering is much more safe in comparison with energy sources using tradidional fossile fuel.

  16. IEEE [Institute of Electrical and Electronics Engineers] standards and nuclear software quality engineering

    International Nuclear Information System (INIS)

    Daughtrey, T.

    1988-01-01

    Significant new nuclear-specific software standards have recently been adopted under the sponsorship of the American Nuclear Society and the American Society of Mechanical Engineers. The interest of the US Nuclear Regulatory Commission has also been expressed through their issuance of NUREG/CR-4640. These efforts all indicate a growing awareness of the need for thorough, referenceable expressions of the way to build in and evaluate quality in nuclear software. A broader professional perspective can be seen in the growing number of software engineering standards sponsored by the Institute of Electrical and Electronics Engineers (IEEE) Computer Society. This family of standards represents a systematic effort to capture professional consensus on quality practices throughout the software development life cycle. The only omission-the implementation phase-is treated by accepted American National Standards Institute or de facto standards for programming languages

  17. 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.

  18. 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.)

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

    International Nuclear Information System (INIS)

    Talou, P.; Chadwick, M.B.; Dietrich, F.S.; Herman, M.; Kawano, T.; Konig, A.; Oblozinsky, 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.

  20. Current earthquake engineering practice for Japanese nuclear power plants

    International Nuclear Information System (INIS)

    Hofmayer, C.H.; Park, Y.J.; Costello, J.F.

    1992-01-01

    This paper provides a brief overview of seismic research being conducted in Japan and describes USNRC efforts to understand Japanese seismic practice. Current earthquake engineering practice for Japanese nuclear power plants is descried in JEAG 4601-1987, ''Technical Guidelines for Aseismic Design of Nuclear Power Plants.'' The USNRC has sponsored BNL to translate this document into English. Efforts are underway to study and understand JEAG 4601-1987 and make the translation more readily available in the United States

  1. 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.

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

    International Nuclear Information System (INIS)

    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

  3. Upgrade and benchmarking of the NIFS physics-engineering-cost code

    International Nuclear Information System (INIS)

    Dolan, T.J.; Yamazaki, K.

    2004-07-01

    The NIFS Physics-Engineering-Cost (PEC) code for helical and tokamak fusion reactors is upgraded by adding data from three blanket-shield designs, a new cost section based on the ARIES cost schedule, more recent unit costs, and improved algorithms for various computations. The PEC code is also benchmarked by modeling the ARIES-AT (advanced technology) tokamak and the ARIES-SPPS (stellarator power plant system). The PEC code succeeds in predicting many of the pertinent plasma parameters and reactor component masses within about 10%. There are cost differences greater than 10% for some fusion power core components, which may be attributed to differences of unit costs used by the codes. The COEs estimated by the PEC code differ from the COEs of the ARIES-AT and ARIES-SPPS studies by 5%. (author)

  4. Use of reliability engineering tools in safety and risk assessment of nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    Raso, Amanda Laureano; Vasconcelos, Vanderley de; Marques, Raíssa Oliveira; Soares, Wellington Antonio; Mesquita, Amir Zacarias, E-mail: amandaraso@hotmail.com, E-mail: vasconv@cdtn.br, E-mail: raissaomarques@gmail.com, E-mail: soaresw@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Serviço de Tecnologia de Reatores

    2017-07-01

    Safety, reliability and availability are fundamental criteria in design, construction and operation of nuclear facilities, as nuclear power plants. Deterministic and probabilistic risk assessments of such facilities are required by regulatory authorities in order to meet licensing regulations, contributing to assure safety, as well as reduce costs and environmental impacts. Probabilistic Risk Assessment has become an important part of licensing requirements of the nuclear power plants in Brazil and in the world. Risk can be defined as a qualitative and/or quantitative assessment of accident sequence frequencies (or probabilities) and their consequences. Risk management is a systematic application of management policies, procedures and practices to identify, analyze, plan, implement, control, communicate and document risks. Several tools and computer codes must be combined, in order to estimate both probabilities and consequences of accidents. Event Tree Analysis (ETA), Fault Tree Analysis (FTA), Reliability Block Diagrams (RBD), and Markov models are examples of evaluation tools that can support the safety and risk assessment for analyzing process systems, identifying potential accidents, and estimating consequences. Because of complexity of such analyzes, specialized computer codes are required, such as the reliability engineering software develop by Reliasoft® Corporation. BlockSim (FTA, RBD and Markov models), RENO (ETA and consequence assessment), Weibull++ (life data and uncertainty analysis), and Xfmea (qualitative risk assessment) are some codes that can be highlighted. This work describes an integrated approach using these tools and software to carry out reliability, safety, and risk assessment of nuclear facilities, as well as, and application example. (author)

  5. Use of reliability engineering tools in safety and risk assessment of nuclear facilities

    International Nuclear Information System (INIS)

    Raso, Amanda Laureano; Vasconcelos, Vanderley de; Marques, Raíssa Oliveira; Soares, Wellington Antonio; Mesquita, Amir Zacarias

    2017-01-01

    Safety, reliability and availability are fundamental criteria in design, construction and operation of nuclear facilities, as nuclear power plants. Deterministic and probabilistic risk assessments of such facilities are required by regulatory authorities in order to meet licensing regulations, contributing to assure safety, as well as reduce costs and environmental impacts. Probabilistic Risk Assessment has become an important part of licensing requirements of the nuclear power plants in Brazil and in the world. Risk can be defined as a qualitative and/or quantitative assessment of accident sequence frequencies (or probabilities) and their consequences. Risk management is a systematic application of management policies, procedures and practices to identify, analyze, plan, implement, control, communicate and document risks. Several tools and computer codes must be combined, in order to estimate both probabilities and consequences of accidents. Event Tree Analysis (ETA), Fault Tree Analysis (FTA), Reliability Block Diagrams (RBD), and Markov models are examples of evaluation tools that can support the safety and risk assessment for analyzing process systems, identifying potential accidents, and estimating consequences. Because of complexity of such analyzes, specialized computer codes are required, such as the reliability engineering software develop by Reliasoft® Corporation. BlockSim (FTA, RBD and Markov models), RENO (ETA and consequence assessment), Weibull++ (life data and uncertainty analysis), and Xfmea (qualitative risk assessment) are some codes that can be highlighted. This work describes an integrated approach using these tools and software to carry out reliability, safety, and risk assessment of nuclear facilities, as well as, and application example. (author)

  6. Development of the next generation code system as an engineering modeling language. (2). Study with prototyping

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Uto, Nariaki; Kasahara, Naoto; Ishikawa, Makoto

    2003-04-01

    In the fast reactor development, numerical simulation using analytical codes plays an important role for complementing theory and experiment. It is necessary that the engineering models and analysis methods can be flexibly changed, because the phenomena to be investigated become more complicated due to the diversity of the needs for research. And, there are large problems in combining physical properties and engineering models in many different fields. Aiming to the realization of the next generation code system which can solve those problems, the authors adopted three methods, (1) Multi-language (SoftWIRE.NET, Visual Basic.NET and Fortran) (2) Fortran 90 and (3) Python to make a prototype of the next generation code system. As this result, the followings were confirmed. (1) It is possible to reuse a function of the existing codes written in Fortran as an object of the next generation code system by using Visual Basic.NET. (2) The maintainability of the existing code written by Fortran 77 can be improved by using the new features of Fortran 90. (3) The toolbox-type code system can be built by using Python. (author)

  7. The School for Nuclear Engineering is 25 years old: Where teachers go to school

    International Nuclear Information System (INIS)

    Knapp, W.

    1986-01-01

    For a quarter of a century the School for Nuclear Engineering at the Karlsruhe Nuclear Research Centre has provided further education in all areas of nuclear engineering. The courses are attended by all kinds of people: school teachers, shift managers in nuclear power stations and engineers from the Third World, for example. (orig.) [de

  8. Nuclear piston engine and pulsed gaseous core reactor power systems

    International Nuclear Information System (INIS)

    Dugan, E.T.

    1976-01-01

    The investigated nuclear piston engines consist of a pulsed, gaseous core reactor enclosed by a moderating-reflecting cylinder and piston assembly and operate on a thermodynamic cycle similar to the internal combustion engine. The primary working fluid is a mixture of uranium hexafluoride, UF 6 , and helium, He, gases. Highly enriched UF 6 gas is the reactor fuel. The helium is added to enhance the thermodynamic and heat transfer characteristics of the primary working fluid and also to provide a neutron flux flattening effect in the cylindrical core. Two and four-stroke engines have been studied in which a neutron source is the counterpart of the sparkplug in the internal combustion engine. The piston motions which have been investigated include pure simple harmonic, simple harmonic with dwell periods, and simple harmonic in combination with non-simple harmonic motion. The results of the conducted investigations indicate good performance potential for the nuclear piston engine with overall efficiencies of as high as 50 percent for nuclear piston engine power generating units of from 10 to 50 Mw(e) capacity. Larger plants can be conceptually designed by increasing the number of pistons, with the mechanical complexity and physical size as the probable limiting factors. The primary uses for such power systems would be for small mobile and fixed ground-based power generation (especially for peaking units for electrical utilities) and also for nautical propulsion and ship power

  9. Development of System Engineering Technology for Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    Kim, Ho Dong; Kim, Sung Ki; Song, Kee Chan

    2010-04-01

    This report is aims to establish design requirements for constructing mock-up system of pyroprocess by 2011 to realize long-term goal of nuclear energy promotion comprehensive plan, which is construction of engineering scale pyroprocess integrated process demonstration facility. The development of efficient process for spent fuel and establishment of system engineering technology to demonstrate the process are required to develop nuclear energy continuously. The detailed contents of research for these are as follows; - Design of Mock-up facility for demonstrate pyroprocess, Construction, Approval, Trial run, Performance test - Development of nuclear material accountancy technology for unit processes of pyroprocess and design of safeguards system - Remote operation of demonstrating pyroprocess / Development of maintenance technology and equipment - Establishment of transportation system and evaluation of pre-safety for interim storage system - Deriving and implementation of a method to improve nuclear transparency for commercialization proliferation resistance nuclear fuel cycle Spent fuel which is the most important pending problem of nuclear power development would be reduced and recycled by developing the system engineering technology of pyroprocess facility by 2010. This technology would contribute to obtain JD for the use of spent fuel between the ROK-US and to amend the ROK-US Atomic Energy Agreement scheduled in 2014

  10. Progress in nuclear well logging modeling using deterministic transport codes

    International Nuclear Information System (INIS)

    Kodeli, I.; Aldama, D.L.; Maucec, M.; Trkov, A.

    2002-01-01

    Further studies in continuation of the work presented in 2001 in Portoroz were performed in order to study and improve the performances, precission and domain of application of the deterministic transport codes with respect to the oil well logging analysis. These codes are in particular expected to complement the Monte Carlo solutions, since they can provide a detailed particle flux distribution in the whole geometry in a very reasonable CPU time. Real-time calculation can be envisaged. The performances of deterministic transport methods were compared to those of the Monte Carlo method. IRTMBA generic benchmark was analysed using the codes MCNP-4C and DORT/TORT. Centric as well as excentric casings were considered using 14 MeV point neutron source and NaI scintillation detectors. Neutron and gamma spectra were compared at two detector positions.(author)

  11. Problem solving in nuclear engineering using supercomputers

    International Nuclear Information System (INIS)

    Schmidt, F.; Scheuermann, W.; Schatz, A.

    1987-01-01

    The availability of supercomputers enables the engineer to formulate new strategies for problem solving. One such strategy is the Integrated Planning and Simulation System (IPSS). With the integrated systems, simulation models with greater consistency and good agreement with actual plant data can be effectively realized. In the present work some of the basic ideas of IPSS are described as well as some of the conditions necessary to build such systems. Hardware and software characteristics as realized are outlined. (orig.) [de

  12. Engineering effects of underground nuclear explosions

    Energy Technology Data Exchange (ETDEWEB)

    Boardman, Charles R [CER Geonuclear Corporation, Las Vegas, NV (United States)

    1970-05-01

    Useful effects of contained underground nuclear explosions are discussed in light of today's most promising potential applications. Relevant data obtained through exploration of explosion environments of nine U.S. tests in competent rock are summarized and presented as a practical basis for estimating magnitudes of effects. Effects discussed include chimney configuration, permeability, and volume as well as rubble particle size distributions and extents of permeability change in the chimney wall rock. Explosion mediums include shale, granite, dolomite, and salt. (author)

  13. Engineering effects of underground nuclear explosions

    International Nuclear Information System (INIS)

    Boardman, Charles R.

    1970-01-01

    Useful effects of contained underground nuclear explosions are discussed in light of today's most promising potential applications. Relevant data obtained through exploration of explosion environments of nine U.S. tests in competent rock are summarized and presented as a practical basis for estimating magnitudes of effects. Effects discussed include chimney configuration, permeability, and volume as well as rubble particle size distributions and extents of permeability change in the chimney wall rock. Explosion mediums include shale, granite, dolomite, and salt. (author)

  14. Should Nuclear Safety Care About Resilience Engineering?

    International Nuclear Information System (INIS)

    Paries, J.

    2016-01-01

    The current nuclear industry safety paradigm is based on the deterministic and/or probabilistic anticipation of all potential situations, and the predetermination of all the (safe) responses. Even the defense in depth concept, which is the core of the nuclear safety strategy and is intended to handle situations in which part of the control is lost, heavily relies on detailed anticipations. In other words, nuclear safety is mainly expected from the real world’s conformity to a designed-to-be-safe world, i.e., a well controlled world, where organizations, processes, hardware, teams, and individuals comply with their rationally predetermined behaviors. In this “command and control” perspective, risk is seen as mainly generated by deviations and variations from rules, procedures, norms, and expectations. However, real operations are complex, even in normal situations, which means that they include some unpredictable events and adaptation behaviors. The traditional “command and control” perspective fail to properly acknowledge the limits to predictability inherent to a complex adaptive system. It actually strives to reduce complexity through tighter compliance to specifications and to improve predictions capabilities through a tighter monitoring of “weak signals” and “precursors”. But in a complex world, precursors are usually obvious after the event, while not identifiable before. And the efforts made to reduce complexity may also simultaneously tighten couplings between system’s components— hence increase complexity — and reduce the diversity and flexibility needed to respond to it.

  15. Notes on nuclear reactor core analysis code: CITATION

    International Nuclear Information System (INIS)

    Cepraga, D.G.

    1980-01-01

    The method which has evolved over the years for making power reactor calculations is the multigroup diffusion method. The CITATION code is designed to solve multigroup neutronics problems with application of the finite-difference diffusion theory approximation to neutron transport in up to three-dimensional geometry. The first part of this paper presents information about the mathematical equations programmed along with background material and certain displays to convey the nature of some of the formulations. The results obtained with the CITATION code regarding the neutron and burnup core analysis for a typical PWR reactor are presented in the second part of this paper. (author)

  16. Integrated computer codes for nuclear power plant severe accident analysis

    International Nuclear Information System (INIS)

    Jordanov, I.; Khristov, Y.

    1995-01-01

    This overview contains a description of the Modular Accident Analysis Program (MAAP), ICARE computer code and Source Term Code Package (STCP). STCP is used to model TMLB sample problems for Zion Unit 1 and WWER-440/V-213 reactors. Comparison is made of STCP implementation on VAX and IBM systems. In order to improve accuracy, a double precision version of MARCH-3 component of STCP is created and the overall thermal hydraulics is modelled. Results of modelling the containment pressure, debris temperature, hydrogen mass are presented. 5 refs., 10 figs., 2 tabs

  17. Integrated computer codes for nuclear power plant severe accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Jordanov, I; Khristov, Y [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika

    1996-12-31

    This overview contains a description of the Modular Accident Analysis Program (MAAP), ICARE computer code and Source Term Code Package (STCP). STCP is used to model TMLB sample problems for Zion Unit 1 and WWER-440/V-213 reactors. Comparison is made of STCP implementation on VAX and IBM systems. In order to improve accuracy, a double precision version of MARCH-3 component of STCP is created and the overall thermal hydraulics is modelled. Results of modelling the containment pressure, debris temperature, hydrogen mass are presented. 5 refs., 10 figs., 2 tabs.

  18. Title list of selected Soviet reports in the fields of nuclear research and nuclear engineering. 9

    International Nuclear Information System (INIS)

    Schumann, G.; Sube, R.

    1979-03-01

    140 titles of reports issued by Soviet institutes (KFTI, EFI, IAE, IFVE, ITEF, NIIAR, NIIEFA, FEI, RI, SFTI) in the fields of nuclear research and nuclear engineering have been arranged according to the INIS subject scope. The reports are available on a loan basis from ZfK Rossendorf, Information Department, Dresden

  19. Title list of selected Soviet reports in the fields of nuclear research and nuclear engineering. 7

    International Nuclear Information System (INIS)

    Schumann, G.; Sube, R.

    1979-03-01

    139 titles of reports issued by Soviet institutes (KFTI, EFI, IAE, IFVE, ITEF, NIIAR, NIIEFA, FEI, RI, SFTI) in the fields of nuclear research and nuclear engineering have been arranged according to the INIS subject scope. The reports are available on a loan basis from ZfK Rossendorf, Information Department, Dresden

  20. Title list of selected Soviet reports in the fields of nuclear research and nuclear engineering. 13

    International Nuclear Information System (INIS)

    Schumann, G.; Sube, R.

    1979-12-01

    136 titles of reports issued by Soviet institutes (KFTI, EFI, IAE, IFVE, ITEF, NIIAR, NIIEFA, FEI, RI, SFTI) in the fields of nuclear research and nuclear engineering have been arranged according to the INIS subject scope. The reports are available on a loan basis from ZfK Rossendorf, Information Department, Dresden

  1. Title list of selected Soviet reports in the fields of nuclear research and nuclear engineering. 14

    International Nuclear Information System (INIS)

    Schumann, G.; Sube, R.

    1979-12-01

    136 titles of reports issued by Soviet institutes (KFTI, EFI, IAE, IFVE, ITEF, NIIAR, NIIEFA, FEI, RI, SFTI) in the fields of nuclear research and nuclear engineering have been arranged according to the INIS subject scope. The reports are available on a loan basis from ZfK Rossendorf, Information Department, Dresden

  2. Title list of selected Soviet reports in the fields of nuclear research and nuclear engineering. 11

    International Nuclear Information System (INIS)

    Schumann, G.; Sube, R.

    1979-07-01

    158 titles of reports issued by Soviet institutes (KFTI, EFI, IAE, IFVE, ITEF, NIIAR, NIIEFA, FEI, RI, SFTI) in the fields of nuclear research and nuclear engineering have been arranged according to the INIS subject scope. The reports are available on a loan basis from ZfK Rossendorf, Information Department, Dresden

  3. Title list of selected Soviet reports in the fields of nuclear research and nuclear engineering. 6

    International Nuclear Information System (INIS)

    Schumann, G.; Sube, R.

    1979-03-01

    135 titles of reports issued by Soviet institutes (KFTI, EFI, IAE, IFVE, ITEF, NIIAR, NIIEFA, FEI, RI, SFTI) in the fields of nuclear research and nuclear engineering have been arranged according to the INIS subject scope. The reports are available on a loan basis from ZfK Rossendorf, Information Department, Dresden

  4. Title list of selected Soviet reports in the fields of nuclear research and nuclear engineering. 5

    International Nuclear Information System (INIS)

    Schumann, G.; Sube, R.

    1979-03-01

    133 titles of reports issued by Soviet institutes (KFTI, EFI, IAE, IFVE, ITEF, NIIAR, NIIEFA, FEI, RI, SFTI) in the fields of nuclear research and nuclear engineering have been arranged according to the INIS subject scope. The reports are available on a loan basis from ZfK Rossendorf, Information Department, Dresden

  5. Title list of selected Soviet reports in the fields of nuclear research and nuclear engineering. 12

    International Nuclear Information System (INIS)

    Schumann, G.; Sube, R.

    1979-12-01

    136 titles of reports issued by Soviet institutes (KFTI, EFI, IAE, IFVE, ITEF, NIIAR, NIIEFA, FEI, RI, SFTI) in the fields of nuclear research and nuclear engineering have been arranged according to the INIS subject scope. The reports are available on a loan basis from ZfK Rossendorf, Information Department, Dresden

  6. Title list of selected Soviet reports in the fields of nuclear research and nuclear engineering. 8

    International Nuclear Information System (INIS)

    Schumann, G.; Sube, R.

    1979-03-01

    141 titles of reports issued by Soviet institutes (KFTI, EFI, IAE, IFVE, ITEF, NIIAR, NIIEFA, FEI, RI, SFTI) in the fields of nuclear research and nuclear engineering have been arranged according to the INIS subject scope. The reports are available on a loan basis from ZfK Rossendorf, Information Department, Dresden

  7. Title list of selected Soviet reports in the fields of nuclear research and nuclear engineering. 4

    International Nuclear Information System (INIS)

    Schumann, G.; Sube, R.

    1979-03-01

    135 titles of reports issued by Soviet institutes (KFTI, EFI, IAE, IFVE, ITEF, NIIAR, NIIEFA, FEI, RI, SFTI) in the fields of nuclear research and nuclear engineering have been arranged according to the INIS subject scope. The reports are available on a loan basis from ZfK Rossendorf, Information Department, Dresden

  8. Title list of selected Soviet reports in the fields of nuclear research and nuclear engineering. 15

    International Nuclear Information System (INIS)

    Schumann, G.; Sube, R.

    1979-12-01

    137 titles of reports issued by Soviet institutes (KFTI, EFI, IAE, IFVE, ITEF, NIIAR, NIIEFA, FEI, RI, SFTI) in the fields of nuclear research and nuclear engineering have been arranged according to the INIS subject scope. The reports are available on a loan basis from ZfK Rossendorf, Information Department, Dresden

  9. Title list of selected Soviet reports in the fields of nuclear research and nuclear engineering. 10

    International Nuclear Information System (INIS)

    Schumann, G.; Sube, R.

    1979-07-01

    142 titles of reports issued by Soviet institutes (KFTI, EFI, IAE, IFVE, ITEF, NIIAR, NIIEFA, FEI, RI, SFTI) in the fields of nuclear research and nuclear engineering have been arranged according to the INIS subject scope. The reports are available on a loan basis from ZfK Rossendorf, Information Department, Dresden

  10. Title list of selected Soviet reports in the fields of nuclear research and nuclear engineering. 0

    International Nuclear Information System (INIS)

    Schumann, G.; Sube, R.

    1977-09-01

    161 titles of reports issued by Soviet institutes (KFTI, EFI, IAE, IFVE, ITEF, NIIAR, NIIEFA, FEI, RI, SFTI) in the fields of nuclear research and nuclear engineering have been arranged according to the INIS subject scope. The reports are availabl on a loan basis from ZfK Rossendorf, Information Department, Dresden

  11. CASKETSS: a computer code system for thermal and structural analysis of nuclear fuel shipping casks

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1989-02-01

    A computer program CASKETSS has been developed for the purpose of thermal and structural analysis of nuclear fuel shipping casks. CASKETSS measn a modular code system for CASK Evaluation code system Thermal and Structural Safety. Main features of CASKETSS are as follow; (1) Thermal and structural analysis computer programs for one-, two-, three-dimensional geometries are contained in the code system. (2) Some of the computer programs in the code system has been programmed to provide near optimal speed on vector processing computers. (3) Data libralies fro thermal and structural analysis are provided in the code system. (4) Input data generator is provided in the code system. (5) Graphic computer program is provided in the code system. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)

  12. The CCONE Code System and its Application to Nuclear Data Evaluation for Fission and Other Reactions

    Science.gov (United States)

    Iwamoto, O.; Iwamoto, N.; Kunieda, S.; Minato, F.; Shibata, K.

    2016-01-01

    A computer code system, CCONE, was developed for nuclear data evaluation within the JENDL project. The CCONE code system integrates various nuclear reaction models needed to describe nucleon, light charged nuclei up to alpha-particle and photon induced reactions. The code is written in the C++ programming language using an object-oriented technology. At first, it was applied to neutron-induced reaction data on actinides, which were compiled into JENDL Actinide File 2008 and JENDL-4.0. It has been extensively used in various nuclear data evaluations for both actinide and non-actinide nuclei. The CCONE code has been upgraded to nuclear data evaluation at higher incident energies for neutron-, proton-, and photon-induced reactions. It was also used for estimating β-delayed neutron emission. This paper describes the CCONE code system indicating the concept and design of coding and inputs. Details of the formulation for modelings of the direct, pre-equilibrium and compound reactions are presented. Applications to the nuclear data evaluations such as neutron-induced reactions on actinides and medium-heavy nuclei, high-energy nucleon-induced reactions, photonuclear reaction and β-delayed neutron emission are mentioned.

  13. The CCONE Code System and its Application to Nuclear Data Evaluation for Fission and Other Reactions

    Energy Technology Data Exchange (ETDEWEB)

    Iwamoto, O., E-mail: iwamoto.osamu@jaea.go.jp; Iwamoto, N.; Kunieda, S.; Minato, F.; Shibata, K.

    2016-01-15

    A computer code system, CCONE, was developed for nuclear data evaluation within the JENDL project. The CCONE code system integrates various nuclear reaction models needed to describe nucleon, light charged nuclei up to alpha-particle and photon induced reactions. The code is written in the C++ programming language using an object-oriented technology. At first, it was applied to neutron-induced reaction data on actinides, which were compiled into JENDL Actinide File 2008 and JENDL-4.0. It has been extensively used in various nuclear data evaluations for both actinide and non-actinide nuclei. The CCONE code has been upgraded to nuclear data evaluation at higher incident energies for neutron-, proton-, and photon-induced reactions. It was also used for estimating β-delayed neutron emission. This paper describes the CCONE code system indicating the concept and design of coding and inputs. Details of the formulation for modelings of the direct, pre-equilibrium and compound reactions are presented. Applications to the nuclear data evaluations such as neutron-induced reactions on actinides and medium-heavy nuclei, high-energy nucleon-induced reactions, photonuclear reaction and β-delayed neutron emission are mentioned.

  14. Improved predictions of nuclear reaction rates with the TALYS reaction code for astrophysical applications

    International Nuclear Information System (INIS)

    Goriely, S.; Hilaire, S.; Koning, A.J

    2008-01-01

    Context. Nuclear reaction rates of astrophysical applications are traditionally determined on the basis of Hauser-Feshbach reaction codes. These codes adopt a number of approximations that have never been tested, such as a simplified width fluctuation correction, the neglect of delayed or multiple-particle emission during the electromagnetic decay cascade, or the absence of the pre-equilibrium contribution at increasing incident energies. Aims. The reaction code TALYS has been recently updated to estimate the Maxwellian-averaged reaction rates that are of astrophysical relevance. These new developments enable the reaction rates to be calculated with increased accuracy and reliability and the approximations of previous codes to be investigated. Methods. The TALYS predictions for the thermonuclear rates of relevance to astrophysics are detailed and compared with those derived by widely-used codes for the same nuclear ingredients. Results. It is shown that TALYS predictions may differ significantly from those of previous codes, in particular for nuclei for which no or little nuclear data is available. The pre-equilibrium process is shown to influence the astrophysics rates of exotic neutron-rich nuclei significantly. For the first time, the Maxwellian- averaged (n, 2n) reaction rate is calculated for all nuclei and its competition with the radiative capture rate is discussed. Conclusions. The TALYS code provides a new tool to estimate all nuclear reaction rates of relevance to astrophysics with improved accuracy and reliability. (authors)

  15. 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.

  16. Ground motion prediction needs for nuclear engineering design

    International Nuclear Information System (INIS)

    Hadjian, A.H.

    1985-01-01

    The basic design philosophy of nuclear power plants stipulates that the risk to the public be as low as reasonably achievable. As a result of this philosophy, the seismic design of nuclear power plants has tended, over time, to diverge from that of other engineered structures. The emphasis at the present time is to specify ground motion at a nuclear facility site as realistically as possible and to design all safety-related structures to respond to the specified ground motion in the elastic range. The characteristics of this realistic design ground motion are discussed and present prediction needs identified

  17. IAEA program for the preparation of safety codes and guides for nuclear power plants

    International Nuclear Information System (INIS)

    1975-01-01

    On the 13th of September, 1974, the IAEA Governors' Council has given its consent to the programme for the establishment of safety codes and guides (annex VII to IAEA document G.C. (XVIII/526)). The programme envisages the establishment of one code of practice for each of the issues governmental organization, siting, design, operation and quality assurance and also of about 50 safety guides between 1975 and 1980. These codes will contain the minimum requirements for the safety of the nuclear power stations, their systems and components. The guides will recommend methods to achieve the aims stated in the codes. It is the purpose of these IAEA activities to provide recommendations and guiding rules which may serve as standards for the assessment of the safety of nuclear power stations for all nations which may become participants in the peaceful use of nuclear energy within the next few years. (orig./AK) [de

  18. 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)

  19. 4+DTM Soft Power for Nuclear Systems Engineering

    International Nuclear Information System (INIS)

    Suh, Kune Y.

    2006-01-01

    Nuclear Power Plants (NPPs) of a bulky and complex structure consisting of hundreds of thousands of parts require huge volume of data during the design, construction, operation, maintenance and decommissioning stages. The systems engineering thus calls for a fully automated way of managing the information spanning their life cycle, i. e. from cradle to grave. In line with practice in disciplines of naval architecture, aerospace engineering, and automotive manufacturing, the paper proposes total digital systems engineering based on three-dimensional (3D) computer-aided design (CAD) models. The signature in the proposal lies with the four-plus-dimensional (4 + D) TM Soft Power engineering, a critical technology for digital management. The technology proposed in the 3D space and time plus cost coordinates, i. e. 4 + D TM coordinates, constitutes, the backbone of digital engineering in the nuclear systems design. This solution will help the preliminary simulation capability for NPP to supply the vital information not only for the design and management of the engineered structures but also for the online maintenance. NPP can be built utilizing the optimized construction schedule and the structural design. The efficiency of project management will also be improved by dynamically storing voluminous information in the advanced database. The 4 + D TM digital engineering will eventually lead to paperless design and construction planing in the global marker place

  20. Reliability engineering for nuclear and other high technology systems

    International Nuclear Information System (INIS)

    Lakner, A.A.; Anderson, R.T.

    1985-01-01

    This book is written for the reliability instructor, program manager, system engineer, design engineer, reliability engineer, nuclear regulator, probability risk assessment (PRA) analyst, general manager and others who are involved in system hardware acquisition, design and operation and are concerned with plant safety and operational cost-effectiveness. It provides criteria, guidelines and comprehensive engineering data affecting reliability; it covers the key aspects of system reliability as it relates to conceptual planning, cost tradeoff decisions, specification, contractor selection, design, test and plant acceptance and operation. It treats reliability as an integrated methodology, explicitly describing life cycle management techniques as well as the basic elements of a total hardware development program, including: reliability parameters and design improvement attributes, reliability testing, reliability engineering and control. It describes how these elements can be defined during procurement, and implemented during design and development to yield reliable equipment. (author)