WorldWideScience

Sample records for reactor physics methods

  1. Advanced reactor physics methods for heterogeneous reactor cores

    Science.gov (United States)

    Thompson, Steven A.

    To maintain the economic viability of nuclear power the industry has begun to emphasize maximizing the efficiency and output of existing nuclear power plants by using longer fuel cycles, stretch power uprates, shorter outage lengths, mixed-oxide (MOX) fuel and more aggressive operating strategies. In order to accommodate these changes, while still satisfying the peaking factor and power envelope requirements necessary to maintain safe operation, more complexity in commercial core designs have been implemented, such as an increase in the number of sub-batches and an increase in the use of both discrete and integral burnable poisons. A consequence of the increased complexity of core designs, as well as the use of MOX fuel, is an increase in the neutronic heterogeneity of the core. Such heterogeneous cores introduce challenges for the current methods that are used for reactor analysis. New methods must be developed to address these deficiencies while still maintaining the computational efficiency of existing reactor analysis methods. In this thesis, advanced core design methodologies are developed to be able to adequately analyze the highly heterogeneous core designs which are currently in use in commercial power reactors. These methodological improvements are being pursued with the goal of not sacrificing the computational efficiency which core designers require. More specifically, the PSU nodal code NEM is being updated to include an SP3 solution option, an advanced transverse leakage option, and a semi-analytical NEM solution option.

  2. Sensitivity and Uncertainty Analysis of Coupled Reactor Physics Problems : Method Development for Multi-Physics in Reactors

    NARCIS (Netherlands)

    Perkó, Z.

    2015-01-01

    This thesis presents novel adjoint and spectral methods for the sensitivity and uncertainty (S&U) analysis of multi-physics problems encountered in the field of reactor physics. The first part focuses on the steady state of reactors and extends the adjoint sensitivity analysis methods well

  3. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  4. A bibliography on finite element and related methods analysis in reactor physics computations (1971--1997)

    Energy Technology Data Exchange (ETDEWEB)

    Carpenter, D.C.

    1998-01-01

    This bibliography provides a list of references on finite element and related methods analysis in reactor physics computations. These references have been published in scientific journals, conference proceedings, technical reports, thesis/dissertations and as chapters in reference books from 1971 to the present. Both English and non-English references are included. All references contained in the bibliography are sorted alphabetically by the first author`s name and a subsort by date of publication. The majority of the references relate to reactor physics analysis using the finite element method. Related topics include the boundary element method, the boundary integral method, and the global element method. All aspects of reactor physics computations relating to these methods are included: diffusion theory, deterministic radiation and neutron transport theory, kinetics, fusion research, particle tracking in finite element grids, and applications. For user convenience, many of the listed references have been categorized. The list of references is not all inclusive. In general, nodal methods were purposely excluded, although a few references do demonstrate characteristics of finite element methodology using nodal methods (usually as a non-conforming element basis). This area could be expanded. The author is aware of several other references (conferences, thesis/dissertations, etc.) that were not able to be independently tracked using available resources and thus were not included in this listing.

  5. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

    Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

  6. Study on improvement of reactor physics analysis method for FBRs with various core concept

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, Toshihisa; Kitada, Takanori; Tagawa, Akihiro; Maruyama, Manabu; Takeda, Toshikazu [Osaka Univ., Suita (Japan). Dept. of Nuclear Engineering

    2000-02-01

    Investigation was made on the following three themes as a part of the improvement of reactor physics analysis method for FBR with various core concepts. Part 1: Investigation of Error Estimation of Neutron Spectra in FBR and Suggestions to Improve the Accuracy. In order to improve the spectrum unfolding method used in the fast experimental reactor JOYO, a trial was made to evaluate the error in the estimated neutron spectrum, cause by cause. And the evaluated errors were summed up to obtain the most probable and reasonable error as possible. The summed up error was found relatively small compared to the error caused by the uncertainty of cross section data: most of the error in the spectrum unfolding method can be attributed to the error in cross sections. It was also found that the error due to the fission spectrum causes a considerable error in the high energy neutron spectrum which is above several MeV. Part 2: Study on Reactor Physics Analysis Method for Gas-Cooled FBR. In gas-cooled FBR, the portion of coolant channels in core volume is larger than in sodium-cooled FBR. This leads to strong neutron streaming effects. For sodium-cooled FBR, several methods were proposed to evaluate the neutron streaming effect, however, these methods can not be used directly to gas-cooled reactor because the direction dependent diffusion coefficient becomes infinitive along the direction parallel to the coolant channel. In this study, a new method is proposed to evaluate the neutron streaming effect, based on the method taking the axial buckling into consideration, which method was originally proposed by Koehler. Part 3: Study on Reactor Physics Analysis Method for Water-Cooled FBR. An investigation was made on low-moderated water-cooled FBR, on the point that the ordinary used analysis method for FBR may give considerable difference in results in such a core. In light water cooled thermal reactors, it is well known that the space dependence of self-shielding effect of heavy

  7. Contribution of reactor physics in past and future. Is reactor physics useful?

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Toshikazu [Osaka Univ. (Japan); Kosaka, Shinya [TEPCO Systems Co. (Japan); Tatsumi, Masahiro [Nuclear Fuel Industries Ltd., Tokyo (Japan)] (and others)

    2003-02-01

    Reactor Physics is a science to create rector and to play an important role in application to calculation science and safety evaluation. This feature articles contains topics, interested problems and development problems in the following field of reactor physics such as theory and experiment of reactor physics, core control, safety evaluation, criticality safety, accelerator driven subcritical reactor (ADS), new type reactor and evaluation of reactor physics. An original nuclear calculation method developed in Japan has been applied to design and analysis of fast breeder reactor. Interested problems are a proposal of fundamental principles of progressive reactor, development of calculation science, new knowledge by application of best estimate method to safety evaluation and investigation of complicated phenomena of criticality safety. (S.Y.)

  8. Assessment of uncertainty in full core reactor physics calculations using statistical methods

    Energy Technology Data Exchange (ETDEWEB)

    McEwan, C., E-mail: mcewac2@mcmaster.ca [McMaster Univ., Hamilton, Ontario (Canada)

    2012-07-01

    The best estimate method of safety analysis involves choosing a realistic set of input parameters for a proposed safety case and evaluating the uncertainty in the results. Determining the uncertainty in code outputs remains a challenge and is the subject of a benchmarking exercise proposed by the Organization for Economic Cooperation and Development. The work proposed in this paper will contribute to this benchmark by assessing the uncertainty in a depletion calculation of the final nuclide concentrations for an experiment performed in the Fukushima-2 reactor. This will be done using lattice transport code DRAGON and a tool known as DINOSAUR. (author)

  9. Reactor antineutrinos and nuclear physics

    Energy Technology Data Exchange (ETDEWEB)

    Balantekin, A.B. [University of Wisconsin, Department of Physics, Madison, WI (United States)

    2016-11-15

    Short-baseline reactor neutrino experiments successfully measured the neutrino parameters they set out to measure, but they also identified a shape distortion in the 5-7 MeV range as well as a reduction from the predicted value of the flux. Nuclear physics input into the calculations of reactor antineutrino spectra needs to be better refined if this anomaly is to be interpreted as due to sterile neutrino states. (orig.)

  10. Fundamentals of Nuclear Reactor Physics

    CERN Document Server

    Lewis, E E

    2008-01-01

    This new streamlined text offers a one-semester treatment of the essentials of how the fission nuclear reactor works, the various approaches to the design of reactors, and their safe and efficient operation. The book includes numerous worked-out examples and end-of-chapter questions to help reinforce the knowledge presented. This textbook offers an engineering-oriented introduction to nuclear physics, with a particular focus on how those physics are put to work in the service of generating nuclear-based power, particularly the importance of neutron reactions and neutron behavior. Engin

  11. Application of invariant embedding to reactor physics

    CERN Document Server

    Shimizu, Akinao; Parsegian, V L

    1972-01-01

    Application of Invariant Embedding to Reactor Physics describes the application of the method of invariant embedding to radiation shielding and to criticality calculations of atomic reactors. The authors intend to show how this method has been applied to realistic problems, together with the results of applications which will be useful to shielding design. The book is organized into two parts. Part A deals with the reflection and transmission of gamma rays by slabs. The chapters in this section cover topics such as the reflection and transmission problem of gamma rays; formulation of the probl

  12. Hilbert space method for the numerical solution of reactor physics problems

    Energy Technology Data Exchange (ETDEWEB)

    Ackroyd, R.T. (UKAEA Risley Nuclear Power Development Labs.)

    1983-01-01

    A Hilbert space approach is used to give a unified treatment of neutron transport by finite element methods. Global solutions can be found by least squares, variational and weighted residual methods stemming from an identity. Bounds for local characteristics of solutions are found by a bi-variational method.

  13. Reactor Physics Analysis Models for a CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok

    2007-10-15

    Canada deuterium uranium (CANDU) reactor physics analysis is typically performed in three steps. At first, macroscopic cross-sections of the reference lattice is produced by modeling the reference fuel channel. Secondly macroscopic cross-sections of reactivity devices in the reactor are generated. The macroscopic cross-sections of a reactivity device are calculated as incremental cross-sections by subtracting macroscopic cross-sections of a three-dimensional lattice without reactivity device from those of a three-dimensional lattice with a reactivity device. Using the macroscopic cross-sections of the reference lattice and incremental cross-sections of the reactivity devices, reactor physics calculations are performed. This report summarizes input data of typical CANDU reactor physics codes, which can be utilized for the future CANDU reactor physics analysis.

  14. Reactor Physics Methods and Preconceptual Core Design Analyses for Conversion of the Advanced Test Reactor to Low-Enriched Uranium Fuel Annual Report for Fiscal Year 2012

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Sean R. Morrell

    2012-09-01

    Under the current long-term DOE policy and planning scenario, both the ATR and the ATRC will be reconfigured at an appropriate time within the next several years to operate with low-enriched uranium (LEU) fuel. This will be accomplished under the auspices of the Reduced Enrichment Research and Test Reactor (RERTR) Program, administered by the DOE National Nuclear Security Administration (NNSA). At a minimum, the internal design and composition of the fuel element plates and support structure will change, to accommodate the need for low enrichment in a manner that maintains total core excess reactivity at a suitable level for anticipated operational needs throughout each cycle while respecting all control and shutdown margin requirements and power distribution limits. The complete engineering design and optimization of LEU cores for the ATR and the ATRC will require significant multi-year efforts in the areas of fuel design, development and testing, as well as a complete re-analysis of the relevant reactor physics parameters for a core composed of LEU fuel, with possible control system modifications. Ultimately, revalidation of the computational physics parameters per applicable national and international standards against data from experimental measurements for prototypes of the new ATR and ATRC core designs will also be required for Safety Analysis Report (SAR) changes to support routine operations with LEU. This report is focused on reactor physics analyses conducted during Fiscal Year (FY) 2012 to support the initial development of several potential preconceptual fuel element designs that are suitable candidates for further study and refinement during FY-2013 and beyond. In a separate, but related, effort in the general area of computational support for ATR operations, the Idaho National Laboratory (INL) is conducting a focused multiyear effort to introduce modern high-fidelity computational reactor physics software and associated validation protocols to replace

  15. The physics of nuclear reactors

    CERN Document Server

    Marguet, Serge

    2017-01-01

    This comprehensive volume offers readers a progressive and highly detailed introduction to the complex behavior of neutrons in general, and in the context of nuclear power generation. A compendium and handbook for nuclear engineers, a source of teaching material for academic lecturers as well as a graduate text for advanced students and other non-experts wishing to enter this field, it is based on the author’s teaching and research experience and his recognized expertise in nuclear safety. After recapping a number of points in nuclear physics, placing the theoretical notions in their historical context, the book successively reveals the latest quantitative theories concerning: •   The slowing-down of neutrons in matter •   The charged particles and electromagnetic rays •   The calculation scheme, especially the simplification hypothesis •   The concept of criticality based on chain reactions •   The theory of homogeneous and heterogeneous reactors •   The problem of self-shielding �...

  16. Current status of fast reactor physics

    Energy Technology Data Exchange (ETDEWEB)

    Hummel, H.H.

    1979-01-01

    The subject of calculation of reactivity coefficients for fast reactors is developed, starting with a discussion of the status of relevant nuclear data and proceeding to the subjects of group cross section generation and of methods of obtaining reactivity coefficients from group cross sections. Reactivity coefficients measured in critical experiments are compared with calculated values. Dependence of reactivity coefficients on reactor design is discussed. Finally, results of the recent international comparison of calculated reactivity coefficients are presented.

  17. Current Reactor Physics Benchmark Activities at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Margaret A. Marshall; Mackenzie L. Gorham; Joseph Christensen; James C. Turnbull; Kim Clark

    2011-11-01

    The International Reactor Physics Experiment Evaluation Project (IRPhEP) [1] and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) [2] were established to preserve integral reactor physics and criticality experiment data for present and future research. These valuable assets provide the basis for recording, developing, and validating our integral nuclear data, and experimental and computational methods. These projects are managed through the Idaho National Laboratory (INL) and the Organisation for Economic Co-operation and Development Nuclear Energy Agency (OECD-NEA). Staff and students at the Department of Energy - Idaho (DOE-ID) and INL are engaged in the development of benchmarks to support ongoing research activities. These benchmarks include reactors or assemblies that support Next Generation Nuclear Plant (NGNP) research, space nuclear Fission Surface Power System (FSPS) design validation, and currently operational facilities in Southeastern Idaho.

  18. EBR-II Reactor Physics Benchmark Evaluation Report

    Energy Technology Data Exchange (ETDEWEB)

    Pope, Chad L. [Idaho State Univ., Pocatello, ID (United States); Lum, Edward S [Idaho State Univ., Pocatello, ID (United States); Stewart, Ryan [Idaho State Univ., Pocatello, ID (United States); Byambadorj, Bilguun [Idaho State Univ., Pocatello, ID (United States); Beaulieu, Quinton [Idaho State Univ., Pocatello, ID (United States)

    2017-12-28

    This report provides a reactor physics benchmark evaluation with associated uncertainty quantification for the critical configuration of the April 1986 Experimental Breeder Reactor II Run 138B core configuration.

  19. The physics of accelerator driven sub-critical reactors

    Indian Academy of Sciences (India)

    Keywords. Accelerator driven systems; nuclear waste transmutation; computer codes; reactor physics; reactor noise; kinetics; burnup; transport theory; Monte Carlo; thorium utilization; neutron multiplication; sub-criticality; sub-critical facilities.

  20. Computational mathematics and physics of fusion reactors.

    Science.gov (United States)

    Garabedian, Paul R

    2003-11-25

    Theory has contributed significantly to recent advances in magnetic fusion research. New configurations have been found for a stellarator experiment by computational methods. Solutions of a free-boundary problem are applied to study the performance of the plasma and look for islands in the magnetic surfaces. Mathematical analysis and numerical calculations have been used to study equilibrium, stability, and transport of optimized fusion reactors.

  1. Computational mathematics and physics of fusion reactors

    Science.gov (United States)

    Garabedian, Paul R.

    2003-01-01

    Theory has contributed significantly to recent advances in magnetic fusion research. New configurations have been found for a stellarator experiment by computational methods. Solutions of a free-boundary problem are applied to study the performance of the plasma and look for islands in the magnetic surfaces. Mathematical analysis and numerical calculations have been used to study equilibrium, stability, and transport of optimized fusion reactors. PMID:14614129

  2. Reactor and method for production of nanostructures

    Energy Technology Data Exchange (ETDEWEB)

    Sunkara, Mahendra Kumar; Kim, Jeong H.; Kumar, Vivekanand

    2017-04-25

    A reactor and method for production of nanostructures, including metal oxide nanowires or nanoparticles, are provided. The reactor includes a regulated metal powder delivery system in communication with a dielectric tube; a plasma-forming gas inlet, whereby a plasma-forming gas is delivered substantially longitudinally into the dielectric tube; a sheath gas inlet, whereby a sheath gas is delivered into the dielectric tube; and a microwave energy generator coupled to the dielectric tube, whereby microwave energy is delivered into a plasma-forming gas. The method for producing nanostructures includes providing a reactor to form nanostructures and collecting the formed nanostructures, optionally from a filter located downstream of the dielectric tube.

  3. VISWAM. A computer code package for thermal reactor physics computations

    Energy Technology Data Exchange (ETDEWEB)

    Jagannathan, V.; Thiyagarajan, T.K.; Ganesan, S.; Jain, R.P.; Pal, U. [Bhabha Atomic Research Centre, Mumbai (India); Karthikeyan, R. [Ecole Polytechnique de Montreal, Montreal, Quebec (Canada)

    2004-07-01

    The nuclear cross section data and reactor physics design methods developed over the past three decades have attained a high degree of reliability for thermal power reactor design and analysis. This is borne out from the analysis of physics commissioning experiments and several reactor-years of operational experience of two types of Indian thermal power reactors, viz. BWRs and PHWRs. Our computational tools were also developed and tested against a large number of IAEA CRP benchmarks on in-core fuel management code package validation for the modern BWR, PWR, VVER and PHWR. Though the computational algorithms are well tested, their mode of use has remained rather obsolete since the codes were developed when the modern high-speed large memory computers were not available. The use of Fortran language limits their potential use for varied applications. We are developing specific Visual Interface Software as the Work Aid support for effective Man-Machine interface (VISWAM). The VISWAM package when fully developed and tested will enable handling the input description of complex fuel assembly and the reactor core geometry with immaculate ease. Selective display of the three dimensional distribution of multi-group fluxes, power distribution and hot spots will provide a good insight into the analysis and also enable inter comparison of different nuclear datasets and methods. Since the new package will be user-friendly, training of requisite human resource for the expanding Indian nuclear power programme will be rendered easier and the gap between an expert and a novice will be greatly reduced. (author)

  4. New applications of neutron noise theory in power reactor physics

    Energy Technology Data Exchange (ETDEWEB)

    Arzhanov, Vasiliy

    2000-04-01

    The present thesis deals with neutron noise theory as applied to three comparatively different topics (or problems) in power reactor physics. Namely they are: theoretical investigation of the possibility to use a newly proposed current-flux (C/F) detector in Pressurized Water Reactors (PWRs) for the localisation of anomalies; both definition and studies on the point kinetic and adiabatic approximations for the relatively recently proposed Accelerator Driven Systems (ADS); development of the general theory of linear reactor kinetics and neutron noise in systems with varying size. One important practical problem is to detect and localise a vibrating control rod pin. The significance comes from the operational experience which indicates that individual pins can execute excessive mechanical vibrations that may lead to damage. Such mechanical vibrations induce neutron noise that can be detected. While the detection is relatively easy, the localisation of a vibrating control rod is much more complicated because only one measuring position is available and one needs to have at least three measured quantities. Therefore it has currently been proposed that the fluctuations of the neutron current vector, called the current noise, can be used in addition to the scalar noise in reactor diagnostic problems. The thesis investigates the possibility of the localization of a vibrating control rod pin in a PWR control assembly by using the scalar neutron noise and the 2-D radial current noise as measured at one central point in the control assembly. An explicit localisation technique is elaborated in which the searched position is determined as the absolute minimum of a minimisation function. The technique is investigated in numerical simulations. The results of the simulation tests show the potential applicability of the method. By design accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and

  5. METHOD OF OPERATING A NEUTRONIC REACTOR

    Science.gov (United States)

    Turkevich, A.

    1963-01-22

    This patent relates to one step in a method of operating a neutronic reactor consisting of a slurry of fissionable material in heavy water. Deuterium gas is passed through the slurry to sweep gaseous fission products therefrom and the deuterium is then separated from the gaseous fission products. (AEC)

  6. Reactor core stability monitoring method for BWR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kanemoto, Shigeru; Ebata, Shigeo.

    1992-09-01

    In an operation for a BWR type reactor, reactor power is usually increased or decreased by controlling both of control rods and reactor core flow rate. Under a certain condition, the reactor core is made unstable by the coupling of nuclear and thermohydrodynamic characteristics in the reactor. Therefore, the reactor power and the reactor core flow rate are changed within a range predetermined by a design calculation. However, if reactor core stability can be always measured and monitored, it is useful for safe operation, as well as an existent operation range can be extended to enable more effective operation. That is, autoregressive a coefficient is determined successively on real time based on fluctuation components of neutron flux signals. Based on the result, an amplification ratio, as a typical measure of the reactor core stability, is determined on a real time. A time constant of the successive calculation for the autoregressive coefficient can be made variable by the amplification ratio. Then, the amplification ratio is estimated at a constant accuracy. With such procedures, the reactor core stability can be monitored successively in an ON-line manner at a high accuracy, thereby enabling to improve the operation performance. (I.S.).

  7. Research on the reactor physics and reactor safety of VVER reactors. AER Symposium 2016

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, S.

    2017-09-15

    The selected paperscan be attributed to the following main subjects: Reactor start-up tests and use of corresponding data for code validation, code development and application, approaches for safety analyses, closure of nuclear fuel cycle, prospective reactor concepts.

  8. The research reactors their contribution to the reactors physics; Les reacteurs de recherche leur apport sur la physique des reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Barral, J.C. [Electricite de France (EDF), 75 - Paris (France); Zaetta, A. [CEA/Cadarache, Direction des Reacteurs Nucleaires, DRN, 13 - Saint-Paul-lez-Durance (France); Johner, J. [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee (DRFC), 13 - Saint Paul lez Durance (France); Mathoniere, G. [CEA/Saclay, DEN, 91 - Gif sur Yvette (France)] [and others

    2000-07-01

    The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

  9. Reactor Section standard analytical methods. Part 1

    Energy Technology Data Exchange (ETDEWEB)

    Sowden, D.

    1954-07-01

    the Standard Analytical Methods manual was prepared for the purpose of consolidating and standardizing all current analytical methods and procedures used in the Reactor Section for routine chemical analyses. All procedures are established in accordance with accepted practice and the general analytical methods specified by the Engineering Department. These procedures are specifically adapted to the requirements of the water treatment process and related operations. The methods included in this manual are organized alphabetically within the following five sections which correspond to the various phases of the analytical control program in which these analyses are to be used: water analyses, essential material analyses, cotton plug analyses boiler water analyses, and miscellaneous control analyses.

  10. Nuclear reactor flow control method and apparatus

    Science.gov (United States)

    Church, J.P.

    1993-03-30

    Method and apparatus for improving coolant flow in a nuclear reactor during accident as well as nominal conditions. The reactor has a plurality of fuel elements in sleeves and a plenum above the fuel and through which the sleeves penetrate. Holes are provided in the sleeve so that coolant from the plenum can enter the sleeve and cool the fuel. The number and size of the holes are varied from sleeve to sleeve with the number and size of holes being greater for sleeves toward the center of the core and less for sleeves toward the periphery of the core. Preferably the holes are all the same diameter and arranged in rows and columns, the rows starting from the bottom of every sleeve and fewer rows in peripheral sleeves and more rows in the central sleeves.

  11. Method of Operating a Neutronic Reactor

    Science.gov (United States)

    Fermi, Enrico; Szilard, Leo

    moderators such as light water or beryllium. In particular, the ratio is given of the absorption cross section to the scattering cross section for several moderators. Procedures for the purification of uranium are described as well. Several methods (i.e., the exponential pile or the "shotgun" method; see Patent No. 2.969,307) are reported for testing the purity against neutron absorption of different materials. The effect of the boron and vanadium impurities in the graphite and light water in the heavy water are considered. Different cooling systems for the reactors are considered and compared in the Patent, based on the circulation of a gas (typically, air) or a liquid (light or heavy water, diphenyl, etc.). The principles and practice for the construction, functioning and control of several kinds of reactors are reported in detail. One reactor considered in the present Patent is a low power uranium-graphite one without cooling system, where the active part consists in (small) cylinders of metallic uranium or pseudo-spheres of uranium oxide (or cylinders of U3O8). The control rods are made of steel with boron inserts, while limitation and safety rods are made of cadmium. In addition, an uranium-graphite pile cooled by air or even by water or diphenyl is considered. It is pointed out that dyphenil should usually be preferred with respect to water, due to its lower absorption of neutrons and to its higher boiling temperature, but the disadvantage related to its use is mainly due to the closed pumping system required and to the possible occurrence of polymerization which makes the fluid viscous. Another kind of reactor described in detail is made of uranium (vertical) bars immersed in heavy water. When, during the operation, heavy water is dissociated into D2 and O2, these two gaseous elements are carried by an inert gas (helium) into a recombination device. The control and safety rods are made of cadmium. Hybrid reactors composed of different lattices in the same neutronic

  12. New Reactor Physics Benchmark Data in the March 2012 Edition of the IRPhEP Handbook

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; J. Blair Briggs; Jim Gulliford

    2012-11-01

    The International Reactor Physics Experiment Evaluation Project (IRPhEP) was established to preserve integral reactor physics experimental data, including separate or special effects data for nuclear energy and technology applications. Numerous experiments that have been performed worldwide, represent a large investment of infrastructure, expertise, and cost, and are valuable resources of data for present and future research. These valuable assets provide the basis for recording, development, and validation of methods. If the experimental data are lost, the high cost to repeat many of these measurements may be prohibitive. The purpose of the IRPhEP is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. Contributors from around the world collaborate in the evaluation and review of selected benchmark experiments for inclusion in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) [1]. Several new evaluations have been prepared for inclusion in the March 2012 edition of the IRPhEP Handbook.

  13. Nuclear data and reactor physics activities in Indonesia

    Energy Technology Data Exchange (ETDEWEB)

    Liem, P.H. [National Atomic Energy Agency, Tangerang (Indonesia). Center for Multipurpose Reactor

    1998-03-01

    The nuclear data and reactor physics activities in Indonesia, especially, in the National Atomic Energy Agency are presented. In the nuclear data field, the Agency is now taking the position of a user of the main nuclear data libraries such as JENDL and ENDF/B. These nuclear data libraries become the main sources for producing problem dependent cross section sets that are needed by cell calculation codes or transport codes for design, analysis and safety evaluation of research reactors. In the reactor physics field, besides utilising the existing core analysis codes obtained from bilateral and international co-operation, the Agency is putting much effort to self-develop Batan`s codes for reactor physics calculations, in particular, for research reactor and high temperature reactor design, analysis and fuel management. Under the collaboration with JAERI, Monte Carlo criticality calculations on the first criticality of RSG GAS (MPR-30) first core were done using JAERI continuous energy, vectorized Monte Carlo code, MVP, with JENDL-3.1 and JENDL-3.2 nuclear data libraries. The results were then compared with the experiment data collected during the commissioning phase. Monte Carlo calculations with both JENDL-3.1 and -3.2 libraries produced k{sub eff} values with excellent agreement with experiment data, however, systematically, JENDL-3.2 library showed slightly higher k{sub eff} values than JENDL-3.1 library. (author)

  14. Educational reactor-physics experiments with the critical assemble TCA

    Energy Technology Data Exchange (ETDEWEB)

    Tsutsui, Hiroaki; Okubo, Masaaki; Igashira, Masayuki [Tokyo Inst. of Tech. (Japan); Horiki, Oichiro; Suzaki, Takenori

    1997-10-01

    The Tank-Type Critical Assembly (TCA) of Japan Atomic Energy Research Institute is research equipment for light water reactor physics. In the present report, the lectures given to the graduate students of Tokyo Institute of Technology who participated in the educational experiment course held on 26-30 August at TCA are rearranged to provide useful information for those who will implement educational basic experiments with TCA in the future. This report describes the principles, procedures, and data analyses for (1) Critical approach and Exponential experiment, (2) Measurement of neutron flux distribution, (3) Measurement of power distribution, (4) Measurement of fuel rod worth distribution, and (5) Measurement of safety plate worth by the rod drop method. (author)

  15. Modelization of physical phenomena in research reactors with the help of new developments in transport methods, and methodology validation with experimental data; Modelisation des phenomenes physiques dans les reacteurs de recherche a l'aide de developpements realises dans les methodes de transport et qualification

    Energy Technology Data Exchange (ETDEWEB)

    Rauck, St

    2000-10-01

    The aim of this work is to develop a scheme for experimental reactors, based on transport equations. This type of reactors is characterized by a small core, a complex, very heterogeneous geometry and a large leakage. The possible insertion of neutron beams in the reflector and the presence of absorbers in the core increase the difficulty of the 3D-geometrical description and the physical modeling of the component parameters of the reactor. The Orphee reactor has been chosen for our study. Physical models (homogenization, collapsing cross section in few groups, albedo multigroup condition) have been developed in the APOLLO2 and CRONOS2 codes to calculate flux and power maps in a 3D-geometry, with different burnup and through transport equations. Comparisons with experimental measurements have shown the interest of taking into account anisotropy, steep flux gradients by using Sn methods, and on the other hand using a 12-group cross section library. The modeling of neutron beams has been done outside the core modeling through Monte Carlo calculations and with the total geometry, including a large thickness of heavy water. Thanks to this calculations, one can evaluate the neutron beams anti-reactivity and determinate the core cycle. We assure these methods more accurate than usual transport-diffusion calculations will be used for the conception of new research reactors. (author)

  16. Neutrino Physics with Accelerator Driven Subcritical Reactors

    Science.gov (United States)

    Ciuffoli, Emilio

    2017-09-01

    Accelerator Driven Subcritical System (ADS) reactors are being developed around the world, to produce energy and, at the same time, to provide an efficient way to dispose of and to recycle nuclear waste. Used nuclear fuel, by itself, cannot sustain a chain reaction; however in ADS reactors the additional neutrons which are required will be supplied by a high-intensity accelerator. This accelerator will produce, as a by-product, a large quantity of {\\bar{ν }}μ via muon Decay At Rest (µDAR). Using liquid scintillators, it will be possible to to measure the CP-violating phase δCP and to look for experimental signs of the presence of sterile neutrinos in the appearance channel, testing the LSND and MiniBooNE anomalies. Even in the first stage of the project, when the beam energy will be lower, it will be possible to produce {\\bar{ν }}e via Isotope Decay At Rest (IsoDAR), which can be used to provide competitive bounds on sterile neutrinos in the disappearance channel. I will consider several experimental setups in which the antineutrinos are created using accelerators that will be constructed as part of the China-ADS program.

  17. Molten Salt Breeder Reactor Analysis Methods

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jinsu; Jeong, Yongjin; Lee, Deokjung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2015-05-15

    Utilizing the uranium-thorium fuel cycle shows considerable potential for the possibility of MSR. The concept of MSBR should be revised because of molten salt reactor's advantage such as outstanding neutron economy, possibility of continuous online reprocessing and refueling, a high level of inherent safety, and economic benefit by keeping off the fuel fabrication process. For the development of MSR research, this paper provides the MSBR single-cell, two-cell and whole core model for computer code input, and several calculation results including depletion calculation of each models. The calculations are carried out by using MCNP6, a Monte Carlo computer code, which has CINDER90 for depletion calculation using ENDF-VII nuclear data. From the calculation results of various reactor design parameters, the temperature coefficients are all negative at the initial state and MTC becomes positive at the equilibrium state. From the results of core rod worth, the graphite control rod alone cannot makes the core subcritical at initial state. But the equilibrium state, the core can be made subcritical state only by graphite control rods. Through the comparison of the results of each models, the two-cell method can represent the MSBR core model more accurately with a little more computational resources than the single-cell method. Many of the thermal spectrum MSR have adopted a multi-region single-fluid strategy.

  18. Methods of experimental physics

    CERN Document Server

    Williams, Dudley

    1962-01-01

    Methods of Experimental Physics, Volume 3: Molecular Physics focuses on molecular theory, spectroscopy, resonance, molecular beams, and electric and thermodynamic properties. The manuscript first considers the origins of molecular theory, molecular physics, and molecular spectroscopy, as well as microwave spectroscopy, electronic spectra, and Raman effect. The text then ponders on diffraction methods of molecular structure determination and resonance studies. Topics include techniques of electron, neutron, and x-ray diffraction and nuclear magnetic, nuclear quadropole, and electron spin reson

  19. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    Energy Technology Data Exchange (ETDEWEB)

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-10-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

  20. Consistent neutron-physical and thermal-physical calculations of fuel rods of VVER type reactors

    Directory of Open Access Journals (Sweden)

    Tikhomirov Georgy

    2017-01-01

    Full Text Available For modeling the isotopic composition of fuel, and maximum temperatures at different moments of time, one can use different algorithms and codes. In connection with the development of new types of fuel assemblies and progress in computer technology, the task makes important to increase accuracy in modeling of the above characteristics of fuel assemblies during the operation. Calculations of neutronphysical characteristics of fuel rods are mainly based on models using averaged temperature, thermal conductivity factors, and heat power density. In this paper, complex approach is presented, based on modern algorithms, methods and codes to solve separate tasks of thermal conductivity, neutron transport, and nuclide transformation kinetics. It allows to perform neutron-physical and thermal-physical calculation of the reactor with detailed temperature distribution, with account of temperature-depending thermal conductivity and other characteristics. It was applied to studies of fuel cell of the VVER-1000 reactor. When developing new algorithms and programs, which should improve the accuracy of modeling the isotopic composition and maximum temperature in the fuel rod, it is necessary to have a set of test tasks for verification. The proposed approach can be used for development of such verification base for testing calculation of fuel rods of VVER type reactors

  1. Fail-safe reactivity compensation method for a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nygaard, Erik T.; Angelo, Peter L.; Aase, Scott B.

    2018-01-23

    The present invention relates generally to the field of compensation methods for nuclear reactors and, in particular to a method for fail-safe reactivity compensation in solution-type nuclear reactors. In one embodiment, the fail-safe reactivity compensation method of the present invention augments other control methods for a nuclear reactor. In still another embodiment, the fail-safe reactivity compensation method of the present invention permits one to control a nuclear reaction in a nuclear reactor through a method that does not rely on moving components into or out of a reactor core, nor does the method of the present invention rely on the constant repositioning of control rods within a nuclear reactor in order to maintain a critical state.

  2. ADVANCED SEISMIC BASE ISOLATION METHODS FOR MODULAR REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    E. Blanford; E. Keldrauk; M. Laufer; M. Mieler; J. Wei; B. Stojadinovic; P.F. Peterson

    2010-09-20

    Advanced technologies for structural design and construction have the potential for major impact not only on nuclear power plant construction time and cost, but also on the design process and on the safety, security and reliability of next generation of nuclear power plants. In future Generation IV (Gen IV) reactors, structural and seismic design should be much more closely integrated with the design of nuclear and industrial safety systems, physical security systems, and international safeguards systems. Overall reliability will be increased, through the use of replaceable and modular equipment, and through design to facilitate on-line monitoring, in-service inspection, maintenance, replacement, and decommissioning. Economics will also receive high design priority, through integrated engineering efforts to optimize building arrangements to minimize building heights and footprints. Finally, the licensing approach will be transformed by becoming increasingly performance based and technology neutral, using best-estimate simulation methods with uncertainty and margin quantification. In this context, two structural engineering technologies, seismic base isolation and modular steel-plate/concrete composite structural walls, are investigated. These technologies have major potential to (1) enable standardized reactor designs to be deployed across a wider range of sites, (2) reduce the impact of uncertainties related to site-specific seismic conditions, and (3) alleviate reactor equipment qualification requirements. For Gen IV reactors the potential for deliberate crashes of large aircraft must also be considered in design. This report concludes that base-isolated structures should be decoupled from the reactor external event exclusion system. As an example, a scoping analysis is performed for a rectangular, decoupled external event shell designed as a grillage. This report also reviews modular construction technology, particularly steel-plate/concrete construction using

  3. Micro-Reactor Physics of MOX-Fueled Core

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, T.

    2001-06-17

    Recently, fuel assemblies of light water reactors have become complicated because of the extension of fuel burnup and the use of high-enriched Gd and mixed-oxide (MOX) fuel, etc. In conventional assembly calculations, the detailed flux distribution, spectrum distribution, and space dependence of self-shielding within a fuel pellet are not directly taken into account. The experimental and theoretical study of investigating these microscopic properties is named micro-reactor physics. The purpose of this work is to show the importance of micro-reactor physics in the analysis of MOX fuel assemblies. Several authors have done related studies; however, their studies are limited to fuel pin cells, and they are never mentioned with regard to burnup effect, which is important for actual core design.

  4. Multiscale Methods for Nuclear Reactor Analysis

    Science.gov (United States)

    Collins, Benjamin S.

    The ability to accurately predict local pin powers in nuclear reactors is necessary to understand the mechanisms that cause fuel pin failure during steady state and transient operation. In the research presented here, methods are developed to improve the local solution using high order methods with boundary conditions from a low order global solution. Several different core configurations were tested to determine the improvement in the local pin powers compared to the standard techniques, that use diffusion theory and pin power reconstruction (PPR). Two different multiscale methods were developed and analyzed; the post-refinement multiscale method and the embedded multiscale method. The post-refinement multiscale methods use the global solution to determine boundary conditions for the local solution. The local solution is solved using either a fixed boundary source or an albedo boundary condition; this solution is "post-refinement" and thus has no impact on the global solution. The embedded multiscale method allows the local solver to change the global solution to provide an improved global and local solution. The post-refinement multiscale method is assessed using three core designs. When the local solution has more energy groups, the fixed source method has some difficulties near the interface: however the albedo method works well for all cases. In order to remedy the issue with boundary condition errors for the fixed source method, a buffer region is used to act as a filter, which decreases the sensitivity of the solution to the boundary condition. Both the albedo and fixed source methods benefit from the use of a buffer region. Unlike the post-refinement method, the embedded multiscale method alters the global solution. The ability to change the global solution allows for refinement in areas where the errors in the few group nodal diffusion are typically large. The embedded method is shown to improve the global solution when it is applied to a MOX/LEU assembly

  5. An Analytical Method For The Solution Of Reactor Dynamic Equations

    African Journals Online (AJOL)

    In this paper, an analytical method for the solution of nuclear reactor dynamic equations is presented. The method is applied to a linearised high-order deterministic model of a pressurised water reactor plant driven by step-reactivity insertion. A comparison of this method with two other techniques (the matrix exponential and ...

  6. Developments in Sensitivity Methodologies and the Validation of Reactor Physics Calculations

    Directory of Open Access Journals (Sweden)

    Giuseppe Palmiotti

    2012-01-01

    Full Text Available The sensitivity methodologies have been a remarkable story when adopted in the reactor physics field. Sensitivity coefficients can be used for different objectives like uncertainty estimates, design optimization, determination of target accuracy requirements, adjustment of input parameters, and evaluations of the representativity of an experiment with respect to a reference design configuration. A review of the methods used is provided, and several examples illustrate the success of the methodology in reactor physics. A new application as the improvement of nuclear basic parameters using integral experiments is also described.

  7. Method for filling a reactor with a catalyst

    DEFF Research Database (Denmark)

    2013-01-01

    The invention relates to a method for filling a reactor with a catalyst for the carbonylation of carbonylated compounds in the gas phase. According to said method, a SILP catalyst is covered with a filling agent which is liquid under normal conditions and is volatile under carbonylation reaction...... conditions, and a thus-treated catalyst is introduced into the reactor and the reactor is sealed....

  8. Validation of High-Fidelity Reactor Physics Models for Support of the KJRR Experimental Campaign in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nigg, David W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nielsen, Joseph W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Norman, Daren R. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-07-01

    The Korea Atomic Energy Research Institute is currently in the process of qualifying a Low-Enriched Uranium fuel element design for the new Ki-Jang Research Reactor (KJRR). As part of this effort, a prototype KJRR fuel element was irradiated for several operating cycles in the Northeast Flux Trap of the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The KJRR fuel element contained a very large quantity of fissile material (618g 235U) in comparison with historical ATR experiment standards (<1g 235U), and its presence in the ATR flux trap was expected to create a neutronic configuration that would be well outside of the approved validation envelope for the reactor physics analysis methods used to support ATR operations. Accordingly it was necessary, prior to high-power irradiation of the KJRR fuel element in the ATR, to conduct an extensive set of new low-power physics measurements with the KJRR fuel element installed in the ATR Critical Facility (ATRC), a companion facility to the ATR that is located in an immediately adjacent building, sharing the same fuel handling and storage canal. The new measurements had the objective of expanding the validation envelope for the computational reactor physics tools used to support ATR operations and safety analysis to include the planned KJRR irradiation in the ATR and similar experiments that are anticipated in the future. The computational and experimental results demonstrated that the neutronic behavior of the KJRR fuel element in the ATRC is well-understood, both in terms of its general effects on core excess reactivity and fission power distributions, its effects on the calibration of the core lobe power measurement system, as well as in terms of its own internal fission rate distribution and total fission power per unit ATRC core power. Taken as a whole, these results have significantly extended the ATR physics validation envelope, thereby enabling an entire new class of irradiation experiments.

  9. Physics of reactor safety. Quarterly report, April--June 1977. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-09-01

    The work in the Applied Physics Division includes reports on reactor safety program by members of the Reactor Safety Appraisals Group, Monte Carlo analysis of safety-related critical assembly experiments by members of the Theoretical Fast Reactor Physics Group, and planning of safety-related critical experiments by members of the Zero Power Reactor (ZPR) Planning and Experiments Group. Work on Reactor core thermal-hydraulic code development performed in the Components Technology Division is also included in the report.

  10. Development status of reactor physics codes in cosine project

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Yixue; Yu, Hui; Liu, Zhanquan; Zhang, Bin; Qiu, Chunhua; Wang, Changhui; Wang, Su; Hu, Xiaoyu; Xu, Hua; Li, Shuo; Liu, Zhiyan; Shen, Feng; Yang, Yanhua [State Nuclear Power Software Development Center, Beijing (China)

    2012-03-15

    COre and System Integrated Engine for design and analysis (COSINE) is the newly approved NPP software R and D key project in China. This paper gives an overview of the COSINE code package, especially the reactor physics subsystem. The code architecture, the theoretical physics model and the current progress are introduced. A brief summary behind the code package is outlined. The detailed discussion here is limited to the lattice physics code (LATC), core analysis code (CORE) and neutron kinetics code (KIND) which are currently under development.

  11. Cascading pressure reactor and method for solar-thermochemical reactions

    Energy Technology Data Exchange (ETDEWEB)

    Ermanoski, Ivan

    2017-11-14

    Reactors and methods for solar thermochemical reactions are disclosed. The reactors and methods include a cascade of reduction chambers at successively lower pressures that leads to over an order of magnitude pressure decrease compared to a single-chambered design. The resulting efficiency gains are substantial, and represent an important step toward practical and efficient solar fuel production on a large scale.

  12. Yale High Energy Physics Research: Precision Studies of Reactor Antineutrinos

    Energy Technology Data Exchange (ETDEWEB)

    Heeger, Karsten M. [Yale Univ., New Haven, CT (United States)

    2014-09-13

    This report presents experimental research at the intensity frontier of particle physics with particular focus on the study of reactor antineutrinos and the precision measurement of neutrino oscillations. The experimental neutrino physics group of Professor Heeger and Senior Scientist Band at Yale University has had leading responsibilities in the construction and operation of the Daya Bay Reactor Antineutrino Experiment and made critical contributions to the discovery of non-zero$\\theta_{13}$. Heeger and Band led the Daya Bay detector management team and are now overseeing the operations of the antineutrino detectors. Postdoctoral researchers and students in this group have made leading contributions to the Daya Bay analysis including the prediction of the reactor antineutrino flux and spectrum, the analysis of the oscillation signal, and the precision determination of the target mass yielding unprecedented precision in the relative detector uncertainty. Heeger's group is now leading an R\\&D effort towards a short-baseline oscillation experiment, called PROSPECT, at a US research reactor and the development of antineutrino detectors with advanced background discrimination.

  13. Assessing reactor physics codes capabilities to simulate fast reactors on the example of the BN-600 benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, Vladimir [Scientific and Engineering Centre for Nuclear and Radiation Safety (SES NRS), Moscow (Russian Federation); Bousquet, Jeremy [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

    2016-11-15

    This work aims to assess the capabilities of reactor physics codes (initially validated for thermal reactors) to simulate fast sodium cooled reactors. The BFS-62-3A critical experiment from the BN-600 Hybrid Core Benchmark Analyses was chosen for the investigation. Monte-Carlo codes (KENO from SCALE and SERPENT 2.1.23) and the deterministic diffusion code DYN3D-MG are applied to calculate the neutronic parameters. It was found that the multiplication factor and reactivity effects calculated by KENO and SERPENT using the ENDF/B-VII.0 continuous energy library are in a good agreement with each other and with the measured benchmark values. Few-groups macroscopic cross sections, required for DYN3D-MG, were prepared in applying different methods implemented in SCALE and SERPENT. The DYN3D-MG results of a simplified benchmark show reasonable agreement with results from Monte-Carlo calculations and measured values. The former results are used to justify DYN3D-MG implementation for sodium cooled fast reactors coupled deterministic analysis.

  14. Transport, Convective Equilibrium, and Reactor Physics in Stellarator Type Devices.

    Science.gov (United States)

    Ho, Darwin Dao-Man

    In Part 1 of this thesis, the neoclassical transport in stellarator reactors is studied in detail. It is found that the electron energy confinement time is in general comparable to that of the ions regardless of the size of the machine. Although the neoclassical losses are large, numerical examples show that ignition can be achieved in a reasonably sized machine. The kinetic calculation for the ion transport with the effect of collisionless detrapping/entrapping has not been carried out. This would be a good subject for later investigation. The energy transfer from thermonuclear (alpha) -particles to the background plasma is calculated in Part 2. It is found that (alpha)-particles can transfer most of their energy into the background plasma before collisionally scatter into the trapping region and are lost. In Part 3, the convective equilibrium hypothesis is proposed for high (beta) reactors which have regions where the plasma (beta) exceeds the critical (beta). Although the convective transport cannot be calculated precisely, it is shown that the density and temperature profiles in the convective region can still be estimated. A simple mixing-length theory shows that the convective transport is highly efficient. A detailed study of the nonlinear behavior of convective cells is currently being investigated. A novel power cycle for direct conversion of (alpha) -particle energy into electricity is proposed for an ignited plasma in a stellarator reactor in Part 4. In analyzing the physics of the cycle, there appears to be no major physical or engineering obstacle that would make the cycle impractical. This power cycle may provide an alternative scheme for extracting energy from D-T fueled reactors and may become an important scheme for energy conversion for advanced neutron-lean fueled reactors. By operating two or more reactors in tandem, the cycle can be made self -sustaining. The dynamics of a coupled reactor reactor system will be the subject of a later study and

  15. Production test-080, physics testing at D reactor deactivation

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, G.F.

    1967-06-15

    The purpose of this test is to provide a set of experimental data to test a compute code frequently used in nuclear safety analyses and to explore certain experimental techniques which may prove extremely valuable in the future. In addition, some basic physics parameters which will be measured may be used in an assessment of the feasibility of using a deactivated Hanford reactor for space-dependent transient tests.

  16. Summary of ORSphere Critical and Reactor Physics Measurements

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, Margaret A.; Bess, John D.

    2016-09-01

    In the early 1970s Dr. John T. Mihalczo (team leader), J. J. Lynn, and J. R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. This critical configuration has been evaluated. Preliminary results were presented at ND2013. Since then, the evaluation was finalized and judged to be an acceptable benchmark experiment for the International Criticality Safety Benchmark Experiment Project (ICSBEP). Additionally, reactor physics measurements were performed to determine surface button worths, central void worth, delayed neutron fraction, prompt neutron decay constant, fission density and neutron importance. These measurements have been evaluated and found to be acceptable experiments and are discussed in full detail in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. The purpose of this paper is summary summarize all the critical and reactor physics measurements evaluations and, when possible, to compare them to GODIVA experiment results.

  17. METHOD AND APPARATUS FOR CONTROL OF A NUCLEAR REACTOR

    Science.gov (United States)

    Cawley, W.E.

    1962-12-11

    A method and apparatus are described for controlling an overmoderated nuclear reactor containing columns of fuel elements aligned in a plurality of coolant tubes in a stream of coolant water. The invention includes means for adjusting the distance between halves of the fuel element column to vary the relative proportion of fuel and moderator at the center of the reactor. (AEC)

  18. 78 FR 31821 - Physical Protection of Shipments of Irradiated Reactor Fuel

    Science.gov (United States)

    2013-05-28

    ... COMMISSION 10 CFR Part 73 RIN 3150-AI64 Physical Protection of Shipments of Irradiated Reactor Fuel AGENCY... (NRC) is issuing Revision 2 of NUREG-0561, ``Physical Protection of Shipments of Irradiated Reactor... regulations for the transport of irradiated reactor fuel at Sec. 73.37 of Title 10 of the Code of Federal...

  19. Overview of the 2014 Edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook)

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; J. Blair Briggs; Jim Gulliford; Ian Hill

    2014-10-01

    The International Reactor Physics Experiment Evaluation Project (IRPhEP) is a widely recognized world class program. The work of the IRPhEP is documented in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Integral data from the IRPhEP Handbook is used by reactor safety and design, nuclear data, criticality safety, and analytical methods development specialists, worldwide, to perform necessary validations of their calculational techniques. The IRPhEP Handbook is among the most frequently quoted reference in the nuclear industry and is expected to be a valuable resource for future decades.

  20. Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    1992-04-01

    This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

  1. Method for automatically scramming a nuclear reactor

    Science.gov (United States)

    Ougouag, Abderrafi M.; Schultz, Richard R.; Terry, William K.

    2005-12-27

    An automatically scramming nuclear reactor system. One embodiment comprises a core having a coolant inlet end and a coolant outlet end. A cooling system operatively associated with the core provides coolant to the coolant inlet end and removes heated coolant from the coolant outlet end, thus maintaining a pressure differential therebetween during a normal operating condition of the nuclear reactor system. A guide tube is positioned within the core with a first end of the guide tube in fluid communication with the coolant inlet end of the core, and a second end of the guide tube in fluid communication with the coolant outlet end of the core. A control element is positioned within the guide tube and is movable therein between upper and lower positions, and automatically falls under the action of gravity to the lower position when the pressure differential drops below a safe pressure differential.

  2. Alpha Particle Physics Experiments in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Budny, R.V.; Darrow, D.S.; Medley, S.S.; Nazikian, R.; Zweben, S.J.; et al.

    1998-12-14

    Alpha particle physics experiments were done on the Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium (DT) run from 1993-1997. These experiments utilized several new alpha particle diagnostics and hundreds of DT discharges to characterize the alpha particle confinement and wave-particle interactions. In general, the results from the alpha particle diagnostics agreed with the classical single-particle confinement model in magnetohydrodynamic (MHD) quiescent discharges. Also, the observed alpha particle interactions with sawteeth, toroidal Alfvén eigenmodes (TAE), and ion cyclotron resonant frequency (ICRF) waves were roughly consistent with theoretical modeling. This paper reviews what was learned and identifies what remains to be understood.

  3. Advanced Mesh-Enabled Monte carlo capability for Multi-Physics Reactor Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, Paul; Evans, Thomas; Tautges, Tim

    2012-12-24

    This project will accumulate high-precision fluxes throughout reactor geometry on a non- orthogonal grid of cells to support multi-physics coupling, in order to more accurately calculate parameters such as reactivity coefficients and to generate multi-group cross sections. This work will be based upon recent developments to incorporate advanced geometry and mesh capability in a modular Monte Carlo toolkit with computational science technology that is in use in related reactor simulation software development. Coupling this capability with production-scale Monte Carlo radiation transport codes can provide advanced and extensible test-beds for these developments. Continuous energy Monte Carlo methods are generally considered to be the most accurate computational tool for simulating radiation transport in complex geometries, particularly neutron transport in reactors. Nevertheless, there are several limitations for their use in reactor analysis. Most significantly, there is a trade-off between the fidelity of results in phase space, statistical accuracy, and the amount of computer time required for simulation. Consequently, to achieve an acceptable level of statistical convergence in high-fidelity results required for modern coupled multi-physics analysis, the required computer time makes Monte Carlo methods prohibitive for design iterations and detailed whole-core analysis. More subtly, the statistical uncertainty is typically not uniform throughout the domain, and the simulation quality is limited by the regions with the largest statistical uncertainty. In addition, the formulation of neutron scattering laws in continuous energy Monte Carlo methods makes it difficult to calculate adjoint neutron fluxes required to properly determine important reactivity parameters. Finally, most Monte Carlo codes available for reactor analysis have relied on orthogonal hexahedral grids for tallies that do not conform to the geometric boundaries and are thus generally not well

  4. Nodal methods in numerical reactor calculations

    Energy Technology Data Exchange (ETDEWEB)

    Hennart, J.P. [UNAM, IIMAS, A.P. 20-726, 01000 Mexico D.F. (Mexico)]. e-mail: jean_hennart@hotmail.com; Valle, E. del [National Polytechnic Institute, School of Physics and Mathematics, Department of Nuclear Engineering, Mexico, D.F. (Mexico)

    2004-07-01

    The present work describes the antecedents, developments and applications started in 1972 with Prof. Hennart who was invited to be part of the staff of the Nuclear Engineering Department at the School of Physics and Mathematics of the National Polytechnic Institute. Since that time and up to 1981, several master theses based on classical finite element methods were developed with applications in point kinetics and in the steady state as well as the time dependent multigroup diffusion equations. After this period the emphasis moved to nodal finite elements in 1, 2 and 3D cartesian geometries. All the thesis were devoted to the numerical solution of the neutron multigroup diffusion and transport equations, few of them including the time dependence, most of them related with steady state diffusion equations. The main contributions were as follows: high order nodal schemes for the primal and mixed forms of the diffusion equations, block-centered finite-differences methods, post-processing, composite nodal finite elements for hexagons, and weakly and strongly discontinuous schemes for the transport equation. Some of these are now being used by several researchers involved in nuclear fuel management. (Author)

  5. 78 FR 69139 - Physical Security-Design Certification and Operating Reactors

    Science.gov (United States)

    2013-11-18

    ... COMMISSION Physical Security--Design Certification and Operating Reactors AGENCY: Nuclear Regulatory... Operating Reactors.'' The public comment period was originally scheduled to close on October 30, 2013. The... this document. FOR FURTHER INFORMATION CONTACT: Wesley Held, Office of New Reactors, U.S. Nuclear...

  6. 78 FR 50313 - Physical Protection of Irradiated Reactor Fuel in Transit

    Science.gov (United States)

    2013-08-19

    ... 3150-AI64 Physical Protection of Irradiated Reactor Fuel in Transit AGENCY: Nuclear Regulatory... Transportation Orders to certain NRC power plant licensees, non-power reactor licensees, special nuclear material... Protection of Irradiated Reactor Fuel in Transit'' (RIN 3150-AI64; NRC-2009-0163). The final rule...

  7. Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle Fuel in a Pressurized Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Michael A. Pope; R. Sonat Sen; Abderrafi M. Ougouag; Gilles Youinou; Brian Boer

    2012-04-01

    Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU)-only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.

  8. Neutron spectrometric methods for core inventory verification in research reactors

    CERN Document Server

    Ellinger, A; Hansen, W; Knorr, J; Schneider, R

    2002-01-01

    In consequence of the Non-Proliferation Treaty safeguards, inspections are periodically made in nuclear facilities by the IAEA and the EURATOM Safeguards Directorate. The inspection methods are permanently improved. Therefore, the Core Inventory Verification method is being developed as an indirect method for the verification of the core inventory and to check the declared operation of research reactors.

  9. Multi-physics design and analyses of long life reactors for lunar outposts

    Science.gov (United States)

    Schriener, Timothy M.

    event of a launch abort accident. Increasing the amount of fuel in the reactor core, and hence its operational life, would be possible by launching the reactor unfueled and fueling it on the Moon. Such a reactor would, thus, not be subject to launch criticality safety requirements. However, loading the reactor with fuel on the Moon presents a challenge, requiring special designs of the core and the fuel elements, which lend themselves to fueling on the lunar surface. This research investigates examples of both a solid core reactor that would be fueled at launch as well as an advanced concept which could be fueled on the Moon. Increasing the operational life of a reactor fueled at launch is exercised for the NaK-78 cooled Sectored Compact Reactor (SCoRe). A multi-physics design and analyses methodology is developed which iteratively couples together detailed Monte Carlo neutronics simulations with 3-D Computational Fluid Dynamics (CFD) and thermal-hydraulics analyses. Using this methodology the operational life of this compact, fast spectrum reactor is increased by reconfiguring the core geometry to reduce neutron leakage and parasitic absorption, for the same amount of HEU in the core, and meeting launch safety requirements. The multi-physics analyses determine the impacts of the various design changes on the reactor's neutronics and thermal-hydraulics performance. The option of increasing the operational life of a reactor by loading it on the Moon is exercised for the Pellet Bed Reactor (PeBR). The PeBR uses spherical fuel pellets and is cooled by He-Xe gas, allowing the reactor core to be loaded with fuel pellets and charged with working fluid on the lunar surface. The performed neutronics analyses ensure the PeBR design achieves a long operational life, and develops safe launch canister designs to transport the spherical fuel pellets to the lunar surface. The research also investigates loading the PeBR core with fuel pellets on the Moon using a transient Discrete

  10. Exploring Stochastic Sampling in Nuclear Data Uncertainties Assessment for Reactor Physics Applications and Validation Studies

    Directory of Open Access Journals (Sweden)

    Alexander Vasiliev

    2016-12-01

    Full Text Available The quantification of uncertainties of various calculation results, caused by the uncertainties associated with the input nuclear data, is a common task in nuclear reactor physics applications. Modern computation resources and improved knowledge on nuclear data allow nowadays to significantly advance the capabilities for practical investigations. Stochastic sampling is the method which has received recently a high momentum for its use and exploration in the domain of reactor design and safety analysis. An application of a stochastic sampling based tool towards nuclear reactor dosimetry studies is considered in the given paper with certain exemplary test evaluations. The stochastic sampling not only allows the input nuclear data uncertainties propagation through the calculations, but also an associated correlation analysis performance with no additional computation costs and for any parameters of interest can be done. Thus, an example of assessment of the Pearson correlation coefficients for several models, used in practical validation studies, is shown here. As a next step, the analysis of the obtained information is proposed for discussion, with focus on the systems similarities assessment. The benefits of the employed method and tools with respect to practical reactor dosimetry studies are consequently outlined.

  11. Physical aspects of liquid-impelled loop reactors

    NARCIS (Netherlands)

    Sonsbeek, van H.

    1992-01-01

    The liquid-impelled loop reactor (LLR) is a reactor that consists of two parts : the main tube and the circulation tube. Both parts are in open connection at the bottom and at the top. The reactor is filled with a liquid phase: the continuous phase. Another liquid phase is injected in the

  12. [Physical methods and molecular biology].

    Science.gov (United States)

    Serdiuk, I N

    2009-01-01

    The review is devoted to the description of the current state of physical and chemical methods used for studying the structural and functional bases of living processes. Special attention is focused on the physical methods that have opened a new page in the research of the structure of biological macromolecules. They include primarily the methods of detecting and manipulating single molecules using optical and magnetic traps. New physical methods, such as two-dimensional infrared spectroscopy, fluorescence correlation spectroscopy and magnetic resonance microscopy are also analyzed briefly in the review. The path that physics and biology have passed for the latest 55 years shows that there is no single method providing all necessary information on macromolecules and their interactions. Each method provides its space-time view of the system. All physical methods are complementary. It is just complementarity that is the fundamental idea justifying the existence in practice of all physical methods, whose description is the aim of the review.

  13. Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Mueller, Donald E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division

    2016-08-31

    Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7LiF-BeF2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

  14. The mechanics in the reactors physics; La mecanique dans la physique des reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Barral, J.C. [Electricite de France (EDF), 75 - Paris (France); Zaetta, A. [CEA/Cadarache, Dept. d' Etudes des Reacteurs, DER/SPRC, 13 - Saint-Paul-lez-Durance (France); Johner, J. [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee, DRFC, 13 - Saint-Paul-lez-Durance (France)] [and others

    1998-12-22

    This meeting of the 24 november 1998, took place in Paris and was organized by the SFEN. After three plenary sessions a technical meeting dealt on the mechanics in reactors physics. The plenary papers presented the state of the art in the PWR type reactors and fast neutron reactors systems and in the thermonuclear reactors system. Five more technical papers presented the seismic behavior of the reactors cores, the fuel-cladding interactions, the defects harmfulness in the fracture mechanics and the fuel rods control system wear. (A.L.B.)

  15. Opportunities for applied measurements using the PROSPECT antineutrino detector: reactor physics and safeguards

    Science.gov (United States)

    Bowden, Nathaniel; Prospect Collaboration

    2015-10-01

    Disagreement of reactor antineutrino spectrum and flux measurements with updated predictions indicates that we have much to learn about the complicated processes underlying antineutrino production in reactors, as well as hinting at new physics. A number of new efforts seek to address these questions, including the PROSPECT experiment planned at the HFIR research reactor. In addition to greatly advancing our understanding of reactor antineutrino emissions, PROSPECT can support a rich applied physics program. The detection technology developed for PROSPECT will enable precision antineutrino spectrum measurements close to essentially any reactor type. Here we describe how such measurements provide opportunities to probe fissile isotope and fission daughter distributions, and their potential use for reactor physics and safeguards applications. LLNL-ABS-673983. Prepared by LLNL under Contract DE-AC52-07NA27344.

  16. Recent improvements of reactor physics codes in MHI

    Science.gov (United States)

    Kosaka, Shinya; Yamaji, Kazuya; Kirimura, Kazuki; Kamiyama, Yohei; Matsumoto, Hideki

    2015-12-01

    This paper introduces recent improvements for reactor physics codes in Mitsubishi Heavy Industries, Ltd(MHI). MHI has developed a new neutronics design code system Galaxy/Cosmo-S(GCS) for PWR core analysis. After TEPCO's Fukushima Daiichi accident, it is required to consider design extended condition which has not been covered explicitly by the former safety licensing analyses. Under these circumstances, MHI made some improvements for GCS code system. A new resonance calculation model of lattice physics code and homogeneous cross section representative model for core simulator have been developed to apply more wide range core conditions corresponding to severe accident status such like anticipated transient without scram (ATWS) analysis and criticality evaluation of dried-up spent fuel pit. As a result of these improvements, GCS code system has very wide calculation applicability with good accuracy for any core conditions as far as fuel is not damaged. In this paper, the outline of GCS code system is described briefly and recent relevant development activities are presented.

  17. Recent improvements of reactor physics codes in MHI

    Energy Technology Data Exchange (ETDEWEB)

    Kosaka, Shinya, E-mail: shinya-kosaka@mhi.co.jp; Yamaji, Kazuya; Kirimura, Kazuki; Kamiyama, Yohei; Matsumoto, Hideki [Mitsubishi Heavy Industries, Ltd. (Japan)

    2015-12-31

    This paper introduces recent improvements for reactor physics codes in Mitsubishi Heavy Industries, Ltd(MHI). MHI has developed a new neutronics design code system Galaxy/Cosmo-S(GCS) for PWR core analysis. After TEPCO’s Fukushima Daiichi accident, it is required to consider design extended condition which has not been covered explicitly by the former safety licensing analyses. Under these circumstances, MHI made some improvements for GCS code system. A new resonance calculation model of lattice physics code and homogeneous cross section representative model for core simulator have been developed to apply more wide range core conditions corresponding to severe accident status such like anticipated transient without scram (ATWS) analysis and criticality evaluation of dried-up spent fuel pit. As a result of these improvements, GCS code system has very wide calculation applicability with good accuracy for any core conditions as far as fuel is not damaged. In this paper, the outline of GCS code system is described briefly and recent relevant development activities are presented.

  18. Measurements of effective delayed neutron fraction in a fast neutron reactor using the perturbation method

    Science.gov (United States)

    Zhou, Hao-Jun; Yin, Yan-Peng; Fan, Xiao-Qiang; Li, Zheng-Hong; Pu, Yi-Kang

    2016-06-01

    A perturbation method is proposed to obtain the effective delayed neutron fraction β eff of a cylindrical highly enriched uranium reactor. Based on reactivity measurements with and without a sample at a specified position using the positive period technique, the reactor reactivity perturbation Δρ of the sample in β eff units is measured. Simulations of the perturbation experiments are performed using the MCNP program. The PERT card is used to provide the difference dk of effective neutron multiplication factors with and without the sample inside the reactor. Based on the relationship between the effective multiplication factor and the reactivity, the equation β eff = dk/Δρ is derived. In this paper, the reactivity perturbations of 13 metal samples at the designable position of the reactor are measured and calculated. The average β eff value of the reactor is given as 0.00645, and the standard uncertainty is 3.0%. Additionally, the perturbation experiments for β eff can be used to evaluate the reliabilities of the delayed neutron parameters. This work shows that the delayed neutron data of 235U and 238U from G.R. Keepin’s publication are more reliable than those from ENDF-B6.0, ENDF-B7.0, JENDL3.3 and CENDL2.2. Supported by Foundation of Key Laboratory of Neutron Physics, China Academy of Engineering Physics (2012AA01, 2014AA01), National Natural Science Foundation (11375158, 91326104)

  19. RSMASS-D models: An improved method for estimating reactor and shield mass for space reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, A.C.

    1997-10-01

    Three relatively simple mathematical models have been developed to estimate minimum reactor and radiation shield masses for liquid-metal-cooled reactors (LMRs), in-core thermionic fuel element (TFE) reactors, and out-of-core thermionic reactors (OTRs). The approach was based on much of the methodology developed for the Reactor/Shield Mass (RSMASS) model. Like the original RSMASS models, the new RSMASS-derivative (RSMASS-D) models use a combination of simple equations derived from reactor physics and other fundamental considerations, along with tabulations of data from more detailed neutron and gamma transport theory computations. All three models vary basic design parameters within a range specified by the user to achieve a parameter choice that yields a minimum mass for the power level and operational time of interest. The impact of critical mass, fuel damage, and thermal limitations are accounted for to determine the required fuel mass. The effect of thermionic limitations are also taken into account for the thermionic reactor models. All major reactor component masses are estimated, as well as instrumentation and control (I&C), boom, and safety system masses. A new shield model was developed and incorporated into all three reactor concept models. The new shield model is more accurate and simpler to use than the approach used in the original RSMASS model. The estimated reactor and shield masses agree with the mass predictions from separate detailed calculations within 15 percent for all three models.

  20. A fast and flexible reactor physics model for simulating neutron spectra and depletion in fast reactors

    Science.gov (United States)

    Recktenwald, Geoff; Deinert, Mark

    2010-03-01

    Determining the time dependent concentration of isotopes within a nuclear reactor core is central to the analysis of nuclear fuel cycles. We present a fast, flexible tool for determining the time dependent neutron spectrum within fast reactors. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to simulate the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. While originally developed for LWR simulations, the model is shown to produce fast reactor spectra that show high degree of fidelity to available fast reactor benchmarks.

  1. Reactor physics experiments related to transmutation in the KUCA

    Energy Technology Data Exchange (ETDEWEB)

    Shiroya, Seiji [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst.

    1997-11-01

    At the Kyoto University Critical Assembly (KUCA), {sup 237}Np/{sup 235}U fission rate ratios are being measured using the back-to-back type double fission chamber to examine the nuclear data and the computational method for the transmutation of minor actinides (MA) in light water reactors (LWRs). The neutron spectra of cores are systematically being varied by changing the moderator-to-fuel volume ratio (V{sub m}/V{sub f}). The measured data are being compared with the calculated results by SRAC with three different nuclear data files. It has been indicated that the calculated results with JENDL-3.2 agreed better with the measured ones than those with JENDL-3.1 and ENDF/B-VI, although the calculated results underestimated the measured ones by around 10%. (author)

  2. Experimental investigation of a new method for advanced fast reactor shutdown cooling

    Science.gov (United States)

    Pakholkov, V. V.; Kandaurov, A. A.; Potseluev, A. I.; Rogozhkin, S. A.; Sergeev, D. A.; Troitskaya, Yu. I.; Shepelev, S. F.

    2017-07-01

    We consider a new method for fast reactor shutdown cooling using a decay heat removal system (DHRS) with a check valve. In this method, a coolant from the decay heat exchanger (DHX) immersed into the reactor upper plenum is supplied to the high-pressure plenum and, then, inside the fuel subassemblies (SAs). A check valve installed at the DHX outlet opens by the force of gravity after primary pumps (PP-1) are shut down. Experimental studies of the new and alternative methods of shutdown cooling were performed at the TISEY test facility at OKBM. The velocity fields in the upper plenum of the reactor model were obtained using the optical particle image velocimetry developed at the Institute of Applied Physics (Russian Academy of Sciences). The study considers the process of development of natural circulation in the reactor and the DHRS models and the corresponding evolution of the temperature and velocity fields. A considerable influence of the valve position in the displacer of the primary pump on the natural circulation of water in the reactor through the DHX was discovered (in some modes, circulation reversal through the DHX was obtained). Alternative DHRS designs without a shell at the DHX outlet with open and closed check valve are also studied. For an open check valve, in spite of the absence of a shell, part of the flow is supplied through the DHX pipeline and then inside the SA simulators. When simulating power modes of the reactor operation, temperature stratification of the liquid was observed, which increased in the cooling mode via the DHRS. These data qualitatively agree with the results of tests at BN-600 and BN-800 reactors.

  3. Software for neutron physical calculations of fast neutron reactors. [PRIDAN, MED, ANALIT

    Energy Technology Data Exchange (ETDEWEB)

    Jordanov, J.; Antonov, N. (Bylgarska Akademiya na Naukite, Sofia. Inst. za Yadrena Izsledvaniya i Yadrena Energetika)

    1983-01-01

    A set of programs has been developed for neutron physical calculations of fast neutron reactors. The set includes the PRIDAN program for calculating the effective cross-sections of the media, characteristics of the fast reactor systems, the MED program - a single dimension multigroup program for calculating fast reactors in multigroup diffusion approximation and the ANALIT program for calculating the criticality in plane geometry. PRIDAN uses the formalism of the self-shielding factors and prepares the effective cross-sections using an iterative procedure. The values of the multigroup cross-sections obtained for different temperatures are used directly as input data for the other programs. MED calculates the critical dimensions, the coefficient of effective multiplication, the real and conjugated neutron fluxes for each enegry group and each spatial zone, the distribution of the neutron fission sources. The maximum number of enegry groups - 26 and the maximum number of spatial points - 170, are distributed in seven spatial zones at most. The ANALIT program solves under given conditions the diffusion multigroup equation analytically using a method proposed by the authors. The body of mathematic pertaining to the case is presented. The results obtained, which demonstrate how functional the programs are and how applicable they are for neutron-physical analysis, are presented in tabular form.

  4. Health physics aspects of advanced reactor licensing reviews

    Energy Technology Data Exchange (ETDEWEB)

    Hinson, C.S. [Nuclear Regulatory Commission, Washington, DC (United States)

    1995-03-01

    The last Construction Permit to be issued by the U.S. Nuclear Regulatory Commission (NRC) for a U.S. light water reactor (LWR) was granted in the late 1970s. In 1989 the NRC issued 10 CFR Part 52 which is intended to serve as a framework for the licensing of future reactor designs. The NRC is currently reviewing four different future on {open_quotes}next-generation{close_quotes} reactor designs. Two of these designs are classified as evolutionary designs (modified versions of current generation LWRs) and two are advanced designs (reactors incorporating simplified designs and passive means for accident mitigation). These {open_quotes}next-generation{close_quotes} reactor designs incorporate many innovative design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four {open_quotes}next-generation{close_quotes} reactor designs incorporate many innovative design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four {open_quotes}next-generation{close_quotes} reactor designs currently being reviewed by the NRC.

  5. METHODS VERIFICATION FOR ELECTROTHERMAL CALCULATIONS OF ELECTRIC REACTORS WITHOUT STEEL

    Directory of Open Access Journals (Sweden)

    V. F. Ivankov

    2015-12-01

    Full Text Available Based on the example of the reactor without steel, type ROM-510/26 with electromagnetic shields, verification of analytical and numeral finite-element methods is carried out by the calculation results comparison. For the purpose of corrected analytical calculation, horizontal and vertical shields of the reactor are represented by the system of shortcircuited elements to consider their final dimensions. Calculation is performed as to their inductances, distribution of currents and losses in the shields, magnetic-field and losses in winding, calculation of winding heating by means of the «overheating» empirical method. It is illustrated that analytical calculations correspond to the researches using numeral methods of the electromagnetic and thermal CFD-analysis with sufficient accuracy. For the purpose of practical application in industrial designing of the equipment, the methods with approved and checked measurement results are recommended

  6. A special topic from nuclear reactor dynamics for the undergraduate physics curriculum

    Energy Technology Data Exchange (ETDEWEB)

    Sevenich, R.A.

    1977-04-01

    An example from the dynamics of nuclear reactors is presented for possible inclusion as a special topic in the undergraduate physics curriculum. An intuitive derivation of the point reactor equations is followed by formulation of equations for inverse and direct kinetics which are readily programmed on a digital computer. Several specific computer simulations involving the effect of control rod motion on reactor power are suggested as programming exercises for the student.

  7. System and method for temperature control in an oxygen transport membrane based reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Sean M.

    2017-02-21

    A system and method for temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

  8. IAEA Coordinated Research Project on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bostelmann, F. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal-hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, modelling and computational algorithms. (The interested reader is referred to the large body of published SA and UA literature for a more complete overview of the various types of uncertainties, methodologies and results obtained). SA is helpful for ranking the various sources of uncertainty and error in the results of core analyses. SA and UA are required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. SA and UA can guide experimental, modelling, and algorithm research and development. Current SA and UA rely either on derivative-based methods such as stochastic sampling methods or on generalized perturbation theory to obtain sensitivity coefficients. Neither approach addresses all needs. In order to benefit from recent advances in modelling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more robust and well validated calculation tools to meet design target accuracies. In February 2009, the Technical Working Group on Gas-Cooled Reactors (TWG-GCR) of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Program (CRP) on

  9. Core Physics of Pebble Bed High Temperature Nuclear Reactors

    NARCIS (Netherlands)

    Auwerda, G.J.

    2014-01-01

    To more accurately predict the temperature distribution inside the reactor core of pebble bed type high temperature reactors, in this thesis we investigated the stochastic properties of randomly stacked beds and the effects of the non-homogeneity of these beds on the neutronics and

  10. Pebble Bed Reactor: core physics and fuel cycle analysis

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Worley, B.A.

    1979-10-01

    The Pebble Bed Reactor is a gas-cooled, graphite-moderated high-temperature reactor that is continuously fueled with small spherical fuel elements. The projected performance was studied over a broad range of reactor applicability. Calculations were done for a burner on a throwaway cycle, a converter with recycle, a prebreeder and breeder. The thorium fuel cycle was considered using low, medium (denatured), and highly enriched uranium. The base calculations were carried out for electrical energy generation in a 1200 MW/sub e/ plant. A steady-state, continuous-fueling model was developed and one- and two-dimensional calculations were used to characterize performance. Treating a single point in time effects considerable savings in computer time as opposed to following a long reactor history, permitting evaluation of reactor performance over a broad range of design parameters and operating modes.

  11. Methods and strategies for future reactor safety goals

    Science.gov (United States)

    Arndt, Steven Andrew

    -informed analyses and discussions. This dissertation examines potential approaches to updating the safety goals that include the establishment of new quantitative safety goal associated with the comparative risk of generating electricity by viable competing technologies and modifications of the goals to account for multi-plant reactor sites, and issues associated with the use of safety goals in both initial licensing and operational decision making. This research develops a new quantitative health objective that uses a comparable benefit risk metric based on the life-cycle risk of the construction, operation and decommissioning of a comparable non-nuclear electric generation facility, as well as the risks associated with mining and transportation. This dissertation also evaluates the effects of using various methods for aggregating site risk as a safety metric, as opposed to using single plant safety goals. Additionally, a number of important assumptions inherent in the current safety goals, including the effect of other potential negative societal effects such as the generation of greenhouse gases (e.g., carbon dioxide) have on the risk of electric power production and their effects on the setting of safety goals, is explored. Finally, the role risk perception should play in establishing safety goals has been explored. To complete this evaluation, a new method to analytically compare alternative technologies of generating electricity was developed, including development of a new way to evaluate risk perception, and a new method was developed for evaluating the risk at multiple units on a single site. To test these modifications to the safety goals a number of possible reactor designs and configurations were evaluated using these new proposed safety goals to determine the goals' usefulness and utility. The results of the analysis showed that the modifications provide measures that more closely evaluate the potential risk to the public from the operation of nuclear power plants than

  12. Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPL

    Directory of Open Access Journals (Sweden)

    A. Rais

    2015-01-01

    Full Text Available In order to analyze the steady state and transient behavior of the CROCUS reactor, several methods and models need to be developed in the areas of reactor physics, thermal-hydraulics, and multiphysics coupling. The long-term objectives of this project are to work towards the development of a modern method for the safety analysis of research reactors and to update the Final Safety Analysis Report of the CROCUS reactor. A first part of the paper deals with generation of a core simulator nuclear data library for the CROCUS reactor using the Serpent 2 Monte Carlo code and also with reactor core modeling using the PARCS code. PARCS eigenvalue, radial power distribution, and control rod reactivity worth results were benchmarked against Serpent 2 full-core model results. Using the Serpent 2 model as reference, PARCS eigenvalue predictions were within 240 pcm, radial power was within 3% in the central region of the core, and control rod reactivity worth was within 2%. A second part reviews the current methodology used for the safety analysis of the CROCUS reactor and presents the envisioned approach for the multiphysics modeling of the reactor.

  13. Methods of modern mathematical physics

    CERN Document Server

    Reed, Michael

    1980-01-01

    This book is the first of a multivolume series devoted to an exposition of functional analysis methods in modern mathematical physics. It describes the fundamental principles of functional analysis and is essentially self-contained, although there are occasional references to later volumes. We have included a few applications when we thought that they would provide motivation for the reader. Later volumes describe various advanced topics in functional analysis and give numerous applications in classical physics, modern physics, and partial differential equations.

  14. Initial Testing of the Microscopic Depletion Implementation in the MAMMOTH Reactor Physics Application

    Energy Technology Data Exchange (ETDEWEB)

    Ortensi, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wang, Y. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schunert, S. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ganapol, B. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gleicher, F. N. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Baker, B. [Idaho National Lab. (INL), Idaho Falls, ID (United States); DeHart, M. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    Present and new nuclear fuels that will be tested at the Transient Reactor Test (TREAT) facility will be analyzed with the MAMMOTH reactor physics application, currently under development, at Idaho National Laboratory. MAMMOTH natively couples the BISON, RELAP-7, and Rattlesnake applications within the MOOSE framework. This system allows the irradiation of fuel from beginning of life in a nuclear reactor until it is placed in TREAT for fuel testing within the same analysis mesh and, thus, retaining a very high level of resolution and fidelity. The calculation of the isotopic distribution in fuel requires the solution to the decay and transmutation equations coupled to the neutron transport equation. The Chebyshev Rational Approximation Method (CRAM) is the current state-of-the-art in the field, as was chosen to be the solver for the decay and transmutation equations. This report shows that the implementation of the CRAM solver within MAMMOTH is correct with various analytic benchmarks for decay and transmutation of nuclides. The results indicate that the solutions with CRAM order 16 achieve the level of precision of the benchmark. The CRAM solutions show little sensitivity to the time step size and consistently produce a high level of accuracy for isotopic decay for time steps of 1x10^11 years. Comparisons to DRAGON5 with 297 isotopes yield comparable results, but some differences need to be further analyzed.

  15. Ceramic oxygen transport membrane array reactor and reforming method

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Sean M.; Christie, Gervase Maxwell; Robinson, Charles; Wilson, Jamie R.; Gonzalez, Javier E.; Doraswami, Uttam R.

    2016-11-08

    The invention relates to a commercially viable modular ceramic oxygen transport membrane reforming reactor configured using repeating assemblies of oxygen transport membrane tubes and catalytic reforming reactors.

  16. Reactor physics studies in the GCFR Phase III critical assembly

    Energy Technology Data Exchange (ETDEWEB)

    Morman, J A [ed.

    1980-03-01

    The third phase of the gas cooled fast reactor (GCFR) program, ZPR-9 Assembly 30, is based on a multi-zoned core of PuO/sub 2/-UO/sub 2/ with radial and axial blankets of UO/sub 2/. Studies performed in this assembly will be compared to the previous phases of the GCFR program and will help to define parameters in this power-flattened demonstration plant-type core. Measurements in the Phase III program included small sample reactivity worths of various materials, central reaction rates and reaction rate distributions, absorption-to-fission ratios and the central point conversion ratio and the worth of steam entry into a small central zone. The reactivity change associated with the construction of a central pin zone in the core and axial blanket was measured. Reaction rate and steam entry measurements were repeated in the pin environment. Standard analysis methods using ENDF/B-IV data are described and the results are compared to measurements performed during the program.

  17. Overview of Experiments for Physics of Fast Reactors from the International Handbooks of Evaluated Criticality Safety Benchmark Experiments and Evaluated Reactor Physics Benchmark Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Bess, J. D.; Briggs, J. B.; Gulliford, J.; Ivanova, T.; Rozhikhin, E. V.; Semenov, M. Yu.; Tsibulya, A. M.; Koscheev, V. N.

    2017-07-01

    Overview of Experiments to Study the Physics of Fast Reactors Represented in the International Directories of Critical and Reactor Experiments John D. Bess Idaho National Laboratory Jim Gulliford, Tatiana Ivanova Nuclear Energy Agency of the Organisation for Economic Cooperation and Development E.V.Rozhikhin, M.Yu.Sem?nov, A.M.Tsibulya Institute of Physics and Power Engineering The study the physics of fast reactors traditionally used the experiments presented in the manual labor of the Working Group on Evaluation of sections CSEWG (ENDF-202) issued by the Brookhaven National Laboratory in 1974. This handbook presents simplified homogeneous model experiments with relevant experimental data, as amended. The Nuclear Energy Agency of the Organization for Economic Cooperation and Development coordinates the activities of two international projects on the collection, evaluation and documentation of experimental data - the International Project on the assessment of critical experiments (1994) and the International Project on the assessment of reactor experiments (since 2005). The result of the activities of these projects are replenished every year, an international directory of critical (ICSBEP Handbook) and reactor (IRPhEP Handbook) experiments. The handbooks present detailed models of experiments with minimal amendments. Such models are of particular interest in terms of the settlements modern programs. The directories contain a large number of experiments which are suitable for the study of physics of fast reactors. Many of these experiments were performed at specialized critical stands, such as BFS (Russia), ZPR and ZPPR (USA), the ZEBRA (UK) and the experimental reactor JOYO (Japan), FFTF (USA). Other experiments, such as compact metal assembly, is also of interest in terms of the physics of fast reactors, they have been carried out on the universal critical stands in Russian institutes (VNIITF and VNIIEF) and the US (LANL, LLNL, and others.). Also worth mentioning

  18. Reactor pressure vessel head vents and methods of using the same

    Science.gov (United States)

    Gels, John L; Keck, David J; Deaver, Gerald A

    2014-10-28

    Internal head vents are usable in nuclear reactors and include piping inside of the reactor pressure vessel with a vent in the reactor upper head. Piping extends downward from the upper head and passes outside of the reactor to permit the gas to escape or be forcibly vented outside of the reactor without external piping on the upper head. The piping may include upper and lowers section that removably mate where the upper head joins to the reactor pressure vessel. The removable mating may include a compressible bellows and corresponding funnel. The piping is fabricated of nuclear-reactor-safe materials, including carbon steel, stainless steel, and/or a Ni--Cr--Fe alloy. Methods install an internal head vent in a nuclear reactor by securing piping to an internal surface of an upper head of the nuclear reactor and/or securing piping to an internal surface of a reactor pressure vessel.

  19. Analysis of Nigeria research reactor-1 thermal power calibration methods

    Energy Technology Data Exchange (ETDEWEB)

    Agbo, Sunday Arome; Ahmed, Yusuf Aminu; Ewa, Ita Okon; Jibrin, Yahaya [Ahmadu Bello University, Zaria (Nigeria)

    2016-06-15

    This paper analyzes the accuracy of the methods used in calibrating the thermal power of Nigeria Research Reactor-1 (NIRR-1), a low-power miniature neutron source reactor located at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria, Nigeria. The calibration was performed at three different power levels: low power (3.6 kW), half power (15 kW), and full power (30 kW). Two methods were used in the calibration, namely, slope and heat balance methods. The thermal power obtained by the heat balance method at low power, half power, and full power was 3.7 ± 0.2 kW, 15.2 ± 1.2 kW, and 30.7 ± 2.5 kW, respectively. The thermal power obtained by the slope method at half power and full power was 15.8 ± 0.7 kW and 30.2 ± 1.5 kW, respectively. It was observed that the slope method is more accurate with deviations of 4% and 5% for calibrations at half and full power, respectively, although the linear fit (slope method) on average temperature-rising rates during the thermal power calibration procedure at low power (3.6 kW) is not fitting. As such, the slope method of power calibration is not suitable at lower power for NIRR-1.

  20. Systems and methods for dismantling a nuclear reactor

    Science.gov (United States)

    Heim, Robert R; Adams, Scott Ryan; Cole, Matthew Denver; Kirby, William E; Linnebur, Paul Damon

    2014-10-28

    Systems and methods for dismantling a nuclear reactor are described. In one aspect the system includes a remotely controlled heavy manipulator ("manipulator") operatively coupled to a support structure, and a control station in a non-contaminated portion of a workspace. The support structure provides the manipulator with top down access into a bioshield of a nuclear reactor. At least one computing device in the control station provides remote control to perform operations including: (a) dismantling, using the manipulator, a graphite moderator, concrete walls, and a ceiling of the bioshield, the manipulator being provided with automated access to all internal portions of the bioshield; (b) loading, using the manipulator, contaminated graphite blocks from the graphite core and other components from the bioshield into one or more waste containers; and (c) dispersing, using the manipulator, dust suppression and contamination fixing spray to contaminated matter.

  1. Fault Diagnosis of Batch Reactor Using Machine Learning Methods

    Directory of Open Access Journals (Sweden)

    Sujatha Subramanian

    2014-01-01

    Full Text Available Fault diagnosis of a batch reactor gives the early detection of fault and minimizes the risk of thermal runaway. It provides superior performance and helps to improve safety and consistency. It has become more vital in this technical era. In this paper, support vector machine (SVM is used to estimate the heat release (Qr of the batch reactor both normal and faulty conditions. The signature of the residual, which is obtained from the difference between nominal and estimated faulty Qr values, characterizes the different natures of faults occurring in the batch reactor. Appropriate statistical and geometric features are extracted from the residual signature and the total numbers of features are reduced using SVM attribute selection filter and principle component analysis (PCA techniques. artificial neural network (ANN classifiers like multilayer perceptron (MLP, radial basis function (RBF, and Bayes net are used to classify the different types of faults from the reduced features. It is observed from the result of the comparative study that the proposed method for fault diagnosis with limited number of features extracted from only one estimated parameter (Qr shows that it is more efficient and fast for diagnosing the typical faults.

  2. The Dynamic Monte Carlo Method for Transient Analysis of Nuclear Reactors

    NARCIS (Netherlands)

    Sjenitzer, B.L.

    2013-01-01

    In this thesis a new method for the analysis of power transients in a nuclear reactor is developed, which is more accurate than the present state-of-the-art methods. Transient analysis is important tool when designing nuclear reactors, since they predict the behaviour of a reactor during changing

  3. PT IP-754 physics testing of H Reactor: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Vaughn, A.D.; Masche, G.C.

    1965-10-07

    Significant changes in fuel design, operating levels, loading geometry, and moderator condition have taken place since the Hanford reactors started operation. Changes which affect nuclear safety parameters require continuing analysis of the possible effects on safety system strengths and maximum potential reactivities. For this reason the shutdown of a Hanford reactor offered an excellent opportunity to measure many of the above factors without a significant production loss. Therefore a test program was proposed and carried out at H Reactor starting April 22, 1965 and ending May 5, 1965. All of the tests were designed to provide information concerning parameters which affect nuclear safety and control calculation limits. This document is a summary of the deactivation test results without application and recommendations to specific nuclear safety questions. Rather, it is intended that the results listed herein will serve as a basis for further study and application to nuclear safety technology.

  4. Testing the applicability of the k0-NAA method at the MINT's TRIGA MARK II reactor

    Science.gov (United States)

    Siong, Wee Boon; Dung, Ho Manh; Wood, Ab. Khalik; Salim, Nazaratul Ashifa Abd.; Elias, Md. Suhaimi

    2006-08-01

    The Analytical Chemistry Laboratory at MINT is using the NAA technique since 1980s and is the only laboratory in Malaysia equipped with a research reactor, namely the TRIGA MARK II. Throughout the years the development of NAA technique has been very encouraging and was made applicable to a wide range of samples. At present, the k0 method has become the preferred standardization method of NAA ( k0-NAA) due to its multi-elemental analysis capability without using standards. Additionally, the k0 method describes NAA in physically and mathematically understandable definitions and is very suitable for computer evaluation. Eventually, the k0-NAA method has been adopted by MINT in 2003, in collaboration with the Nuclear Research Institute (NRI), Vietnam. The reactor neutron parameters ( α and f) for the pneumatic transfer system and for the rotary rack at various locations, as well as the detector efficiencies were determined. After calibration of the reactor and the detectors, the implemented k0 method was validated by analyzing some certified reference materials (including IAEA Soil 7, NIST 1633a, NIST 1632c, NIST 1646a and IAEA 140/TM). The analysis results of the CRMs showed an average u score well below the threshold value of 2 with a precision of better than ±10% for most of the elemental concentrations obtained, validating herewith the introduction of the k0-NAA method at the MINT.

  5. Statistical methods for physical science

    CERN Document Server

    Stanford, John L

    1994-01-01

    This volume of Methods of Experimental Physics provides an extensive introduction to probability and statistics in many areas of the physical sciences, with an emphasis on the emerging area of spatial statistics. The scope of topics covered is wide-ranging-the text discusses a variety of the most commonly used classical methods and addresses newer methods that are applicable or potentially important. The chapter authors motivate readers with their insightful discussions, augmenting their material withKey Features* Examines basic probability, including coverage of standard distributions, time s

  6. The Multi一physics Research on I ron一Core Vibration Noise of Power Reactor

    Directory of Open Access Journals (Sweden)

    LI U Ja

    2017-02-01

    Full Text Available On the basis of theoretical research releted to the magnetostriction and maxwell’.s equations,the fi- nite element coupling in the transient electromagnetic field coupling,structure and sound field coupling has been developed In thts paper by using the flnlte element sOftWare CO}IS01., Whleh establish a serles three-phase COT’e re- actor model, to analyzing the power frequency magnetic field distribution,core magnetostrictive displacement,max- well force displacement and sound pressure level of the three-phase series core reactor under the power frequency working state. According to transient magnetic field distribution in the simulation of the reactor,the magnetic flux density distribution inside the reactor and the vibration displacement distribution are calculated,the acoustic field distribution is measured alao. It is shown that physical field simulation results and measured data are basically in consisent by experiment,it is proved multi-physics coupling is an effective method for forecast of noise.

  7. An overview of the current status of resonance theory in reactor physics applications

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, R.N.

    1993-12-31

    The neutron resonance phenomena constitute one of the most fundamental subjects in nuclear physics as well as in reactor physics. It is the area where the concepts of nuclear interaction and the treatment of the neutronic balance in reactor lattices become intertwined. The later requires the detailed knowledge of resonance structures of many nuclide of practical interest to the development of nuclear energy. The key issue of the resonance treatment in reactor applications is directly associated with the use of the microscopic cross sections in the macroscopic reactor cells with a wide range of composition, temperature,and geometric configurations. It gives rise to the so called self-shielding effect. The accurate estimations of such a effect is essential not only in the calculation of the criticality of a reactor but also from the point of view of safety considerations. The latter manifests through the Doppler effect particularly crucial to the fast reactor development. The task of accurate treatment of the self-shielding effect, however, is by no means simple. In fact, it is perhaps the most complicated problem in neutron physics which, strictly speaking, requires the dependence of many physical variables. Two important elements of particular interest are : (1) a concise description of the resonance cross sections as a function of energy and temperature; (2) accurate estimation of the corresponding neutron flux where appropriate. These topics will be discussed from both the historical as well as the state-of-art perspectives.

  8. Material characteristics and construction methods for a typical research reactor concrete containment in Iran

    Energy Technology Data Exchange (ETDEWEB)

    Ebrahimia, Mahsa; Suha, Kune Y. [Seoul National Univ., Seoul (Korea, Republic of); Eghbalic, Rahman; Jahan, Farzaneh Asadi malek [School of Architecture and Urbanism, Qazvin (Iran, Islamic Republic of)

    2012-10-15

    Generally selecting an appropriate material and also construction style for a concrete containment due to its function and special geometry play an important role in applicability and also construction cost and duration decrease in a research reactor (RR) project. The reactor containment enclosing the reactor vessel comprises physical barriers reflecting the safety design and construction codes, regulations and standards so as to prevent the community and the environment from uncontrolled release of radioactive materials. It is the third and the last barrier against radioactivity release. It protects the reactor vessel from such external events as earthquake and aircraft crash as well. Thus, it should be designed and constructed in such a manner as to withstand dead and live loads, ground and seismic loads, missiles and aircraft loads, and thermal and shrinkage loads. This study aims to present a construction method for concrete containment of a typical RR in Iran. The work also presents an acceptable characteristic for concrete and reinforcing re bar of a typical concrete containment. The current study has evaluated the various types of the RR containments. The most proper type was selected in accordance with the current knowledge and technology of Iran.

  9. Analysis of Nigeria Research Reactor-1 Thermal Power Calibration Methods

    Directory of Open Access Journals (Sweden)

    Sunday Arome Agbo

    2016-06-01

    Full Text Available This paper analyzes the accuracy of the methods used in calibrating the thermal power of Nigeria Research Reactor-1 (NIRR-1, a low-power miniature neutron source reactor located at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria, Nigeria. The calibration was performed at three different power levels: low power (3.6 kW, half power (15 kW, and full power (30 kW. Two methods were used in the calibration, namely, slope and heat balance methods. The thermal power obtained by the heat balance method at low power, half power, and full power was 3.7 ± 0.2 kW, 15.2 ± 1.2 kW, and 30.7 ± 2.5 kW, respectively. The thermal power obtained by the slope method at half power and full power was 15.8 ± 0.7 kW and 30.2 ± 1.5 kW, respectively. It was observed that the slope method is more accurate with deviations of 4% and 5% for calibrations at half and full power, respectively, although the linear fit (slope method on average temperature-rising rates during the thermal power calibration procedure at low power (3.6 kW is not fitting. As such, the slope method of power calibration is not suitable at lower power for NIRR-1.

  10. High power ring methods and accelerator driven subcritical reactor application

    Energy Technology Data Exchange (ETDEWEB)

    Tahar, Malek Haj [Univ. of Grenoble (France)

    2016-08-07

    High power proton accelerators allow providing, by spallation reaction, the neutron fluxes necessary in the synthesis of fissile material, starting from Uranium 238 or Thorium 232. This is the basis of the concept of sub-critical operation of a reactor, for energy production or nuclear waste transmutation, with the objective of achieving cleaner, safer and more efficient process than today’s technologies allow. Designing, building and operating a proton accelerator in the 500-1000 MeV energy range, CW regime, MW power class still remains a challenge nowadays. There is a limited number of installations at present achieving beam characteristics in that class, e.g., PSI in Villigen, 590 MeV CW beam from a cyclotron, SNS in Oakland, 1 GeV pulsed beam from a linear accelerator, in addition to projects as the ESS in Europe, a 5 MW beam from a linear accelerator. Furthermore, coupling an accelerator to a sub-critical nuclear reactor is a challenging proposition: some of the key issues/requirements are the design of a spallation target to withstand high power densities as well as ensure the safety of the installation. These two domains are the grounds of the PhD work: the focus is on the high power ring methods in the frame of the KURRI FFAG collaboration in Japan: upgrade of the installation towards high intensity is crucial to demonstrate the high beam power capability of FFAG. Thus, modeling of the beam dynamics and benchmarking of different codes was undertaken to validate the simulation results. Experimental results revealed some major losses that need to be understood and eventually overcome. By developing analytical models that account for the field defects, one identified major sources of imperfection in the design of scaling FFAG that explain the important tune variations resulting in the crossing of several betatron resonances. A new formula is derived to compute the tunes and properties established that characterize the effect of the field imperfections on the

  11. High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems

    Science.gov (United States)

    Mahadevan, Vijay S.; Merzari, Elia; Tautges, Timothy; Jain, Rajeev; Obabko, Aleksandr; Smith, Michael; Fischer, Paul

    2014-01-01

    An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework. PMID:24982250

  12. High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems.

    Science.gov (United States)

    Mahadevan, Vijay S; Merzari, Elia; Tautges, Timothy; Jain, Rajeev; Obabko, Aleksandr; Smith, Michael; Fischer, Paul

    2014-08-06

    An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework.

  13. Geometric Methods in Physics XXXV

    CERN Document Server

    Odzijewicz, Anatol; Previato, Emma

    2018-01-01

    This book features a selection of articles based on the XXXV Białowieża Workshop on Geometric Methods in Physics, 2016. The series of Białowieża workshops, attended by a community of experts at the crossroads of mathematics and physics, is a major annual event in the field. The works in this book, based on presentations given at the workshop, are previously unpublished, at the cutting edge of current research, typically grounded in geometry and analysis, and with applications to classical and quantum physics. In 2016 the special session "Integrability and Geometry" in particular attracted pioneers and leading specialists in the field. Traditionally, the Białowieża Workshop is followed by a School on Geometry and Physics, for advanced graduate students and early-career researchers, and the book also includes extended abstracts of the lecture series.

  14. Test tasks for verification of program codes for calculation of neutron-physical characteristics of the BN series reactors

    Science.gov (United States)

    Tikhomirov, Georgy; Ternovikh, Mikhail; Smirnov, Anton; Saldikov, Ivan; Bahdanovich, Rynat; Gerasimov, Alexander

    2017-09-01

    System of test tasks is presented with the fast reactor BN-1200 with nitride fuel as prototype. The system of test tasks includes three test based on different geometric models. Model of fuel element in homogeneous and in heterogeneous form, model of fuel assembly in height-heterogeneous and full heterogeneous form, and modeling of the active core of BN-1200 reactor. Cross-verification of program codes was performed. Transition from simple geometry to more complex one allows to identify the causes of discrepancies in the results during the early stage of cross-verification of codes. This system of tests can be applied for certification of engineering programs based on the method of Monte Carlo to the calculation of full-scale models of the reactor core of the BN series. The developed tasks take into account the basic layout and structural features of the reactor BN-1200. They are intended for study of neutron-physical characteristics, estimation of influence of heterogeneous structure and influence of diffusion approximation. The development of system of test tasks allowed to perform independent testing of programs for calculation of neutron-physical characteristics: engineering programs JARFR and TRIGEX, and codes MCU, TDMCC, and MMK based on the method of Monte Carlo.

  15. Catalyst support structure, catalyst including the structure, reactor including a catalyst, and methods of forming same

    Science.gov (United States)

    Van Norman, Staci A.; Aston, Victoria J.; Weimer, Alan W.

    2017-05-09

    Structures, catalysts, and reactors suitable for use for a variety of applications, including gas-to-liquid and coal-to-liquid processes and methods of forming the structures, catalysts, and reactors are disclosed. The catalyst material can be deposited onto an inner wall of a microtubular reactor and/or onto porous tungsten support structures using atomic layer deposition techniques.

  16. Substantiation of physical concepts of fast reactors in Russia: experience and prospects

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, P.N. [Russian Research Center ' Kurchatov Institute' (RRC KI), 1, Kurchatov Sq., Moscow, 123182 (Russian Federation); Vasiliev, B.A. [Experimental Design Bureau of Machine Building (OKBM) 15, Burnakovskiy Pr., N. Novgorod, 603074 (Russian Federation); Kormilitsyn, M.V. [State Scientific Center of Russian Federation - Research Institute of Atomic Reactors (NIIAR) Dimitrovgrad-10, Ulianovsk Reg., 433510 (Russian Federation); Lopatkin, A.V. [N.A. Dollezhal Research and Development Institute of Power Engineering (NIKIET) 2/8, M. Krasnoselskaya Str., Moscow, 107140 (Russian Federation); Seleznev, E.F. [All-Russian Research Institute for Nuclear Power Plant Operation (VNIIAES) 25, Ferganskaya, Moscow, 109507 (Russian Federation); Khomyakov, Yu.S.; Tsybulia, A.M. [State Scientific Center of the Russian Federation - A. I. Leypunsky Institute for Physics and Power Engineering (SSC RF- IPPE) 1, Bondarenko Sq., Obninsk, Kaluga Reg., 249033 (Russian Federation); Tocheny, L.V. [International Science and Technology Center (ISTC) 32-34 Krasnoproletarskaya Ulitsa, Moscow, 127473 (Russian Federation)

    2008-07-01

    The fast reactor concept in Russia has accumulated unique experience, since its advent in the 1950's and up to the present, from the creation of the first experimental installation BR-1, experimental reactors BR-5 and BOR-60, the pilot industrial reactors BN-350 in Kazakhstan and up to the BN-600 at Beloyarsk Atomic Power Station. Investigations on the first experimental installations BR-1 and BR-5/-10 proved the propriety of the idea that it is possible to create nuclear reactors that can produce more nuclear fuel than they consume, i.e. the idea of breeding. The architecture of such reactors was also designed, producing a current leader among fast reactors with sodium coolant and oxide uranium-plutonium fuel. Operational experience of BOR-60, BN-350 and, particularly, BN-600 confirmed the engineering and technical feasibility of the concept of fast reactors, the possibility for its realization both for power production and for certain other purposes as well, such as desalinisation of sea water (BN-350) and for radionuclide production (BN-350, BN-600), and it enabled the development and verification of different models, computer methods and codes. The paper presents a review of experience in the creation of plants with fast reactors, scientific research on these installations, principal results, the current status of experimental data analysis, and prospective directions in the development of fast reactors and the corresponding experimental basis in Russia. (authors)

  17. Integral reactor system and method for fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Fernandes, Neil Edward; Brown, Michael S.; Cheekatamaria, Praveen; Deng, Thomas; Dimitrakopoulos, James; Litka, Anthony F.

    2017-03-07

    A reactor system is integrated internally within an anode-side cavity of a fuel cell. The reactor system is configured to convert higher hydrocarbons to smaller species while mitigating the lower production of solid carbon. The reactor system may incorporate one or more of a pre-reforming section, an anode exhaust gas recirculation device, and a reforming section.

  18. 10 CFR Appendix D to Part 73 - Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Physical Protection of Irradiated Reactor Fuel in Transit... Irradiated Reactor Fuel in Transit, Training Program Subject Schedule Pursuant to the provision of § 73.37 of... reactor fuel is required to assure that individuals used as shipment escorts have completed a training...

  19. Reactor physics ideas to design novel reactors with faster fissile growth

    Energy Technology Data Exchange (ETDEWEB)

    Jagannathan, V. [Light Water Reactors Physics Section, Reactor Physics Design Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India)], E-mail: vjagan@barc.gov.in; Pal, Usha; Karthikeyan, R.; Raj, Devesh; Srivastava, Argala [Light Water Reactors Physics Section, Reactor Physics Design Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Khan, Suhail Ahmad [Reactor Project Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India)

    2008-08-15

    There are several types of fission reactors operating in the world adopting generally the open fuel cycle which considers the naturally available fissile nuclide, viz., {sup 235}U. The accumulated discharged fuel is considered as waste in some countries. However the discharged fuel contains the precious man-made fissile plutonium which would provide the sole means of harnessing the nuclear energy from either depleted uranium or the natural thorium in future. It must be emphasized that the present day power reactors use just about 0.5% of the mined uranium and it would be imprudent to discard the rest of the mass as waste. It is therefore necessary to explore ways and means of exploiting the fertile mass which has the potential of providing the energy without the green house effects for millennia to come. This has to be done by innovating means of large scale fertile to fissile conversion and then using the man-made fissile material for sustenance as well as growth of fission nuclear power. This paper attempts to give a broad picture of the available options and the challenges in realizing the theoretical possibilities.

  20. Physical particularities of nuclear reactors using heavy moderators of neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Kulikov, G. G., E-mail: ggkulikov@mephi.ru; Shmelev, A. N. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute) (Russian Federation)

    2016-12-15

    In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using {sup 233}U as a fissile nuclide and {sup 232}Th and {sup 231}Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.

  1. Physical aspects of the Canadian generation IV supercritical water-cooled pressure tube reactor plant design

    Energy Technology Data Exchange (ETDEWEB)

    Gaudet, M.; Yetisir, M.; Haque, Z. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    The form of the containment building is a function of the requirements imposed by various systems. In order to provide sufficient driving force for naturally-circulated emergency cooling systems, as well as providing a gravity-driven core flooding pool function, the Canadian SCWR reactor design relies on elevation differences between the reactor and the safety systems. These elevation differences, the required cooling pool volumes and the optimum layout of safety-related piping are major factors influencing the plant design. As a defence-in-depth, the containment building and safety systems also provide successive barriers to the unplanned release of radioactive materials, while providing a path for heat flow to the ultimate heat sink, the atmosphere. Access to the reactor for refuelling is from the top of the reactor, with water used as shielding during the refuelling operations. The accessibility to the reactor and protection of the environment are additional factors influencing the plant design. This paper describes the physical implementation of the major systems of the Canadian SCWR within the reactor building, and the position of major plant services relative to the reactor building. (author)

  2. Specification of the Advanced Burner Test Reactor Multi-Physics Coupling Demonstration Problem

    Energy Technology Data Exchange (ETDEWEB)

    Shemon, E. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Grudzinski, J. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Thomas, J. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Yu, Y. Q. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-12-21

    This document specifies the multi-physics nuclear reactor demonstration problem using the SHARP software package developed by NEAMS. The SHARP toolset simulates the key coupled physics phenomena inside a nuclear reactor. The PROTEUS neutronics code models the neutron transport within the system, the Nek5000 computational fluid dynamics code models the fluid flow and heat transfer, and the DIABLO structural mechanics code models structural and mechanical deformation. The three codes are coupled to the MOAB mesh framework which allows feedback from neutronics, fluid mechanics, and mechanical deformation in a compatible format.

  3. Fast burner reactor benchmark results from the NEA working party on physics of plutonium recycle

    Energy Technology Data Exchange (ETDEWEB)

    Hill, R.N.; Wade, D.C. [Argonne National Lab., IL (United States); Palmiotti, G. [CEA - Cadarache, Saint-Paul-Les-Durance (France)

    1995-12-01

    As part of a program proposed by the OECD/NEA Working Party on Physics of Plutonium Recycling (WPPR) to evaluate different scenarios for the use of plutonium, fast reactor physics benchmarks were developed; fuel cycle scenarios using either PUREX/TRUEX (oxide fuel) or pyrometallurgical (metal fuel) separation technologies were specified. These benchmarks were designed to evaluate the nuclear performance and radiotoxicity impact of a transuranic-burning fast reactor system. International benchmark results are summarized in this paper; and key conclusions are highlighted.

  4. Multi-Physics Demonstration Problem with the SHARP Reactor Simulation Toolkit

    Energy Technology Data Exchange (ETDEWEB)

    Merzari, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Shemon, E. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Yu, Y. Q. [Argonne National Lab. (ANL), Argonne, IL (United States); Thomas, J. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Obabko, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Jain, Rajeev [Argonne National Lab. (ANL), Argonne, IL (United States); Mahadevan, Vijay [Argonne National Lab. (ANL), Argonne, IL (United States); Tautges, Timothy [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Solberg, Jerome [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ferencz, Robert Mark [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Whitesides, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-12-21

    This report describes to employ SHARP to perform a first-of-a-kind analysis of the core radial expansion phenomenon in an SFR. This effort required significant advances in the framework Multi-Physics Demonstration Problem with the SHARP Reactor Simulation Toolkit used to drive the coupled simulations, manipulate the mesh in response to the deformation of the geometry, and generate the necessary modified mesh files. Furthermore, the model geometry is fairly complex, and consistent mesh generation for the three physics modules required significant effort. Fully-integrated simulations of a 7-assembly mini-core test problem have been performed, and the results are presented here. Physics models of a full-core model of the Advanced Burner Test Reactor have also been developed for each of the three physics modules. Standalone results of each of the three physics modules for the ABTR are presented here, which provides a demonstration of the feasibility of the fully-integrated simulation.

  5. Introduction to the neutron kinetics of nuclear power reactors

    CERN Document Server

    Tyror, J G; Grant, P J

    2013-01-01

    An Introduction to the Neutron Kinetics of Nuclear Power Reactors introduces the reader to the neutron kinetics of nuclear power reactors. Topics covered include the neutron physics of reactor kinetics, feedback effects, water-moderated reactors, fast reactors, and methods of plant control. The reactor transients following faults are also discussed, along with the use of computers in the study of power reactor kinetics. This book is comprised of eight chapters and begins with an overview of the reactor physics characteristics of a nuclear power reactor and their influence on system design and

  6. A simplified method for limit conversion calculation in membrane reactors

    NARCIS (Netherlands)

    Gallucci, F.; De Falco, Marcello; Basile, Angelo

    2010-01-01

    Membrane reactors (MRs) are often used to carry out equilibrium limited reactions. This is because the thermodynamic equilibrium is a strong constrain for traditional systems. Even with very active catalysts, traditional reactors (TRs) cannot give conversions higher than those allowed by the

  7. Physical acoustics principles and methods

    CERN Document Server

    Mason, Warren P

    1964-01-01

    Physical Acoustics: Principles and Methods, Volume l-Part A focuses on high frequency sound waves in gases, liquids, and solids that have been proven as powerful tools in analyzing the molecular, defect, domain wall, and other types of motions. The selection first tackles wave propagation in fluids and normal solids and guided wave propagation in elongated cylinders and plates. Discussions focus on fundamentals of continuum mechanics; small-amplitude waves in a linear viscoelastic medium; representation of oscillations and waves; and special effects associated with guided elastic waves in plat

  8. Students' Assessment of Interactive Distance Experimentation in Nuclear Reactor Physics Laboratory Education

    Science.gov (United States)

    Malkawi, Salaheddin; Al-Araidah, Omar

    2013-01-01

    Laboratory experiments develop students' skills in dealing with laboratory instruments and physical processes with the objective of reinforcing the understanding of the investigated subject. In nuclear engineering, where research reactors play a vital role in the practical education of students, the high cost and long construction time of research…

  9. Compatibility analysis of DUPIC fuel (Part II) - Reactor physics design and analysis

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Choi, Hang Bok; Rhee, Bo Wook; Roh, Gyu Hong; Kim, Do Hun [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    The compatibility analysis of the DUPIC fuel in a CANDU reactor has been assessed. This study includes the fuel composition adjustment, comparison of lattice properties, performance analysis of reactivity devices, determination of regional over-power (ROP) trip setpoint, and uncertainty estimation of core performance parameters. For the DUPIC fuel composition adjustment, three options have been proposed, which can produce uniform neutronic characteristics of the DUPIC fuel. The lattice analysis has shown that the characteristics of the DUPIC fuel is compatible with those of natural uranium fuel. The reactivity devices of the CANDU-6 reactor maintain their functional requirements even for the DUPIC fuel system. The ROP analysis has shown that the trip setpoint is not sacrificed for the DUPIC fuel system owing to the power shape that enhances more thermal margin. The uncertainty analysis of the core performance parameter has shown that the uncertainty associated with the fuel composition variation is reduced appreciably, which is primarily due to the fuel composition adjustment and secondly the on-power refueling feature and spatial control function of the CANDU reactor. The reactor physics calculation has also shown that it is feasible to use spent PWR fuel directly in CANDU reactors without deteriorating the CANDU-6 core physics design requirements. 29 refs., 67 figs., 60 tabs. (Author)

  10. ARES: A Parallel Discrete Ordinates Transport Code for Radiation Shielding Applications and Reactor Physics Analysis

    Directory of Open Access Journals (Sweden)

    Yixue Chen

    2017-01-01

    Full Text Available ARES is a multidimensional parallel discrete ordinates particle transport code with arbitrary order anisotropic scattering. It can be applied to a wide variety of radiation shielding calculations and reactor physics analysis. ARES uses state-of-the-art solution methods to obtain accurate solutions to the linear Boltzmann transport equation. A multigroup discretization is applied in energy. The code allows multiple spatial discretization schemes and solution methodologies. ARES currently provides diamond difference with or without linear-zero flux fixup, theta weighted, directional theta weighted, exponential directional weighted, and linear discontinuous finite element spatial differencing schemes. Discrete ordinates differencing in angle and spherical harmonics expansion of the scattering source are adopted. First collision source method is used to eliminate or mitigate the ray effects. Traditional source iteration and Krylov iterative method preconditioned with diffusion synthetic acceleration are applied to solve the linear system of equations. ARES uses the Koch-Baker-Alcouffe parallel sweep algorithm to obtain high parallel efficiency. Verification and validation for the ARES transport code system have been done by lots of benchmarks. In this paper, ARES solutions to the HBR-2 benchmark and C5G7 benchmarks are in excellent agreement with published results. Numerical results are presented which demonstrate the accuracy and efficiency of these methods.

  11. Reactor Physics Assessment of Thick Silicon Carbide Clad PWR Fuels

    Science.gov (United States)

    2013-06-01

    Be was modeled in SERPENT ; the depletion of Be at 60 MWd/kg in 5.5% 235 U enriched fuel was negligible as the difference between the SERPENT predicted...SIMULATE in the evaluation of core physics performance. 77 Comparison of ENDF-VI based CASMO results with ENDF-VII based SERPENT results for PuO2

  12. Physically-Based Power-Level Control for Modular High Temperature Gas-Cooled Reactors

    Science.gov (United States)

    Dong, Zhe

    2012-10-01

    Because of its strong inherent safety, the modular high temperature gas-cooled nuclear reactor (MHTGR) has been regarded as the central part of the next generation nuclear plants (NGNPs). Power-level control is one of the key techniques which provide safe, stable and efficient operation for the MHTGRs. The physically-based regulation theory is definitely a promising trend of modern control theory and provides a control design method that can suppress the unstable part of the system dynamics and remain the stable part. Usually, the control law designed by the physically-based control theory has a simple form and high performance. Stimulated by this, a novel nonlinear dynamic output feedback power-level control is established in this paper for the MHTGR based upon its own dynamic features. This newly-built control strategy guarantees the globally asymptotic stability and provides a satisfactory transient performance through properly adjusting the feedback gains. Simulation results not only verify the correctness of the theoretical results but also illustrate the high control performance.

  13. Group theoretical methods in Physics

    Energy Technology Data Exchange (ETDEWEB)

    Olmo, M.A. del; Santander, M.; Mateos Guilarte, J.M. (eds.) (Universidad de Valladolid. Facultad de Ciencias. Valladolid (Spain))

    1993-01-01

    The meeting had 102 papers. These was distributed in following areas: -Quantum groups,-Integrable systems,-Physical Applications of Group Theory,-Mathematical Results,-Geometry, Topology and Quantum Field Theory,-Super physics,-Super mathematics,-Atomic, Molecular and Condensed Matter Physics. Nuclear and Particle Physics,-Symmetry and Foundations of classical and Quantum mechanics.

  14. Ceramic oxygen transport membrane array reactor and reforming method

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Sean M.; Christie, Gervase Maxwell; Rosen, Lee J.; Robinson, Charles; Wilson, Jamie R.; Gonzalez, Javier E.; Doraswami, Uttam R.

    2016-09-27

    A commercially viable modular ceramic oxygen transport membrane reforming reactor for producing a synthesis gas that improves the thermal coupling of reactively-driven oxygen transport membrane tubes and catalyst reforming tubes required to efficiently and effectively produce synthesis gas.

  15. Technical Basis for Physical Fidelity of NRC Control Room Training Simulators for Advanced Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Minsk, Brian S.; Branch, Kristi M.; Bates, Edward K.; Mitchell, Mark R.; Gore, Bryan F.; Faris, Drury K.

    2009-10-09

    The objective of this study is to determine how simulator physical fidelity influences the effectiveness of training the regulatory personnel responsible for examination and oversight of operating personnel and inspection of technical systems at nuclear power reactors. It seeks to contribute to the U.S. Nuclear Regulatory Commission’s (NRC’s) understanding of the physical fidelity requirements of training simulators. The goal of the study is to provide an analytic framework, data, and analyses that inform NRC decisions about the physical fidelity requirements of the simulators it will need to train its staff for assignment at advanced reactors. These staff are expected to come from increasingly diverse educational and experiential backgrounds.

  16. INTEGRAL BENCHMARKS AVAILABLE THROUGH THE INTERNATIONAL REACTOR PHYSICS EXPERIMENT EVALUATION PROJECT AND THE INTERNATIONAL CRITICALITY SAFETY BENCHMARK EVALUATION PROJECT

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; Lori Scott; Enrico Sartori; Yolanda Rugama

    2008-09-01

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) continue to expand their efforts and broaden their scope to identify, evaluate, and provide integral benchmark data for method and data validation. Benchmark model specifications provided by these two projects are used heavily by the international reactor physics, nuclear data, and criticality safety communities. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. The status of the IRPhEP and ICSBEP is discussed in this paper, and the future of the two projects is outlined and discussed. Selected benchmarks that have been added to the IRPhEP and ICSBEP handbooks since PHYSOR’06 are highlighted, and the future of the two projects is discussed.

  17. Beam transient analyses of Accelerator Driven Subcritical Reactors based on neutron transport method

    Energy Technology Data Exchange (ETDEWEB)

    He, Mingtao; Wu, Hongchun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China); Zheng, Youqi, E-mail: yqzheng@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China); Wang, Kunpeng [Nuclear and Radiation Safety Center, PO Box 8088, Beijing 100082 (China); Li, Xunzhao; Zhou, Shengcheng [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China)

    2015-12-15

    Highlights: • A transport-based kinetics code for Accelerator Driven Subcritical Reactors is developed. • The performance of different kinetics methods adapted to the ADSR is investigated. • The impacts of neutronic parameters deteriorating with fuel depletion are investigated. - Abstract: The Accelerator Driven Subcritical Reactor (ADSR) is almost external source dominated since there is no additional reactivity control mechanism in most designs. This paper focuses on beam-induced transients with an in-house developed dynamic analysis code. The performance of different kinetics methods adapted to the ADSR is investigated, including the point kinetics approximation and space–time kinetics methods. Then, the transient responds of beam trip and beam overpower are calculated and analyzed for an ADSR design dedicated for minor actinides transmutation. The impacts of some safety-related neutronics parameters deteriorating with fuel depletion are also investigated. The results show that the power distribution varying with burnup leads to large differences in temperature responds during transients, while the impacts of kinetic parameters and feedback coefficients are not very obvious. Classification: Core physic.

  18. An alternative method of determining the neutrino mass ordering in reactor neutrino experiments

    Directory of Open Access Journals (Sweden)

    S.M. Bilenky

    2017-09-01

    Full Text Available We discuss a novel alternative method of determining the neutrino mass ordering in medium baseline experiments with reactor antineutrinos. Results on the potential sensitivity of the new method are also presented.

  19. An alternative method of determining the neutrino mass ordering in reactor neutrino experiments

    Science.gov (United States)

    Bilenky, S. M.; Capozzi, F.; Petcov, S. T.

    2017-09-01

    We discuss a novel alternative method of determining the neutrino mass ordering in medium baseline experiments with reactor antineutrinos. Results on the potential sensitivity of the new method are also presented.

  20. An overview of modeling methods for thermal mixing and stratification in large enclosures for reactor safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Per F. Peterson

    2010-10-01

    Thermal mixing and stratification phenomena play major roles in the safety of reactor systems with large enclosures, such as containment safety in current fleet of LWRs, long-term passive containment cooling in Gen III+ plants including AP-1000 and ESBWR, the cold and hot pool mixing in pool type sodium cooled fast reactor systems (SFR), and reactor cavity cooling system behavior in high temperature gas cooled reactors (HTGR), etc. Depending on the fidelity requirement and computational resources, 0-D steady state models (heat transfer correlations), 0-D lumped parameter based transient models, 1-D physical-based coarse grain models, and 3-D CFD models are available. Current major system analysis codes either have no models or only 0-D models for thermal stratification and mixing, which can only give highly approximate results for simple cases. While 3-D CFD methods can be used to analyze simple configurations, these methods require very fine grid resolution to resolve thin substructures such as jets and wall boundaries. Due to prohibitive computational expenses for long transients in very large volumes, 3-D CFD simulations remain impractical for system analyses. For mixing in stably stratified large enclosures, UC Berkeley developed 1-D models basing on Zuber’s hierarchical two-tiered scaling analysis (HTTSA) method where the ambient fluid volume is represented by 1-D transient partial differential equations and substructures such as free or wall jets are modeled with 1-D integral models. This allows very large reductions in computational effort compared to 3-D CFD modeling. This paper will present an overview on important thermal mixing and stratification phenomena in large enclosures for different reactors, major modeling methods and their advantages and limits, potential paths to improve simulation capability and reduce analysis uncertainty in this area for advanced reactor system analysis tools.

  1. Calculation of reactor antineutrino spectra in TEXONO

    CERN Document Server

    Chen Dong Liang; Mao Ze Pu; Wong, T H

    2002-01-01

    In the low energy reactor antineutrino physics experiments, either for the researches of antineutrino oscillation and antineutrino reactions, or for the measurement of abnormal magnetic moment of antineutrino, the flux and the spectra of reactor antineutrino must be described accurately. The method of calculation of reactor antineutrino spectra was discussed in detail. Furthermore, based on the actual circumstances of NP2 reactors and the arrangement of detectors, the flux and the spectra of reactor antineutrino in TEXONO were worked out

  2. Optimization of reload of nuclear power plants using ACO together with the GENES reactor physics code

    Energy Technology Data Exchange (ETDEWEB)

    Lima, Alan M.M. de; Freire, Fernando S.; Nicolau, Andressa S.; Schirru, Roberto, E-mail: alan@lmp.ufrj.br, E-mail: andressa@lmp.ufrj.br, E-mail: schirru@lmp.ufrj.br, E-mail: ffreire@eletronuclear.gov.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Eletrobras Termonuclear S.A. (ELETRONUCLEAR), Rio de Janeiro, RJ (Brazil)

    2017-11-01

    The Nuclear reload of a Pressurized Water Reactor (PWR) occurs whenever the burning of the fuel elements can no longer maintain the criticality of the reactor, that is, it cannot maintain the Nuclear power plant operates within its nominal power. Nuclear reactor reload optimization problem consists of finding a loading pattern of fuel assemblies in the reactor core in order to minimize the cost/benefit ratio, trying to obtain maximum power generation with a minimum of cost, since in all reloads an average of one third of the new fuel elements are purchased. This loading pattern must also satisfy constraints of symmetry and security. In practice, it consists of the placing 121 fuel elements in 121 core positions, in the case of the Angra 1 Brazilian Nuclear Power Plant (NPP), making this new arrangement provide the best cost/benefit ratio. It is an extremely complex problem, since it has around 1% of great places. A core of 121 fuel elements has approximately 10{sup 13} combinations and 10{sup 11} great locations. With this number of possible combinations it is impossible to test all, in order to choose the best. In this work a system called ACO-GENES is proposed in order to optimization the Nuclear Reactor Reload Problem. ACO is successfully used in combination problems, and it is expected that ACO-GENES will show a robust optimization system, since in addition to optimizing ACO, it allows important prior knowledge such as K infinite, burn, etc. After optimization by ACO-GENES, the best results will be validated by a licensed reactor physics code and will be compared with the actual results of the cycle. (author)

  3. Multi-physical developments for safety related investigations of low moderated boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Schlenker, Markus Thomas

    2014-12-19

    The main objective of this dissertation is the development and optimization of a low moderated boiling water reactor (BWR) core with improved fuel utilization to be incorporated in a Gen II BWR nuclear power plant. The assessment of the new core design is done by comparing it with a full MOX BWR core design regarding neutron physical and thermal-hydraulic design and safety criteria (e.g. inherent reactivity coefficients) and different sustainability parameters (e.g. conversion ratio).

  4. Development of a coupling approach for multi-physics analyses of fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Qiu, Yuefeng

    2016-05-12

    An integrated multi-physics coupling system has been developed for fusion reactor systems analyses. This system has an advanced Monte Carlo (MC) modeling approach for converting complex CAD models to MC models with hybrid constructive solid and unstructured mesh geometries, and a high-fidelity coupling approach for data mapping from MC to thermal hydraulics and structural mechanics codes. The system was proven to be reliable, robust and efficient through verification calculations.

  5. Control of plutonium content using a concept of physical accounting method for adjusted fissile enrichment

    Energy Technology Data Exchange (ETDEWEB)

    Nakae, Nobuo; Ito, Masanori; Mishima, Tsuyoshi; Shiina, Akira.

    1989-06-01

    In Fast Breeder Reactor (FBR), plutonium is used as main fissile material. Plutonium has the following characteristics with respect to reactor reactivity. (1) Plutonium is obtained by reprocessing of spent fuels. The plutonium isotope ratio depends on the burn-up of the spent fuels. (2) Pu-241 which is one of the major Plutonium isotopes lose reactivity quickly with the disintegration to Am-241. (3) Each plutonium isotope has considerably different effect on reactor reactivity. A concept of physical accounting method for adjusted fissile enrichment has been introduced in controlling plutonium content of the fuel loaded into PNC's Monju reactor. The method enables exact control of the fresh fuel's reactivity. In this report basic concept of the method, plutonium content control in a fuel fabrication line based on the method and the simulation result are explained. (author).

  6. Methods of Mathematical Physics, 2

    CERN Document Server

    Courant, Richard

    1989-01-01

    Since the first volume of this work came out in Germany in 1937, this book, together with its first volume, has remained standard in the field. Courant and Hilbert's treatment restores the historically deep connections between physical intuition and mathematical development, providing the reader with a unified approach to mathematical physics. The present volume represents Richard Courant's final revision of 1961.

  7. Methods of Mathematical Physics, 1

    CERN Document Server

    Courant, Richard

    1989-01-01

    Since the first volume of this work came out in Germany in 1924, this book, together with its second volume, has remained standard in the field. Courant and Hilbert's treatment restores the historically deep connections between physical intuition and mathematical development, providing the reader with a unified approach to mathematical physics. The present volume represents Richard Courant's second and final revision of 1953.

  8. THERMAL FISSION REACTOR COMPOSITIONS AND METHOD OF FABRICATING SAME

    Science.gov (United States)

    Blainey, A.

    1959-10-01

    A body is presented for use in a thermal fission reactor comprising a sintered compressed mass of a substance of the group consisting of uranium, thorium, and oxides and carbides of uranium and thorium, enclosed in an envelope of a sintered, compacted, heat-conductive material of the group consisting of beryllium, zirconium, and oxides and carbides of beryllium and zirconium.

  9. Mathematical methods of classical physics

    CERN Document Server

    Cortés, Vicente

    2017-01-01

    This short primer, geared towards students with a strong interest in mathematically rigorous approaches, introduces the essentials of classical physics, briefly points out its place in the history of physics and its relation to modern physics, and explains what benefits can be gained from a mathematical perspective. As a starting point, Newtonian mechanics is introduced and its limitations are discussed. This leads to and motivates the study of different formulations of classical mechanics, such as Lagrangian and Hamiltonian mechanics, which are the subjects of later chapters. In the second part, a chapter on classical field theories introduces more advanced material. Numerous exercises are collected in the appendix.

  10. Applicability of a multivariable autoregressive method to boiling water reactor core stability estimation

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, S.; Fukunishi, K.; Kishi, S.; Yoshimoto, Y.; Kishimoto, K.

    1986-08-01

    A multivariable autoregressive (MAR) method is applied to the core stability estimation of a boiling water reactor-5 operation. Noise data measured during steady-state operations at startup tests are used. In this method, the closed loop transfer function from reactor pressure to reactor power is identified from reactor noise data and transformed into an impulse response function. The decay ratio representing stability characteristics is evaluated from this function. The variation range of decay ratio estimates obtained by this method is sufficiently small, if the analyzing conditions are appropriately selected. The value of the decay ratio is 0.23 during natural circulation and decreases with core flow, reaching close to zero at the rated power. A similar power dependence for the decay ratio is seen in results from a core stability analysis code. The MAR method is a useful tool for stability estimation, even if no external disturbance tests are conducted.

  11. An Evaluation of liquid metal leak detection methods for the Clinch River Breeder Reactor Plant

    Energy Technology Data Exchange (ETDEWEB)

    Morris, C.J.; Doctor, S.R.

    1977-12-01

    This report documents an independent review and evaluation of sodium leak detection methods described in the Clinch River Breeder Reactor Preliminary Safety Analysis Report. Only information in publicly available documents was used in making the assessments.

  12. Computational Neutronics Methods and Transmutation Performance Analyses for Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    R. Ferrer; M. Asgari; S. Bays; B. Forget

    2007-03-01

    The once-through fuel cycle strategy in the United States for the past six decades has resulted in an accumulation of Light Water Reactor (LWR) Spent Nuclear Fuel (SNF). This SNF contains considerable amounts of transuranic (TRU) elements that limit the volumetric capacity of the current planned repository strategy. A possible way of maximizing the volumetric utilization of the repository is to separate the TRU from the LWR SNF through a process such as UREX+1a, and convert it into fuel for a fast-spectrum Advanced Burner Reactor (ABR). The key advantage in this scenario is the assumption that recycling of TRU in the ABR (through pyroprocessing or some other approach), along with a low capture-to-fission probability in the fast reactor’s high-energy neutron spectrum, can effectively decrease the decay heat and toxicity of the waste being sent to the repository. The decay heat and toxicity reduction can thus minimize the need for multiple repositories. This report summarizes the work performed by the fuel cycle analysis group at the Idaho National Laboratory (INL) to establish the specific technical capability for performing fast reactor fuel cycle analysis and its application to a high-priority ABR concept. The high-priority ABR conceptual design selected is a metallic-fueled, 1000 MWth SuperPRISM (S-PRISM)-based ABR with a conversion ratio of 0.5. Results from the analysis showed excellent agreement with reference values. The independent model was subsequently used to study the effects of excluding curium from the transuranic (TRU) external feed coming from the LWR SNF and recycling the curium produced by the fast reactor itself through pyroprocessing. Current studies to be published this year focus on analyzing the effects of different separation strategies as well as heterogeneous TRU target systems.

  13. Ceramic oxygen transport membrane array reactor and reforming method

    Science.gov (United States)

    Kelly, Sean M.; Christie, Gervase Maxwell; Robinson, Charles; Wilson, Jamie R; Gonzalez, Javier E.; Doraswami, Uttam R.

    2017-10-03

    The invention relates to a commercially viable modular ceramic oxygen transport membrane system for utilizing heat generated in reactively-driven oxygen transport membrane tubes to generate steam, heat process fluid and/or provide energy to carry out endothermic chemical reactions. The system provides for improved thermal coupling of oxygen transport membrane tubes to steam generation tubes or process heater tubes or reactor tubes for efficient and effective radiant heat transfer.

  14. Statistical methods in radiation physics

    CERN Document Server

    Turner, James E; Bogard, James S

    2012-01-01

    This statistics textbook, with particular emphasis on radiation protection and dosimetry, deals with statistical solutions to problems inherent in health physics measurements and decision making. The authors begin with a description of our current understanding of the statistical nature of physical processes at the atomic level, including radioactive decay and interactions of radiation with matter. Examples are taken from problems encountered in health physics, and the material is presented such that health physicists and most other nuclear professionals will more readily understand the application of statistical principles in the familiar context of the examples. Problems are presented at the end of each chapter, with solutions to selected problems provided online. In addition, numerous worked examples are included throughout the text.

  15. Statistical methods in physical mapping

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, David O. [Univ. of California, Berkeley, CA (United States)

    1995-05-01

    One of the great success stories of modern molecular genetics has been the ability of biologists to isolate and characterize the genes responsible for serious inherited diseases like fragile X syndrome, cystic fibrosis and myotonic muscular dystrophy. This dissertation concentrates on constructing high-resolution physical maps. It demonstrates how probabilistic modeling and statistical analysis can aid molecular geneticists in the tasks of planning, execution, and evaluation of physical maps of chromosomes and large chromosomal regions. The dissertation is divided into six chapters. Chapter 1 provides an introduction to the field of physical mapping, describing the role of physical mapping in gene isolation and ill past efforts at mapping chromosomal regions. The next two chapters review and extend known results on predicting progress in large mapping projects. Such predictions help project planners decide between various approaches and tactics for mapping large regions of the human genome. Chapter 2 shows how probability models have been used in the past to predict progress in mapping projects. Chapter 3 presents new results, based on stationary point process theory, for progress measures for mapping projects based on directed mapping strategies. Chapter 4 describes in detail the construction of all initial high-resolution physical map for human chromosome 19. This chapter introduces the probability and statistical models involved in map construction in the context of a large, ongoing physical mapping project. Chapter 5 concentrates on one such model, the trinomial model. This chapter contains new results on the large-sample behavior of this model, including distributional results, asymptotic moments, and detection error rates. In addition, it contains an optimality result concerning experimental procedures based on the trinomial model. The last chapter explores unsolved problems and describes future work.

  16. Some mathematical methods of physics

    CERN Document Server

    Goertzel, Gerald

    2014-01-01

    This well-rounded, thorough treatment for advanced undergraduates and graduate students introduces basic concepts of mathematical physics involved in the study of linear systems. The text emphasizes eigenvalues, eigenfunctions, and Green's functions. Prerequisites include differential equations and a first course in theoretical physics.The three-part presentation begins with an exploration of systems with a finite number of degrees of freedom (described by matrices). In part two, the concepts developed for discrete systems in previous chapters are extended to continuous systems. New concepts u

  17. GROWTH OF THE INTERNATIONAL CRITICALITY SAFETY AND REACTOR PHYSICS EXPERIMENT EVALUATION PROJECTS

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; John D. Bess; Jim Gulliford

    2011-09-01

    Since the International Conference on Nuclear Criticality Safety (ICNC) 2007, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) have continued to expand their efforts and broaden their scope. Eighteen countries participated on the ICSBEP in 2007. Now, there are 20, with recent contributions from Sweden and Argentina. The IRPhEP has also expanded from eight contributing countries in 2007 to 16 in 2011. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments1' have increased from 442 evaluations (38000 pages), containing benchmark specifications for 3955 critical or subcritical configurations to 516 evaluations (nearly 55000 pages), containing benchmark specifications for 4405 critical or subcritical configurations in the 2010 Edition of the ICSBEP Handbook. The contents of the Handbook have also increased from 21 to 24 criticality-alarm-placement/shielding configurations with multiple dose points for each, and from 20 to 200 configurations categorized as fundamental physics measurements relevant to criticality safety applications. Approximately 25 new evaluations and 150 additional configurations are expected to be added to the 2011 edition of the Handbook. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments2' have increased from 16 different experimental series that were performed at 12 different reactor facilities to 53 experimental series that were performed at 30 different reactor facilities in the 2011 edition of the Handbook. Considerable effort has also been made to improve the functionality of the searchable database, DICE (Database for the International Criticality Benchmark Evaluation Project) and verify the accuracy of the data contained therein. DICE will be discussed in separate papers at ICNC 2011. The status of the

  18. Reactor physics modelling of accident tolerant fuel for LWRs using ANSWERS codes

    Directory of Open Access Journals (Sweden)

    Lindley Benjamin A.

    2016-01-01

    Full Text Available The majority of nuclear reactors operating in the world today and similarly the majority of near-term new build reactors will be LWRs. These currently accommodate traditional Zr clad UO2/PuO2 fuel designs which have an excellent performance record for normal operation. However, the events at Fukushima culminated in significant hydrogen production and hydrogen explosions, resulting from high temperature Zr/steam interaction following core uncovering for an extended period. These events have resulted in increased emphasis towards developing more accident tolerant fuels (ATFs-clad systems, particularly for current and near-term build LWRs. R&D programmes are underway in the US and elsewhere to develop ATFs and the UK is engaging in these international programmes. Candidate advanced fuel materials include uranium nitride (UN and uranium silicide (U3Si2. Candidate cladding materials include advanced stainless steel (FeCrAl and silicon carbide. The UK has a long history in industrial fuel manufacture and fabrication for a wide range of reactor systems including LWRs. This is supported by a national infrastructure to perform experimental and theoretical R&D in fuel performance, fuel transient behaviour and reactor physics. In this paper, an analysis of the Integral Inherently Safe LWR design (I2S-LWR, a reactor concept developed by an international collaboration led by the Georgia Institute of Technology, within a US DOE Nuclear Energy University Program (NEUP Integrated Research Project (IRP is considered. The analysis is performed using the ANSWERS reactor physics code WIMS and the EDF Energy core simulator PANTHER by researchers at the University of Cambridge. The I2S-LWR is an advanced 2850 MWt integral PWR with inherent safety features. In order to enhance the safety features, the baseline fuel and cladding materials that were chosen for the I2S-LWR design are U3Si2 and advanced stainless steel respectively. In addition, the I2S-LWR design

  19. Oxygen transport membrane system and method for transferring heat to catalytic/process reactors

    Science.gov (United States)

    Kelly, Sean M; Kromer, Brian R; Litwin, Michael M; Rosen, Lee J; Christie, Gervase Maxwell; Wilson, Jamie R; Kosowski, Lawrence W; Robinson, Charles

    2014-01-07

    A method and apparatus for producing heat used in a synthesis gas production is provided. The disclosed method and apparatus include a plurality of tubular oxygen transport membrane elements adapted to separate oxygen from an oxygen containing stream contacting the retentate side of the membrane elements. The permeated oxygen is combusted with a hydrogen containing synthesis gas stream contacting the permeate side of the tubular oxygen transport membrane elements thereby generating a reaction product stream and radiant heat. The present method and apparatus also includes at least one catalytic reactor containing a catalyst to promote the stream reforming reaction wherein the catalytic reactor is surrounded by the plurality of tubular oxygen transport membrane elements. The view factor between the catalytic reactor and the plurality of tubular oxygen transport membrane elements radiating heat to the catalytic reactor is greater than or equal to 0.5.

  20. Updates to the Generation of Physics Data Inputs for MAMMOTH Simulations of the Transient Reactor Test Facility - FY2016

    Energy Technology Data Exchange (ETDEWEB)

    Ortensi, Javier [Idaho National Lab. (INL), Idaho Falls, ID (United States); Baker, Benjamin Allen [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schunert, Sebastian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wang, Yaqi [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gleicher, Frederick Nathan [Idaho National Lab. (INL), Idaho Falls, ID (United States); DeHart, Mark David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-06-01

    The INL is currently evolving the modeling and simulation (M&S) capability that will enable improved core operation as well as design and analysis of TREAT experiments. This M&S capability primarily uses MAMMOTH, a reactor physics application being developed under Multi-physics Object Oriented Simulation Environment (MOOSE) framework. MAMMOTH allows the coupling of a number of other MOOSE-based applications. This second year of work has been devoted to the generation of a deterministic reference solution for the full core, the preparation of anisotropic diffusion coefficients, the testing of the SPH equivalence method, and the improvement of the control rod modeling. In addition, this report includes the progress made in the modeling of the M8 core configuration and experiment vehicle since January of this year.

  1. Chemical reactor and method for chemically converting a first material into a second material

    Science.gov (United States)

    Kong, Peter C.

    2008-04-08

    A chemical reactor and method for converting a first material into a second material is disclosed and wherein the chemical reactor is provided with a feed stream of a first material which is to be converted into a second material; and wherein the first material is combusted in the chemical reactor to produce a combustion flame, and a resulting gas; and an electrical arc is provided which is passed through or superimposed upon the combustion flame and the resulting gas to facilitate the production of the second material.

  2. A method of reactor power decrease by 2DOF control system during BWR power oscillation

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, Nobuyuki; Suzuki, Katsuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-09-01

    Occurrence of power oscillation events caused by void feedback effects in BWRs operated at low-flow and high-power condition has been reported. After thoroughly examining these events, BWRs have been equipped with the SRI (Selected Rod Insertion) system to avoid the power oscillation by decreasing the power under such reactor condition. This report presents a power control method for decreasing the reactor power stably by a two degree of freedom (2DOF) control. Performing a numerical simulation by utilizing a simple reactor dynamics model, it is found that the control system designed attains a satisfactory control performance of power decrease from a viewpoint of setting time and oscillation. (author)

  3. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  4. Validation and Benchmarking of Westinghouse BWR lattice physics methods

    OpenAIRE

    Luszczek, Karol

    2015-01-01

    A lattice physics code is a vital tool, forming a base of reactor coreanalysis. It enables the neutronic properties of the fuel assembly to becalculated and generates a proper set of data to be used by a 3-D full coresimulator. Due to advancement and complexity of modern Boiling WaterReactor assembly designs, a new deterministic lattice physics codeis being developed at Westinghouse Sweden AB, namely PHOENIX5.Each time a new code is written, its methodology of solving the neutrontransport equ...

  5. FN approximation of the solution to a singular integral equation of classical reactor physics

    Energy Technology Data Exchange (ETDEWEB)

    Ganapol, B.D. [Department of Aerospace and Mechanical Engineering, University of Arizona, AME Building, Tucson, AZ 85721 (United States)]. E-mail: ganapol@ame.arizona.edu

    2004-11-01

    The iterated FN method is applied to a singular integral equation arising from a classical problem of reactor physics to determine the distribution of fissile material giving a spatially uniform flux. The FN iterations are accelerated toward convergence through the Wynn-algorithm - but first - Happy Birthday 'Fast Eddie' Larsen Why do I refer to the well known, most proper and exquisitely accomplished Edward W. Larsen as 'Fast Eddie'. Well our story begins in a small back bar room in the lobby of one of Los Alamos' finest and most luxurious hotels. Two young men were having a transport theoretic discussion while they were engaged in a serious game of pool with monetary benefits going to the winner. In addition, the two were sipping their most favorite lavation in rather large quantities - one, a short stocky man with thinning hair, was sipping to forget the cost of his recent divorce, and the other, a shorter stockier man also with thinning hair, was drinking, well because he liked to drink and it just made him silly. As they continued their transport discussion, one stocky man turned to the other and said, 'I wonder what 'Fast Eddie' Larsen would say to our transport question'. The other stocky man immediately thought the 'Fast Eddie' reference was to Paul Newman who played 'Fast Eddie', an expert at applied particle transport theory (a pool player) in the movie the Hustler and asked if indeed this was the case. The first stocky man said 'No. I call everyone with the name Ed 'Fast Eddie' ' - and that's the story of how 'Fast Eddie' Larsen got his name. Happy 60th Ed and thanks for all the great transport theory - from one of your biggest fans.

  6. Physics design of initial and approach to equilibrium cores of a reactor concept for thorium utilization

    Energy Technology Data Exchange (ETDEWEB)

    Pal, Usha [Light Water Reactor Physics Section, Reactor Physics Design Division, Bhabha Atomic Research Centre, A-5-15, Central Complex, Mumbai, Maharashtra (India)], E-mail: ushapal@barc.gov.in; Jagannathan, V. [Light Water Reactor Physics Section, Reactor Physics Design Division, Bhabha Atomic Research Centre, A-5-15, Central Complex, Mumbai, Maharashtra (India)], E-mail: vjagan@barc.gov.in

    2008-07-15

    A thermal reactor concept 'a thorium breeder reactor' (ATBR) was conceived and reported by the authors during 1998. The distinctive physical characteristics of ATBR core with different types of seed fuels have been discussed in subsequent publications. The equilibrium core of ATBR with Pu seed was shown to exhibit a flat and low excess reactivity for a fuel cycle duration of two years. Notably this is achieved by no conventional burnable poison but by intrinsic balancing of reactivity between fissile and fertile zones. In this paper we present the design of the initial core and the refueling strategy for subsequent fuel cycles to enable a smooth transition to the equilibrium core. Three fuel types with characteristics similar to the three batch fuels of equilibrium core were designed for the initial core. Fuel requirement for the initial core is 4673 kg of reactor grade (RG) Pu for a cycle length of two years at 1875 MWt as against the 2200 kg needed for each fuel cycle of equilibrium core for same quantum of energy. The core reactivity variation during the first fuel cycle is monotonic fall and is relatively high ({approx}40 mk) but gradually diminishes to {+-}5 mk for fuel cycles 5-8.

  7. On the accuracy of reactor physics calculations for square HPLWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Jatuff, F. [Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland)]. E-mail: fabian.jatuff@psi.ch; Macku, K. [Ecole Polytechnique Federale de Lausanne (EPFL), CH-1015 Lausanne (Switzerland); Chawla, R. [Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne (EPFL), CH-1015 Lausanne (Switzerland)

    2006-01-15

    Although the supercritical-pressure or high-performance light water reactor (HPLWR) concept is largely based on the well-established technological experience available with conventional light water reactors, there is still no consensus on various key design features such as an optimal layout for the fuel assembly. This results mainly from the very large density variations of supercritical-pressure water in the core, which render it difficult to ensure reliable values for parameters such as power peaking factors and reactivity worths. The present paper describes studies carried out to compare deterministic and Monte Carlo codes for analysing a representative square HPLWR lattice with uniform 5%-enriched UO{sub 2} fuel. The main purpose has been to assess the prediction accuracies achievable for integral parameters such as the multiplication factor, control absorber effectiveness, moderator/coolant density reactivity feedback and pin power distributions. The results show good agreement between the deterministic and stochastic calculations for the unperturbed lattice. However, for certain perturbed situations involving, for example, local coolant density changes in the assembly or control absorber insertion, the observed discrepancies are large enough to question the basic viability of the reactor physics design, e.g. with respect to the thermal performance.

  8. Plasma flow reactor for steady state monitoring of physical and chemical processes at high temperatures

    Science.gov (United States)

    Koroglu, Batikan; Mehl, Marco; Armstrong, Michael R.; Crowhurst, Jonathan C.; Weisz, David G.; Zaug, Joseph M.; Dai, Zurong; Radousky, Harry B.; Chernov, Alex; Ramon, Erick; Stavrou, Elissaios; Knight, Kim; Fabris, Andrea L.; Cappelli, Mark A.; Rose, Timothy P.

    2017-09-01

    We present the development of a steady state plasma flow reactor to investigate gas phase physical and chemical processes that occur at high temperature (1000 < T < 5000 K) and atmospheric pressure. The reactor consists of a glass tube that is attached to an inductively coupled argon plasma generator via an adaptor (ring flow injector). We have modeled the system using computational fluid dynamics simulations that are bounded by measured temperatures. In situ line-of-sight optical emission and absorption spectroscopy have been used to determine the structures and concentrations of molecules formed during rapid cooling of reactants after they pass through the plasma. Emission spectroscopy also enables us to determine the temperatures at which these dynamic processes occur. A sample collection probe inserted from the open end of the reactor is used to collect condensed materials and analyze them ex situ using electron microscopy. The preliminary results of two separate investigations involving the condensation of metal oxides and chemical kinetics of high-temperature gas reactions are discussed.

  9. A New In-core Production Method of Co-60 in CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lyu, Jinqi; Kim, Woosong; Kim, Yonghee [KAIST, Daejeon (Korea, Republic of); Park, Younwon [BEES Inc, Daejeon (Korea, Republic of)

    2016-05-15

    This study introduces an innovative method for Co-60 production in the CANDU6 core. In this new scheme, the central fuel element is replaced by a Co-59 target and Co-60 is obtained after the fuel bundle is discharged. It has been shown that the new method can produce significantly higher amount of Co-60 than the conventional Co production method in CANDU6 reactors without compromising the fuel burnup by removing some (<50%) of the adjuster rods in the whole core. The coolant void reactivity is noticeably reduced when a Co-59 target is loaded into the central pin of the fuel bundle. Meanwhile, the peak power in a fuel bundle is just a little higher due to the central Co-59 target than in conventional CANDU6 fuel design. The basic technology for Co-60 producing was developed by MDS Nordion and Atomic Energy of Canada Limited (AECL) in 1946 and the same technology was adapted and applied in CANDU6 power reactors. The standard CANDU6 reactor has 21 adjuster rods which are fully inserted into the core during normal operation. The stainless steel adjuster rods are replaced with neutronically-equivalent Co-59 adjusters to produce Co-60. Nowadays, the roles of the adjuster rods are rather vague since nuclear reactors cannot be quickly restarted after a sudden reactor trip due to more stringent regulations. In some Canadian CANDU6 reactors, some or all the adjuster rods are removed from the core to maximize the uranium utilization.

  10. Nuclear reactor design

    CERN Document Server

    2014-01-01

    This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

  11. 75 FR 6413 - Office of New Reactors; Proposed Revision to Standard Review Plan, Section 14.3.12 on Physical...

    Science.gov (United States)

    2010-02-09

    ... COMMISSION Office of New Reactors; Proposed Revision to Standard Review Plan, Section 14.3.12 on Physical... NUREG-0800, ``Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,'' on a proposed Revision 1 to Standard Review Plan (SRP), Section 14.3.12 on ``Physical Security...

  12. The New Cold Neutron Radiography Facility (CNRF) at the Mianyang Research Reactor of the China Academy of Engineering Physics

    Science.gov (United States)

    Bin, Tang; Heyong, Huo; Ke, Tang; Rogers, John; Haste, Martin; Christodoulou, Marios

    A new cold neutron radiography beamline has been designed and constructed for the Mianyang reactor at the Institute of Nuclear Physics and Chemistry of the China Academy of Engineering Physics. This paper describes the components of the system and demonstrates the achievable image resolution.

  13. A Roadmap and Discussion of Issues for Physics Analyses Required to Support Plutonium Disposition in VVER-1000 Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Primm, R.T.; Drischler, J.D.; Pavlovichev, A.M. Styrine, Y.A.

    2000-06-01

    The purpose of this report is to document the physics analyses that must be performed to successfully disposition weapons-usable plutonium in VVER-1000 reactors in the Russian Federation. The report is a document to support programmatic and financial planning. It does not include documentation of the technical procedures by which physics analyses are performed, nor are the results of any analyses included.

  14. Core damage severity evaluation for pressurized water reactors by artificial intelligence methods

    Science.gov (United States)

    Mironidis, Anastasios Pantelis

    1998-12-01

    During the course of nuclear power evolution, accidents have occurred. However, in the western world, none of them had a severe impact on the public because of the design features of nuclear plants. In nuclear reactors, barriers constitute physical obstacles to uncontrolled fission product releases. These barriers are an important factor in safety analysis. During an accident, reactor safety systems become actuated to prevent the barriers from been breached. In addition, operators are required to take specified actions, meticulously depicted in emergency response procedures. In an accident, on-the-spot knowledge regarding the condition of the core is necessary. In order to make the right decisions toward mitigating the accident severity and its consequences, we need to know the status of the core [1, 3]. However, power plant instrumentation that can provide a direct indication of the status of the core during the time when core damage is a potential outcome, does not exist. Moreover, the information from instruments may have large uncertainty of various types. Thus, a very strong potential for misinterpreting incoming information exists. This research endeavor addresses the problem of evaluating the core damage severity of a Pressurized Water Reactor during a transient or an accident. An expert system has been constructed, that incorporates knowledge and reasoning of human experts. The expert system's inference engine receives incoming plant data that originate in the plethora of core-related instruments. Its knowledge base relies on several massive, multivariate fuzzy logic rule-sets, coupled with several artificial neural networks. These mathematical models have encoded information that defines possible core states, based on correlations of parameter values. The inference process classifies the core as intact, or as experiencing clad damage and/or core melting. If the system detects a form of core damage, a quantification procedure will provide a numerical

  15. Systems and methods for managing shared-path instrumentation and irradiation targets in a nuclear reactor

    Science.gov (United States)

    Heinold, Mark R.; Berger, John F.; Loper, Milton H.; Runkle, Gary A.

    2015-12-29

    Systems and methods permit discriminate access to nuclear reactors. Systems provide penetration pathways to irradiation target loading and offloading systems, instrumentation systems, and other external systems at desired times, while limiting such access during undesired times. Systems use selection mechanisms that can be strategically positioned for space sharing to connect only desired systems to a reactor. Selection mechanisms include distinct paths, forks, diverters, turntables, and other types of selectors. Management methods with such systems permits use of the nuclear reactor and penetration pathways between different systems and functions, simultaneously and at only distinct desired times. Existing TIP drives and other known instrumentation and plant systems are useable with access management systems and methods, which can be used in any nuclear plant with access restrictions.

  16. Theoretical and experimental physical methods of neutron-capture therapy

    Science.gov (United States)

    Borisov, G. I.

    2011-09-01

    This review is based to a substantial degree on our priority developments and research at the IR-8 reactor of the Russian Research Centre Kurchatov Institute. New theoretical and experimental methods of neutron-capture therapy are developed and applied in practice; these are: A general analytical and semi-empiric theory of neutron-capture therapy (NCT) based on classical neutron physics and its main sections (elementary theories of moderation, diffuse, reflection, and absorption of neutrons) rather than on methods of mathematical simulation. The theory is, first of all, intended for practical application by physicists, engineers, biologists, and physicians. This theory can be mastered by anyone with a higher education of almost any kind and minimal experience in operating a personal computer.

  17. Modelling dynamic processes in a nuclear reactor by state change modal method

    Science.gov (United States)

    Avvakumov, A. V.; Strizhov, V. F.; Vabishchevich, P. N.; Vasilev, A. O.

    2017-12-01

    Modelling of dynamic processes in nuclear reactors is carried out, mainly, using the multigroup neutron diffusion approximation. The basic model includes a multidimensional set of coupled parabolic equations and ordinary differential equations. Dynamic processes are modelled by a successive change of the reactor states. It is considered that the transition from one state to another occurs promptly. In the modal method the approximate solution is represented as eigenfunction expansion. The numerical-analytical method is based on the use of dominant time-eigenvalues of a group diffusion model taking into account delayed neutrons.

  18. Method for regenerating, replacing or treating the catalyst in a hydroprocessing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Epperly, W.R.; Sprague, B.N.; Kelso, D.T.; Bowers, W.E.

    1993-06-01

    A method is described for regenerating, replacing or treating the catalyst in a hydroprocessing reactor, which catalyst comprises a platinum group metal on a support, the method comprising admixing with the feedstock, recycle stream or hydrogen stream of the reactor an additive which comprises a nonionic, organometallic platinum group metal coordination composition wherein said composition (a) has a breakdown temperature between about 40 C. and about 570 C.; and (b) is substantially free from a disadvantageous amount of phosphorus, arsenic, sulfur, antimony or halides, wherein platinum group metal is caused to be deposited on said catalyst.

  19. A physical description of fission product behavior fuels for advanced power reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Kaganas, G.; Rest, J.; Nuclear Engineering Division; Florida International Univ.

    2007-10-18

    The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuels under varying operating conditions. Key sources include the FASTGRASS code with an application to UO{sub 2} power reactor fuel and the Dispersion Analysis Research Tool (DART ) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and ?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the {alpha}-, intermediate- and {gamma}-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile ?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.

  20. Ultrasonic methods in solid state physics

    CERN Document Server

    Truell, John; Elbaum, Charles

    1969-01-01

    Ultrasonic Methods in Solid State Physics is devoted to studies of energy loss and velocity of ultrasonic waves which have a bearing on present-day problems in solid-state physics. The discussion is particularly concerned with the type of investigation that can be carried out in the megacycle range of frequencies from a few megacycles to kilomegacycles; it deals almost entirely with short-duration pulse methods rather than with standing-wave methods. The book opens with a chapter on a classical treatment of wave propagation in solids. This is followed by separate chapters on methods and techni

  1. Development of multi-physics code systems based on the reactor dynamics code DYN3D

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany); Schuetze, Jochen [ANSYS Germany GmbH, Darmstadt (Germany); Frank, Thomas [ANSYS Germany GmbH, Otterfing (Germany); Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany)

    2011-07-15

    The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)

  2. Mathematical methods for physical and analytical chemistry

    CERN Document Server

    Goodson, David Z

    2011-01-01

    Mathematical Methods for Physical and Analytical Chemistry presents mathematical and statistical methods to students of chemistry at the intermediate, post-calculus level. The content includes a review of general calculus; a review of numerical techniques often omitted from calculus courses, such as cubic splines and Newton's method; a detailed treatment of statistical methods for experimental data analysis; complex numbers; extrapolation; linear algebra; and differential equations. With numerous example problems and helpful anecdotes, this text gives chemistry students the mathematical

  3. Passive and Active Radiation Measurements Capability at the INL Zero Power Physics Reactor (ZPPR) Facility

    Energy Technology Data Exchange (ETDEWEB)

    Robert Neibert; John Zabriskie; Collin Knight; James L. Jones

    2010-12-01

    The Zero Power Physics Reactor (ZPPR) facility is a Department of Energy facility located in the Idaho National Laboratory’s (INL) Materials and Fuels Complex. It contains various nuclear and non-nuclear materials that are available to support many radiation measurement assessments. User-selected, single material, nuclear and non-nuclear materials can be readily utilized with ZPPR clamshell containers with almost no criticality concerns. If custom, multi-material configurations are desired, the ZPPR clamshell or an approved aluminum Inspection Object (IO) Box container may be utilized, yet each specific material configuration will require a criticality assessment. As an example of the specialized material configurations possible, the National Nuclear Security Agency’s Office of Nuclear Verification (NNSA/NA 243) has sponsored the assembly of six material configurations. These are shown in the Appendixes and have been designated for semi-permanent storage that can be available to support various radiation measurement applications.

  4. Lessons learned for participation in recent OECD-NEA reactor physics and thermalhydraulic benchmarks

    Energy Technology Data Exchange (ETDEWEB)

    Novog, D.R.; Leung, K.H.; Ball, M. [McMaster Univ., Dept. of Engineering Physics, Hamilton, Ontario (Canada)

    2013-07-01

    Over the last 6 years the OECD-NEA has initiated a series of computational benchmarks in the fields of reactor physics and thermalhydraulics. Within this context McMaster university has been a key contributor and applied several state of the art tools including TSUNAMI, DRAGON, ASSERT, STAR-CCM+, RELAP and TRACE. Considering the tremendous amount of international participation in these benchmarks, there were many lessons of both technical and non-technical that should be shared. This paper presents a summary of the benchmarks, the results and contributions from McMaster, and the authors opinion on the overall conclusions gained from these extensive benchmarks. The benchmarks discussed in this paper include the Uncertainty Analysis in Modelling (UAM), the BWR fine mesh bundle test (BFBT), the PWR Subchannel Boiling Test (PSBT), the MATiS mixing experiment and the IAEA super critical water benchmarks on heat transfer and stability. (author)

  5. Evaluation method for core thermohydraulics during natural circulation in fast reactors. Numerical predictions of inter-wrapper flow

    Energy Technology Data Exchange (ETDEWEB)

    Kamide, Hideki; Kimura, Nobuyuki; Miyakoshi, Hiroyuki [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Nagasawa, Kazuyoshi [Nuclear Energy System Incorporation, Oarai Office, Oarai, Ibaraki (Japan)

    2002-08-01

    Decay heat removal using natural circulation is one of significant functions for a reactor. As the decay heat removal system, a direct reactor auxiliary cooling system has been selected in current designs of fast reactors. In this system, cold sodium is provided in an upper plenum of reactor vessel and it covers the reactor core outlet. The cold sodium can penetrate into the gap region between the subassemblies. This gap flow is referred as inter-wrapper flow (IWF). A numerical estimation method for such phenomena was developed, which modeled each subassembly as a rectangular duct with gap region and also the upper plenum. This numerical simulation method was verified by a sodium test and also a water test. We applied this method to the natural circulation in a 600 MWe class fast reactor. The temperature in the core strongly depended on IWF, flow redistribution in the core, and inter-subassembly heat transfer. (author)

  6. High-Order Homogenization Method in Diffusion Theory for Reactor Core Simulation and Design Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Farzad Rahnema

    2003-09-30

    Most modern nodal methods in use by the reactor vendors and utilities are based on the generalized equivalence theory (GET) that uses homogenized cross sections and flux discontinuity factors. These homogenized parameters, referred to as infinite medium parameters, are precomputed by performing single bundle fine-mesh calculations with zero current boundary conditions. It is known that for configurations in which the node-to-node leakage (e.g., surface current-to-flux ratio) is large the use of the infinite medium parameters could lead to large errors in the nodal solution. This would be the case for highly heterogeneous core configurations, typical of modern reactor core designs.

  7. Application of the Combined Reactors Method for Analysis of Steelmaking Process

    Science.gov (United States)

    Lekakh, Simon N.; Robertson, D. G. C.

    A new integrated CFD-combined reactors approach is proposed for the description of processes in metallurgical vessels. CFD simulations were used to obtain the melt flow pattern in the vessels (ladle, tundish, and continuous caster mold). From these simulations, the characteristic curves were derived: (i) the residence time distribution curves (RTD) for flow-through systems (at tundish exit or at dendrite coherency surface in the mold) and (ii) the mixing curves for closed systems (ladle). In the next step, the melt flow was represented in a "combined reactors" system consisting of a combination of unit reactors (Plug Flow, Mixer, and Recirculated Volume). An inverse simulation was used to define the volumes of the reactor units and the melt flow rates between them by fitting to the characteristic curves from both methods (CFD and combined reactors). The suggested approach is demonstrated for multiple designs of Ar-stirred ladles, tundish, and SEN. This methodology can be used to enhance traditional post-processing CFD analysis and also as a tool for on-line process control.

  8. Method for acquiring grain-shaped growth of a microorganism in a reactor

    NARCIS (Netherlands)

    Heijnen, J.J.; Van Loosdrecht, M.C.M.

    1998-01-01

    The invention relates to a method of acquiring granular growth of a microorganism in a reactor containing a liquid medium. Surprisingly, according to the invention, aerobic microorganisms also can be induced to granular growth by maintaining specific culture conditions. During a first step an

  9. Advanced analysis methods in particle physics

    Energy Technology Data Exchange (ETDEWEB)

    Bhat, Pushpalatha C.; /Fermilab

    2010-10-01

    Each generation of high energy physics experiments is grander in scale than the previous - more powerful, more complex and more demanding in terms of data handling and analysis. The spectacular performance of the Tevatron and the beginning of operations of the Large Hadron Collider, have placed us at the threshold of a new era in particle physics. The discovery of the Higgs boson or another agent of electroweak symmetry breaking and evidence of new physics may be just around the corner. The greatest challenge in these pursuits is to extract the extremely rare signals, if any, from huge backgrounds arising from known physics processes. The use of advanced analysis techniques is crucial in achieving this goal. In this review, I discuss the concepts of optimal analysis, some important advanced analysis methods and a few examples. The judicious use of these advanced methods should enable new discoveries and produce results with better precision, robustness and clarity.

  10. Interface of nanocatalysis and microfluidic reactors for green chemistry methods

    CSIR Research Space (South Africa)

    Makgwane, PR

    2013-10-01

    Full Text Available The development of green catalytic methods for chemical synthesis and energy generation based on nanocoated catalyst microfluidic systems is a growing area of innovative research. The interface between heterogeneous catalysis and microchannel...

  11. Reactor physics analysis for the design of nuclear fuel lattices with burnable poisons

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa-Paredes, G. [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico, D.F. (Mexico); Guzman, Juan R., E-mail: maestro_juan_rafael@hotmail.com [Departamento de Fisica y Matematicas, Instituto Politecnico Nacional, Adolfo Lopez Mateos, San Pedro Zacatenco, 07738 Mexico, D.F. (Mexico)

    2011-12-15

    Highlights: Black-Right-Pointing-Pointer A fuel rod optimization for the coupled bundle-core design in a BWR is developed. Black-Right-Pointing-Pointer An algorithm to minimize the rod power peaking factor is used. Black-Right-Pointing-Pointer The fissile content is divided in two factors. Black-Right-Pointing-Pointer A reactor physics analysis of these factors is performed. Black-Right-Pointing-Pointer The algorithm is applied to a typical BWR fuel lattice. - Abstract: The main goals in nuclear fuel lattice design are: (1) minimizing the rod power peaking factor (PPF) in order that the power level distribution is the most uniform; (2) obtaining a prescribed target value for the multiplication factor (k) at the end of the irradiation in order that the fuel lattice reaches the desired reactivity; and (3) obtaining a prescribed target value for the k at the beginning of the irradiation in order that the reactivity excess is neither a high value (to ease the maneuvering of the control systems) nor a low value (to avoid the penalization of the high cost of the burnable poison content). In this work a simple algorithm to design the burnable poison bearing nuclear fuel lattice is presented. This algorithm is based on a reactor physics analysis. The algorithm is focused on finding the radial distribution of the fuel rods having different fissile and burnable poison contents in order to obtain: (1) an adequate minimum PPF; (2) a prescribed target value of the k at the end of the irradiation; and (3) a prescribed target value of the k at the beginning of the irradiation. This algorithm is based on the factorization of the fissile and burnable poison contents of each fuel rod and on the application of the first-order perturbation theory. The performance of the algorithm is demonstrated with the design of a fuel lattice composed of uranium dioxide (UO{sub 2}) and gadolinium dioxide (Gd{sub 2}O{sub 3}) for boiling water reactors (BWR). This algorithm has been accomplished

  12. A new MC-based method to evaluate the fission fraction uncertainty at reactor neutrino experiment

    CERN Document Server

    Ma, X B; Chen, Y X

    2016-01-01

    Uncertainties of fission fraction is an important uncertainty source for the antineutrino flux prediction in a reactor antineutrino experiment. A new MC-based method of evaluating the covariance coefficients between isotopes was proposed. It was found that the covariance coefficients will varying with reactor burnup and which may change from positive to negative because of fissioning balance effect, for example, the covariance coefficient between $^{235}$U and $^{239}$Pu changes from 0.15 to -0.13. Using the equation between fission fraction and atomic density, the consistent of uncertainty of fission fraction and the covariance matrix were obtained. The antineutrino flux uncertainty is 0.55\\% which does not vary with reactor burnup, and the new value is about 8.3\\% smaller.

  13. Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

    Science.gov (United States)

    Cisneros, Anselmo Tomas, Jr.

    and PEBBED for a high temperature gas cooled pebble bed reactor. Three parametric studies were performed for exploring the design space of the PB-FHR---to select a fuel design for the PB-FHR] to select a core configuration; and to optimize the PB-FHR design. These parametric studies investigated trends in the dependence of important reactor performance parameters such as burnup, temperature reactivity feedback, radiation damage, etc on the reactor design variables and attempted to understand the underlying reactor physics responsible for these trends. A pebble fuel parametric study determined that pebble fuel should be designed with a carbon to heavy metal ratio (C/HM) less than 400 to maintain negative coolant temperature reactivity coefficients. Seed and thorium blanket-, seed and inert pebble reflector- and seed only core configurations were investigated for annular FHR PBRs---the C/HM of the blanket pebbles and discharge burnup of the thorium blanket pebbles were additional design variable for core configurations with thorium blankets. Either a thorium blanket or graphite pebble reflector is required to shield the outer graphite reflector enough to extend its service lifetime to 60 EFPY. The fuel fabrication costs and long cycle lengths of the thorium blanket fuel limit the potential economic advantages of using a thorium blanket. Therefore, the seed and pebble reflector core configuration was adopted as the baseline core configuration. Multi-objective optimization with respect to economics was performed for the PB-FHR accounting for safety and other physical design constraints derived from the high-level safety regulatory criteria. These physical constraints were applied along in a design tool, Nuclear Application Value Estimator, that evaluated a simplified cash flow economics model based on estimates of reactor performance parameters calculated using correlations based on the results of parametric design studies for a specific PB-FHR design and a set of

  14. Advanced Analysis Methods in High Energy Physics

    Energy Technology Data Exchange (ETDEWEB)

    Pushpalatha C. Bhat

    2001-10-03

    During the coming decade, high energy physics experiments at the Fermilab Tevatron and around the globe will use very sophisticated equipment to record unprecedented amounts of data in the hope of making major discoveries that may unravel some of Nature's deepest mysteries. The discovery of the Higgs boson and signals of new physics may be around the corner. The use of advanced analysis techniques will be crucial in achieving these goals. The author discusses some of the novel methods of analysis that could prove to be particularly valuable for finding evidence of any new physics, for improving precision measurements and for exploring parameter spaces of theoretical models.

  15. Physical acoustics v.8 principles and methods

    CERN Document Server

    Mason, Warren P

    1971-01-01

    Physical Acoustics: Principles and Methods, Volume VIII discusses a number of themes on physical acoustics that are divided into seven chapters. Chapter 1 describes the principles and applications of a tool for investigating phonons in dielectric crystals, the spin phonon spectrometer. The next chapter discusses the use of ultrasound in investigating Landau quantum oscillations in the presence of a magnetic field and their relation to the strain dependence of the Fermi surface of metals. The third chapter focuses on the ultrasonic measurements that are made by pulsing methods with velo

  16. Calculation system for physical analysis of boiling water reactors; Modelisation des phenomenes physiques specifiques aux reacteurs a eau bouillante, notamment le couplage neutronique-thermohydraulique

    Energy Technology Data Exchange (ETDEWEB)

    Bouveret, F

    2001-07-01

    Although Boiling Water Reactors generate a quarter of worldwide nuclear electricity, they have been only little studied in France. A certain interest now shows up for these reactors. So, the aim of the work presented here is to contribute to determine a core calculation methodology with CEA (Commissariat a l'Energie Atomique) codes. Vapour production in the reactor core involves great differences in technological options from pressurised water reactor. We analyse main physical phenomena for BWR and offer solutions taking them into account. BWR fuel assembly heterogeneity causes steep thermal flux gradients. The two dimensional collision probability method with exact boundary conditions makes possible to calculate accurately the flux in BWR fuel assemblies using the APOLLO-2 lattice code but induces a very long calculation time. So, we determine a new methodology based on a two-level flux calculation. Void fraction variations in assemblies involve big spectrum changes that we have to consider in core calculation. We suggest to use a void history parameter to generate cross-sections libraries for core calculation. The core calculation code has also to calculate the depletion of main isotopes concentrations. A core calculation associating neutronics and thermal-hydraulic codes lays stress on points we still have to study out. The most important of them is to take into account the control blade in the different calculation stages. (author)

  17. Geometric Methods in Physics : XXXIII Workshop

    CERN Document Server

    Bieliavsky, Pierre; Odzijewicz, Anatol; Schlichenmaier, Martin; Voronov, Theodore

    2015-01-01

    This book presents a selection of papers based on the XXXIII Białowieża Workshop on Geometric Methods in Physics, 2014. The Białowieża Workshops are among the most important meetings in the field and attract researchers from both mathematics and physics. The articles gathered here are mathematically rigorous and have important physical implications, addressing the application of geometry in classical and quantum physics. Despite their long tradition, the workshops remain at the cutting edge of ongoing research. For the last several years, each Białowieża Workshop has been followed by a School on Geometry and Physics, where advanced lectures for graduate students and young researchers are presented; some of the lectures are reproduced here. The unique atmosphere of the workshop and school is enhanced by its venue, framed by the natural beauty of the Białowieża forest in eastern Poland. The volume will be of interest to researchers and graduate students in mathematical physics, theoretical physics and m...

  18. An Idea of 20% test of the Initial Core Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Kyung Ho; Yang, Sung Tae; Jung, Ji Eun [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2012-05-15

    Many tests have been performed on the OPR1000 and APR1400 before commercial operation. Some of these tests were performed at reactor power levels of 20% and 50%. The CPC (Core Protection Calculator) power distribution test is one of these tests. It is performed to assure the reliability of the Core Protection Calculator System (CPCS). Through this test, SAM1 is calculated using the snapshots2. The test takes about nine hours at a reactor power level of 20% and about thirty hours at a reactor power level of 50%. SAM used at each reactor power level is as follows: 1. Reactor power of 0% {approx} 20%: Designed SAM (DSAM) 2. Reactor power of 20% {approx} 50%: SAM calculated (C-SAM) at a reactor power of 20% 3. Reactor power 50% {approx} End of Cycle : SAM calculated at a reactor power of 50% As mentioned earlier, SAM is calculated and punched into CPC to assure the reliability of CPCS. Therefore, CPC is operated having penalties with D-SAM until3 reaching a reactor power of 20%. That is, the penalty of CPC will be removed when SAM is calculated and punched into the CPC at a reactor power of 20%. But these penalties are considered to be removed after a reactor power of 50% test in order to maintain the conservatism of the CPC. This is done because the final values calculated using C-SAM, in contrast to those calculated using SAM, a reactor power of 50%, are not correct. This paper began from an idea, 'If so, what would happen if we removed the CPC power distribution test at a reactor power of 20%?'

  19. Oxygen transport membrane reactor based method and system for generating electric power

    Science.gov (United States)

    Kelly, Sean M.; Chakravarti, Shrikar; Li, Juan

    2017-02-07

    A carbon capture enabled system and method for generating electric power and/or fuel from methane containing sources using oxygen transport membranes by first converting the methane containing feed gas into a high pressure synthesis gas. Then, in one configuration the synthesis gas is combusted in oxy-combustion mode in oxygen transport membranes based boiler reactor operating at a pressure at least twice that of ambient pressure and the heat generated heats steam in thermally coupled steam generation tubes within the boiler reactor; the steam is expanded in steam turbine to generate power; and the carbon dioxide rich effluent leaving the boiler reactor is processed to isolate carbon. In another configuration the synthesis gas is further treated in a gas conditioning system configured for carbon capture in a pre-combustion mode using water gas shift reactors and acid gas removal units to produce hydrogen or hydrogen-rich fuel gas that fuels an integrated gas turbine and steam turbine system to generate power. The disclosed method and system can also be adapted to integrate with coal gasification systems to produce power from both coal and methane containing sources with greater than 90% carbon isolation.

  20. New modelling method for fast reactor neutronic behaviours analysis; Nouvelles methodes de modelisation neutronique des reacteurs rapides de quatrieme Generation

    Energy Technology Data Exchange (ETDEWEB)

    Jacquet, P.

    2011-05-23

    Due to safety rules running on fourth generation reactors' core development, neutronics simulation tools have to be as accurate as never before. First part of this report enumerates every step of fast reactor's neutronics simulation implemented in current reference code: ECCO. Considering the field of fast reactors that meet criteria of fourth generation, ability of models to describe self-shielding phenomenon, to simulate neutrons leakage in a lattice of fuel assemblies and to produce representative macroscopic sections is evaluated. The second part of this thesis is dedicated to the simulation of fast reactors' core with steel reflector. These require the development of advanced methods of condensation and homogenization. Several methods are proposed and compared on a typical case: the ZONA2B core of MASURCA reactor. (author) [French] Les criteres de surete qui regissent le developpement de coeurs de reacteurs de quatrieme generation implique l'usage d'outils de calcul neutronique performants. Une premiere partie de la these reprend toutes les etapes de modelisation neutronique des reacteurs rapides actuellement d'usage dans le code de reference ECCO. La capacite des modeles a decrire le phenomene d'autoprotection, a representer les fuites neutroniques au niveau d'un reseau d'assemblages combustibles et a generer des sections macroscopiques representatives est appreciee sur le domaine des reacteurs rapides innovants respectant les criteres de quatrieme generation. La deuxieme partie de ce memoire se consacre a la modelisation des coeurs rapides avec reflecteur acier. Ces derniers necessitent le developpement de methodes avancees de condensation et d'homogenisation. Plusieurs methodes sont proposees et confrontees sur un probleme de modelisation typique: le coeur ZONA2B du reacteur maquette MASURCA

  1. Physics and mathematical tools methods and examples

    CERN Document Server

    Alastuey, Angel; Magro, Marc; Pujol, Pierre

    2016-01-01

    This book presents mathematical methods and tools which are useful for physicists and engineers: response functions, Kramers-Kronig relations, Green's functions, saddle point approximation. The derivations emphasize the underlying physical arguments and interpretations without any loss of rigor. General introductions describe the main features of the methods, while connections and analogies between a priori different problems are discussed. They are completed by detailed applications in many topics including electromagnetism, hydrodynamics, statistical physics, quantum mechanics, etc. Exercises are also proposed, and their solutions are sketched. A self-contained reading of the book is favored by avoiding too technical derivations, and by providing a short presentation of important tools in the appendices. It is addressed to undergraduate and graduate students in physics, but it can also be used by teachers, researchers and engineers.

  2. Particle identification methods in High Energy Physics

    Energy Technology Data Exchange (ETDEWEB)

    Va' Vra, J.

    2000-01-27

    This paper deals with two major particle identification methods: dE/dx and Cherenkov detection. In the first method, the authors systematically compare existing dE/dx data with various predictions available in the literature, such as the Particle Data group recommendation, and judge the overall consistency. To my knowledge, such comparison was not done yet in a published form for the gaseous detectors used in High-Energy physics. As far as the second method, there are two major Cherenkov light detection techniques: the threshold and the Ring imaging methods. The authors discuss the recent trend in these techniques.

  3. Apparatus for suppressing formation of vortices in the coolant fluid of a nuclear reactor and associated method

    Science.gov (United States)

    Ekeroth, Douglas E.; Garner, Daniel C.; Hopkins, Ronald J.; Land, John T.

    1993-01-01

    An apparatus and method are provided for suppressing the formation of vortices in circulating coolant fluid of a nuclear reactor. A vortex-suppressing plate having a plurality of openings therein is suspended within the lower plenum of a reactor vessel below and generally parallel to the main core support of the reactor. The plate is positioned so as to intersect vortices which may form in the circulating reactor coolant fluid. The intersection of the plate with such vortices disrupts the rotational flow pattern of the vortices, thereby disrupting the formation thereof.

  4. Molten metal reactor and method of forming hydrogen, carbon monoxide and carbon dioxide using the molten alkaline metal reactor

    Science.gov (United States)

    Bingham, Dennis N.; Klingler, Kerry M.; Turner, Terry D.; Wilding, Bruce M.

    2012-11-13

    A molten metal reactor for converting a carbon material and steam into a gas comprising hydrogen, carbon monoxide, and carbon dioxide is disclosed. The reactor includes an interior crucible having a portion contained within an exterior crucible. The interior crucible includes an inlet and an outlet; the outlet leads to the exterior crucible and may comprise a diffuser. The exterior crucible may contain a molten alkaline metal compound. Contained between the exterior crucible and the interior crucible is at least one baffle.

  5. Landslides as weathering reactors; links between physical erosion and weathering in rapidly eroding mountain belts

    Science.gov (United States)

    Emberson, R.; Hovius, N.; Galy, A.

    2014-12-01

    The link between physical erosion and chemical weathering is generally modelled with a surface-blanketing weathering zone, where the supply of fresh minerals is tied to the average rate of denudation. In very fast eroding environments, however, sediment production is dominated by landsliding, which acts in a stochastic fashion across the landscape, contrasting strongly with more uniform denudation models. If physical erosion is a driver of weathering at the highest erosion rates, then an alternative weathering model is required. Here we show that landslides can be effective 'weathering reactors'. Previous work modelling the effect of landslides on chemical weathering (Gabet 2007) considered the fresh bedrock surfaces exposed in landslide scars. However, fracturing during the landslide motion generates fresh surfaces, the total surface area of which exceeds that of the exposed scar by many orders of magnitude. Moreover, landslides introduce concavity into hillslopes, which acts to catch precipitation. This is funnelled into a deposit of highly fragmented rock mass with large reactive surface area and limited hydraulic conductivity (Lo et al. 2007). This allows percolating water reaction time for chemical weathering; any admixture of macerated organic debris could yield organic acid to further accelerate weathering. In the South island of New Zealand, seepage from recent landslide deposits has systematically high solute concentrations, far outstripping concentration in runoff from locations where soils are present. River total dissolved load in the western Southern Alps is highly correlated with the rate of recent (<35yrs) landsliding, suggesting that landslides are the dominant locus of weathering in this rapidly eroding landscape. A tight link between landsliding and weathering implies that localized weathering migrates through the landscape with physical erosion; this contrasts with persistent and ubiquitous weathering associated with soil production. Solute

  6. PHYSICAL METHODS IN AGRO-FOOD CHAIN

    Directory of Open Access Journals (Sweden)

    ANNA ALADJADJIYAN

    2009-06-01

    Full Text Available Chemical additives (fertilizers and plant protection preparations are largely used for improving the production yield of food produce. Their application often causes the contamination of raw materials for food production, which can be dangerous for the health of consumers. Alternative methods are developed and implemented to improve and ensure the safety of on-farm production. The substitution of chemical fertilizers and soil additives with alternative treatment methods, such as irradiation, ultrasound and the use of electromagnetic energy are discussed. Successful application of physical methods in different stages of food-preparation is recommended.

  7. Evaluation of methods to assess physical activity

    Science.gov (United States)

    Leenders, Nicole Y. J. M.

    Epidemiological evidence has accumulated that demonstrates that the amount of physical activity-related energy expenditure during a week reduces the incidence of cardiovascular disease, diabetes, obesity, and all-cause mortality. To further understand the amount of daily physical activity and related energy expenditure that are necessary to maintain or improve the functional health status and quality of life, instruments that estimate total (TDEE) and physical activity-related energy expenditure (PAEE) under free-living conditions should be determined to be valid and reliable. Without evaluation of the various methods that estimate TDEE and PAEE with the doubly labeled water (DLW) method in females there will be eventual significant limitations on assessing the efficacy of physical activity interventions on health status in this population. A triaxial accelerometer (Tritrac-R3D, (TT)), an uniaxial (Computer Science and Applications Inc., (CSA)) activity monitor, a Yamax-Digiwalker-500sp°ler , (YX-stepcounter), by measuring heart rate responses (HR method) and a 7-d Physical Activity Recall questionnaire (7-d PAR) were compared with the "criterion method" of DLW during a 7-d period in female adults. The DLW-TDEE was underestimated on average 9, 11 and 15% using 7-d PAR, HR method and TT. The underestimation of DLW-PAEE by 7-d PAR was 21% compared to 47% and 67% for TT and YX-stepcounter. Approximately 56% of the variance in DLW-PAEE*kgsp{-1} is explained by the registration of body movement with accelerometry. A larger proportion of the variance in DLW-PAEE*kgsp{-1} was explained by jointly incorporating information from the vertical and horizontal movement measured with the CSA and Tritrac-R3D (rsp2 = 0.87). Although only a small amount of variance in DLW-PAEE*kgsp{-1} is explained by the number of steps taken per day, because of its low cost and ease of use, the Yamax-stepcounter is useful in studies promoting daily walking. Thus, studies involving the

  8. Evaluation method for core thermohydraulics during natural circulation in fast reactors numerical predictions of inter-wrapper flow

    Energy Technology Data Exchange (ETDEWEB)

    Kamide, H.; Kimura, N.; Miyakoshi, H. [Japan Nuclear Cycle Development Institute, Reactor Engineering Group, O-arai Engineering Center, Ibaraki (Japan); Nagasawa, K. [Nuclear Energy System Incorporation, O-arai Office, Ibaraki (Japan)

    2001-07-01

    Decay heat removal using natural circulation is one of the important functions for the safety of fast reactors. As a decay heat removal system, direct reactor auxiliary cooling system has been selected in current designs of fast reactors. In this design, dumped heat exchanger provides cold sodium and it covers the reactor core outlet. The cold sodium can penetrate into the gap region between the subassemblies. This gap flow is referred as inter-wrapper flow (IWF). A numerical estimation method for such natural circulation phenomena in a reactor core has been developed, which models each subassembly as a rectangular duct with gap region between the subassemblies and also the upper plenum in a reactor vessel. This numerical simulation method was verified based on experimental data of a sodium test using 7- subassembly core model and also a water test which simulates IWF using the 1/12 sector model of a reactor core. We applied the estimation method to the natural circulation in a 600 MW class fast reactor. The temperature in the core strongly depended on IWF, flow redistribution in the core, and inter-subassembly heat transfer. It is desired for prediction methods on the natural circulation to simulate these phenomena. (author)

  9. Simplified method for measuring the response time of scram release electromagnet in a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Patri, Sudheer, E-mail: patri@igcar.gov.in; Mohana, M.; Kameswari, K.; Kumar, S. Suresh; Narmadha, S.; Vijayshree, R.; Meikandamurthy, C.; Venkatesan, A.; Palanisami, K.; Murthy, D. Thirugnana; Babu, B.; Prakash, V.; Rajan, K.K.

    2015-04-15

    Highlights: • An alternative method for estimating the electromagnet clutch release time. • A systematic approach to develop a computer based measuring system. • Prototype tests on the measurement system. • Accuracy of the method is ±6% and repeatability error is within 2%. - Abstract: The delay time in electromagnet clutch release during a reactor trip (scram action) is an important safety parameter, having a bearing on the plant safety during various design basis events. Generally, it is measured using current decay characteristics of electromagnet coil and its energising circuit. A simplified method of measuring the same in a Sodium cooled fast reactors (SFR) is proposed in this paper. The method utilises the position data of control rod to estimate the delay time in electromagnet clutch release. A computer based real time measurement system for measuring the electromagnet clutch delay time is developed and qualified for retrofitting in prototype fast breeder reactor. Various stages involved in the development of the system are principle demonstration, experimental verification of hardware capabilities and prototype system testing. Tests on prototype system have demonstrated the satisfactory performance of the system with intended accuracy and repeatability.

  10. Mathematical methods in engineering and physics

    CERN Document Server

    Felder, Gary N

    2016-01-01

    This text is intended for the undergraduate course in math methods, with an audience of physics and engineering majors. As a required course in most departments, the text relies heavily on explained examples, real-world applications and student engagement. Supporting the use of active learning, a strong focus is placed upon physical motivation combined with a versatile coverage of topics that can be used as a reference after students complete the course. Each chapter begins with an overview that includes a list of prerequisite knowledge, a list of skills that will be covered in the chapter, and an outline of the sections. Next comes the motivating exercise, which steps the students through a real-world physical problem that requires the techniques taught in each chapter.

  11. V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

    Energy Technology Data Exchange (ETDEWEB)

    Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

    2010-07-15

    V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

  12. Molecular Physics: Theoretical Principles and Experimental Methods

    Science.gov (United States)

    Demtröder, Wolfgang

    2005-11-01

    The richly illustrated book comprehensively explains the important principles of diatomic and polyatomic molecules and their spectra in two separate, distinct parts. The first part concentrates on the theoretical aspects of molecular physics, such as the vibration, rotation, electronic states, potential curves, and spectra of molecules. The different methods of approximation for the calculation of electronic wave functions and their energy are also covered. The introduction of basics terms used in group theory and their meaning in molecular physics enables an elegant description of polyatomic molecules and their symmetries. Molecular spectra and the dynamic processes involved in their excited states are given its own chapter. The theoretical part then concludes with a discussion of the field of Van der Waals molecules and clusters. The second part is devoted entirely to experimental techniques, such as laser, Fourier, NMR, and ESR spectroscopies, used in the fields of physics, chemistry, biology, and material science. Time-resolved measurements and the influence of chemical reactions by coherent controls are also treated. A list of general textbooks and specialized literature is provided for further reading. With specific examples, definitions, and notes integrated within the text to aid understanding, this is suitable for undergraduates and graduates in physics and chemistry with a knowledge of atomic physics and familiar with the basics of quantum mechanics.

  13. Ion transport membrane reactor systems and methods for producing synthesis gas

    Science.gov (United States)

    Repasky, John Michael

    2015-05-12

    Embodiments of the present invention provide cost-effective systems and methods for producing a synthesis gas product using a steam reformer system and an ion transport membrane (ITM) reactor having multiple stages, without requiring inter-stage reactant injections. Embodiments of the present invention also provide techniques for compensating for membrane performance degradation and other changes in system operating conditions that negatively affect synthesis gas production.

  14. The use of active learning strategies in the instruction of Reactor Physics concepts

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, Michael A.

    2000-01-01

    Each of the Active Learning strategies employed to teach Reactor Physics material has been or promises to be instructionally successful. The Cooperative Group strategy has demonstrated a statistically significant increase in student performance on the unit exam in teaching conceptually difficult, transport and diffusion theory material. However, this result was achieved at the expense of a modest increase in class time. The Tutorial CBI programs have enabled learning equally as well as classroom lectures without the direct intervention of an instructor. Thus, the Tutorials have been successful as homework assignments, releasing classroom time for other instruction. However, the time required for development of these tools was large, on the order of two hundred hours per hour of instruction. The initial introduction of the Case-Based strategy was roughly as effective as the traditional classroom instruction. Case-Based learning could well, after important modifications, perform better than traditional instruction. A larger percentage of the students prefer active learning strategies than prefer traditional lecture presentations. Student preferences for the active strategies were particularly strong when they believed that the strategies helped them learn the material better than they would have by using a lecture format. In some cases, students also preferred the active strategies because they were different from traditional instruction, a change of pace. Some students preferred lectures to CBI instruction, primarily because the CBI did not afford them the opportunity to question the instructor during the presentation.

  15. Activity report of working party on reactor physics of accelerator-driven system. July 1999 to March 2001

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-02-01

    Under the Research Committee on Reactor Physics, the Working Party on Reactor Physics of Accelerator-Driven System (ADS-WP) was set in July 1999 to review and investigate special subjects related to reactor physics research for the Accelerator-Driven Subcritical System (ADS). The ADS-WP, at the first meeting, discussed a guideline of its activity for two years and decided to concentrate upon three subjects: (1) neutron transport calculations in high energy range, (2) static and kinetic (safety-related) characteristics of subcritical system, and (3) system design including ADS concepts and elemental technology developments required. The activity of ADS-WP continued from July 1999 to March 2001. In this duration, the members of ADS-WP met together four times and discussed the above subjects. In addition, the ADS-WP conducted a questionnaire on requests and proposals for the plan of Transmutation Physics Experimental Facility in the High-Intensity Proton Accelerator Project, which is a joint project between JAERI and KEK (High Energy Accelerator Research Organization). This report summarizes the results obtained by the above ADS-WP activity. (author)

  16. Applied methods for mitigation of damage by stress corrosion in BWR type reactors; Metodos aplicados para la mitigacion del dano por corrosion bajo esfuerzo en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez C, R.; Diaz S, A.; Gachuz M, M.; Arganis J, C. [Instituto Nacional de Investigaciones Nucleares, Gerencia de Ciencia de Materiales, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1998-07-01

    The Boiling Water nuclear Reactors (BWR) have presented stress corrosion problems, mainly in components and pipes of the primary system, provoking negative impacts in the performance of energy generator plants, as well as the increasing in the radiation exposure to personnel involucred. This problem has caused development of research programs, which are guided to find solution alternatives for the phenomena control. Among results of greater relevance the control for the reactor water chemistry stands out particularly in the impurities concentration and oxidation of radiolysis products; as well as the supervision in the materials selection and the stresses levels reduction. The present work presents the methods which can be applied to diminish the problems of stress corrosion in BWR reactors. (Author)

  17. Analysis of neutron flux distribution for the validation of computational methods for the optimization of research reactor utilization.

    Science.gov (United States)

    Snoj, L; Trkov, A; Jaćimović, R; Rogan, P; Zerovnik, G; Ravnik, M

    2011-01-01

    In order to verify and validate the computational methods for neutron flux calculation in TRIGA research reactor calculations, a series of experiments has been performed. The neutron activation method was used to verify the calculated neutron flux distribution in the TRIGA reactor. Aluminium (99.9 wt%)-Gold (0.1 wt%) foils (disks of 5mm diameter and 0.2mm thick) were irradiated in 33 locations; 6 in the core and 27 in the carrousel facility in the reflector. The experimental results were compared to the calculations performed with Monte Carlo code MCNP using detailed geometrical model of the reactor. The calculated and experimental normalized reaction rates in the core are in very good agreement for both isotopes indicating that the material and geometrical properties of the reactor core are modelled well. In conclusion one can state that our computational model describes very well the neutron flux and reaction rate distribution in the reactor core. In the reflector however, the accuracy of the epithermal and thermal neutron flux distribution and attenuation is lower, mainly due to lack of information about the material properties of the graphite reflector surrounding the core, but the differences between measurements and calculations are within 10%. Since our computational model properly describes the reactor core it can be used for calculations of reactor core parameters and for optimization of research reactor utilization. Copyright © 2010 Elsevier Ltd. All rights reserved.

  18. Molten salt rolling bubble column, reactors utilizing same and related methods

    Science.gov (United States)

    Turner, Terry D.; Benefiel, Bradley C.; Bingham, Dennis N.; Klinger, Kerry M.; Wilding, Bruce M.

    2015-11-17

    Reactors for carrying out a chemical reaction, as well as related components, systems and methods are provided. In accordance with one embodiment, a reactor is provided that includes a furnace and a crucible positioned for heating by the furnace. The crucible may contain a molten salt bath. A downtube is disposed at least partially within the interior crucible along an axis. The downtube includes a conduit having a first end in communication with a carbon source and an outlet at a second end of the conduit for introducing the carbon material into the crucible. At least one opening is formed in the conduit between the first end and the second end to enable circulation of reaction components contained within the crucible through the conduit. An oxidizing material may be introduced through a bottom portion of the crucible in the form of gas bubbles to react with the other materials.

  19. Application of preconditioned conjugate gradient-like methods to reactor kinetics

    Energy Technology Data Exchange (ETDEWEB)

    Yang, D.Y.; Chen, G.S.; Chou, H.P. (National Tsing Hua Univ., Hsinchu, Taiwan (China). Dept. of Nuclear Engineering)

    1993-01-01

    Several conjugate gradient-like (CG-like) methods are applied to solve the nonsymmetric linear systems of equations derived from the time-dependent two-dimensional two-energy-group neutron diffusion equations by a finite difference approximation. The methods are: the generalized conjugate residual method; the generalized conjugate gradient least-square method; the generalized minimal residual method (GMRES); the conjugate gradient square method; and a variant of bi-conjugate gradient method (Bi-CGSTAB). In order to accelerate these methods, six preconditioning techniques are investigated. Two are based on pointwise incomplete factorization: the incomplete LU (ILU) and the modified incomplete LU (MILU) decompositions; two, based on the block tridiagonal structure of the coefficient matrix, are blockwise and modified blockwise incomplete factorizations, BILU and MBILU; two are the alternating-direction implicit and symmetric successive overrelaxation (SSOR) preconditioners, derived from the basic iterative schemes. Comparisons are made by using CG-like methods combined with different preconditioners to solve a sequence of time-step reactor transient problems. Numerical tests indicate that preconditioned BI-CGSTAB with the preconditioner MBILU requires less CPU time and fewer iterations than other methods. The preconditioned CG-like methods are less sensitive to the time-step size used than the Chebyshev semi-iteration method and line SOR method. The indication is that the CGS, Bi-CGSTAB and GMRES methods are, on average, better than the other methods in reactor kinetics computation and that a good preconditioner is more important than the choice of CG-like methods. The MILU decomposition based on the conventional row-sum criterion has difficulty yielding a convergent solution and an improved version is introduced. (author).

  20. Methods and performance of the three-dimensional pressurized water reactor core dynamics SIMTRAN on-line code

    Energy Technology Data Exchange (ETDEWEB)

    Aragones, J.M.; Ahnert, C.; Cabellos, O. [Polytechnic Univ., Madrid (Spain)

    1996-09-01

    New reactor physics and computation methods have been developed in the three-dimensional pressurized water reactor (PWR) core dynamics SIMTRAN code for on-line surveillance and prediction. The accuracy of the coupled neutronic thermal-hydraulic solution is improved, and its scope is extended to provide, mainly, the calculation of the fission reaction rates at the in-core mini detectors, the responses at the ex-core detectors, and the in-vessel coolant flow and temperature distributions. The functional capabilities implemented in the on-line SIMTRAN code include on-line surveillance, in-core-ex-core calibration, evaluation of peak power factors and thermal margins, nominal cycle follow, prediction of maneuvers, and diagnosis of fast transients and oscillations. The new code has been operating on-line at the Vandellos-II PWR unit in Spain since the startup of its cycle 7 in mid-June 1994, including the machine-man interfaces for on-line acquisition of measured data and interactive graphical utilization. The agreement of the simulations with the measurements, along the full cycle 7 and the first months of cycle 8 operation, is well within the accuracy requirements. The performance and usefulness for operational support shown during the demo and routine use phases have proved that the on-line SIMTRAN code has the qualities for the accurate, reliable, comprehensive, and user-friendly on-line core surveillance and prediction.

  1. Physics principles to achieve comparable fission power from fertile and fissile rods of the conceptual ATBR/FTBR reactors

    Energy Technology Data Exchange (ETDEWEB)

    Pal, Usha [Light Water Reactors Physics Section, Reactor Physics Design Division, Bhabha Atomic Research Centre, Mumbai - 400 085 (India)], E-mail: ushapal@barc.gov.in; Jagannathan, V. [Light Water Reactors Physics Section, Reactor Physics Design Division, Bhabha Atomic Research Centre, Mumbai - 400 085 (India)], E-mail: vjagan@barc.gov.in

    2008-09-15

    Loading of seedless fertile rods has been used as the central principle to maximize fertile to fissile conversion in the two thorium breeder reactor concepts, viz. ATBR and FTBR [Jagannathan, V., Pal, Usha, Karthikeyan, R., Ganesan, S., Jain, R.P., Kamat, S.U., 2001. ATBR - a thorium breeder reactor concept for an early induction of thorium in an enriched uranium reactor. Nuclear Technology 133, 1-32; Jagannathan, V., Pal, Usha, Karthikeyan, R., Raj Devesh, Srivastava, Argala, Ahmad Khan, Suhail, 2007. Reactor physics ideas to design novel reactors with faster fissile growth. In: Paper accepted for oral presentation in 'ICENES 2007 - 13th International Conference on Emerging Nuclear Energy Systems, 3--8 June 2007, Istanbul, Turkey]. At fresh state the seedless thoria rods will produce practically no fission power, or nearly thousand times less fission rate compared to the seed fuel rods. Hence it is conceived that the fuel assembly would be constituted by assembling the fresh seed rods with one fuel cycle irradiated fertile thoria rods. Even in this state there is a wide disparity between the fissile content of these rods. By judicious choice of the rod dimensions and their relative locations, a degree of balance in the fission rate is achieved in the fresh state of seeded rods. Remarkably as the burnup proceeds the initially seedless fertile rods have a continuous growth of fissile content up to an asymptotic value for a given spectrum and the fissile content in seeded rods monotonically decreases. If the discharge burnup is sufficiently large by design, it is seen that the power share of the initially seedless fertile rods can even exceed that of the seed fuel rods. The physics principles of achieving this characteristic are presented in this paper.

  2. Fundamental Thermal Fluid Physics of High Temperature Flows in Advanced Reactor Systems - Nuclear Energy Research Initiative Program Interoffice Work Order (IWO) MSF99-0254 Final Report for Period 1 August 1999 to 31 December 2002

    Energy Technology Data Exchange (ETDEWEB)

    McEligot, D.M.; Condie, K.G.; Foust, T.D.; McCreery, G.E.; Pink, R.J.; Stacey, D.E. (INEEL); Shenoy, A.; Baccaglini, G. (General Atomics); Pletcher, R.H. (Iowa State U.); Wallace, J.M.; Vukoslavcevic, P. (U. Maryland); Jackson, J.D. (U. Manchester, UK); Kunugi, T. (Kyoto U., Japan); Satake, S.-i. (Tokyo U. Science, Japan)

    2002-12-31

    The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of advanced reactors for higher efficiency and enhanced safety and for deployable reactors for electrical power generation, process heat utilization and hydrogen generation. While key applications would be advanced gas-cooled reactors (AGCRs) using the closed Brayton cycle (CBC) for higher efficiency (such as the proposed Gas Turbine - Modular Helium Reactor (GT-MHR) of General Atomics [Neylan and Simon, 1996]), results of the proposed research should also be valuable in reactor systems with supercritical flow or superheated vapors, e.g., steam. Higher efficiency leads to lower cost/kwh and reduces life-cycle impacts of radioactive waste (by reducing waters/kwh). The outcome will also be useful for some space power and propulsion concepts and for some fusion reactor concepts as side benefits, but they are not the thrusts of the investigation. The objective of the project is to provide fundamental thermal fluid physics knowledge and measurements necessary for the development of the improved methods for the applications.

  3. Optimal Protection of Reactor Hall Under Nuclear Fuel Container Drop Using Simulation Methods

    Directory of Open Access Journals (Sweden)

    Králik Juraj

    2014-12-01

    Full Text Available This paper presents of the optimal design of the damping devices cover of reactor hall under impact of nuclear fuel container drop of type TK C30. The finite element idealization of nuclear power plant structure is used in software ANSYS. The steel pipe damper system is proposed for dissipation of the kinetic energy of the container free fall in comparison with the experimental results. The probabilistic and sensitivity analysis of the damping devices was considered on the base of the simulation methods in program AntHill using the Monte Carlo method.

  4. Neutron transport in hexagonal reactor cores modeled by trigonal-geometry diffusion and simplified P{sub 3} nodal methods

    Energy Technology Data Exchange (ETDEWEB)

    Duerigen, Susan

    2013-05-15

    The superior advantage of a nodal method for reactor cores with hexagonal fuel assemblies discretized as cells consisting of equilateral triangles is its mesh refinement capability. In this thesis, a diffusion and a simplified P{sub 3} (or SP{sub 3}) neutron transport nodal method are developed based on trigonal geometry. Both models are implemented in the reactor dynamics code DYN3D. As yet, no other well-established nodal core analysis code comprises an SP{sub 3} transport theory model based on trigonal meshes. The development of two methods based on different neutron transport approximations but using identical underlying spatial trigonal discretization allows a profound comparative analysis of both methods with regard to their mathematical derivations, nodal expansion approaches, solution procedures, and their physical performance. The developed nodal approaches can be regarded as a hybrid NEM/AFEN form. They are based on the transverse-integration procedure, which renders them computationally efficient, and they use a combination of polynomial and exponential functions to represent the neutron flux moments of the SP{sub 3} and diffusion equations, which guarantees high accuracy. The SP{sub 3} equations are derived in within-group form thus being of diffusion type. On this basis, the conventional diffusion solver structure can be retained also for the solution of the SP{sub 3} transport problem. The verification analysis provides proof of the methodological reliability of both trigonal DYN3D models. By means of diverse hexagonal academic benchmark and realistic detailed-geometry full-transport-theory problems, the superiority of the SP{sub 3} transport over the diffusion model is demonstrated in cases with pronounced anisotropy effects, which is, e.g., highly relevant to the modeling of fuel assemblies comprising absorber material.

  5. Heat Transfer Analysis of Methane Hydrate Sediment Dissociation in a Closed Reactor by a Thermal Method

    Directory of Open Access Journals (Sweden)

    Mingjun Yang

    2012-05-01

    Full Text Available The heat transfer analysis of hydrate-bearing sediment involved phase changes is one of the key requirements of gas hydrate exploitation techniques. In this paper, experiments were conducted to examine the heat transfer performance during hydrate formation and dissociation by a thermal method using a 5L volume reactor. This study simulated porous media by using glass beads of uniform size. Sixteen platinum resistance thermometers were placed in different position in the reactor to monitor the temperature differences of the hydrate in porous media. The influence of production temperature on the production time was also investigated. Experimental results show that there is a delay when hydrate decomposed in the radial direction and there are three stages in the dissociation period which is influenced by the rate of hydrate dissociation and the heat flow of the reactor. A significant temperature difference along the radial direction of the reactor was obtained when the hydrate dissociates and this phenomenon could be enhanced by raising the production temperature. In addition, hydrate dissociates homogeneously and the temperature difference is much smaller than the other conditions when the production temperature is around the 10 °C. With the increase of the production temperature, the maximum of ΔToi grows until the temperature reaches 40 °C. The period of ΔToi have a close relation with the total time of hydrate dissociation. Especially, the period of ΔToi with production temperature of 10 °C is twice as much as that at other temperatures. Under these experimental conditions, the heat is mainly transferred by conduction from the dissociated zone to the dissociating zone and the production temperature has little effect on the convection of the water in the porous media.

  6. Supplemental Reactor Physics Calculations and Analysis of ELF Mk 1A Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Pope, Michael A. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-10-01

    These calculations supplement previous the reactor physics work evaluating the Enhanced Low Enriched Uranium (LEU) Fuel (ELF) Mk 1A element. This includes various additional comparisons between the current Highly Enriched Uranium (HEU) and LEU along with further characterization of the performance of the ELF fuel. The excess reactivity to be held down at BOC for ELF Mk 1A fuel is estimated to be approximately $2.75 greater than with HEU for a typical cycle. This is a combined effect of the absence of burnable poison in the ELF fuel and the reduced neck shim worth in LEU fuel compared to HEU. Burnable poison rods were conceptualized for use in the small B positions containing Gd2O3 absorber. These were shown to provide $2.37 of negative reactivity at BOC and to burn out in less than half of a cycle. The worth of OSCCs is approximately the same between HEU and ELF Mk 1A (LEU) fuels in the representative loading evaluated. This was evaluated by rotating all banks simultaneously. The safety rod worth is relatively unchanged between HEU and ELF Mk 1A (LEU) fuels in the representative loading evaluated. However, this should be reevaluated with different loadings. Neutron flux, both total and fast (>1 MeV), is either the same or reduced upon changing from HEU to ELF Mk 1A (LEU) fuels in the representative loading evaluated. This is consistent with the well-established trend of lower neutron fluxes for a given power in LEU than HEU.The IPT loop void reactivity is approximately the same or less positive with ELF Mk 1A (LEU) fuel than HEU in the representative loading evaluated.

  7. KIT multi-physics tools for the analysis of design and beyond design basis accidents of light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, Victor Hugo; Miassoedov, Alexei; Steinbrueck, M.; Tromm, W. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany)

    2016-05-15

    This paper describes the KIT numerical simulation tools under extension and validation for the analysis of design and beyond design basis accidents (DBA) of Light Water Reactors (LWR). The description of the complex thermal hydraulic, neutron kinetics and chemo-physical phenomena going on during off-normal conditions requires the development of multi-physics and multi-scale simulations tools which are fostered by the rapid increase in computer power nowadays. The KIT numerical tools for DBA and beyond DBA are validated using experimental data of KIT or from abroad. The developments, extensions, coupling approaches and validation work performed at KIT are shortly outlined and discussed in this paper.

  8. M and c'99 : Mathematics and computation, reactor physics and environmental analysis in nuclear applications, Madrid, September 27-30, 1999

    Energy Technology Data Exchange (ETDEWEB)

    Aragones, J. M.; Ahnert, C.; Cabellos, O.

    1999-07-01

    The international conference on mathematics and computation, reactor physics and environmental analysis in nuclear applications in the biennial topical meeting of the mathematics and computation division of the American Nuclear Society. (Author)

  9. Experimental research in neutron physic and thermal-hydraulic at the CDTN Triga reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z.; Souza, Rose Mary G.P.; Ferreira, Andrea V.; Pinto, Antonio J.; Costa, Antonio C.L.; Rezende, Hugo C., E-mail: amir@cdtn.b, E-mail: souzarm@cdtn.b, E-mail: avf@cdtn.b, E-mail: ajp@cdtn.b, E-mail: aclc@cdtn.b, E-mail: hcr@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The IPR-R1 TRIGA (Training, Research, Isotopes production, General Atomics) at Nuclear Technology Development Center (CDTN) is a pool type reactor cooled by natural circulation of light water and an open surface. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world and characterized by inherent safety. The IPR-R1 is the only Brazilian nuclear research reactor available and able to perform experiments in which interaction between neutronic and thermal-hydraulic areas occurs. The IPR-R1 has started up on November 11th, 1960. At that time the maximum thermal power was 30 kW. The present forced cooling system was built in the 70th and the power was upgraded to 100 kW. Recently the core configuration and instrumentation was upgraded again to 250 kW at steady state, and is awaiting the license of CNEN to operate definitely at this new power. This paper describes the experimental research project carried out in the IPR-R1 reactor that has as objective evaluate the behaviour of the reactor operational parameters, and mainly to investigate the influence of temperature on the neutronic variables. The research was supported by Research Support Foundation of the State of Minas Gerais (FAPEMIG) and Brazilian Council for Scientific and Technological Development (CNPq). The research project meets the recommendations of the IAEA, for safety, modernization and development of strategic plan for research reactors utilization. This work is in line with the strategic objectives of Brazil, which aims to design and construct the Brazilian Multipurpose research Reactor (RMB). (author)

  10. Report on Reactor Physics Assessment of Candidate Accident Tolerant Fuel Cladding Materials in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); George, Nathan [Univ. of Tennessee, Knoxville, TN (United States); Maldonado, G. Ivan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-08-28

    This work focuses on ATF concepts being researched at Oak Ridge National Laboratory (ORNL), expanding on previous studies of using alternate cladding materials in pressurized water reactors (PWRs). The neutronic performance of two leading alternate cladding materials were assessed in boiling water reactors (BWRs): iron-chromium-aluminum (FeCrAl) cladding, and silicon carbide (SiC)-based composite cladding. This report fulfills ORNL Milestone M3FT-15OR0202332 within the fiscal year 2015 (FY15)

  11. Development of Subspace-based Hybrid Monte Carlo-Deterministric Algorithms for Reactor Physics Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Abdel-Khalik, Hany S. [North Carolina State Univ., Raleigh, NC (United States); Zhang, Qiong [North Carolina State Univ., Raleigh, NC (United States)

    2014-05-20

    The development of hybrid Monte-Carlo-Deterministic (MC-DT) approaches, taking place over the past few decades, have primarily focused on shielding and detection applications where the analysis requires a small number of responses, i.e. at the detector locations(s). This work further develops a recently introduced global variance reduction approach, denoted by the SUBSPACE approach is designed to allow the use of MC simulation, currently limited to benchmarking calculations, for routine engineering calculations. By way of demonstration, the SUBSPACE approach is applied to assembly level calculations used to generate the few-group homogenized cross-sections. These models are typically expensive and need to be executed in the order of 103 - 105 times to properly characterize the few-group cross-sections for downstream core-wide calculations. Applicability to k-eigenvalue core-wide models is also demonstrated in this work. Given the favorable results obtained in this work, we believe the applicability of the MC method for reactor analysis calculations could be realized in the near future.

  12. Very High-Temperature Reactor (VHTR) Proliferation Resistance and Physical Protection (PR&PP)

    Energy Technology Data Exchange (ETDEWEB)

    Moses, David Lewis [ORNL

    2011-10-01

    This report documents the detailed background information that has been compiled to support the preparation of a much shorter white paper on the design features and fuel cycles of Very High-Temperature Reactors (VHTRs), including the proposed Next-Generation Nuclear Plant (NGNP), to identify the important proliferation resistance and physical protection (PR&PP) aspects of the proposed concepts. The shorter white paper derived from the information in this report was prepared for the Department of Energy Office of Nuclear Science and Technology for the Generation IV International Forum (GIF) VHTR Systems Steering Committee (SSC) as input to the GIF Proliferation Resistance and Physical Protection Working Group (PR&PPWG) (http://www.gen-4.org/Technology/horizontal/proliferation.htm). The short white paper was edited by the GIF VHTR SCC to address their concerns and thus may differ from the information presented in this supporting report. The GIF PR&PPWG will use the derived white paper based on this report along with other white papers on the six alternative Generation IV design concepts (http://www.gen-4.org/Technology/systems/index.htm) to employ an evaluation methodology that can be applied and will evolve from the earliest stages of design. This methodology will guide system designers, program policy makers, and external stakeholders in evaluating the response of each system, to determine each system's resistance to proliferation threats and robustness against sabotage and terrorism threats, and thereby guide future international cooperation on ensuring safeguards in the deployment of the Generation IV systems. The format and content of this report is that specified in a template prepared by the GIF PR&PPWG. Other than the level of detail, the key exception to the specified template format is the addition of Appendix C to document the history and status of coated-particle fuel reprocessing technologies, which fuel reprocessing technologies have yet to be

  13. Mathematical methods of studying physical phenomena

    Science.gov (United States)

    Man'ko, Margarita A.

    2013-03-01

    In recent decades, substantial theoretical and experimental progress was achieved in understanding the quantum nature of physical phenomena that serves as the foundation of present and future quantum technologies. Quantum correlations like the entanglement of the states of composite systems, the phenomenon of quantum discord, which captures other aspects of quantum correlations, quantum contextuality and, connected with these phenomena, uncertainty relations for conjugate variables and entropies, like Shannon and Rényi entropies, and the inequalities for spin states, like Bell inequalities, reflect the recently understood quantum properties of micro and macro systems. The mathematical methods needed to describe all quantum phenomena mentioned above were also the subject of intense studies in the end of the last, and beginning of the new, century. In this section of CAMOP 'Mathematical Methods of Studying Physical Phenomena' new results and new trends in the rapidly developing domain of quantum (and classical) physics are presented. Among the particular topics under discussion there are some reviews on the problems of dynamical invariants and their relations with symmetries of the physical systems. In fact, this is a very old problem of both classical and quantum systems, e.g. the systems of parametric oscillators with time-dependent parameters, like Ermakov systems, which have specific constants of motion depending linearly or quadratically on the oscillator positions and momenta. Such dynamical invariants play an important role in studying the dynamical Casimir effect, the essence of the effect being the creation of photons from the vacuum in a cavity with moving boundaries due to the presence of purely quantum fluctuations of the electromagnetic field in the vacuum. It is remarkable that this effect was recently observed experimentally. The other new direction in developing the mathematical approach in physics is quantum tomography that provides a new vision of

  14. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

  15. Reactor based plutonium disposition - physics and fuel behaviour benchmark studies of an OECD/NEA experts group

    Energy Technology Data Exchange (ETDEWEB)

    D' Hondt, P. [SCK.CEN, Mol (Belgium); Gehin, J. [ORNL, Oak Ridge, TN (United States); Na, B.C.; Sartori, E. [Organisation for Economic Co-Operation and Development, Nuclear Energy Agency, 92 - Issy les Moulineaux (France); Wiesenack, W. [Organisation for Economic Co-Operation and Development/HRP, Halden (Norway)

    2001-07-01

    One of the options envisaged for disposing of weapons grade plutonium, declared surplus for national defence in the Russian Federation and Usa, is to burn it in nuclear power reactors. The scientific/technical know-how accumulated in the use of MOX as a fuel for electricity generation is of great relevance for the plutonium disposition programmes. An Expert Group of the OECD/Nea is carrying out a series of benchmarks with the aim of facilitating the use of this know-how for meeting this objective. This paper describes the background that led to establishing the Expert Group, and the present status of results from these benchmarks. The benchmark studies cover a theoretical reactor physics benchmark on a VVER-1000 core loaded with MOX, two experimental benchmarks on MOX lattices and a benchmark concerned with MOX fuel behaviour for both solid and hollow pellets. First conclusions are outlined as well as future work. (author)

  16. Aspects of the physics and chemistry of water radiolysis by fast neutrons and fast electrons in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    McCracken, D.R. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Tsang, K.T. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Laughton, P.J

    1998-09-01

    Detailed radiation physics calculations of energy deposition have been done for the coolant of CANDU reactors and Pressurized Water Reactors (PWRs). The geometry of the CANDU fuel channel was modelled in detail. Fluxes and energy-deposition rates for neutrons, recoil ions, photons, and fast electrons have been calculated using MCNP4B, WIMS-AECL, and specifically derived energy-transfer factors. These factors generate the energy/flux spectra of recoil ions from fast-neutron energy/flux spectra. The energy spectrum was divided into 89 discrete ranges (energy bins).The production of oxidizing species and net coolant radiolysis can be suppressed by the addition of hydrogen to the coolant of nuclear reactors. It is argued that the net dissociation of coolant by gamma rays is suppressed by lower levels of excess hydrogen than when dissociation is by ion recoils. This has consequences for the modelling of coolant radiolysis by homogeneous kinetics. More added hydrogen is required to stop water radiolysis by recoil ions acting alone than if recoil ions and gamma rays acted concurrently in space and time. Homogeneous kinetic models and experimental data suggest that track overlap is very inefficient in providing radicals from gamma-ray tracks to recombine molecular products in ion-recoil tracks. An inhomogeneous chemical model is needed that incorporates ionizing-particle track structure and track overlap. Such a model does not yet exist, but a number of limiting cases using homogeneous kinetics are discussed. There are sufficient uncertainties and contradictions in the data relevant to the radiolysis of reactor coolant that the relatively high CHC's (critical hydrogen concentration) observed in NRU reactor experiments (compared to model predictions) may be explainable by errors in fundamental data and understanding of water radiolysis under reactor conditions. The radiation chemistry program at CRL has been focused to generate quantitative water-radiolysis data in a

  17. Nuclear reactor transient analysis via a quasi-static kinetics Monte Carlo method

    Energy Technology Data Exchange (ETDEWEB)

    Jo, YuGwon; Cho, Bumhee; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr [Korea Advanced Institute of Science and Technology 291 Daehak-ro, Yuseong-gu, Daejeon, Korea 305-701 (Korea, Republic of)

    2015-12-31

    The predictor-corrector quasi-static (PCQS) method is applied to the Monte Carlo (MC) calculation for reactor transient analysis. To solve the transient fixed-source problem of the PCQS method, fission source iteration is used and a linear approximation of fission source distributions during a macro-time step is introduced to provide delayed neutron source. The conventional particle-tracking procedure is modified to solve the transient fixed-source problem via MC calculation. The PCQS method with MC calculation is compared with the direct time-dependent method of characteristics (MOC) on a TWIGL two-group problem for verification of the computer code. Then, the results on a continuous-energy problem are presented.

  18. A computationally efficient method for full-core conjugate heat transfer modeling of sodium fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Rui, E-mail: rhu@anl.gov; Yu, Yiqi

    2016-11-15

    Highlights: • Developed a computationally efficient method for full-core conjugate heat transfer modeling of sodium fast reactors. • Applied fully-coupled JFNK solution scheme to avoid the operator-splitting errors. • The accuracy and efficiency of the method is confirmed with a 7-assembly test problem. • The effects of different spatial discretization schemes are investigated and compared to the RANS-based CFD simulations. - Abstract: For efficient and accurate temperature predictions of sodium fast reactor structures, a 3-D full-core conjugate heat transfer modeling capability is developed for an advanced system analysis tool, SAM. The hexagon lattice core is modeled with 1-D parallel channels representing the subassembly flow, and 2-D duct walls and inter-assembly gaps. The six sides of the hexagon duct wall and near-wall coolant region are modeled separately to account for different temperatures and heat transfer between coolant flow and each side of the duct wall. The Jacobian Free Newton Krylov (JFNK) solution method is applied to solve the fluid and solid field simultaneously in a fully coupled fashion. The 3-D full-core conjugate heat transfer modeling capability in SAM has been demonstrated by a verification test problem with 7 fuel assemblies in a hexagon lattice layout. Additionally, the SAM simulation results are compared with RANS-based CFD simulations. Very good agreements have been achieved between the results of the two approaches.

  19. Optimization of an auto-thermal ammonia synthesis reactor using cyclic coordinate method

    Science.gov (United States)

    A-N Nguyen, T.; Nguyen, T.-A.; Vu, T.-D.; Nguyen, K.-T.; K-T Dao, T.; P-H Huynh, K.

    2017-06-01

    The ammonia synthesis system is an important chemical process used in the manufacture of fertilizers, chemicals, explosives, fibers, plastics, refrigeration. In the literature, many works approaching the modeling, simulation and optimization of an auto-thermal ammonia synthesis reactor can be found. However, they just focus on the optimization of the reactor length while keeping the others parameters constant. In this study, the other parameters are also considered in the optimization problem such as the temperature of feed gas enters the catalyst zone, the initial nitrogen proportion. The optimal problem requires the maximization of an objective function which is multivariable function and subject to a number of equality constraints involving the solution of coupled differential equations and also inequality constraint. The cyclic coordinate search was applied to solve the multivariable-optimization problem. In each coordinate, the golden section method was applied to find the maximum value. The inequality constraints were treated using penalty method. The coupled differential equations system was solved using Runge-Kutta 4th order method. The results obtained from this study are also compared to the results from the literature.

  20. Neutron fluxes in test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Youinou, Gilles Jean-Michel [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-01-01

    Communicate the fact that high-power water-cooled test reactors such as the Advanced Test Reactor (ATR), the High Flux Isotope Reactor (HFIR) or the Jules Horowitz Reactor (JHR) cannot provide fast flux levels as high as sodium-cooled fast test reactors. The memo first presents some basics physics considerations about neutron fluxes in test reactors and then uses ATR, HFIR and JHR as an illustration of the performance of modern high-power water-cooled test reactors.

  1. Improved method in distance teaching of physics

    Science.gov (United States)

    Gustafsson, Peter

    2004-03-01

    Results of introducing cooperative working methods on a distance learning course in physics are reported. This has increased the throughput of students in the course as measured in the number of ECTS points generated by the students. There is no significant indication that students more experienced in academic studies manage to complete the course more often than those with less experience. In student groups where the cooperative concept was fully realized a larger gain of knowledge was achieved, as measured by the force concept inventory test. Hence, it is important for the tutor to monitor activities in the groups by follow-up questions during the course and to stress the importance of all students participating actively.

  2. Solar energy utilization by physical methods.

    Science.gov (United States)

    Wolf, M

    1974-04-19

    On the basis of the estimated contributions of these differing methods of the utilization of solar energy, their total energy delivery impact on the projected U.S. energy economy (9) can be evaluated (Fig. 5). Despite this late energy impact, the actual sales of solar energy utilization equipment will be significant at an early date. Potential sales in photovoltaic arrays alone could exceed $400 million by 1980, in order to meet the projected capacity buildup (10). Ultimately, the total energy utilization equipment industry should attain an annual sales volume of several tens of billion dollars in the United States, comparable to that of several other energy related industries. Varying amounts of technology development are required to assure the technical and economic feasibility of the different solar energy utilization methods. Several of these developments are far enough along that the paths can be analyzed from the present time to the time of demonstration of technical and economic feasibility, and from there to production and marketing readiness. After that point, a period of market introduction will follow, which will differ in duration according to the type of market addressed. It may be noted that the present rush to find relief from the current energy problem, or to be an early leader in entering a new market, can entail shortcuts in sound engineering practice, particularly in the areas of design for durability and easy maintenance, or of proper application engineering. The result can be loss of customer acceptance, as has been experienced in the past with various products, including solar water heaters. Since this could cause considerable delay in achieving the expected total energy impact, it will be important to spend adequate time at this stage for thorough development. Two other aspects are worth mentioning. The first is concerned with the economic impacts. Upon reflection on this point, one will observe that largescale solar energy utilization will

  3. Advanced Instrumentation and Control Methods for Small and Medium Reactors with IRIS Demonstration

    Energy Technology Data Exchange (ETDEWEB)

    J. Wesley Hines; Belle R. Upadhyaya; J. Michael Doster; Robert M. Edwards; Kenneth D. Lewis; Paul Turinsky; Jamie Coble

    2011-05-31

    Development and deployment of small-scale nuclear power reactors and their maintenance, monitoring, and control are part of the mission under the Small Modular Reactor (SMR) program. The objectives of this NERI-consortium research project are to investigate, develop, and validate advanced methods for sensing, controlling, monitoring, diagnosis, and prognosis of these reactors, and to demonstrate the methods with application to one of the proposed integral pressurized water reactors (IPWR). For this project, the IPWR design by Westinghouse, the International Reactor Secure and Innovative (IRIS), has been used to demonstrate the techniques developed under this project. The research focuses on three topical areas with the following objectives. Objective 1 - Develop and apply simulation capabilities and sensitivity/uncertainty analysis methods to address sensor deployment analysis and small grid stability issues. Objective 2 - Develop and test an autonomous and fault-tolerant control architecture and apply to the IRIS system and an experimental flow control loop, with extensions to multiple reactor modules, nuclear desalination, and optimal sensor placement strategy. Objective 3 - Develop and test an integrated monitoring, diagnosis, and prognosis system for SMRs using the IRIS as a test platform, and integrate process and equipment monitoring (PEM) and process and equipment prognostics (PEP) toolboxes. The research tasks are focused on meeting the unique needs of reactors that may be deployed to remote locations or to developing countries with limited support infrastructure. These applications will require smaller, robust reactor designs with advanced technologies for sensors, instrumentation, and control. An excellent overview of SMRs is described in an article by Ingersoll (2009). The article refers to these as deliberately small reactors. Most of these have modular characteristics, with multiple units deployed at the same plant site. Additionally, the topics focus

  4. Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

    OpenAIRE

    Cisneros, Anselmo Tomas

    2013-01-01

    The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids - flibe (33%7Li2F-67%BeF) - from molten salt reactors. This combination of fuel and coolant enables FHRs to operate i...

  5. Guidebook to nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Nero, A.V. Jr.

    1976-05-01

    A general introduction to reactor physics and theory is followed by descriptions of commercial nuclear reactor types. Future directions for nuclear power are also discussed. The technical level of the material is suitable for laymen.

  6. A mixed flow reactor method to synthesize amorphous calcium carbonate under controlled chemical conditions.

    Science.gov (United States)

    Blue, Christina R; Rimstidt, J Donald; Dove, Patricia M

    2013-01-01

    This study describes a new procedure to synthesize amorphous calcium carbonate (ACC) from well-characterized solutions that maintain a constant supersaturation. The method uses a mixed flow reactor to prepare ACC in significant quantities with consistent compositions. The experimental design utilizes a high-precision solution pump that enables the reactant solution to continuously flow through the reactor under constant mixing and allows the precipitation of ACC to reach steady state. As a proof of concept, we produced ACC with controlled Mg contents by regulating the Mg/Ca ratio of the input solution and the carbonate concentration and pH. Our findings show that the Mg/Ca ratio of the reactant solution is the primary control for the Mg content in ACC, as shown in previous studies, but ACC composition is further regulated by the carbonate concentration and pH of the reactant solution. The method offers promise for quantitative studies of ACC composition and properties and for investigating the role of this phase as a reactive precursor to biogenic minerals. © 2013 Elsevier Inc. All rights reserved.

  7. Physical Analysis of the Initial Core and Running-In Phase for Pebble-Bed Reactor HTR-PM

    Directory of Open Access Journals (Sweden)

    Jingyu Zhang

    2017-01-01

    Full Text Available The pebble-bed reactor HTR-PM is being built in China and is planned to be critical in one or two years. At present, one emphasis of engineering design is to determine the fuel management scheme of the initial core and running-in phase. There are many possible schemes, and many factors need to be considered in the process of scheme evaluation and analysis. Based on the experience from the constructed or designed pebble-bed reactors, the fuel enrichment and the ratio of fuel spheres to graphite spheres are important. In this paper, some relevant physical considerations of the initial core and running-in phase of HTR-PM are given. Then a typical scheme of the initial core and running-in phase is proposed and simulated with VSOP code, and some key physical parameters, such as the maximum power per fuel sphere, the maximum fuel temperature, the refueling rate, and the discharge burnup, are calculated. Results of the physical parameters all satisfy the relevant design requirements, which means the proposed scheme is safe and reliable and can provide support for the fuel management of HTR-PM in the future.

  8. Oklo: The fossil nuclear reactors. Physics study - Translation of chapters 6, 13 and conclusions

    Energy Technology Data Exchange (ETDEWEB)

    Naudet, R. [CEA, Paris (France)

    1996-09-01

    Three parts of the 1991 book `Oklo: reacteurs nucleaires fossiles. Etude physique` have been translated in this report. The chapters bear the titles `Study of criticality`(45 p.), `Some problems with the overall functioning of the reactor zones`(45 p.) and `Conclusions` (15 p.), respectively.

  9. Collaborative Physical and Biological Dosimetry Studies for Neutron Capture Therapy at the RA-1 Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    Nigg, D.W.; Schwint, A.E.; Hartwell, J.K.; Heber, E.M.; Trivillin, V.; Castillo, J.; Wentzeis, L.; Sloan, P.; Wemple, C.A.

    2004-10-04

    Initial physical dosimetry measurements have been completed using activation spectrometry and thermoluminiscent dosimeters to characterize the BNCT irradiation facility developed at the RA-1 research reactor operated by the Argentine National Atomic Energy Commission in Buenos Aires. Some biological scoping irradiations have also been completed using a small-animal (hamster) oral mucosa tumor model. Results indicate that the RA-1 neutron source produces useful dose rates but that some improvements in the initial configuration will be needed to optimize the spectrum for thermal-neutron BNCT research applications.

  10. Collaborative Physical and Biological Dosimetry Studies for Neutron Capture Therapy at the RA-1 Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Amanda E. Schwint; John K. Hartwell; Elisa M. Heber; Veronica Trivillin; Jorge Castillo; Luis Wentzeis; Patrick Sloan; Charles A. Wemple

    2004-10-01

    Initial physical dosimetry measurements have been completed using activation spectrometry and thermoluminiscent dosimeters to characterize the BNCT irradiation facility developed at the RA-1 research reactor operated by the Argentine National Atomic Energy Commission in Buenos Aires. Some biological scoping irradiations have also been completed using a small-animal (hamster) oral mucosa tumor model. Results indicate that the RA-1 neutron source produces useful dose rates but that some improvements in the initial configuration will be needed to optimize the spectrum for thermal-neutron BNCT research applications.

  11. A new method to determine in situ the transmission of a neutron-guide system at a reactor source

    CERN Document Server

    Haan, V O D; Gommers, R M; Labohm, F; Well, A A V; De Leege, P F A; Schebetov, A; Pusenkov, V

    2002-01-01

    In this paper, a description of a new method to determine the transmission of neutron guides after they are installed in a beam-tube at a reactor source is given. The method is based on activation measurements of gold foils at the entrance of the beam-tube and at the exit of the neutron guides compared to Monte-Carlo calculations. In this method, a quality factor is defined as the ratio between the actual transmission and the theoretical maximum attainable transmission. This method is used to determine the quality of an optimised neutron-guide system developed for beam-tube R2 of the HOR. The HOR is a pool-type nuclear research reactor at the Interfaculty Reactor Institute of the Delft University of Technology. It is shown that the quality factors of the newly installed neutron guides are between 0.49 and 0.63.

  12. Statistical analysis using the Bayesian nonparametric method for irradiation embrittlement of reactor pressure vessels

    Science.gov (United States)

    Takamizawa, Hisashi; Itoh, Hiroto; Nishiyama, Yutaka

    2016-10-01

    In order to understand neutron irradiation embrittlement in high fluence regions, statistical analysis using the Bayesian nonparametric (BNP) method was performed for the Japanese surveillance and material test reactor irradiation database. The BNP method is essentially expressed as an infinite summation of normal distributions, with input data being subdivided into clusters with identical statistical parameters, such as mean and standard deviation, for each cluster to estimate shifts in ductile-to-brittle transition temperature (DBTT). The clusters typically depend on chemical compositions, irradiation conditions, and the irradiation embrittlement. Specific variables contributing to the irradiation embrittlement include the content of Cu, Ni, P, Si, and Mn in the pressure vessel steels, neutron flux, neutron fluence, and irradiation temperatures. It was found that the measured shifts of DBTT correlated well with the calculated ones. Data associated with the same materials were subdivided into the same clusters even if neutron fluences were increased.

  13. Apparatus and method for quantitative assay of generic transuranic wastes from nuclear reactors

    Science.gov (United States)

    Caldwell, J.T.; Kunz, W.E.; Atencio, J.D.

    1982-03-31

    A combination of passive and active neutron measurements which yields quantitative information about the isotopic composition of transuranic wastes from nuclear power or weapons material manufacture reactors is described. From the measurement of prompt and delayed neutron emission and the incidence of two coincidentally emitted neutrons from induced fission of fissile material in the sample, one can quantify /sup 233/U, /sup 235/U and /sup 239/Pu isotopes in waste samples. Passive coincidence counting, including neutron multiplicity measurement and determination of the overall passive neutron flux additionally enables the separate quantitative evaluation of spontaneous fission isotopes such as /sup 240/Pu, /sup 244/Cm and /sup 252/Cf, and the spontaneous alpha particle emitter /sup 241/Am. These seven isotopes are the most important constituents of wastes from nuclear power reactors and once the mass of each isotope present is determined by the apparatus and method of the instant invention, the overall alpha particle activity can be determined to better than 1 nCi/g from known radioactivity data. Therefore, in addition to the quantitative analysis of the waste sample useful for later reclamation purposes, the alpha particle activity can be determined to decide whether permanent low-level burial is appropriate for the waste sample.

  14. Mathematical methods in physics distributions, Hilbert space operators, variational methods, and applications in quantum physics

    CERN Document Server

    Blanchard, Philippe

    2015-01-01

    The second edition of this textbook presents the basic mathematical knowledge and skills that are needed for courses on modern theoretical physics, such as those on quantum mechanics, classical and quantum field theory, and related areas.  The authors stress that learning mathematical physics is not a passive process and include numerous detailed proofs, examples, and over 200 exercises, as well as hints linking mathematical concepts and results to the relevant physical concepts and theories.  All of the material from the first edition has been updated, and five new chapters have been added on such topics as distributions, Hilbert space operators, and variational methods.   The text is divided into three main parts. Part I is a brief introduction to distribution theory, in which elements from the theories of ultradistributions and hyperfunctions are considered in addition to some deeper results for Schwartz distributions, thus providing a comprehensive introduction to the theory of generalized functions. P...

  15. High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems

    National Research Council Canada - National Science Library

    Mahadevan, Vijay S; Merzari, Elia; Tautges, Timothy; Jain, Rajeev; Obabko, Aleksandr; Smith, Michael; Fischer, Paul

    2014-01-01

    ...-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently...

  16. Validation of reactor noise linear stability methods by means of advanced stochastic differential equation models

    Energy Technology Data Exchange (ETDEWEB)

    Munoz-Cobo, J.L., E-mail: jlcobos@iqn.upv.es [Instituto de Ingenieri' a Energetica, Universidad Politecnica de Valencia, Camino de Vera 14, Valencia 46022 (Spain); Montesinos, M.E. [Instituto de Ingenieri' a Energetica, Universidad Politecnica de Valencia, Camino de Vera 14, Valencia 46022 (Spain); Pena, J. [Departamento de Matematica Aplicada, Universidad Politecnica de Valencia, Valencia (Spain); Escriva, A.; Gonzalez, C. [Instituto de Ingenieri' a Energetica, Universidad Politecnica de Valencia, Camino de Vera 14, Valencia 46022 (Spain); Melara, J. [IBERDROLA Ingenieri' a y Construccion, Avenida Manoteras 20, Madrid 28050 (Spain)

    2011-07-15

    Highlights: > We study validation methods for stability monitoring of BWR. > We generate synthetic signals from BWR reduced order models with previously known decay ratio. > Parametric and non parametric methods are used for the prediction of the stability. > We show the optimal method for filtering in the evaluation of the decay ratio. - Abstract: The aim of this paper is to show a validation method of a stability monitor using a BWR model with multiple Wiener noise sources, of additive and multiplicative nature. This model is solved using the modern methods to integrate stochastic differential equation systems, that are based on the stochastic Ito-Taylor expansion, and developed by Kloeden and Platen (1995), Kloeden et al. (1994). The synthetic signals generated with this BWR reduced order model with multiple Wiener processes are then used to obtain what are the optimal ways of filtering the signals for the different methods to estimate the decay ration (DR) and the natural frequency ({omega}) of the system. Also, for each DR estimation method, we study what is the optimal combination of algorithms to obtain the order and coefficients of the AR model that yields the best prediction of the reactor stability parameters for a broad range of DR values.

  17. The long-term future for civilian nuclear power generation in France: The case for breeder reactors. Breeder reactors: The physical and physical chemistry parameters, associate material thermodynamics and mechanical engineering: Novelties and issues

    Science.gov (United States)

    Dautray, Robert

    2011-06-01

    The author firstly gives a summary overview of the knowledge base acquired since the first breeder reactors became operational in the 1950s. "Neutronics", thermal phenomena, reactor core cooling, various coolants used and envisioned for this function, fuel fabrication from separated materials, main equipment (pumps, valves, taps, waste cock, safety circuits, heat exchange units, etc.) have now attained maturity, sufficient to implement sodium cooling circuits. Notwithstanding, the use of metallic sodium still raises certain severe questions in terms of safe handling (i.e. inflammability) and other important security considerations. The structural components, both inside the reactor core and outside (i.e. heat exchange devices) are undergoing in-depth research so as to last longer. The fuel cycle, notably the refabrication of fuel elements and fertile elements, the case of transuranic elements, etc., call for studies into radiation induced phenomena, chemistry separation, separate or otherwise treatments for materials that have different radioactive, physical, thermodynamical, chemical and biological properties. The concerns that surround the definitive disposal of certain radioactive wastes could be qualitatively improved with respect to the pressurized water reactors (PWRs) in service today. Lastly, the author notes that breeder reactors eliminate the need for an isotope separation facility, and this constitutes a significant contribution to contain nuclear proliferation. Among the priorities for a fully operational system (power station - the fuel cycle - operation-maintenance - the spent fuel pool and its cooling system-emergency cooling system-emergency electric power-transportation movements-equipment handling - final disposal of radioactive matter, independent safety barriers), the author includes materials (fabrication of targets, an irradiation and inspection instrument), the chemistry of all sorting processes, equipment "refabrication" or rehabilitation

  18. Research on friction coefficient of nuclear Reactor Vessel Internals Hold Down Spring: Stress coefficient test analysis method

    Energy Technology Data Exchange (ETDEWEB)

    Linjun, Xie, E-mail: linjunx@zjut.edu.cn [College of Mechanical Engineering, Zhejiang University of Technology, Hangzhou 310014 (China); Guohong, Xue; Ming, Zhang [Shanghai Nuclear Engineering Research & Design Institute, Shanghai 200233 (China)

    2016-08-01

    Graphical abstract: HDS stress coefficient test apparatus. - Highlights: • This paper performs mathematic deduction to the physical model of Hold Down Spring (HDS), establishes a mathematic model of axial load P and stress, stress coefficient and friction coefficient and designs a set of test apparatuses for simulating the pretightening process of the HDS for the first time according to a model similarity criterion. • The mathematical relation between the load and the strain is obtained about the HDS, and the mathematical model of the stress coefficient and the friction coefficient is established. So, a set of test apparatuses for obtaining the stress coefficient is designed according to the model scaling criterion and the friction coefficient of the K1000 HDS is calculated to be 0.336 through the obtained stress coefficient. • The relation curve between the theoretical load and the friction coefficient is obtained through analysis and indicates that the change of the friction coefficient f would influence the pretightening load under the condition of designed stress. The necessary pretightening load in the design process is calculated to be 5469 kN according to the obtained friction coefficient. Therefore, the friction coefficient and the pretightening load under the design conditions can provide accurate pretightening data for the analysis and design of the reactor HDS according to the operations. - Abstract: This paper performs mathematic deduction to the physical model of Hold Down Spring (HDS), establishes a mathematic model of axial load P and stress, stress coefficient and friction coefficient and designs a set of test apparatuses for simulating the pretightening process of the HDS for the first time according to a model similarity criterion. By carrying out tests and researches through a stress testing technique, P–σ curves in loading and unloading processes of the HDS are obtained and the stress coefficient k{sub f} of the HDS is obtained. So, the

  19. Applied Mathematical Methods in Theoretical Physics

    Science.gov (United States)

    Masujima, Michio

    2005-04-01

    All there is to know about functional analysis, integral equations and calculus of variations in a single volume. This advanced textbook is divided into two parts: The first on integral equations and the second on the calculus of variations. It begins with a short introduction to functional analysis, including a short review of complex analysis, before continuing a systematic discussion of different types of equations, such as Volterra integral equations, singular integral equations of Cauchy type, integral equations of the Fredholm type, with a special emphasis on Wiener-Hopf integral equations and Wiener-Hopf sum equations. After a few remarks on the historical development, the second part starts with an introduction to the calculus of variations and the relationship between integral equations and applications of the calculus of variations. It further covers applications of the calculus of variations developed in the second half of the 20th century in the fields of quantum mechanics, quantum statistical mechanics and quantum field theory. Throughout the book, the author presents over 150 problems and exercises -- many from such branches of physics as quantum mechanics, quantum statistical mechanics, and quantum field theory -- together with outlines of the solutions in each case. Detailed solutions are given, supplementing the materials discussed in the main text, allowing problems to be solved making direct use of the method illustrated. The original references are given for difficult problems. The result is complete coverage of the mathematical tools and techniques used by physicists and applied mathematicians Intended for senior undergraduates and first-year graduates in science and engineering, this is equally useful as a reference and self-study guide.

  20. A reverse method for the determination of the radiological inventory of irradiated graphite at reactor scale

    Energy Technology Data Exchange (ETDEWEB)

    Nicaise, Gregory [Institut de Radioprotection et de Surete Nucleaire, Fontenay-aux-roses (France); Poncet, Bernard [EDF-DP2D, Lyon (France)

    2016-11-15

    Irradiated graphite waste will be produced from the decommissioning of the six gas-cooled nuclear reactors operated by Electricite De France (EDF). Determining the radionuclide content of this waste is an important legal commitment for both safety reasons and in order to determine the best suited management strategy. As evidenced by numerous studies nuclear graphite is a very pure material, however, it cannot be considered from an analytical viewpoint as a usual homogeneous material. Because of graphite high purity, radionuclide measurements in irradiated graphite exhibit very high discrepancies especially when corresponding to precursors at trace level. Therefore the assessment of a radionuclide inventory only based on few number of radiochemical measurements leads in most of cases to a gross over or under-estimation that can be detrimental to graphite waste management. A reverse method using an identification calculation-measurement process is proposed in order to assess the radionuclide inventory as precisely as possible.

  1. Systems and methods for harvesting and storing materials produced in a nuclear reactor

    Science.gov (United States)

    Heinold, Mark R.; Dayal, Yogeshwar; Brittingham, Martin W.

    2016-04-05

    Systems produce desired isotopes through irradiation in nuclear reactor instrumentation tubes and deposit the same in a robust facility for immediate shipping, handling, and/or consumption. Irradiation targets are inserted and removed through inaccessible areas without plant shutdown and placed in the harvesting facility, such as a plurality of sealable and shipping-safe casks and/or canisters. Systems may connect various structures in a sealed manner to avoid release of dangerous or unwanted matter throughout the nuclear plant, and/or systems may also automatically decontaminate materials to be released. Useable casks or canisters can include plural barriers for containment that are temporarily and selectively removable with specially-configured paths inserted therein. Penetrations in the facilities may limit waste or pneumatic gas escape and allow the same to be removed from the systems without over-pressurization or leakage. Methods include processing irradiation targets through such systems and securely delivering them in such harvesting facilities.

  2. Systems and methods for enhancing isolation of high-temperature reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Per F.

    2017-09-26

    A high-temperature containment-isolation system for transferring heat from a nuclear reactor containment to a high-pressure heat exchanger is presented. The system uses a high-temperature, low-volatility liquid coolant such as a molten salt or a liquid metal, where the coolant flow path provides liquid free surfaces a short distance from the containment penetrations for the reactor hot-leg and the cold-leg, where these liquid free surfaces have a cover gas maintained at a nearly constant pressure and thus prevent high-pressures from being transmitted into the reactor containment, and where the reactor vessel is suspended within a reactor cavity with a plurality of refractory insulator blocks disposed between an actively cooled inner cavity liner and the reactor vessel.

  3. Optimization of hydrodynamic cavitations reactor efficiency for biodiesel production by response surface methods (Case study: Sunflower oil

    Directory of Open Access Journals (Sweden)

    H Javadikia

    2017-05-01

    Full Text Available Introduction Biofuels are considered as one of the largest sources of renewable fuels or replacement of fossil fuels. Combustion of plant-based fuels is the indirect use of solar energy. Biofuels significantly have less pollution than other fossil fuels and can easily generate from residual plant material. Waste and residues of foods and wastewater can also be a good source for biofuel production. Transesterification method (one of biodiesel production methods is the most common forms to produce mono-alkyl esters from vegetable oil and animal fats. The procedure aims are reduction the oil viscosity during the reaction between triglycerides and alcohol in the presence of a catalyst or without it. In this study, the method of transesterification with alkaline catalysts is used that it is the most common and most commercial biodiesel production method. In this study, configurations of made hydrodynamic cavitation reactor were studied to measure biodiesel fuel quality and enhanced device performance with optimum condition. The Design Expert software and response surface methodology were used to get this purpose. Materials and Methods Transesterification method was used in this study. The procedure aims were reduction of the oil viscosity during the reaction between triglycerides and alcohol in the presence of a catalyst or without it. Materials needed in the production of biodiesel transesterification method include: vegetable oil, alcohol and catalysts. The used oil in the production of biodiesel was sunflower oil, which was used 0.6 liters per each test in the production process base on titration method. Methanol with purity of 99.8 percent and the molar ratio of 6:1 to oil was used based on titration equation and according to the results of other researchers. The used catalyst in continuous production process was high-purity sodium hydroxide (99% that it is one of alkaline catalysts. Weight of hydroxide was 1% of the used oil weight in the

  4. Physics Research Methods at Jefferson High.

    Science.gov (United States)

    Demchik, Michael

    1989-01-01

    Described are several physics activities developed by a high school: research corner, exploring an invention, construction projects, archives report and archives update, short term research and design projects, essay contest, and special projects. (YP)

  5. Multi-Scale Multi-physics Methods Development for the Calculation of Hot-Spots in the NGNP

    Energy Technology Data Exchange (ETDEWEB)

    Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Seker, Volkan [Univ. of Michigan, Ann Arbor, MI (United States)

    2013-04-30

    Radioactive gaseous fission products are released out of the fuel element at a significantly higher rate when the fuel temperature exceeds 1600°C in high-temperature gas-cooled reactors (HTGRs). Therefore, it is of paramount importance to accurately predict the peak fuel temperature during all operational and design-basis accident conditions. The current methods used to predict the peak fuel temperature in HTGRs, such as the Next-Generation Nuclear Plant (NGNP), estimate the average fuel temperature in a computational mesh modeling hundreds of fuel pebbles or a fuel assembly in a pebble-bed reactor (PBR) or prismatic block type reactor (PMR), respectively. Experiments conducted in operating HTGRs indicate considerable uncertainty in the current methods and correlations used to predict actual temperatures. The objective of this project is to improve the accuracy in the prediction of local "hot" spots by developing multi-scale, multi-physics methods and implementing them within the framework of established codes used for NGNP analysis.The multi-scale approach which this project will implement begins with defining suitable scales for a physical and mathematical model and then deriving and applying the appropriate boundary conditions between scales. The macro scale is the greatest length that describes the entire reactor, whereas the meso scale models only a fuel block in a prismatic reactor and ten to hundreds of pebbles in a pebble bed reactor. The smallest scale is the micro scale--the level of a fuel kernel of the pebble in a PBR and fuel compact in a PMR--which needs to be resolved in order to calculate the peak temperature in a fuel kernel.

  6. Analysis of JSI TRIGA MARK II reactor physical parameters calculated with TRIPOLI and MCNP.

    Science.gov (United States)

    Henry, R; Tiselj, I; Snoj, L

    2015-03-01

    New computational model of the JSI TRIGA Mark II research reactor was built for TRIPOLI computer code and compared with existing MCNP code model. The same modelling assumptions were used in order to check the differences of the mathematical models of both Monte Carlo codes. Differences between the TRIPOLI and MCNP predictions of keff were up to 100pcm. Further validation was performed with analyses of the normalized reaction rates and computations of kinetic parameters for various core configurations. Copyright © 2014 Elsevier Ltd. All rights reserved.

  7. Atmospheric Physics Background – Methods – Trends

    CERN Document Server

    2012-01-01

    On the occasion of the 50th anniversary of the Institute of Atmospheric Physics of the German Aerospace Center (DLR), this book presents more than 50 chapters highlighting results of the institute’s research. The book provides an up-to-date, in-depth survey across the entire field of atmospheric science, including atmospheric dynamics, radiation, cloud physics, chemistry, climate, numerical simulation, remote sensing, instruments and measurements, as well as atmospheric acoustics. The authors have provided a readily comprehensible and self-contained presentation of the complex field of atmospheric science. The topics are of direct relevance for aerospace science and technology. Future research challenges are identified.

  8. The effect of carbon crystal structure on treat reactor physics calculations

    Energy Technology Data Exchange (ETDEWEB)

    Swanson, R.W.; Harrison, L.J.

    1988-01-01

    The Transient Reactor Test Facility (TREAT) at Argonne National Laboratory-West (ANL-W) is fueled with urania in a graphite and carbon mixture. This fuel was fabricated from a mixture of graphite flour, thermax (a thermatomic carbon produced by ''cracking'' natural gas), coal-tar resin and U/sub 3/O/sub 8/. During the fabrication process, the fuel was baked to dissociate the resin, but the high temperature necessary to graphitize the carbon in the thermax and in the resin was avoided. Therefore, the carbon crystal structure is a complex mixture of graphite particles in a nongraphitized elemental carbon matrix. Results of calculations using macroscopic carbon cross sections obtained by mixing bound-kernel graphite cross sections for the graphitized carbon and free-gas carbon cross sections for the remainder of the carbon and calculations using only bound-kernel graphite cross sections are compared to experimental data. It is shown that the use of the hybridized cross sections which reflect the allotropic mixture of the carbon in the TREAT fuel results in a significant improvement in the accuracy of calculated neutronics parameters for the TREAT reactor. 6 refs., 2 figs., 3 tabs.

  9. Phenomena-based Uncertainty Quantification in Predictive Coupled- Physics Reactor Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Adams, Marvin [Texas A & M Univ., College Station, TX (United States)

    2017-06-12

    This project has sought to develop methodologies, tailored to phenomena that govern nuclearreactor behavior, to produce predictions (including uncertainties) for quantities of interest (QOIs) in the simulation of steady-state and transient reactor behavior. Examples of such predictions include, for each QOI, an expected value as well as a distribution around this value and an assessment of how much of the distribution stems from each major source of uncertainty. The project has sought to test its methodologies by comparing against measured experimental outcomes. The main experimental platform has been a 1-MW TRIGA reactor. This is a flexible platform for a wide range of experiments, including steady state with and without temperature feedback, slow transients with and without feedback, and rapid transients with strong feedback. The original plan was for the primary experimental data to come from in-core neutron detectors. We made considerable progress toward this goal but did not get as far along as we had planned. We have designed, developed, installed, and tested vertical guide tubes, each able to accept a detector or stack of detectors that can be moved axially inside the tube, and we have tested several new detector designs. One of these shows considerable promise.

  10. Experimental Investigations of Physical and Chemical Properties for Microalgae HTL Bio-Crude Using a Large Batch Reactor

    Directory of Open Access Journals (Sweden)

    Farhad M. Hossain

    2017-04-01

    Full Text Available As a biofuel feedstock, microalgae has good scalability and potential to supply a significant proportion of world energy compared to most types of biofuel feedstock. Hydrothermal liquefaction (HTL is well-suited to wet biomass (such as microalgae as it greatly reduces the energy requirements associated with dewatering and drying. This article presents experimental analyses of chemical and physical properties of bio-crude oil produced via HTL using a high growth-rate microalga Scenedesmus sp. in a large batch reactor. The overarching goal was to investigate the suitability of microalgae HTL bio-crude produced in a large batch reactor for direct application in marine diesel engines. To this end we characterized the chemical and physical properties of the bio-crudes produced. HTL literature mostly reports work using very small batch reactors which are preferred by researchers, so there are few experimental and parametric measurements for bio-crude physical properties, such as viscosity and density. In the course of this study, a difference between traditionally calculated values and measured values was noted. In the parametric study, the bio-crude viscosity was significantly closer to regular diesel and biodiesel standards than transesterified (FAME microalgae biodiesel. Under optimised conditions, HTL bio-crude’s high density (0.97–1.04 kg·L−1 and its high viscosity (70.77–73.89 mm2·s−1 had enough similarity to marine heavy fuels. although the measured higher heating value, HHV, was lower (29.8 MJ·kg−1. The reaction temperature was explored in the range 280–350 °C and bio-crude oil yield and HHV reached their maxima at the highest temperature. Slurry concentration was explored between 15% and 30% at this temperature and the best HHV, O:C, and N:C were found to occur at 25%. Two solvents (dichloromethane and n-hexane were used to recover the bio-crude oil, affecting the yield and chemical composition of the bio-crude.

  11. Application of the Monte Carlo method to estimate doses due to neutron activation of different materials in a nuclear reactor

    Science.gov (United States)

    Ródenas, José

    2017-11-01

    All materials exposed to some neutron flux can be activated independently of the kind of the neutron source. In this study, a nuclear reactor has been considered as neutron source. In particular, the activation of control rods in a BWR is studied to obtain the doses produced around the storage pool for irradiated fuel of the plant when control rods are withdrawn from the reactor and installed into this pool. It is very important to calculate these doses because they can affect to plant workers in the area. The MCNP code based on the Monte Carlo method has been applied to simulate activation reactions produced in the control rods inserted into the reactor. Obtained activities are introduced as input into another MC model to estimate doses produced by them. The comparison of simulation results with experimental measurements allows the validation of developed models. The developed MC models have been also applied to simulate the activation of other materials, such as components of a stainless steel sample introduced into a training reactors. These models, once validated, can be applied to other situations and materials where a neutron flux can be found, not only nuclear reactors. For instance, activation analysis with an Am-Be source, neutrography techniques in both medical applications and non-destructive analysis of materials, civil engineering applications using a Troxler, analysis of materials in decommissioning of nuclear power plants, etc.

  12. Numerical Simulation of Particle Flow Motion in a Two-Dimensional Modular Pebble-Bed Reactor with Discrete Element Method

    Directory of Open Access Journals (Sweden)

    Guodong Liu

    2013-01-01

    Full Text Available Modular pebble-bed nuclear reactor (MPBNR technology is promising due to its attractive features such as high fuel performance and inherent safety. Particle motion of fuel and graphite pebbles is highly associated with the performance of pebbled-bed modular nuclear reactor. To understand the mechanism of pebble’s motion in the reactor, we numerically studied the influence of number ratio of fuel and graphite pebbles, funnel angle of the reactor, height of guide ring on the distribution of pebble position, and velocity by means of discrete element method (DEM in a two-dimensional MPBNR. Velocity distributions at different areas of the reactor as well as mixing characteristics of fuel and graphite pebbles were investigated. Both fuel and graphite pebbles moved downward, and a uniform motion was formed in the column zone, while pebbles motion in the cone zone was accelerated due to the decrease of the cross sectional flow area. The number ratio of fuel and graphite pebbles and the height of guide ring had a minor influence on the velocity distribution of pebbles, while the variation of funnel angle had an obvious impact on the velocity distribution. Simulated results agreed well with the work in the literature.

  13. Methods of assessing the effects of interface oxide growth in Magnox and advanced gas-cooled reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    McLauchlin, I.R.; Wooton, M.R.; Morgan, J.D.; Watson, L.H.

    1986-10-01

    Growth of oxide at interfaces between structural steel components in CO/sub 2/-cooled reactors can deform fastenings such as bolts and welds. The mechanical response of joint members to oxide growth is discussed, and methods of assessment are outlined which contribute to procedures for ensuring continued structural integrity.

  14. Method for carrying out biotechnological processes by means of a multi-phase system in a loop reactor.

    NARCIS (Netherlands)

    Tramper, J.

    1987-01-01

    A method for carrying out biotechnological processes by means of multiphase system in a loop reactor, which system comprises an aqueous phase (11) and at least one organic solvent (9, 10) which is immiscible with water and which has a different density from water; one of the liquid components is

  15. The Tokamak Fusion Test Reactor decontamination and decommissioning project and the Tokamak Physics Experiment at the Princeton Plasma Physics Laboratory. Environmental Assessment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-05-27

    If the US is to meet the energy needs of the future, it is essential that new technologies emerge to compensate for dwindling supplies of fossil fuels and the eventual depletion of fissionable uranium used in present-day nuclear reactors. Fusion energy has the potential to become a major source of energy for the future. Power from fusion energy would provide a substantially reduced environmental impact as compared with other forms of energy generation. Since fusion utilizes no fossil fuels, there would be no release of chemical combustion products to the atmosphere. Additionally, there are no fission products formed to present handling and disposal problems, and runaway fuel reactions are impossible due to the small amounts of deuterium and tritium present. The purpose of the TPX Project is to support the development of the physics and technology to extend tokamak operation into the continuously operating (steady-state) regime, and to demonstrate advances in fundamental tokamak performance. The purpose of TFTR D&D is to ensure compliance with DOE Order 5820.2A ``Radioactive Waste Management`` and to remove environmental and health hazards posed by the TFTR in a non-operational mode. There are two proposed actions evaluated in this environmental assessment (EA). The actions are related because one must take place before the other can proceed. The proposed actions assessed in this EA are: the decontamination and decommissioning (D&D) of the Tokamak Fusion Test Reactor (TFTR); to be followed by the construction and operation of the Tokamak Physics Experiment (TPX). Both of these proposed actions would take place primarily within the TFTR Test Cell Complex at the Princeton Plasma Physics Laboratory (PPPL). The TFTR is located on ``D-site`` at the James Forrestal Campus of Princeton University in Plainsboro Township, Middlesex County, New Jersey, and is operated by PPPL under contract with the United States Department of Energy (DOE).

  16. An Advanced Integrated Diffusion/Transport Method for the Design, Analysis and Optimization of the Very-High-Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Farzad Rahnema; Dingkang Zhang; Abderrafi Ougouag; Frederick Gleicher

    2011-04-04

    The main objective of this research is to develop an integrated diffusion/transport (IDT) method to substantially improve the accuracy of nodal diffusion methods for the design and analysis of Very High Temperature Reactors (VHTR). Because of the presence of control rods in the reflector regions in the Pebble Bed Reactor (PBR-VHTR), traditional nodal diffusion methods do not accurately model these regions, within which diffusion theory breaks down in the vicinity of high neutron absorption and steep flux gradients. The IDT method uses a local transport solver based on a new incident flux response expansion method in the controlled nodes. Diffusion theory is used in the rest of the core. This approach improves the accuracy of the core solution by generating transport solutions of controlled nodes while maintaining computational efficiency by using diffusion solutions in nodes where such a treatment is sufficient. The transport method is initially developed and coupled to the reformulated 3-D nodal diffusion model in the CYNOD code for PBR core design and fuel cycle analysis. This method is also extended to the prismatic VHTR. The new method accurately captures transport effects in highly heterogeneous regions with steep flux gradients. The calculations of these nodes with transport theory avoid errors associated with spatial homogenization commonly used in diffusion methods in reactor core simulators

  17. A survey on the human reliability analysis methods for the design of Korean next generation reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Hee; Lee, J. W.; Park, J. C.; Kwack, H. Y.; Lee, K. Y.; Park, J. K.; Kim, I. S.; Jung, K. W

    2000-03-01

    Enhanced features through applying recent domestic technologies may characterize the safety and efficiency of KNGR(Korea Next Generation Reactor). Human engineered interface and control room environment are expected to be beneficial to the human aspects of KNGR design. However, since the current method for human reliability analysis is not up to date after THERP/SHARP, it becomes hard to assess the potential of human errors due to both of the positive and negative effect of the design changes in KNGR. This is a state of the art report on the human reliability analysis methods that are potentially available for the application to the KNGR design. We surveyed every technical aspects of existing HRA methods, and compared them in order to obtain the requirements for the assessment of human error potentials within KNGR design. We categorized the more than 10 methods into the first and the second generation according to the suggestion of Dr. Hollnagel. THERP was revisited in detail. ATHEANA proposed by US NRC for an advanced design and CREAM proposed by Dr. Hollnagel were reviewed and compared. We conclude that the key requirements might include the enhancement in the early steps for human error identification and the quantification steps with considerations of more extended error shaping factors over PSFs(performance shaping factors). The utilization of the steps and approaches of ATHEANA and CREAM will be beneficial to the attainment of an appropriate HRA method for KNGR. However, the steps and data from THERP will be still maintained because of the continuity with previous PSA activities in KNGR design.

  18. Estimation of in-plant source term release behaviors from Fukushima daiichi reactor cores by forward method and comparison with reverse method

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Won; Rhee, Bo Wook; Song, Jin Ho; Kim, Sung Il; Ha, Kwang Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-06-15

    The purpose of this paper is to confirm the event timings and the magnitude of fission product aerosol release from the Fukushima accident. Over a few hundreds of technical papers have been published on the environmental impact of Fukushima Daiichi accident since the accident occurred on March 11, 2011. However, most of the research used reverse or inverse method based on the monitoring of activities in the remote places and only few papers attempted to estimate the release of fission products from individual reactor core or from individual spent fuel pool. Severe accident analysis code can be used to estimate the radioactive release from which reactor core and from which radionuclide the peaks in monitoring points can be generated. The basic material used for this study are the initial core inventory obtained from the report JAEA-Data/Code 2012–018 and the given accident scenarios provided by Japanese Government or Tokyo Electric Power Company (TEPCO) in official reports. In this research a forward method using severe accident progression code is used as it might be useful for justifying the results of reverse or inverse method or vice versa. The release timing and amounts to the environment are estimated for volatile radioactive fission products such as noble gases, cesium, iodine, and tellurium up to 184 hours (about 7.7 days) after earthquake occurs. The in-plant fission product behaviors and release characteristics to environment are estimated using the severe accident progression analysis code, MELCOR, for Fukushima Daiichi accident. These results are compared with other research results which are summarized in UNSCEAR 2013 Report and other technical papers. Also it may provide the physically based arguments for justifying or suspecting the rationale for the scenarios provided in open literature. The estimated results by MELCOR code simulation of this study indicate that the release amount of volatile fission products to environment from Units 1, 2, and 3

  19. Relevance of β-delayed neutron data for reactor, nuclear physics and astrophysics applications

    Energy Technology Data Exchange (ETDEWEB)

    Kratz, Karl-Ludwig [Fachbereich Chemie, Pharmazie and Geowissenschaften, Universität Mainz, D-55128 Mainz (Germany)

    2015-02-24

    Initially, yields (or abundances) and branching ratios of β-delayed neutrons (βdn) from fission products (P{sub n}-values) have had their main importance in nuclear reactor control. At that time, the six-group mathematical approximation of the time-dependence of βdn-data in terms of the so-called 'Keepin groups' was generally accepted. Later, with the development of high-resolution neutron spectroscopy, βdn data have provided important information on nuclear-structure properties at intermediate excitation energy in nuclei far from stability, as well as in nuclear astrophysics. In this paper, I will present some examples of the βdn-studies performed by the Kernchemie Mainz group during the past three decades. This work has been recognized as an example of 'broad scientific diversity' which has led to my nomination for the 2014 Hans A. Bethe prize.

  20. The Lead Cooled Fast Reactor Benchmark BREST-300:. Analysis with Sensitivity Method

    Science.gov (United States)

    Smirnov, Valery; Orlov, Victor; Mourogov, Alexandre; Lecarpentier, David; Ivanova, Tatiana

    2006-04-01

    Sustainable development of atomic energy will require development of new types of reactors able to exceed the limits of the existing reactor types in terms of optimum use of natural fuel resources, reduction in the production of long-lived radioactive waste, economic and safety competitiveness. Lead cooled fast neutrons reactor is one of the most interesting candidates with a potential to address these needs. BREST-300 is a 300 MWe lead cooled fast reactor developed by the NIKIET (Russia) with a deterministic safety approach which aims to exclude reactivity margins greater than the delayed neutron fraction. The development of innovative reactors (lead coolant, nitride fuel…) and fuel cycles with new constraints such as cycle closure or actinide burning, requires new technologies and new data from various disciplines: fuel types, fuel designs and fuel reprocessing. In this connection, the tool and neutron data used for the calculational analysis of reactor characteristics requires thorough validation, even if computational codes in Russia and France relies to the calculation of fast reactors' parameters and “fast” experiments. NIKIET developed a reactor benchmark fitting of design type calculational tools (including neutron data). In the frame of technical exchanges between the NIKIET and the EDF (Electricité De France), results of this benchmark calculation concerning the principal parameters of fuel evolution and safety parameters has been intercompared, in order to estimate the uncertainties and validate the codes for calculations of these new kind of reactors. Different codes and cross-sections data have been used, and sensitivity studies have been performed to understand and quantify the uncertainties sources.

  1. Minor Actinide Transmutation Physics for Low Conversion Ratio Sodium Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mehdi Asgari; Samuel E. Bays; Benoit Forget; Rodolfo Ferrer

    2007-09-01

    The effects of varying the reprocessing strategy used in the closed cycle of a Sodium Fast Reactor (SNF) prototype are presented in this paper. The isotopic vector from the aqueous separation of transuranic (TRU) elements in Light Water Reactor (LWR) spent nuclear fuel (SNF) is assumed to also vary according to the reprocessing strategy of the closed fuel cycle. The decay heat, gamma energy, and neutron emission of the fuel discharge at equilibrium are found to vary depending on the separation strategy. The SFR core used in this study corresponds to a burner configuration with a conversion ratio of ~0.5 based on the Super-PRISM design. The reprocessing strategies stemming from the choice of either metal or oxide fuel for the SFR are found to have a large impact on the equilibrium discharge decay heat, gamma energy, and neutron emission. Specifically, metal fuel SFR with pyroprocessing of the discharge produces the largest amount of TRU consumption (166 kg per Effective Full Power Year or EFPY), but also the highest decay heat, gamma energy, and neutron emission. On the other hand, an oxide fuel SFR with PUREX reprocessing minimizes the decay heat and related parameters of interest to a minimum, even when compared to thermal Mixed Oxide (MOX) or Inert Matrix Fuel (IMF) on a per mass basis. On an assembly basis, however, the metal SFR discharge has a lower decay heat than an equivalent oxide SFR assembly for similar minor actinide consumptions (~160 kg/EFPY.) Another disadvantage in the oxide PUREX reprocessing scenario is that there is no consumption of americium and curium, since PUREX reprocessing separates these minor actinides (MA) and requires them to be disposed of externally.

  2. Reactor Neutrinos

    CERN Document Server

    Lasserre, T; Lasserre, Thierry; Sobel, Henry W.

    2005-01-01

    We review the status and the results of reactor neutrino experiments, that toe the cutting edge of neutrino research. Short baseline experiments have provided the measurement of the reactor neutrino spectrum, and are still searching for important phenomena such as the neutrino magnetic moment. They could open the door to the measurement of coherent neutrino scattering in a near future. Middle and long baseline oscillation experiments at Chooz and KamLAND have played a relevant role in neutrino oscillation physics in the last years. It is now widely accepted that a new middle baseline disappearance reactor neutrino experiment with multiple detectors could provide a clean measurement of the last undetermined neutrino mixing angle theta13. We conclude by opening on possible use of neutrinos for Society: NonProliferation of Nuclear materials and Geophysics.

  3. Method and apparatus for enhancing reactor air-cooling system performance

    Science.gov (United States)

    Hunsbedt, A.

    1996-03-12

    An enhanced decay heat removal system is disclosed for removing heat from the inert gas-filled gap space between the reactor vessel and the containment vessel of a liquid metal-cooled nuclear reactor. Multiple cooling ducts in flow communication with the inert gas-filled gap space are incorporated to provide multiple flow paths for the inert gas to circulate to heat exchangers which remove heat from the inert gas, thereby introducing natural convection flows in the inert gas. The inert gas in turn absorbs heat directly from the reactor vessel by natural convection heat transfer. 6 figs.

  4. A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Puls, M.P.; Wilkins, B.J.S.; Rigby, G.L. [Whiteshell Labs., Pinawa (Canada)] [and others

    1997-04-01

    A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis.

  5. Mathematical methods in physics and engineering

    CERN Document Server

    Dettman, John W

    2011-01-01

    Intended for college-level physics, engineering, or mathematics students, this volume offers an algebraically based approach to various topics in applied math. It is accessible to undergraduates with a good course in calculus which includes infinite series and uniform convergence. Exercises follow each chapter to test the student's grasp of the material; however, the author has also included exercises that extend the results to new situations and lay the groundwork for new concepts to be introduced later. A list of references for further reading will be found at the end of each chapter. For t

  6. Methods of the physics of porous media

    CERN Document Server

    Wong, Po-Zen; De Graef, Marc

    1999-01-01

    Over the past 25 years, the field of VUV physics has undergone significant developments as new powerful spectroscopic tools, VUV lasers, and optical components have become available. This volume is aimed at experimentalists who are in need of choosing the best type of modern instrumentation in this applied field. In particular, it contains a detailed chapter on laboratory sources. This volume provides an up-to-date description of state-of-the-art equipment and techniques, and a broad reference bibliography. It treats phenomena from the standpoint of an experimental physicist, whereby such topi

  7. [Patients on the move: validated methods to quantify physical activity].

    Science.gov (United States)

    Bakker, Esmée A; Eijsvogels, Thijs M H; de Vegt, Femmie; Busser, Guus S F; Hopman, Maria T E; Verbeek, André L M

    2015-01-01

    Physical activity is an important component in the maintenance and improvement of general health; physical inactivity is, however, an increasing problem in the Netherlands. Requests for advice on physical activity are increasing within the healthcare. Assessment of an individual's physical activity pattern is required to provide tailored advice. There are a number of methods for measuring physical activity; these are divided into subjective and objective methods. Subjective measures include physical activity questionnaires and diaries. Objective measures include indirect calorimetry, measurement with doubly labelled water, heart-rate monitoring and the use of an accelerometer or pedometer. The choice of method depends predominantly on the aim of the measurement, and the availability of personnel, time and financial resources. In clinical practice a validated questionnaire is usually the preferred method, but when measuring effects this should be combined with an objective measurement instrument.

  8. Development of modern methods with respect to neutron transport and uncertainty analyses for reactor core calculations. Interim report; Weiterentwicklung moderner Verfahren zu Neutronentransport und Unsicherheitsanalysen fuer Kernberechnungen. Zwischenbericht

    Energy Technology Data Exchange (ETDEWEB)

    Zwermann, Winfried; Aures, Alexander; Bostelmann, Friederike; Pasichnyk, Ihor; Perin, Yann; Velkov, Kiril; Zilly, Matias

    2016-12-15

    This report documents the status of the research and development goals reached within the reactor safety research project RS1536 ''Development of modern methods with respect to neutron transport and uncertainty analyses for reactor core calculations'' as of the 3{sup rd} quarter of 2016. The superordinate goal of the project is the development, validation, and application of neutron transport methods and uncertainty analyses for reactor core calculations. These calculation methods will mainly be applied to problems related to the core behaviour of light water reactors and innovative reactor concepts, in particular fast reactors cooled by liquid metal. The contributing individual goals are the further optimization and validation of deterministic calculation methods with high spatial and energy resolution, the development of a coupled calculation system using the Monte Carlo method for the neutron transport to describe time-dependent reactor core states, the processing and validation of nuclear data, particularly with regard to covariance data, the development, validation, and application of sampling-based methods for uncertainty and sensitivity analyses, the creation of a platform for performing systematic uncertainty analyses for fast reactor systems, as well as the description of states of severe core damage with the Monte Carlo method. Moreover, work regarding the European NURESAFE project, started in the preceding project RS1503, are being continued and completed.

  9. Research and Development Program in Reactor Diagnostics and Monitoring with Neutron Noise Methods. Stage 11 and 12. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Sunde, C.; Pazsit, I.; Demaziere, C.; Dahl, O.; Mileshina, L. [Chalmers Univ. of Technology, Goeteborg (Sweden). Dept. of Nuclear Engineering

    2006-06-15

    eigenfunctions and eigenvalues in the same model system as the ones in which the semi-analytical calculations were made. Excellent agreement was found between the two methods. Calculations were made in a large and a small system, and the decay of the higher eigenvalues with the order number could be compared. Since the benchmark showed the correct functioning of the simulator, it can be used in the continuation for treating real inhomogeneous systems. The noise simulator was used in a 2-dimensional model of a realistic reactor, supposed to run in a subcritical state, driven by a source. It was meant to simulate a core under loading. For the sake of comparison, a small system was also investigated. The full space-frequency response of the system to fluctuations of the source strength were calculated, and compared to the point kinetic response, also calculated by the simulator. The so-called break frequency method of determining the reactivity was also investigated. It was found that the system behaviour deviated from the point kinetic one quite markedly even in the small system, and that the break frequency method showed a relatively large error, that depended on the position of the detector used. Zero power noise and power reactor noise are two different branches of the field of neutron noise. They depend on different underlying physical processes; they are dominating in two different power regimes (low or 'zero' power and high power, respectively); and last, but not least, they are treated with two completely different methods. A number of new results were obtained. First of all it was shown that the backward equation treatment is not applicable in systems with fluctuating parameters. The concept of criticality had to be generalised to 'criticality in the mean', and it was shown that a system whose state fluctuates between a subcritical and a supercritical state, can be made critical with a given special combination of the system properties in the two states

  10. 10 CFR 73.55 - Requirements for physical protection of licensed activities in nuclear power reactors against...

    Science.gov (United States)

    2010-01-01

    ... activities in nuclear power reactors against radiological sabotage. 73.55 Section 73.55 Energy NUCLEAR... power reactors against radiological sabotage. (a) Introduction. (1) By March 31, 2010, each nuclear... this section as applicable to operating nuclear power reactor facilities. (6) Applicants for an...

  11. Using the Case Study Method in Teaching College Physics

    Science.gov (United States)

    Burko, Lior M.

    2016-01-01

    The case study teaching method has a long history (starting at least with Socrates) and wide current use in business schools, medical schools, law schools, and a variety of other disciplines. However, relatively little use is made of it in the physical sciences, specifically in physics or astronomy. The case study method should be considered by…

  12. Three-dimensional reactor dynamics code for VVER type nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kyrki-Rajamaeki, R. [VTT Energy, Espoo (Finland)

    1995-10-01

    A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.).

  13. GeN-Foam: a novel OpenFOAM{sup ®} based multi-physics solver for 2D/3D transient analysis of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Fiorina, Carlo, E-mail: carlo.fiorina@psi.ch [Paul Scherrer Institut, Nuclear Energy and Safety Department, Laboratory for Reactor Physics and Systems Behaviour – PSI, Villigen 5232 (Switzerland); Clifford, Ivor [Paul Scherrer Institut, Nuclear Energy and Safety Department, Laboratory for Reactor Physics and Systems Behaviour – PSI, Villigen 5232 (Switzerland); Aufiero, Manuele [LPSC-IN2P3-CNRS/UJF/Grenoble INP, 53 avenue des Martyrs, 38026 Grenoble Cedex (France); Mikityuk, Konstantin [Paul Scherrer Institut, Nuclear Energy and Safety Department, Laboratory for Reactor Physics and Systems Behaviour – PSI, Villigen 5232 (Switzerland)

    2015-12-01

    Highlights: • Development of a new multi-physics solver based on OpenFOAM{sup ®}. • Tight coupling of thermal-hydraulics, thermal-mechanics and neutronics. • Combined use of traditional RANS and porous-medium models. • Mesh for neutronics deformed according to the predicted displacement field. • Use of three unstructured meshes, adaptive time step, parallel computing. - Abstract: The FAST group at the Paul Scherrer Institut has been developing a code system for reactor analysis for many years. For transient analysis, this code system is currently based on a state-of-the-art coupled TRACE-PARCS routine. This work presents an attempt to supplement the FAST code system with a novel solver characterized by tight coupling between the different equations, parallel computing capabilities, adaptive time-stepping and more accurate treatment of some of the phenomena involved in a reactor transient. The new solver is based on OpenFOAM{sup ®}, an open-source C++ library for the solution of partial differential equations using finite-volume discretization. It couples together a multi-scale fine/coarse mesh sub-solver for thermal-hydraulics, a multi-group diffusion sub-solver for neutronics, a displacement-based sub-solver for thermal-mechanics and a finite-difference model for the temperature field in the fuel. It is targeted toward the analysis of pin-based reactors (e.g., liquid metal fast reactors or light water reactors) or homogeneous reactors (e.g., fast-spectrum molten salt reactors). This paper presents each “single-physics” sub-solver and the overall coupling strategy, using the sodium-cooled fast reactor as a test case, and essential code verification tests are described.

  14. Development of computational methods for the safety assessment of gas-cooled high-temperature and supercritical light-water reactors. Final report; Rechenmethoden zur Bewertung der Sicherheit von gasgekuehlten Hochtemperaturreaktoren und superkritischen Leichtwasserreaktoren. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Buchholz, S.; Cron, D. von der; Hristov, H.; Lerchl, G.; Papukchiev, A.; Seubert, A.; Sureda, A.; Weis, J.; Weyermann, F.

    2012-12-15

    This report documents developments and results in the frame of the project RS1191 ''Development of computational methods for the safety assessment of gas-cooled high temperature and supercritical light-water reactors''. The report is structured according to the five work packages: 1. Reactor physics modeling of gas-cooled high temperature reactors; 2. Coupling of reactor physics and 3-D thermal hydraulics for the core barrel; 3. Extension of ATHLET models for application to supercritical reactors (HPLWR); 4. Further development of ATHLET for application to HTR; 5. Further development and validation of ANSYS CFX for application to alternative reactor concepts. Chapter 4 describes the extensions made in TORT-TD related to the simulation of pebble-bed HTR, e.g. spectral zone buckling, Iodine-Xenon dynamics, nuclear decay heat calculation and extension of the cross section interpolation algorithms to higher dimensions. For fast running scoping calculations, a time-dependent 3-D diffusion solver has been implemented in TORT-TD. For the PBMR-268 and PBMR-400 as well as for the HTR-10 reactor, appropriate TORT-TD models have been developed. Few-group nuclear cross sections have been generated using the spectral codes MICROX- 2 and DRAGON4. For verification and validation of nuclear cross sections and deterministic reactor models, MCNP models of reactor core and control rod of the HTR-10 have been developed. Comparisons with experimental data have been performed for the HTR-10 first criticality and control rod worth. The development of the coupled 3-D neutron kinetics and thermal hydraulics code system TORT-TD/ATTICA3D is documented in chapter 5. Similar to the couplings with ATHLET and COBRA-TF, the ''internal'' coupling approach has been implemented. Regarding the review of experiments and benchmarks relevant to HTR for validation of the coupled code system, the PBMR-400 benchmarks and the HTR-10 test reactor have been selected

  15. Control method for physical systems and devices

    Science.gov (United States)

    Guckenheimer, John

    1997-01-01

    A control method for stabilizing systems or devices that are outside the control domain of a linear controller is provided. When applied to nonlinear systems, the effectiveness of this method depends upon the size of the domain of stability that is produced for the stabilized equilibrium. If this domain is small compared to the accuracy of measurements or the size of disturbances within the system, then the linear controller is likely to fail within a short period. Failure of the system or device can be catastrophic: the system or device can wander far from the desired equilibrium. The method of the invention presents a general procedure to recapture the stability of a linear controller, when the trajectory of a system or device leaves its region of stability. By using a hybrid strategy based upon discrete switching events within the state space of the system or device, the system or device will return from a much larger domain to the region of stability utilized by the linear controller. The control procedure is robust and remains effective under large classes of perturbations of a given underlying system or device.

  16. Benchmark problems of start-up core physics of High Temperature Engineering Test Reactor (HTTR)

    Energy Technology Data Exchange (ETDEWEB)

    Yamashita, Kiyonobu; Nojiri, Naoki; Fujimoto, Nozomu; Nakano, Masaaki; Ando, Hiroei; Nagao, Yoshiharu; Nagaya, Yasunobu; Akino, Fujiyosi; Takeuchi, Mituo; Fujisaki, Shingo; Shiozawa, Shusaku [Japan Atomic Energy Research Institute JAERI, Ibaraki-ken (Japan)

    1998-09-01

    The experimental data of the HTTRs start-up core physics are useful to verify design codes of commercial HTGRs due to the similarities in the core size and excess reactivity. Form these viewpoints, it is significant to carry out the bench mark tests of design codes by using data of start-up core physics experiments planned for the HTTR. The evaluations of the first criticality, excess reactivity of annular cores, etc., are proposed for the benchmark problem. It was found from our precalculations that diffusion calculations provide larger excess reactivity and small number of fuel columns for the first criticality than Monte Carlo calculations. 19 refs.

  17. State of the art on nuclear heating measurement methods and expected improvements in zero power research reactors

    Directory of Open Access Journals (Sweden)

    Le Guillou Mael

    2017-01-01

    Full Text Available The paper focuses on the recent methodological advances suitable for nuclear heating measurements in zero power research reactors. This bibliographical work is part of an experimental approach currently in progress at CEA Cadarache, aiming at optimizing photon heating measurements in low-power research reactors. It provides an overview of the application fields of the most widely used detectors, namely thermoluminescent dosimeters (TLDs and optically stimulated luminescent dosimeters. Starting from the methodology currently implemented at CEA, the expected improvements relate to the experimental determination of the neutron component, which is a key point conditioning the accuracy of photon heating measurements in mixed n–γ field. A recently developed methodology based on the use of 7Li and 6Li-enriched TLDs, precalibrated both in photon and neutron fields, is a promising approach to deconvolute the two components of nuclear heating. We also investigate the different methods of optical fiber dosimetry, with a view to assess the feasibility of online photon heating measurements, whose primary benefit is to overcome constraints related to the withdrawal of dosimeters from the reactor immediately after irradiation. Moreover, a fibered setup could allow measuring the instantaneous dose rate during irradiation, as well as the delayed photon dose after reactor shutdown. Some insights from potential further developments are given. Obviously, any improvement of the technique has to lead to a measurement uncertainty at least equal to that of the currently used methodology (∼5% at 1σ.

  18. [New physical methods in osteoarthritis treatment].

    Science.gov (United States)

    Janczewska, Katarzyna; Klimkiewicz, Robert; Kubsik-Gidlewska, Anna; Jankowska, Agnieszka; Klimkiewicz, Paulina; Woldańska-Okońska, Marta

    2017-01-01

    Osteoarthritis is a chronic disease in which the pathological processes start from the catabolism of cartilage extracellular matrix and next extend on the whole joint. Therefore, it is important to diagnose the disease and determining treatment, selecting individually for each patient. The main health problems presents by every patients is pain, which decreases the everyday functioning and quality of life. The paper presents the definition of the disease and new therapeutic methods which improve the quality of life, as well as reduce intensity of pain.

  19. Engineering and Physics Optimization of Breed and Burn Fast Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Michael J. Driscoll; Pavel Hejzlar; Peter Yarsky; Dan Wachs; Kevan Weaver; Ken Czerwinski; Michael Pope; Cliff Davis; Theron Marshall; James Parry

    2005-12-09

    This project is organized under four major tasks (each of which has two or more subtasks) with contributions among the three collaborating organizations (MIT, INEEL and ANL-West): Task A: Core Physics and Fuel Cycle; Task B: Core Thermal Hydraulics; Task C: Plant Design Task; and D: Fuel Design.

  20. Evaluation of physical-chemical performance of an UASB reactor in removing pollutants of pig wastewater

    Directory of Open Access Journals (Sweden)

    Fabricio Moterani

    2010-04-01

    Full Text Available Attention has been given by governmental agencies concerning the swine production in confined areas, due to the pollution potential and problems related to epidemiology. Thus, anaerobic treatment system, similar to the one applied in this research, has became very important and raised interest for large scale production and field application. The purpose of this research was to evaluate the UASB reactor behavior considering a hydraulic retention time (HRT of 9.7 hours, the hydraulic loading rate (HLR of 2.5 m3 m-3 d-1, Organic loading rate (OLR of 1.77 kg m3 m-3 d-1 and the average biogas production of 437.08 L d-1. It was found in this work, that the alkalinity in the affluent and effluent were 1,383 mg L-1 and 1,442 mg L-1, respectively. The Ripley alkalinity in the affluent and effluent presented a relation of IA/PA of 1,5 e 1,7, respectively. The CODtotal concentration in the affluent and effluent was 2,705 mg L-1 and 1,849 mg L-1, respectively. The BOD5 concentration in the affluent and effluent was 707 mg L-1 and 317 mg L-1, respectively. The total phosphorus concentration was 1.07 mg L-1 and 1.11 mg L-1 and the concentration of total Kjeldahl nitrogen was 69 mg L-1 and 63 mg L-1, respectively. The CODtotal/BOD5 relation was 0.41 and the efficiency of total, fix and volatile solids removal was 40%, 28%, 48%, respectively. The treatment system presented a good performance and therefore the operational parameters applied could be also useful for large scale systems.

  1. Education and Training on ISIS Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Foulon, F.; Badeau, G.; Lescop, B.; Wohleber, X. [French Atomic Energy and Alternative Energies Commission, Paris (France)

    2013-07-01

    In the frame of academic and vocational programs the National Institute for Nuclear Science and Technology uses the ISIS research reactor as a major tool to ensure a practical and comprehensive understanding of the nuclear reactor physics, principles and operation. A large set of training courses have been developed on ISIS, optimising both the content of the courses and the pedagogical approach. Programs with duration ranging from 3 hours (introduction to reactor operation) to 24 hours (full program for the future operators of research reactors) are carried out on ISIS reactor. The reactor is operated about 350 hours/year for education and training, about 40 % of the courses being carried out in English. Thus, every year about 400 trainees attend training courses on ISIS reactor. We present here the ISIS research reactor and the practical courses that have been developed on ISIS reactor. Emphasis is given to the pedagogical method which is used to focus on the operational and safety aspects, both in normal and incidental operation. We will present the curricula of the academic and vocational courses in which the practical courses are integrated, the courses being targeted to a wide public, including operators of research reactors, engineers involved in the design and operation of nuclear reactors as well as staff of the regulatory body. We address the very positive impact of the courses on the development of the competences and skills of participants. Finally, we describe the Internet Reactor Laboratories (IRL) that are under development and will consist in broadcasting the training courses via internet to remote facilities or institutions.

  2. Mathematical and computational methods in nuclear physics

    Energy Technology Data Exchange (ETDEWEB)

    Dehesa, J.S.; Gomez, J.M.G.; Polls, A.

    1983-01-01

    The lectures, covering various aspects of the many-body problem in nuclei, review present knowledge and include some unpublished material as well. Bohigas and Giannoni discuss the fluctuation properties of spectra of many-body systems by means of random matrix theories, and the attempts to search for quantum mechanical manifestations of classical chaotic motion. The role of spectral distributions (expressed as explicit functions of the microscopic matrix elements of the Hamiltonian) in the statistical spectroscopy of nuclear systems is analyzed by French. Zucker, after a brief review of the theoretical basis of the shell model, discusses a reformulation of the theory of effective interactions and gives a survey of the linked cluster theory. Goeke's lectures center on the mean-field methods, particularly TDHF, used in the investigation of the large-amplitude nuclear collective motion, pointing out both the successes and failures of the theory.

  3. Complex risk analysis for loss of electric power in liquid metal nuclear reactor by system dynamics (SD) method

    Energy Technology Data Exchange (ETDEWEB)

    Woo, Tae Ho [Seoul National Univ. (Korea, Republic of). Dept. of Nuclear Engineering

    2012-07-15

    The power stabilization of the nuclear power plants (NPPs) is investigated in the aspect of the liquid metal coolant. The quantification of the risk analysis is performed by the system dynamics (SD) method which is processed by the feedback and accumulation complex algorithms. The Vensim software package is used for the simulations, which is supported by the Monte-Carlo method. There are 2 kinds of considerations as the economic and safety properties. The result shows the stability of the operations when the power can be decided. This shows the higher efficiency of the reactor. The failure frequency is 16/60 = 27%. In the event of Power Stabilized, the failure event is in the quite lower frequency rate. The commercial use of the reactor is important in the operations. (orig.)

  4. Methods for modeling impact-induced reactivity changes in small reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Tallman, Tyler N.; Radel, Tracy E.; Smith, Jeffrey A.; Villa, Daniel L.; Smith, Brandon M. (U. of Wisconsin, Madison, WI); Radel, Ross F.; Lipinski, Ronald J.; Wilson, Paul Philip Hood (U. of Wisconsin, Madison, WI)

    2010-10-01

    This paper describes techniques for determining impact deformation and the subsequent reactivity change for a space reactor impacting the ground following a potential launch accident or for large fuel bundles in a shipping container following an accident. This technique could be used to determine the margin of subcriticality for such potential accidents. Specifically, the approach couples a finite element continuum mechanics model (Pronto3D or Presto) with a neutronics code (MCNP). DAGMC, developed at the University of Wisconsin-Madison, is used to enable MCNP geometric queries to be performed using Pronto3D output. This paper summarizes what has been done historically for reactor launch analysis, describes the impact criticality analysis methodology, and presents preliminary results using representative reactor designs.

  5. Using the case-study method in teaching college physics

    CERN Document Server

    Burko, Lior M

    2016-01-01

    The case-study teaching method has a long history (starting at least with Socrates), and wide current use in business schools, medical schools, law schools, and a variety of other disciplines. However, relatively little use is made of it in the physical sciences, specifically in physics or astronomy. The case-study method should be considered by physics faculty as part of the effort to transition the teaching of college physics from the traditional frontal-lecture format to other formats that enhance active student participation. In this paper we endeavor to interest physics instructors in the case-study method, and hope that it would also serve as a call for more instructors to produce cases that they use in their own classes and that can also be adopted by other instructors.

  6. Using the Case Study Method in Teaching College Physics

    Science.gov (United States)

    Burko, Lior M.

    2016-10-01

    The case study teaching method has a long history (starting at least with Socrates) and wide current use in business schools, medical schools, law schools, and a variety of other disciplines. However, relatively little use is made of it in the physical sciences, specifically in physics or astronomy. The case study method should be considered by physics faculty as part of the effort to transition the teaching of college physics from the traditional frontal-lecture format to other formats that enhance active student participation. In this paper we endeavor to interest physics instructors in the case study method, and hope that it would also serve as a call for more instructors to produce cases that they use in their own classes and that can also be adopted by other instructors.

  7. Physical Model Method for Seismic Study of Concrete Dams

    Directory of Open Access Journals (Sweden)

    Bogdan Roşca

    2008-01-01

    Full Text Available The study of the dynamic behaviour of concrete dams by means of the physical model method is very useful to understand the failure mechanism of these structures to action of the strong earthquakes. Physical model method consists in two main processes. Firstly, a study model must be designed by a physical modeling process using the dynamic modeling theory. The result is a equations system of dimensioning the physical model. After the construction and instrumentation of the scale physical model a structural analysis based on experimental means is performed. The experimental results are gathered and are available to be analysed. Depending on the aim of the research may be designed an elastic or a failure physical model. The requirements for the elastic model construction are easier to accomplish in contrast with those required for a failure model, but the obtained results provide narrow information. In order to study the behaviour of concrete dams to strong seismic action is required the employment of failure physical models able to simulate accurately the possible opening of joint, sliding between concrete blocks and the cracking of concrete. The design relations for both elastic and failure physical models are based on dimensional analysis and consist of similitude relations among the physical quantities involved in the phenomenon. The using of physical models of great or medium dimensions as well as its instrumentation creates great advantages, but this operation involves a large amount of financial, logistic and time resources.

  8. Physics study on recycling of ThO{sub 2}/UO{sub 2} fuel in CANDU reactors through dry reprocess technology

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Park, Chang Je; Jeong, Chang Joon

    2003-06-01

    The dry process fuel technology has high proliferation-resistance which is one of important goals of Generation-IV (Gen-IV) reactor development. It is expected that the dry process fuel technology can be applied not only to existing nuclear systems but also to future nuclear systems. In this report, the homogeneous ThO{sub 2}-UO{sub 2} fuel cycle option of a CANDU reactor has been studied, including the physics analysis of recycling spent fuel. Reactivity swing and variation of isotopic content with irradiation are reported for various cases of initial uranium loadings. It was found that natural uranium saving increases significantly by recycling thorium/uranium fuel and it is feasible to recycle thorium with the dry process technology in a CANDU reactor. It is, however, required to further investigate the dry process that can be applied to the thorium-abundant dioxide fuel.

  9. NEUTRONIC REACTORS

    Science.gov (United States)

    Wigner, E.P.; Young, G.J.

    1958-10-14

    A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

  10. Exhaled breath analysis: physical methods, instruments, and medical diagnostics

    Science.gov (United States)

    Vaks, V. L.; Domracheva, E. G.; Sobakinskaya, E. A.; Chernyaeva, M. B.

    2014-07-01

    This paper reviews the analysis of exhaled breath, a rapidly growing field in noninvasive medical diagnostics that lies at the intersection of physics, chemistry, and medicine. Current data are presented on gas markers in human breath and their relation to human diseases. Various physical methods for breath analysis are described. It is shown how measurement precision and data volume requirements have stimulated technological developments and identified the problems that have to be solved to put this method into clinical practice.

  11. Measurements Methods for the analysis of Nuclear Reactors Thermal Hydraulic in Water Scaled Facilities

    Science.gov (United States)

    Spaccapaniccia, C.; Planquart, P.; Buchlin, J. M. AB(; ), AC(; )

    2018-01-01

    The Belgian nuclear research institute (SCK•CEN) is developing MYRRHA. MYRRHA is a flexible fast spectrum research reactor, conceived as an accelerator driven system (ADS). The configuration of the primary loop is pool-type: the primary coolant and all the primary system components (core and heat exchangers) are contained within the reactor vessel, while the secondary fluid is circulating in the heat exchangers. The primary coolant is Lead Bismuth Eutectic (LBE). The recent nuclear accident of Fukushima in 2011 changed the requirements for the design of new reactors, which should include the possibility to remove the residual decay heat through passive primary and secondary systems, i.e. natural convection (NC). After the reactor shut down, in the unlucky event of propeller failures, the primary and secondary loops should be able to remove the decay heat in passive way (Natural Convection). The present study analyses the flow and the temperature distribution in the upper plenum by applying laser imaging techniques in a laboratory scaled water model. A parametric study is proposed to study stratification mitigation strategies by varying the geometry of the buffer tank simulating the upper plenum.

  12. Theoretical Study of Palladium Membrane Reactor Performance During Propane Dehydrogenation Using CFD Method

    Directory of Open Access Journals (Sweden)

    Kamran Ghasemzadeh

    2017-04-01

    Full Text Available This study presents a 2D-axisymmetric computational fluid dynamic (CFD model to investigate the performance Pd membrane reactor (MR during propane dehydrogenation process for hydrogen production. The proposed CFD model provided the local information of temperature and component concentration for the driving force analysis. After investigation of mesh independency of CFD model, the validation of CFD model results was carried out by other modeling data and a good agreement between CFD model results and theoretical data was achieved. Indeed, in the present model, a tubular reactor with length of 150 mm was considered, in which the Pt-Sn-K/Al2O3 as catalyst were filled in reaction zone. Hence, the effects of the important operating parameter (reaction temperature on the performances of membrane reactor (MR were studied in terms of propane conversion and hydrogen yield. The CFD results showed that the suggested MR system during propane dehydrogenation reaction presents higher performance with respect to once obtained in the conventional reactor (CR. In particular, by applying Pd membrane, was found that propane conversion can be increased from 41% to 49%. Moreover, the highest value of propane conversion (X = 91% was reached in case of Pd-Ag MR. It was also established that the feed flow rate of the MR is to be the one of the most important factors defining efficiency of the propane dehydrogenation process.

  13. The development and application of advanced analytical methods to commercial ICF reactor chambers. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Cousseau, P.; Engelstad, R.; Henderson, D.L. [and others

    1997-10-01

    Progress is summarized in this report for each of the following tasks: (1) multi-dimensional radiation hydrodynamics computer code development; (2) 2D radiation-hydrodynamic code development; (3) ALARA: analytic and Laplacian adaptive radioactivity analysis -- a complete package for analysis of induced activation; (4) structural dynamics modeling of ICF reactor chambers; and (5) analysis of self-consistent target chamber clearing.

  14. A Method for Evaluating Physical Activity Programs in Schools.

    Science.gov (United States)

    Kelly, Cheryl; Carpenter, Dick; Tucker, Elizabeth; Luna, Carmen; Donovan, John; Behrens, Timothy K

    2017-09-14

    Providing opportunities for students to be physically active during the school day leads to increased academic performance, better focus, and fewer behavioral problems. As schools begin to incorporate more physical activity programming into the school day, evaluators need methods to measure how much physical activity students are being offered through this programming. Because classroom-based physical activity is often offered in 3-minute to 5-minute bouts at various times of the day, depending on the teachers' time to incorporate it, it is a challenge to evaluate this activity. This article describes a method to estimate the number of physical activity minutes provided before, during, and after school. The web-based tool can be used to gather data cost-effectively from a large number of schools. Strategies to increase teacher response rates and assess intensity of activity should be explored.

  15. Computer methods in physics 250 problems with guided solutions

    CERN Document Server

    Landau, Rubin H

    2018-01-01

    Our future scientists and professionals must be conversant in computational techniques. In order to facilitate integration of computer methods into existing physics courses, this textbook offers a large number of worked examples and problems with fully guided solutions in Python as well as other languages (Mathematica, Java, C, Fortran, and Maple). It’s also intended as a self-study guide for learning how to use computer methods in physics. The authors include an introductory chapter on numerical tools and indication of computational and physics difficulty level for each problem.

  16. Study on vibration characteristics and fault diagnosis method of oil-immersed flat wave reactor in Arctic area converter station

    Science.gov (United States)

    Lai, Wenqing; Wang, Yuandong; Li, Wenpeng; Sun, Guang; Qu, Guomin; Cui, Shigang; Li, Mengke; Wang, Yongqiang

    2017-10-01

    Based on long term vibration monitoring of the No.2 oil-immersed fat wave reactor in the ±500kV converter station in East Mongolia, the vibration signals in normal state and in core loose fault state were saved. Through the time-frequency analysis of the signals, the vibration characteristics of the core loose fault were obtained, and a fault diagnosis method based on the dual tree complex wavelet (DT-CWT) and support vector machine (SVM) was proposed. The vibration signals were analyzed by DT-CWT, and the energy entropy of the vibration signals were taken as the feature vector; the support vector machine was used to train and test the feature vector, and the accurate identification of the core loose fault of the flat wave reactor was realized. Through the identification of many groups of normal and core loose fault state vibration signals, the diagnostic accuracy of the result reached 97.36%. The effectiveness and accuracy of the method in the fault diagnosis of the flat wave reactor core is verified.

  17. Modern methodic of power cardio training in students’ physical education

    Directory of Open Access Journals (Sweden)

    Osipov A.Yu.

    2016-12-01

    Full Text Available Purpose: significant increase of students’ physical condition and health level at the account of application of modern power cardio training methodic. Material: 120 students (60 boys and 60 girls participated in the research. The age of the tested was 19 years. The research took one year. We used methodic of power and functional impact on trainees’ organism (HOT IRON. Such methodic is some systems of physical exercises with weights (mini-barbells, to be fulfilled under accompaniment of specially selected music. Results: we showed advantages of power-cardio and fitness trainings in students’ health improvement and in elimination obesity. Control tests showed experimental group students achieved confidently higher physical indicators. Boys demonstrated increase of physical strength and general endurance indicators. Girls had confidently better indicators of physical strength, flexibility and general endurance. Increase of control group students’ body mass can be explained by students’ insufficient physical activity at trainings, conducted as per traditional program. Conclusions: students’ trainings by power-cardio methodic with application HOT IRON exercises facilitate development the following physical qualities: strength and endurance in boys and strength, flexibility and endurance in girls. Besides, it was found that such systems of exercises facilitate normalization of boys’ body mass and correction of girls’ constitution.

  18. Study on disposal method of graphite blocks and storage of spent fuel for modular gas-cooled reactor. Joint research

    Energy Technology Data Exchange (ETDEWEB)

    Sumita, Junya; Sawa, Kazuhiro; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tsuchie, Yasuo; Urakami, Masao [Japan Atomic Power Co., Tokyo (Japan)

    2003-02-01

    This report describes the result of study on disposal method of graphite blocks in future block-type reactor. Present study was carried out within a framework of joint research, ''Research of Modular High Temperature Gas-cooled Reactors (No. 3)'', between Japan Atomic Energy Research Institute (JAERI) and the Japan Atomic Power Company (JAPCO), in 2000. In this study, activities in fuel and reflector graphite blocks were evaluated and were compared with the disposal limits defined as low-level of radioactive waste. As a result, it was found that the activity for only C-14 was higher than disposal limits for the low-level of radioactive waste and that the amount of air in the graphite is important to evaluate precisely of C-14 activity. In addition, spent fuels can be stored in air-cooled condition at least after two years cooling in the storage pool. (author)

  19. Simulation of the spatial distribution of the acoustic pressure in sonochemical reactors with numerical methods: a review.

    Science.gov (United States)

    Tudela, Ignacio; Sáez, Verónica; Esclapez, María Deseada; Díez-García, María Isabel; Bonete, Pedro; González-García, José

    2014-05-01

    Numerical methods for the calculation of the acoustic field inside sonoreactors have rapidly emerged in the last 15 years. This paper summarizes some of the most important works on this topic presented in the past, along with the diverse numerical works that have been published since then, reviewing the state of the art from a qualitative point of view. In this sense, we illustrate and discuss some of the models recently developed by the scientific community to deal with some of the complex events that take place in a sonochemical reactor such as the vibration of the reactor walls and the nonlinear phenomena inherent to the presence of ultrasonic cavitation. In addition, we point out some of the upcoming challenges that must be addressed in order to develop a reliable tool for the proper designing of efficient sonoreactors and the scale-up of sonochemical processes. Copyright © 2013 Elsevier B.V. All rights reserved.

  20. Core map generation for the ITU TRIGA Mark II research reactor using Genetic Algorithm coupled with Monte Carlo method

    Energy Technology Data Exchange (ETDEWEB)

    Türkmen, Mehmet, E-mail: tm@hacettepe.edu.tr [Nuclear Engineering Department, Hacettepe University, Beytepe Campus, Ankara (Turkey); Çolak, Üner [Energy Institute, Istanbul Technical University, Ayazağa Campus, Maslak, Istanbul (Turkey); Ergün, Şule [Nuclear Engineering Department, Hacettepe University, Beytepe Campus, Ankara (Turkey)

    2015-12-15

    Highlights: • Optimum core maps were generated for the ITU TRIGA Mark II Research Reactor. • Calculations were performed using a Monte Carlo based reactor physics code, MCNP. • Single-Objective and Multi-Objective Genetic Algorithms were used for the optimization. • k{sub eff} and ppf{sub max} were considered as the optimization objectives. • The generated core maps were compared with the fresh core map. - Abstract: The main purpose of this study is to present the results of Core Map (CM) generation calculations for the İstanbul Technical University TRIGA Mark II Research Reactor by using Genetic Algorithms (GA) coupled with a Monte Carlo (MC) based-particle transport code. Optimization problems under consideration are: (i) maximization of the core excess reactivity (ρ{sub ex}) using Single-Objective GA when the burned fuel elements with no fresh fuel elements are used, (ii) maximization of the ρ{sub ex} and minimization of maximum power peaking factor (ppf{sub max}) using Multi-Objective GA when the burned fuels with fresh fuels are used. The results were obtained when all the control rods are fully withdrawn. ρ{sub ex} and ppf{sub max} values of the produced best CMs were provided. Core-averaged neutron spectrum, and variation of neutron fluxes with respect to radial distance were presented for the best CMs. The results show that it is possible to find an optimum CM with an excess reactivity of 1.17 when the burned fuels are used. In the case of a mix of burned fuels and fresh fuels, the best pattern has an excess reactivity of 1.19 with a maximum peaking factor of 1.4843. In addition, when compared with the fresh CM, the thermal fluxes of the generated CMs decrease by about 2% while change in the fast fluxes is about 1%.Classification: J. Core physics.

  1. As-Run Physics Analysis for the UCSB-1 Experiment in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nielsen, Joseph Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    The University of California Santa Barbara (UCSB) -1 experiment was irradiated in the A-10 position of the ATR. The experiment was irradiated during cycles 145A, 145B, 146A, and 146B. Capsule 6A was removed from the test train following Cycle 145A and replaced with Capsule 6B. This report documents the as-run physics analysis in support of Post-Irradiation Examination (PIE) of the test. This report documents the as-run fluence and displacements per atom (DPA) for each capsule of the experiment based on as-run operating history of the ATR. Average as-run heating rates for each capsule are also presented in this report to support the thermal analysis.

  2. Equipment for neutron measurements at VR-1 Sparrow training reactor.

    Science.gov (United States)

    Kolros, Antonin; Huml, Ondrej; Kríz, Martin; Kos, Josef

    2010-01-01

    The VR-1 sparrow reactor is an experimental nuclear facility for training, student education and teaching purposes. The sparrow reactor is an educational platform for the basic experiments at the reactor physic and dosimetry. The aim of this article is to describe the new experimental equipment EMK310 features and possibilities for neutron detection by different gas filled detectors at VR-1 reactor. Among the EMK310 equipment typical attributes belong precise set-up, simple control, resistance to electromagnetic interference, high throughput (counting rate), versatility and remote controllability. The methods for non-linearity correction of pulse neutron detection system and reactimeter application are presented. Copyright 2009. Published by Elsevier Ltd.

  3. Development and application of neutron transport methods and uncertainty analyses for reactor core calculations. Technical report; Entwicklung und Einsatz von Neutronentransportmethoden und Unsicherheitsanalysen fuer Reaktorkernberechnungen. Technischer Bericht

    Energy Technology Data Exchange (ETDEWEB)

    Zwermann, W.; Aures, A.; Bernnat, W.; and others

    2013-06-15

    This report documents the status of the research and development goals reached within the reactor safety research project RS1503 ''Development and Application of Neutron Transport Methods and Uncertainty Analyses for Reactor Core Calculations'' as of the 1{sup st} quarter of 2013. The superordinate goal of the project is the development, validation, and application of neutron transport methods and uncertainty analyses for reactor core calculations. These calculation methods will mainly be applied to problems related to the core behaviour of light water reactors and innovative reactor concepts. The contributions of this project towards achieving this goal are the further development, validation, and application of deterministic and stochastic calculation programmes and of methods for uncertainty and sensitivity analyses, as well as the assessment of artificial neutral networks, for providing a complete nuclear calculation chain. This comprises processing nuclear basis data, creating multi-group data for diffusion and transport codes, obtaining reference solutions for stationary states with Monte Carlo codes, performing coupled 3D full core analyses in diffusion approximation and with other deterministic and also Monte Carlo transport codes, and implementing uncertainty and sensitivity analyses with the aim of propagating uncertainties through the whole calculation chain from fuel assembly, spectral and depletion calculations to coupled transient analyses. This calculation chain shall be applicable to light water reactors and also to innovative reactor concepts, and therefore has to be extensively validated with the help of benchmarks and critical experiments.

  4. APPLICATIONS OF LASERS AND OTHER TOPICS IN LASER PHYSICS AND TECHNOLOGY: Hybrid reactor based on laser thermonuclear fusion

    Science.gov (United States)

    Basov, N. G.; Belousov, N. I.; Grishunin, P. A.; Kalmykov, Yu K.; Lebo, I. G.; Rozanov, Vladislav B.; Sklizkov, G. V.; Subbotin, V. I.; Finkel'shteĭn, K. I.; Kharitonov, V. V.; Sherstnev, K. B.

    1987-10-01

    A physicotechnical and parametric analysis is used as the basis for a conceptual design of a thermonuclear inertial-confinement hybrid reactor as a breeder of fuel for fission nuclear power stations. It is proposed to use a laser as a driver in this reactor.

  5. Comparison of four NDT methods for indication of reactor steel degradation by high fluences of neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Tomáš, I., E-mail: tomas@fzu.cz [Institute of Physics ASCR, Na Slovance 2, Prague 18221 (Czech Republic); Vértesy, G. [Research Centre for Natural Sciences, Institute of Technical Physics and Materials Science, Konkoly Thege Miklós út 29-33, H-1121 Budapest (Hungary); Pirfo Barroso, S. [KFKI Atomic Energy Research Institute, Konkoly Thege Miklós út 29-33, H-1121 Budapest (Hungary); The Open University, Walton Hall, MK92BS Milton Keynes (United Kingdom); Kobayashi, S. [Department of Materials Science and Engineering, Faculty of Engineering, Iwate University, Morioka 020-8551 (Japan)

    2013-12-15

    Highlights: • Results of 4 NDT methods on highly irradiated steel are normalized and compared. • Two of the methods (MAT and HV) correlate well with DBTT. • Magnetic Adaptive Testing gives the most sensitive and the best correlated results. • Measurements and sample shapes for an NDT surveillance program are suggested. - Abstract: Results of three magnetic nondestructive methods, Magnetic Barkhausen Emission (MBE), magnetic minor loops Power Scaling Laws (PSL) and Magnetic Adaptive Testing (MAT), and of one reference mechanical measurement, Vickers Hardness (HV), applied on the same series of neutron heavily irradiated nuclear reactor pressure vessel steel materials, were normalized and presented here for the purpose of their straightforward quantitative mutual comparison. It is uncommon to carry out different round-robin testing on irradiated materials, and if not answering all open questions, the comparison alone justifies this paper. The assessment methods were all based on ferromagnetism, although each of them used a different aspect of it. The presented comparison yielded a justified recommendation of the most reliable nondestructive method for indication of the reactor steel irradiation hardening and embrittlement effects. The A533 type B Class 1 steel (JRQ), and the base (15Kh2MFA) and welding (10KhMFT) steels for the WWER 440-type Russian reactors were used for the investigations. The samples were irradiated by high-energy neutrons (>1 MeV) with up to 11.9 × 10{sup 19} n/cm{sup 2} fluences. From all the applied measurements, the results of MAT produced the most satisfactory correlation with independently measured ductile-brittle-transition temperature (DBTT) values of the steel. The other two magnetic methods showed a weaker correlation with DBTT, but some other aspects and information could be assessed by them. As MAT and MBE were sensitive to uncontrolled fluctuation of surface quality of the steel, contact-less ways of testing and more

  6. 75 FR 67636 - Physical Protection of Shipments of Irradiated Reactor Fuel

    Science.gov (United States)

    2010-11-03

    ... the following methods. Federal Rulemaking Web site: Go to http://www.regulations.gov and search for... electronic form will be posted on the NRC Web site and on the Federal rulemaking Web site http://www..., the public can gain entry into ADAMS, which provides text and image files of NRC's public documents...

  7. Identification of active sonochemical zones in a triple frequency ultrasonic reactor via physical and chemical characterization techniques.

    Science.gov (United States)

    Tiong, T Joyce; Liew, Derick K L; Gondipon, Ramona C; Wong, Ryan W; Loo, Yuen Ling; Lok, Matthew S T; Manickam, Sivakumar

    2017-03-01

    Coupling multiple frequencies in ultrasonic systems is one of the highly desired area of research for sonochemists, as it is known for producing synergistic effects on various ultrasonic reactions. In this study, the characteristics of a hexagonal-shaped triple frequency ultrasonic reactor with the combination frequencies of 28, 40 and 70kHz were studied. The results showed that uniform temperature increment was achieved throughout the reactor at all frequency combinations. On the other hand, sonochemiluminescence emission and degradation rate of Rhodamine B varies throughout different areas of the reactor, indicating the presence of acoustic 'hot spots' at certain areas of the reactor. Also, coupling dual and triple frequencies showed a decrease in the hydroxyl radical (OH) production, suggesting probable wave cancelling effect in the system. The results can therefore be served as a guide to optimize the usage of a triple frequency ultrasonic reactor for future applications. Copyright © 2016 Elsevier B.V. All rights reserved.

  8. Interactive Methods of Teaching Physics at Technical Universities

    Science.gov (United States)

    Krišták, L'uboš; Nemec, Miroslav; Danihelová, Zuzana

    2014-01-01

    The paper presents results of "non-traditional" teaching of the basic course of Physics in the first year of study at the Technical University in Zvolen, specifically teaching via interactive method enriched with problem tasks and experiments. This paper presents also research results of the use of the given method in conditions of…

  9. Application of nondestructive methods for qualification of high density fuels in the IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Jose E.R.; Silva, Antonio T.; Domingos, Douglas B.; Terremoto, Luis A.A., E-mail: jersilva@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    IPEN/CNEN-SP manufactures fuels to be used in its research reactor - the IEA-R1. To qualify those fuels, it is necessary to check if they have a good performance under irradiation. As Brazil still does not have nuclear research reactors with high neutron fluxes, or suitable hot cells for carrying out post-irradiation examination of nuclear fuels, IPEN/CNEN-SP has conducted a fuel qualification program based on the use of uranium compounds (U{sub 3}O{sub 8} and U{sub 3}Si{sub 2} dispersed in Al matrix) internationally tested and qualified to be used in research reactors, and has attained experience in the technological development stages for the manufacturing of fuel plates, irradiation and non-destructive post-irradiation testing. Fuel elements containing low volume fractions of fuel in the dispersion were manufactured and irradiated successfully directly in the core of the IEA-R1. However, there are plans at IPEN/CNEN-SP to increase the uranium density of the fuels. Ten fuel miniplates (five containing U{sub 3}O{sub 8}-Al and five containing U{sub 3}Si{sub 2}-Al), with densities of 3.2 gU/cm{sup 3} and 4.8 gU/cm{sup 3} respectively, are being irradiated inside an irradiation device placed in a peripheral position of the IEA-R1 core. Non-destructive methods will be used to evaluate irradiation performance of the fuel miniplates after successive cycles of irradiation, by means: monitoring the reactor parameters during operation; periodic underwater visual inspection of fuel miniplates, eventual sipping test for fuel miniplates suspected of leakage and underwater measuring of the miniplate thickness for assessment of the fuel miniplate swelling. (author)

  10. Evaluation of temperature distribution sensing method for fast reactor using optical fiber

    Energy Technology Data Exchange (ETDEWEB)

    Kimura, Atsushi; Nakazawa, Masaharu [Tokyo Univ. (Japan); Ichige, Satoshi [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1999-12-01

    Optical fiber sensors (OFSs) have many advantages like flexible configuration, intrinsic immunity for electromagnetic fields, and so on. For these reasons, it is very useful to apply OFSs to fast reactor plants for remote inspection and surveillance. However, under irradiation, because of radiation-induced transmission loss of optical fibers, OFSs have radiation-induced errors. Therefore, to apply OFSs to nuclear facilities, we have to estimate and correct the errors. In this report, Raman Distributed Temperature Sensor (RDTS; one of the OFSs) has been installed at the primary coolant loop of the experimental fast reactor JOYO of JNC (Japan Nuclear Cycle Development Institute). Two correction techniques (correction technique with two thermocouples and correction technique with loop arrangement) for radiation-induced errors have been developed and demonstrated. Because of the radiation-induced loss, measured temperature distributions had radiation-induced errors. However, during the continuous measurements with the total dose of more than 8 x 10{sup 3}[C/kg](3 x 10{sup 7}[R]), the radiation induced errors showed a saturation tendency. In case of the temperature distributions with fluorine doped fiber, with one of the correction techniques, the temperature errors reduced to 1{approx}2degC and the feasibility of the loss correction techniques was demonstrated. For these results, it can be said that RDTS can be applied as a temperature distribution monitor in harsh radiation environments like fast reactor plants. (author)

  11. Review and future outlook of Dragon project/signatory organisations' collaborations in the integral reactor physics experiments - December 1973

    Energy Technology Data Exchange (ETDEWEB)

    Gutmann, H.

    1974-10-15

    The paper provides an overview of the collaborative reactor physics experiments conducted in the DRAGON Countries as of December 1973 summarizing those that have been conducted for high enriched uranium/thorium systems, those being conducted for low enriched uranium and plutonium systems, those conducted with irradiated fuel, and those on-going with integral fuel blocks. A list of relevant reports and papers on experiments is provided.

  12. From Talk to Experience: Transforming the Preservice Physics Methods Course

    Directory of Open Access Journals (Sweden)

    Tom Russell

    2010-07-01

    Full Text Available This report of a collaborative self-study describes and interprets our pedagogical approach at the beginning of a preservice physics methods course and outlines the strategy that we used to create a context for productive learning. We focus on our attempt to engage teacher candidates in dialogue about learning physics and learning to teach physics by engaging them in brief teaching experiences in the first month of a preservice teacher education program, before the first practicum placement. Self-study methodologies are used to frame and reframe our perceptions of teaching and learning as we enacted a pedagogy of teacher education that was unfamiliar both to us and to our teacher candidates.Keywords: self-study of teacher education practices, lesson study, teacher education, physics, curriculum methods

  13. Physical methods of nucleic acid transfer: general concepts and applications.

    Science.gov (United States)

    Villemejane, Julien; Mir, Lluis M

    2009-05-01

    Physical methods of gene (and/or drug) transfer need to combine two effects to deliver the therapeutic material into cells. The physical methods must induce reversible alterations in the plasma membrane to allow the direct passage of the molecules of interest into the cell cytosol. They must also bring the nucleic acids in contact with the permeabilized plasma membrane or facilitate access to the inside of the cell. These two effects can be achieved in one or more steps, depending upon the methods employed. In this review, we describe and compare several physical methods: biolistics, jet injection, hydrodynamic injection, ultrasound, magnetic field and electric pulse mediated gene transfer. We describe the physical mechanisms underlying these approaches and discuss the advantages and limitations of each approach as well as its potential application in research or in preclinical and clinical trials. We also provide conclusions, comparisons, and projections for future developments. While some of these methods are already in use in man, some are still under development or are used only within clinical trials for gene transfer. The possibilities offered by these methods are, however, not restricted to the transfer of genes and the complementary uses of these technologies are also discussed. As these methods of gene transfer may bypass some of the side effects linked to viral or biochemical approaches, they may find their place in specific clinical applications in the future.

  14. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

    This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

  15. Coffee research using optical methods of physical and chemical analysis

    OpenAIRE

    Tikhonov B.; Kuznetsov V.

    2016-01-01

    The paper discusses aspects of application of optical methods of physical and chemical analysis to determine the identity of the functional groups of substances extraction at different ways of making coffee. Infrared spectroscopy, diffusion reflectance spectrophotometry and refractometry methods were used for the researches. A comparison of coffee preparation methods was conducted in cupping and aeropresse Aerobie Aeropress Coffee Maker. Temperature of the water used to brew, and the punc...

  16. Massive computation methodology for reactor operation (MACRO)

    Energy Technology Data Exchange (ETDEWEB)

    Gustavsson, Cecilia; Pomp, Stephan; Sjoestrand, Henrik; Wallin, Gustav; Oesterlund, Michael [Division of applied nuclear physics, Department of physics and astronomy, Uppsala University, Laegerhyddsvaegen 1, 751 20 Uppsala (Sweden); Koning, Arjan; Rochman, Dimitri [Nuclear Research and consultancy Group (NRG) Westerduinweg 3, Petten (Netherlands); Bejmer, Klaes-Hakan [Vattenfall Nuclear Fuel AB, Jaemtlandsgatan 99, Vaellingby (Sweden); Henriksson, Hans [Vattenfall Research and Development AB, Jaemtlandsgatan 99, Vaellingby (Sweden)

    2010-07-01

    Today, nuclear data libraries do not handle uncertainties from nuclear data in a consistent manner and the reactor codes do not request uncertainties in nuclear data input. Thus, the output from these codes have unknown uncertainties. The plan is to use a method proposed by Koning and Rochman to investigate the propagation of nuclear data uncertainties into reactor physics codes and macroscopic parameters. A project (acronym MACRO) has started at Uppsala University in collaboration with A. Koning and with financial support from Vattenfall AB and the Swedish Research Council within the GENIUS (Generation IV research in universities of Sweden) project. In the proposed method the uncertainties in nuclear model parameters will be derived from theoretical considerations and comparisons of nuclear model results with experimental cross-section data. Given the probability distribution in the model parameters a large set of random, complete ENDF-formatted nuclear data libraries will be created using the TALYS code. The generated nuclear data libraries will then be used in neutron transport codes to obtain macroscopic reactor parameters. For this, models of reactor systems with proper geometry and elements will be used. This will be done for all data libraries and the variation of the final results will be regarded as a systematic uncertainty in the investigated reactor parameter. The understanding of these systematic uncertainties is especially important for the design and intercomparison of new reactor concepts, i.e., Generation IV, and optimization applications for current generation reactors is envisaged. (authors)

  17. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    Energy Technology Data Exchange (ETDEWEB)

    P. Delmolino

    2005-05-06

    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  18. Recent advances in physics and technology of ion cyclotron resonance heating in view of future fusion reactors

    Science.gov (United States)

    Ongena, J.; Messiaen, A.; Kazakov, Ye O.; Koch, R.; Ragona, R.; Bobkov, V.; Crombé, K.; Durodié, F.; Goniche, M.; Krivska, A.; Lerche, E.; Louche, F.; Lyssoivan, A.; Vervier, M.; Van Eester, D.; Van Schoor, M.; Wauters, T.; Wright, J.; Wukitch, S.

    2017-05-01

    Ion temperatures of over 100 million degrees need to be reached in future fusion reactors for the deuterium-tritium fusion reaction to work. Ion cyclotron resonance heating (ICRH) is a method that has the capability to directly heat ions to such high temperatures, via a resonant interaction between the plasma ions and radiofrequency waves launched in the plasma. This paper gives an overview of recent developments in this field. In particular a novel and recently developed three-ion heating scenario will be highlighted. It is a flexible scheme with the potential to accelerate heavy ions to high energies in high density plasmas as expected for future fusion reactors. New antenna designs will be needed for next step large future devices like DEMO, to deliver steady-state high power levels, cope with fast variations in coupling due to fast changes in the edge density and to reduce the possibility for impurity production. Such a new design is the traveling wave antenna (TWA) consisting of an array of straps distributed around the circumference of the machine, which is intrinsically resilient to edge density variations and has an optimized power coupling to the plasma. The structure of the paper is as follows: to provide the general reader with a basis for a good understanding of the later sections, an overview is given of wave propagation, coupling and RF power absorption in the ion cyclotron range of frequencies, including a brief summary of the traditionally used heating scenarios. A special highlight is the newly developed three-ion scenario together with its promising applications. A next section discusses recent developments to study edge-wave interaction and reduce impurity influx from ICRH: the dedicated devices IShTAR and Aline, field aligned and three-strap antenna concepts. The principles behind and the use of ICRH as an important option for first wall conditioning in devices with a permanent magnetic field is discussed next. The final section presents ongoing

  19. H Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

  20. Flow through in situ reactors with suction lysimeter sampling capability and methods of using

    Science.gov (United States)

    Radtke, Corey W [Idaho Falls, ID; Blackwelder, D Brad [Blackfoot, ID; Hubbell, Joel M [Idaho Falls, ID

    2009-11-17

    An in situ reactor for use in a geological strata includes a liner defining a centrally disposed passageway and a sampling conduit received within the passageway. The sampling conduit may be used to receive a geological speciment derived from geological strata therein and a lysimeter is disposed within the sampling conduit in communication with the geological specimen. Fluid may be added to the geological specimen through the passageway defined by the liner, between an inside surface of the liner and an outside surface of the sampling conduit. A distal portion of the sampling conduit may be in fluid communication with the passageway.

  1. Fast reactor programme in India

    Indian Academy of Sciences (India)

    2015-09-04

    Sep 4, 2015 ... Home; Journals; Pramana – Journal of Physics; Volume 85; Issue 3. Fast reactor programme in India. P Chellapandi P R ... Keywords. Sodium fast reactor; design challenges; construction challenges; emerging safety criteria; passive shutdown and decay heat removal systems; fast breeder reactors in India.

  2. Physical and Biological Release of Poly- and Perfluoroalkyl Substances (PFASs) from Municipal Solid Waste in Anaerobic Model Landfill Reactors.

    Science.gov (United States)

    Allred, B McKay; Lang, Johnsie R; Barlaz, Morton A; Field, Jennifer A

    2015-07-07

    A wide variety of consumer products that are treated with poly- and perfluoroalkyl substances (PFASs) and related formulations are disposed of in landfills. Landfill leachate has significant concentrations of PFASs and acts as secondary point sources to surface water. This study models how PFASs enter leachate using four laboratory-scale anaerobic bioreactors filled with municipal solid waste (MSW) and operated over 273 days. Duplicate reactors were monitored under live and abiotic conditions to evaluate influences attributable to biological activity. The biologically active reactors simulated the methanogenic conditions that develop in all landfills, producing ∼140 mL CH4/dry g refuse. The average total PFAS leaching measured in live reactors (16.7 nmol/kg dry refuse) was greater than the average for abiotic reactors (2.83 nmol/kg dry refuse), indicating biological processes were primarily responsible for leaching. The low-level leaching in the abiotic reactors was primarily due to PFCAs ≤C8 (2.48 nmol/kg dry refuse). Concentrations of known biodegradation intermediates, including methylperfluorobutane sulfonamide acetic acid and the n:2 and n:3 fluorotelomer carboxylates, increased steadily after the onset of methanogenesis, with the 5:3 fluorotelomer carboxylate becoming the single most concentrated PFAS observed in live reactors (9.53 nmol/kg dry refuse).

  3. Fast Spectrum Reactors

    CERN Document Server

    Todd, Donald; Tsvetkov, Pavel

    2012-01-01

    Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...

  4. System and method for the analysis of one or more compounds and/or species produced by a solution-based nuclear reactor

    Science.gov (United States)

    Policke, Timothy A; Nygaard, Eric T

    2014-05-06

    The present invention relates generally to both a system and method for determining the composition of an off-gas from a solution nuclear reactor (e.g., an Aqueous Homogeneous Reactor (AHR)) and the composition of the fissioning solution from those measurements. In one embodiment, the present invention utilizes at least one quadrupole mass spectrometer (QMS) in a system and/or method designed to determine at least one or more of: (i) the rate of production of at least one gas and/or gas species from a nuclear reactor; (ii) the effect on pH by one or more nitrogen species; (iii) the rate of production of one or more fission gases; and/or (iv) the effect on pH of at least one gas and/or gas species other than one or more nitrogen species from a nuclear reactor.

  5. Growth and Expansion of the International Criticality Safety Benchmark Evaluation Project and the Newly Organized International Reactor Physics Experiment Evaluation Project

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; Lori Scott; Yolanda Rugama; Enrico Satori

    2007-05-01

    Since ICNC 2003, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) has continued to expand its efforts and broaden its scope. Criticality-alarm / shielding type benchmarks and fundamental physics measurements that are relevant to criticality safety applications are not only included in the scope of the project, but benchmark data are also included in the latest version of the handbook. A considerable number of improvements have been made to the searchable database, DICE and the criticality-alarm / shielding benchmarks and fundamental physics measurements have been included in the database. There were 12 countries participating on the ICSBEP in 2003. That number has increased to 18 with recent contributions of data and/or resources from Brazil, Czech Republic, Poland, India, Canada, and China. South Africa, Germany, Argentina, and Australia have been invited to participate. Since ICNC 2003, the contents of the “International Handbook of Evaluated Criticality Safety Benchmark Experiments” have increased from 350 evaluations (28,000 pages) containing benchmark specifications for 3070 critical or subcritical configurations to 442 evaluations (over 38,000 pages) containing benchmark specifications for 3957 critical or subcritical configurations, 23 criticality-alarm-placement / shielding configurations with multiple dose points for each, and 20 configurations that have been categorized as fundamental physics measurements that are relevant to criticality safety applications in the 2006 Edition of the ICSBEP Handbook. Approximately 30 new evaluations and 250 additional configurations are expected to be added to the 2007 Edition of the Handbook. Since ICNC 2003, a reactor physics counterpart to the ICSBEP, The International Reactor Physics Experiment Evaluation Project (IRPhEP) was initiated. Beginning in 1999, the IRPhEP was conducted as a pilot activity by the by the Organization of Economic Cooperation and Development (OECD) Nuclear Energy

  6. An introduction to computer simulation methods applications to physical systems

    CERN Document Server

    Gould, Harvey; Christian, Wolfgang

    2007-01-01

    Now in its third edition, this book teaches physical concepts using computer simulations. The text incorporates object-oriented programming techniques and encourages readers to develop good programming habits in the context of doing physics. Designed for readers at all levels , An Introduction to Computer Simulation Methods uses Java, currently the most popular programming language. Introduction, Tools for Doing Simulations, Simulating Particle Motion, Oscillatory Systems, Few-Body Problems: The Motion of the Planets, The Chaotic Motion of Dynamical Systems, Random Processes, The Dynamics of Many Particle Systems, Normal Modes and Waves, Electrodynamics, Numerical and Monte Carlo Methods, Percolation, Fractals and Kinetic Growth Models, Complex Systems, Monte Carlo Simulations of Thermal Systems, Quantum Systems, Visualization and Rigid Body Dynamics, Seeing in Special and General Relativity, Epilogue: The Unity of Physics For all readers interested in developing programming habits in the context of doing phy...

  7. Computational Neutronics Methods and Transmutation Performance Analyses for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    M. Asgari; B. Forget; S. Piet; R. Ferrer; S. Bays

    2007-03-01

    The urgency for addressing repository impacts has grown in the past few years as a result of Spent Nuclear Fuel (SNF) accumulation from commercial nuclear power plants. One obvious path that has been explored by many is to eliminate the transuranic (TRU) inventory from the SNF thus reducing the need for additional long term repository storage sites. One strategy for achieving this is to burn the separated TRU elements in the currently operating U.S. Light Water Reactor (LWR) fleet. Many studies have explored the viability of this strategy by loading a percentage of LWR cores with TRU in the form of either Mixed Oxide (MOX) fuels or Inert Matrix Fuels (IMF). A task was undertaken at INL to establish specific technical capabilities to perform neutronics analyses in order to further assess several key issues related to the viability of thermal recycling. The initial computational study reported here is focused on direct thermal recycling of IMF fuels in a heterogeneous Pressurized Water Reactor (PWR) bundle design containing Plutonium, Neptunium, Americium, and Curium (IMF-PuNpAmCm) in a multi-pass strategy using legacy 5 year cooled LWR SNF. In addition to this initial high-priority analysis, three other alternate analyses with different TRU vectors in IMF pins were performed. These analyses provide comparison of direct thermal recycling of PuNpAmCm, PuNpAm, PuNp, and Pu.

  8. Application of econometric and ecology analysis methods in physics software

    Science.gov (United States)

    Han, Min Cheol; Hoff, Gabriela; Kim, Chan Hyeong; Kim, Sung Hun; Grazia Pia, Maria; Ronchieri, Elisabetta; Saracco, Paolo

    2017-10-01

    Some data analysis methods typically used in econometric studies and in ecology have been evaluated and applied in physics software environments. They concern the evolution of observables through objective identification of change points and trends, and measurements of inequality, diversity and evenness across a data set. Within each analysis area, various statistical tests and measures have been examined. This conference paper summarizes a brief overview of some of these methods.

  9. Internet methods in the study of women's physical activity.

    Science.gov (United States)

    Tsai, Hsiu-Min; Chee, Wonshik; Im, Eun-Ok

    2006-01-01

    Internet self-reporting methods have opened new opportunities in research that focuses on women's physical activity. Understanding the strengths and limitations of this self-report Internet method is critical to conducting a feasible and effective Internet study. The purpose of this paper is to address consideration of the strengths and limitations for researchers undertaking physical activity studies of women utilizing the Internet self-reporting method (Tables 1 and 2). The analysis utilizes a cross-sectional Internet survey regarding physical activity among women. Five major strengths were found including (1) reciprocal communication, (2) reduction of data incompleteness, (3) accuracy of data entry, (4) convenience, and (5) confidentiality and anonymity. Five potential limitations were found including (1) low response rate, (2) recall bias, (3) validity and reliability of Internet-based instruments, (4) sample bias, and (5) indirect measurement. Information in this paper may serve as a future reference for researchers engaged in using a self-report Internet method to estimate women's engagement in physical activity.

  10. An entrepreneurial physics method and its experimental test

    Science.gov (United States)

    Brown, Robert

    2012-02-01

    As faculty in a master's program for entrepreneurial physics and in an applied physics PhD program, I have advised upwards of 40 master and doctoral theses in industrial physics. I have been closely involved with four robust start-up manufacturing companies focused on physics high-technology and I have spent 30 years collaborating with industrial physicists on research and development. Thus I am in a position to reflect on many articles and advice columns centered on entrepreneurship. What about the goals, strategies, resources, skills, and the 10,000 hours needed to be an entrepreneur? What about business plans, partners, financing, patents, networking, salesmanship and regulatory affairs? What about learning new technology, how to solve problems and, in fact, learning innovation itself? At this point, I have my own method to propose to physicists in academia for incorporating entrepreneurship into their research lives. With this method, we do not start with a major invention or discovery, or even with a search for one. The method is based on the training we have, and the teaching we do (even quantum electrodynamics!), as physicists. It is based on the networking we build by 1) providing courses of continuing education for people working in industry and 2) through our undergraduate as well as graduate students who have gone on to work in industry. In fact, if we were to be limited to two words to describe the method, they are ``former students.'' Data from local and international medical imaging manufacturing industry are presented.

  11. Parametric studies by means of uncertainty and sensitivity methods for coupled thermal-hydraulic/neutron-physics application

    Energy Technology Data Exchange (ETDEWEB)

    Jaeger, W.; Sanchez, V.; Cheng, X. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). Inst. for Neutron Physics and Reactor Technology; Monti, L. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). Inst. for Nuclear and Energy Technologies; Hurtado, A. [Technical Univ. of Dresden (Germany). Inst. of Power Engineering

    2011-07-01

    At the Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT), the development and validation of coupled codes systems is one major activity. In this paper, a 2-step method is proposed to perform uncertainty and sensitivity analysis of a nuclear fuel bundle. At first, the SUSA package (Software system for Uncertainty and Sensitivity Analysis), 2 is applied to the thermal hydraulic results of the TRACE (TRACE/RELAP Advanced Computational Engine) code to identify crucial thermal hydraulic parameter combinations which are successively used in the TH/NP coupled system TRACEERANOS to account for the neutronic feedbacks. This 2-step method was applied since the TRACE-ERANOS system runs 1 input in approximately 1 day (depending on the computer configurations). Since the uncertainty and sensitivity analysis requires about 100 runs of the thermal hydraulic input (with altered parameters, running within minutes) an integral TRACE-SUSA-ERANOS analysis would need around 100 days. For this analysis a fuel assembly model of the HPLWR (High Performance Light Water Reactor) was selected. Due to the general structure of the coupling and code communication scripts, the system can be used for any kind of reactor/system which can be described with TRACE and ERANOS (e.g., fast systems) while SUSA can be applied to anything. (orig.)

  12. Effect of different physical activity training methods on overweight adolescents.

    Science.gov (United States)

    Ghatrehsamani, Shohreh; Khavarian, Noushin; Beizaei, Maryam; Ramedan, Reza; Poursafa, Parinaz; Kelishadi, Roya

    2010-01-01

    In view of the growing trend of obesity around the world, including in our country, and the effect of reduced physical activity in increasing the incidence of obesity and overweight in children and adolescents and limitations of families in providing transport for their children to attend exercise classes, as well as time limitations of students in taking part in these classes, accessing appropriate methods for presenting physical activity training seems essential. This non-pharmacological clinical trial was performed during six months from May to November 2007 on 105 children and adolescents aged 6-18 years with obesity, randomly assigned to 3 groups of thirty-five. Nutrition and treatment behavior were the same in all groups, but physical activity training in the first group was taking part in physical activity training classes twice a week, in the second group by providing a training CD, and in the third group via face-to-face training. Before and after the intervention, anthropometric indicators were measured and recorded. Mean body mass index (BMI) of participants in group attended physical activity training classes, and in the group undergone training with CD, after the interventions was significantly lower than that before the intervention. Our findings demonstrated that training using CDs can also be effective in reducing BMI in overweight and obese children and adolescents as much as face-to-face education and participation in physical training classes. Extending such interventions can be effective at the community level.

  13. XXXIV Bialowieza Workshop on Geometric Methods in Physics

    CERN Document Server

    Ali, S; Bieliavsky, Pierre; Odzijewicz, Anatol; Schlichenmaier, Martin; Voronov, Theodore

    2016-01-01

    This book features a selection of articles based on the XXXIV Białowieża Workshop on Geometric Methods in Physics, 2015. The articles presented are mathematically rigorous, include important physical implications and address the application of geometry in classical and quantum physics. Special attention deserves the session devoted to discussions of Gerard Emch's most important and lasting achievements in mathematical physics. The Białowieża workshops are among the most important meetings in the field and gather participants from mathematics and physics alike. Despite their long tradition, the Workshops remain at the cutting edge of ongoing research. For the past several years, the Białowieża Workshop has been followed by a School on Geometry and Physics, where advanced lectures for graduate students and young researchers are presented. The unique atmosphere of the Workshop and School is enhanced by the venue, framed by the natural beauty of the Białowieża forest in eastern Poland.

  14. Applications of Symmetry Methods to the Theory of Plasma Physics

    Directory of Open Access Journals (Sweden)

    Giampaolo Cicogna

    2006-02-01

    Full Text Available The theory of plasma physics offers a number of nontrivial examples of partial differential equations, which can be successfully treated with symmetry methods. We propose three different examples which may illustrate the reciprocal advantage of this "interaction" between plasma physics and symmetry techniques. The examples include, in particular, the complete symmetry analysis of system of two PDE's, with the determination of some conditional and partial symmetries, the construction of group-invariant solutions, and the symmetry classification of a nonlinear PDE.

  15. Academic Training Lecture: Statistical Methods for Particle Physics

    CERN Multimedia

    PH Department

    2012-01-01

    2, 3, 4 and 5 April 2012 Academic Training Lecture  Regular Programme from 11:00 to 12:00 -  Bldg. 222-R-001 - Filtration Plant Statistical Methods for Particle Physics by Glen Cowan (Royal Holloway) The series of four lectures will introduce some of the important statistical methods used in Particle Physics, and should be particularly relevant to those involved in the analysis of LHC data. The lectures will include an introduction to statistical tests, parameter estimation, and the application of these tools to searches for new phenomena.  Both frequentist and Bayesian methods will be described, with particular emphasis on treatment of systematic uncertainties.  The lectures will also cover unfolding, that is, estimation of a distribution in binned form where the variable in question is subject to measurement errors.

  16. U.S. Sodium Fast Reactor Codes and Methods: Current Capabilities and Path Forward

    Energy Technology Data Exchange (ETDEWEB)

    Brunett, A. J.; Fanning, T. H.

    2017-06-26

    The United States has extensive experience with the design, construction, and operation of sodium cooled fast reactors (SFRs) over the last six decades. Despite the closure of various facilities, the U.S. continues to dedicate research and development (R&D) efforts to the design of innovative experimental, prototype, and commercial facilities. Accordingly, in support of the rich operating history and ongoing design efforts, the U.S. has been developing and maintaining a series of tools with capabilities that envelope all facets of SFR design and safety analyses. This paper provides an overview of the current U.S. SFR analysis toolset, including codes such as SAS4A/SASSYS-1, MC2-3, SE2-ANL, PERSENT, NUBOW-3D, and LIFE-METAL, as well as the higher-fidelity tools (e.g. PROTEUS) being integrated into the toolset. Current capabilities of the codes are described and key ongoing development efforts are highlighted for some codes.

  17. Data Collection Methods for Validation of Advanced Multi-Resolution Fast Reactor Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Tokuhiro, Akiro [Univ. of Idaho, Moscow, ID (United States); Ruggles, Art [Univ. of Tennessee, Knoxville, TN (United States); Pointer, David [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-01-22

    In pool-type Sodium Fast Reactors (SFR) the regions most susceptible to thermal striping are the upper instrumentation structure (UIS) and the intermediate heat exchanger (IHX). This project experimentally and computationally (CFD) investigated the thermal mixing in the region exiting the reactor core to the UIS. The thermal mixing phenomenon was simulated using two vertical jets at different velocities and temperatures as prototypic of two adjacent channels out of the core. Thermal jet mixing of anticipated flows at different temperatures and velocities were investigated. Velocity profiles are measured throughout the flow region using Ultrasonic Doppler Velocimetry (UDV), and temperatures along the geometric centerline between the jets were recorded using a thermocouple array. CFD simulations, using COMSOL, were used to initially understand the flow, then to design the experimental apparatus and finally to compare simulation results and measurements characterizing the flows. The experimental results and CFD simulations show that the flow field is characterized into three regions with respective transitions, namely, convective mixing, (flow direction) transitional, and post-mixing. Both experiments and CFD simulations support this observation. For the anticipated SFR conditions the flow is momentum dominated and thus thermal mixing is limited due to the short flow length associated from the exit of the core to the bottom of the UIS. This means that there will be thermal striping at any surface where poorly mixed streams impinge; rather unless lateral mixing is ‘actively promoted out of the core, thermal striping will prevail. Furthermore we note that CFD can be considered a ‘separate effects (computational) test’ and is recommended as part of any integral analysis. To this effect, poorly mixed streams then have potential impact on the rest of the SFR design and scaling, especially placement of internal components, such as the IHX that may see poorly mixed

  18. Monte Carlo methods for medical physics a practical introduction

    CERN Document Server

    Schuemann, Jan; Paganetti, Harald

    2018-01-01

    The Monte Carlo (MC) method, established as the gold standard to predict results of physical processes, is now fast becoming a routine clinical tool for applications that range from quality control to treatment verification. This book provides a basic understanding of the fundamental principles and limitations of the MC method in the interpretation and validation of results for various scenarios. It shows how user-friendly and speed optimized MC codes can achieve online image processing or dose calculations in a clinical setting. It introduces this essential method with emphasis on applications in hardware design and testing, radiological imaging, radiation therapy, and radiobiology.

  19. The under-critical reactors physics for the hybrid systems; La physique des reacteurs sous-critiques des systemes hybrides

    Energy Technology Data Exchange (ETDEWEB)

    Schapira, J.P. [Institut de Physique Nucleaire, IN2P3/CNRS 91 - Orsay (France); Vergnes, J. [Electricite de France, EDF, Direction des Etudes et Recherches, 75 - Paris (France); Zaetta, A. [CEA/Saclay, Direction des Reacteurs Nucleaires, DRN, 91 - Gif-sur-Yvette (France)] [and others

    1998-03-12

    This day, organized by the SFEN, took place at Paris the 12 march 1998. Nine papers were presented. They take stock on the hybrid systems and more specifically the under-critical reactors. One of the major current preoccupation of nuclear industry is the problems of the increase of radioactive wastes produced in the plants and the destruction of the present stocks. To solve these problems a solution is the utilisation of hybrid systems: the coupling of a particle acceleration to an under-critical reactor. Historical aspects, advantages and performances of such hybrid reactors are presented in general papers. More technical papers are devoted to the spallation, the MUSE and the TARC experiments. (A.L.B.)

  20. Optimised k0-instrumental neutron activation method using the TRIGA MARK I IPR-R1 reactor at CDTN/CNEN, Belo Horizonte, Brazil

    Science.gov (United States)

    Menezes, M. Â. B. C.; Jaćimović, R.

    2006-08-01

    The Nuclear Technology Development Centre/Brazilian Commission for Nuclear Energy, CDTN/CNEN, is the only Brazilian Institution to apply the k0-standardisation method of instrumental neutron activation technique determining elements using its own nuclear reactor, TRIGA MARK I IPR-R1. After changes in the reactor core configuration, the reactor neutron flux distribution in typical irradiation channels had to be updated, as well as the parameters f and α, needed to apply the k0-method of neutron activation analysis. The neutron flux distribution in the rotary rack was evaluated through the specific count rate of 198Au and the parameters f and α, were determined in five selected channels applying the "Cd-ratio for multi-monitor" method, using a set of Al-(0.1%)Au and Zr (99.8%) monitors. Several reference materials were analysed, indicating the effectiveness of the improved method.

  1. Utilization of niching methods of genetic algorithms in nuclear reactor problems optimization; A utilizacao dos metodos de nichos dos algoritmos geneticos na otimizacao de problemas de reatores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Sacco, Wagner Figueiredo; Schirru, Roberto [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear

    2000-07-01

    Genetic Algorithms (GAs) are biologically motivated adaptive systems which have been used, with good results, in function optimization. However, traditional GAs rapidly push an artificial population toward convergence. That is, all individuals in the population soon become nearly identical. Niching Methods allow genetic algorithms to maintain a population of diverse individuals. GAs that incorporate these methods are capable of locating multiple, optimal solutions within a single population. The purpose of this study is to test existing niching techniques and two methods introduced herein, bearing in mind their eventual application in nuclear reactor related problems, specially the nuclear reactor core reload one, which has multiple solutions. Tests are performed using widely known test functions and their results show that the new methods are quite promising, specially in real world problems like the nuclear reactor core reload. (author)

  2. Method for the removal of sulphur oxides from a flue gas with a baghouse used as a secondary reactor

    Energy Technology Data Exchange (ETDEWEB)

    Teller, A.J.

    1986-04-08

    A method is described for removing sulfur oxides from a flue gas which includes: (a) introducing the flue gas containing sulfur oxides and particulates into a reactor; (b) contacting the flue gas with a calcium-based reagent to effect a neutralization reaction, thereby forming reaction products, the calcium-based reagent containing between 3-30% by weight of an alkaline metal cation salt based on the calcium compound in the reagent; (c) entraining the reaction products of the neutralization reaction and the particulates as solid in the effluent stream discharged from the quench reactor; (d) contacting the effluent stream with a gaseous stream having target particulates dispersed therein to promote inertial impaction between the entrained solids and the target particulates whereby the submicron solids entrained in the effluent stream are captured by the target particulates; (e) flowing the effluent stream into a collection zone; (f) accumulating the solids on a filter to form a bed of substantially non-tacky solids, some of the solids containing unreacted reagent; and (g) increasing the depth of the solids on the filter to provide at least a 40% removal of the sulfur oxides flowing into the collection zone.

  3. Methods of Efficient Study Habits and Physics Learning

    Science.gov (United States)

    Zettili, Nouredine

    2010-02-01

    We want to discuss the methods of efficient study habits and how they can be used by students to help them improve learning physics. In particular, we deal with the most efficient techniques needed to help students improve their study skills. We focus on topics such as the skills of how to develop long term memory, how to improve concentration power, how to take class notes, how to prepare for and take exams, how to study scientific subjects such as physics. We argue that the students who conscientiously use the methods of efficient study habits achieve higher results than those students who do not; moreover, a student equipped with the proper study skills will spend much less time to learn a subject than a student who has no good study habits. The underlying issue here is not the quantity of time allocated to the study efforts by the students, but the efficiency and quality of actions so that the student can function at peak efficiency. These ideas were developed as part of Project IMPACTSEED (IMproving Physics And Chemistry Teaching in SEcondary Education), an outreach grant funded by the Alabama Commission on Higher Education. This project is motivated by a major pressing local need: A large number of high school physics teachers teach out of field. )

  4. Introduction to methods of approximation in physics and astronomy

    CERN Document Server

    van Putten, Maurice H P M

    2017-01-01

    This textbook provides students with a solid introduction to the techniques of approximation commonly used in data analysis across physics and astronomy. The choice of methods included is based on their usefulness and educational value, their applicability to a broad range of problems and their utility in highlighting key mathematical concepts. Modern astronomy reveals an evolving universe rife with transient sources, mostly discovered - few predicted - in multi-wavelength observations. Our window of observations now includes electromagnetic radiation, gravitational waves and neutrinos. For the practicing astronomer, these are highly interdisciplinary developments that pose a novel challenge to be well-versed in astroparticle physics and data-analysis. The book is organized to be largely self-contained, starting from basic concepts and techniques in the formulation of problems and methods of approximation commonly used in computation and numerical analysis. This includes root finding, integration, signal dete...

  5. Theoretical physics 7 quantum mechanics : methods and applications

    CERN Document Server

    Nolting, Wolfgang

    2017-01-01

    This textbook offers a clear and comprehensive introduction to methods and applications in quantum mechanics, one of the core components of undergraduate physics courses. It follows on naturally from the previous volumes in this series, thus developing the understanding of quantized states further on. The first part of the book introduces the quantum theory of angular momentum and approximation methods. More complex themes are covered in the second part of the book, which describes multiple particle systems and scattering theory. Ideally suited to undergraduate students with some grounding in the basics of quantum mechanics, the book is enhanced throughout with learning features such as boxed inserts and chapter summaries, with key mathematical derivations highlighted to aid understanding. The text is supported by numerous worked examples and end of chapter problem sets.  About the Theoretical Physics series Translated from the renowned and highly successful German editions, the eight volumes of this seri...

  6. Study on the Neutrino Oscillation with a Next Generation Medium-Baseline Reactor Experiment

    OpenAIRE

    Chang Dong Shin; Kyung Kwang Joo

    2014-01-01

    For over fifty years, reactor experiments have played an important role in neutrino physics, in both discoveries and precision measurements. One of the methods to verify the existence of neutrino is the observation of neutrino oscillation phenomena. Electron antineutrinos emitted from a reactor provide the measurement of the small mixing angle θ13 , providing rich programs of neutrino properties, detector development, nuclear monitoring, and application. Using reactor neutrinos, future reacto...

  7. Variational Principles and Methods in Theoretical Physics and Chemistry

    Science.gov (United States)

    Nesbet, Robert K.

    2005-07-01

    Preface; Part I. Classical Mathematics and Physics: 1. History of variational theory; 2. Classical mechanics; 3. Applied mathematics; Part II. Bound States in Quantum Mechanics: 4. Time-independent quantum mechanics; 5. Independent-electron models; 6. Time-dependent theory and linear response; Part III. Continuum States and Scattering Theory: 7. Multiple scattering theory for molecules and solids; 8. Variational methods for continuum states; 9. Electron-impact rovibrational excitation of molecules; Part IV. Field Theories: 10. Relativistic Lagrangian theories.

  8. New applications of renormalization group methods in nuclear physics.

    Science.gov (United States)

    Furnstahl, R J; Hebeler, K

    2013-12-01

    We review recent developments in the use of renormalization group (RG) methods in low-energy nuclear physics. These advances include enhanced RG technology, particularly for three-nucleon forces, which greatly extends the reach and accuracy of microscopic calculations. We discuss new results for the nucleonic equation of state with applications to astrophysical systems such as neutron stars, new calculations of the structure and reactions of finite nuclei, and new explorations of correlations in nuclear systems.

  9. Statistical Methods for Particle Physics (4/4)

    CERN Multimedia

    CERN. Geneva

    2012-01-01

    The series of four lectures will introduce some of the important statistical methods used in Particle Physics, and should be particularly relevant to those involved in the analysis of LHC data. The lectures will include an introduction to statistical tests, parameter estimation, and the application of these tools to searches for new phenomena. Both frequentist and Bayesian methods will be described, with particular emphasis on treatment of systematic uncertainties. The lectures will also cover unfolding, that is, estimation of a distribution in binned form where the variable in question is subject to measurement errors.

  10. Statistical Methods for Particle Physics (3/4)

    CERN Multimedia

    CERN. Geneva

    2012-01-01

    The series of four lectures will introduce some of the important statistical methods used in Particle Physics, and should be particularly relevant to those involved in the analysis of LHC data. The lectures will include an introduction to statistical tests, parameter estimation, and the application of these tools to searches for new phenomena. Both frequentist and Bayesian methods will be described, with particular emphasis on treatment of systematic uncertainties. The lectures will also cover unfolding, that is, estimation of a distribution in binned form where the variable in question is subject to measurement errors.

  11. Statistical Methods for Particle Physics (2/4)

    CERN Multimedia

    CERN. Geneva

    2012-01-01

    The series of four lectures will introduce some of the important statistical methods used in Particle Physics, and should be particularly relevant to those involved in the analysis of LHC data. The lectures will include an introduction to statistical tests, parameter estimation, and the application of these tools to searches for new phenomena. Both frequentist and Bayesian methods will be described, with particular emphasis on treatment of systematic uncertainties. The lectures will also cover unfolding, that is, estimation of a distribution in binned form where the variable in question is subject to measurement errors.

  12. Statistical Methods for Particle Physics (1/4)

    CERN Multimedia

    CERN. Geneva

    2012-01-01

    The series of four lectures will introduce some of the important statistical methods used in Particle Physics, and should be particularly relevant to those involved in the analysis of LHC data. The lectures will include an introduction to statistical tests, parameter estimation, and the application of these tools to searches for new phenomena. Both frequentist and Bayesian methods will be described, with particular emphasis on treatment of systematic uncertainties. The lectures will also cover unfolding, that is, estimation of a distribution in binned form where the variable in question is subject to measurement errors.

  13. Method for Solving Physical Problems Described by Linear Differential Equations

    Science.gov (United States)

    Belyaev, B. A.; Tyurnev, V. V.

    2017-01-01

    A method for solving physical problems is suggested in which the general solution of a differential equation in partial derivatives is written in the form of decomposition in spherical harmonics with indefinite coefficients. Values of these coefficients are determined from a comparison of the decomposition with a solution obtained for any simplest particular case of the examined problem. The efficiency of the method is demonstrated on an example of calculation of electromagnetic fields generated by a current-carrying circular wire. The formulas obtained can be used to analyze paths in the near-field magnetic (magnetically inductive) communication systems working in moderately conductive media, for example, in sea water.

  14. Large Marks-decahedral Pd nanoparticles synthesized by a modified hydrothermal method using a homogeneous reactor

    Science.gov (United States)

    Zhao, Haiqiang; Qi, Weihong; Ji, Wenhai; Wang, Tianran; Peng, Hongcheng; Wang, Qi; Jia, Yanlin; He, Jieting

    2017-05-01

    Fivefold symmetry appears only in small particles and quasicrystals because internal stress in the particles increases with the particle size. However, a typical Marks decahedron with five re-entrant grooves located at the ends of the twin boundaries can further reduce the strain energy. During hydrothermal synthesis, it is difficult to stir the reaction solution contained in a digestion high-pressure tank because of the relatively small size and high-temperature and high-pressure sealed environment. In this work, we optimized a hydrothermal reaction system by replacing the conventional drying oven with a homogeneous reactor to shift the original static reaction solution into a full mixing state. Large Marks-decahedral Pd nanoparticles ( 90 nm) have been successfully synthesized in the optimized hydrothermal synthesis system. Additionally, in the products, round Marks-decahedral Pd particles were also found for the first time. While it remains a challenge to understand the growth mechanism of the fivefold twinned structure, we proposed a plausible growth-mediated mechanism for Marks-decahedral Pd nanoparticles based on observations of the synthesis process.

  15. Determination of Light Water Reactor Fuel Burnup with the Isotope Ratio Method

    Energy Technology Data Exchange (ETDEWEB)

    Gerlach, David C.; Mitchell, Mark R.; Reid, Bruce D.; Gesh, Christopher J.; Hurley, David E.

    2007-11-01

    For the current project to demonstrate that isotope ratio measurements can be extended to zirconium alloys used in LWR fuel assemblies we report new analyses on irradiated samples obtained from a reactor. Zirconium alloys are used for structural elements of fuel assemblies and for the fuel element cladding. This report covers new measurements done on irradiated and unirradiated zirconium alloys, Unirradiated zircaloy samples serve as reference samples and indicate starting values or natural values for the Ti isotope ratio measured. New measurements of irradiated samples include results for 3 samples provided by AREVA. New results indicate: 1. Titanium isotope ratios were measured again in unirradiated samples to obtain reference or starting values at the same time irradiated samples were analyzed. In particular, 49Ti/48Ti ratios were indistinguishably close to values determined several months earlier and to expected natural values. 2. 49Ti/48Ti ratios were measured in 3 irradiated samples thus far, and demonstrate marked departures from natural or initial ratios, well beyond analytical uncertainty, and the ratios vary with reported fluence values. The irradiated samples appear to have significant surface contamination or radiation damage which required more time for SIMS analyses. 3. Other activated impurity elements still limit the sample size for SIMS analysis of irradiated samples. The sub-samples chosen for SIMS analysis, although smaller than optimal, were still analyzed successfully without violating the conditions of the applicable Radiological Work Permit

  16. VALIDATION OF NUMERICAL METHODS TO CALCULATE BYPASS FLOW IN A PRISMATIC GAS-COOLED REACTOR CORE

    Directory of Open Access Journals (Sweden)

    NAM-IL TAK

    2013-11-01

    Full Text Available For thermo-fluid and safety analyses of a High Temperature Gas-cooled Reactor (HTGR, intensive efforts are in progress in the developments of the GAMMA+ code of Korea Atomic Energy Research Institute (KAERI and the AGREE code of the University of Michigan (U of M. One of the important requirements for GAMMA+ and AGREE is an accurate modeling capability of a bypass flow in a prismatic core. Recently, a series of air experiments were performed at Seoul National University (SNU in order to understand bypass flow behavior and generate an experimental database for the validation of computer codes. The main objective of the present work is to validate the GAMMA+ and AGREE codes using the experimental data published by SNU. The numerical results of the two codes were compared with the measured data. A good agreement was found between the calculations and the measurement. It was concluded that GAMMA+ and AGREE can reliably simulate the bypass flow behavior in a prismatic core.

  17. Uncertainty quantification in reactor physics using adjoint/perturbation techniques and adaptive spectral methods

    NARCIS (Netherlands)

    Gilli, L.

    2013-01-01

    This thesis presents the development and the implementation of an uncertainty propagation algorithm based on the concept of spectral expansion. The first part of the thesis is dedicated to the study of uncertainty propagation methodologies and to the analysis of spectral techniques. The concepts

  18. Solution for the nuclear reactor point-kinetics problem via decomposition method; Solucao via metodo da decomposicao do problema de cinetica puntual de um reator nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Vargas, Rubem Mario Figueiro [Pontificia Univ. Catolica do Rio Grande do Sul, Porto Alegre, RS (Brazil). Faculdade de Engenharia. Dept. de Engenharia Quimica]. E-mail: rvargas@pucrs.br; Vilhena, Marco Tullio de [Rio Grande do Sul Univ., Porto Alegre, RS (Brazil). Inst. de Matematica]. E-mail: vilhena@mat.ufrgs.br; Cardona, Augusto Vieira [Pontificia Univ. Catolica do Rio Grande do Sul, Porto Alegre, RS (Brazil). Faculdade de Matematica]. E-mail: acardona@pucrs.br

    2005-07-01

    The decomposition method is a mathematical technique, usually, applied to solve nonlinear problems, but can be an effective procedure for analytical solution of linear problems presenting advantages when compared with others techniques. In this work, an analytical solution for the nuclear reactor point-kinetics equations is developed using the decomposition method. (author)

  19. Maximum likelihood method and Fisher's information in physics and econophysics

    CERN Document Server

    Syska, Jacek

    2012-01-01

    Three steps in the development of the maximum likelihood (ML) method are presented. At first, the application of the ML method and Fisher information notion in the model selection analysis is described (Chapter 1). The fundamentals of differential geometry in the construction of the statistical space are introduced, illustrated also by examples of the estimation of the exponential models. At second, the notions of the relative entropy and the information channel capacity are introduced (Chapter 2). The observed and expected structural information principle (IP) and the variational IP of the modified extremal physical information (EPI) method of Frieden and Soffer are presented and discussed (Chapter 3). The derivation of the structural IP based on the analyticity of the logarithm of the likelihood function and on the metricity of the statistical space of the system is given. At third, the use of the EPI method is developed (Chapters 4-5). The information channel capacity is used for the field theory models cl...

  20. Comparison of Chemical and Physical-chemical Wastewater Discoloring Methods

    Directory of Open Access Journals (Sweden)

    Durašević, V.

    2007-11-01

    Full Text Available Today's chemical and physical-chemical wastewater discoloration methods do not completely meet demands regarding degree of discoloration. In this paper discoloration was performed using Fenton (FeSO4 . 7 H2O + H2O2 + H2SO4 and Fenton-like (FeCl3 . 6 H2O + H2O2 + HCOOH chemical methods and physical-chemical method of coagulation/flocculation (using poly-electrolyte (POEL combining anion active coagulant (modified poly-acrylamides and cationic flocculant (product of nitrogen compounds in combination with adsorption on activated carbon. Suitability of aforementioned methods was investigated on reactive and acid dyes, regarding their most common use in the textile industry. Also, investigations on dyes of different chromogen (anthraquinone, phthalocyanine, azo and xanthene were carried out in order to determine the importance of molecular spatial structure. Oxidative effect of Fenton and Fenton-like reagents resulted in decomposition of colored chromogen and high degree of discoloration. However, the problem is the inability of adding POEL in stechiometrical ratio (also present in physical-chemical methods, when the phenomenon of overdosing coagulants occurs in order to obtain a higher degree of discoloration, creating a potential danger of burdening water with POEL. Input and output water quality was controlled through spectrophotometric measurements and standard biological parameters. In addition, part of the investigations concerned industrial wastewaters obtained from dyeing cotton materials using reactive dye (C. I. Reactive Blue 19, a process that demands the use of vast amounts of electrolytes. Also, investigations of industrial wastewaters was labeled as a crucial step carried out in order to avoid serious misassumptions and false conclusions, which may arise if dyeing processes are only simulated in the laboratory.

  1. Advanced computational methods for the assessment of reactor core behaviour during reactivity initiated accidents. Final report; Fortschrittliche Rechenmethoden zum Kernverhalten bei Reaktivitaetsstoerfaellen. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Pautz, A.; Perin, Y.; Pasichnyk, I.; Velkov, K.; Zwermann, W.; Seubert, A.; Klein, M.; Gallner, L.; Krzycacz-Hausmann, B.

    2012-05-15

    The document at hand serves as the final report for the reactor safety research project RS1183 ''Advanced Computational Methods for the Assessment of Reactor Core Behavior During Reactivity-Initiated Accidents''. The work performed in the framework of this project was dedicated to the development, validation and application of advanced computational methods for the simulation of transients and accidents of nuclear installations. These simulation tools describe in particular the behavior of the reactor core (with respect to neutronics, thermal-hydraulics and thermal mechanics) at a very high level of detail. The overall goal of this project was the deployment of a modern nuclear computational chain which provides, besides advanced 3D tools for coupled neutronics/ thermal-hydraulics full core calculations, also appropriate tools for the generation of multi-group cross sections and Monte Carlo models for the verification of the individual calculational steps. This computational chain shall primarily be deployed for light water reactors (LWR), but should beyond that also be applicable for innovative reactor concepts. Thus, validation on computational benchmarks and critical experiments was of paramount importance. Finally, appropriate methods for uncertainty and sensitivity analysis were to be integrated into the computational framework, in order to assess and quantify the uncertainties due to insufficient knowledge of data, as well as due to methodological aspects.

  2. Optimization of the Gas Turbine-Modular Helium Reactor using statistical methods to maximize performance without compromising system design margins

    Energy Technology Data Exchange (ETDEWEB)

    Lommers, L.J.; Parme, L.L.; Shenoy, A.S. [General Atomics, San Diego, CA (United States)

    1995-12-31

    This paper describes a statistical approach for determining the impact of system performance and design uncertainties on power plant performance. The objectives of this design approach are to ensure that adequate margin is provided, that excess margin is minimized, and that full advantage can be taken of unconsumed margin. It is applicable to any thermal system in which these factors are important. The method is demonstrated using the Gas Turbine-Modular Helium Reactor as an example. The quantitative approach described allows the characterization of plant performance and the specification of the system design requirements necessary to achieve the desired performance with high confidence. Performance variations due to design evolution, in service degradation, and basic performance uncertainties are considered. The impact of all performance variabilities is combined using Monte Carlo analysis to predict the range of expected operation.

  3. Optimization of the gas turbine-modular helium reactor using statistical methods to maximize performance without compromising system design margins

    Energy Technology Data Exchange (ETDEWEB)

    Lommers, L.J.; Parme, L.L.; Shenoy, A.S.

    1995-07-01

    This paper describes a statistical approach for determining the impact of system performance and design uncertainties on power plant performance. The objectives of this design approach are to ensure that adequate margin is provided, that excess margin is minimized, and that full advantage can be taken of unconsumed margin. It is applicable to any thermal system in which these factors are important. The method is demonstrated using the Gas Turbine Modular Helium Reactor as an example. The quantitative approach described allows the characterization of plant performance and the specification of the system design requirements necessary to achieve the desired performance with high confidence. Performance variations due to design evolution, inservice degradation, and basic performance uncertainties are considered. The impact of all performance variabilities is combined using Monte Carlo analysis to predict the range of expected operation.

  4. [Patients on the move: validated methods to quantify physical activity

    NARCIS (Netherlands)

    Bakker, E.A.; Eijsvogels, T.M.H.; Vegt, F. de; Busser, G.S.; Hopman, M.T.E.; Verbeek, A.L.M.

    2015-01-01

    - Physical activity is an important component in the maintenance and improvement of general health; physical inactivity is, however, an increasing problem in the Netherlands.- Requests for advice on physical activity are increasing within the healthcare. - Assessment of an individual's physical

  5. An Innovative Three-Dimensional Heterogeneous Coarse-Mesh Transport Method for Advanced and Generation IV Reactor Core Analysis and Design

    Energy Technology Data Exchange (ETDEWEB)

    Farzad Rahnema

    2009-11-12

    This project has resulted in a highly efficient method that has been shown to provide accurate solutions to a variety of 2D and 3D reactor problems. The goal of this project was to develop (1) an accurate and efficient three-dimensional whole-core neutronics method with the following features: based sollely on transport theory, does not require the use of cross-section homogenization, contains a highly accurate and self-consistent global flux reconstruction procedure, and is applicable to large, heterogeneous reactor models, and to (2) create new numerical benchmark problems for code cross-comparison.

  6. Antineutrino emission and gamma background characteristics from a thermal research reactor

    CERN Document Server

    Bui, V M; Fallot, M; Communeau, V; Cormon, S; Estienne, M; Lenoir, M; Peuvrel, N; Shiba, T; Cucoanes, A S; Elnimr, M; Martino, J; Onillon, A; Porta, A; Pronost, G; Remoto, A; Thiolliere, N; Yermia, F; Zakari-Issoufou, A -A

    2016-01-01

    The detailed understanding of the antineutrino emission from research reactors is mandatory for any high sensitivity experiments either for fundamental or applied neutrino physics, as well as a good control of the gamma and neutron backgrounds induced by the reactor operation. In this article, the antineutrino emission associated to a thermal research reactor: the OSIRIS reactor located in Saclay, France, is computed in a first part. The calculation is performed with the summation method, which sums all the contributions of the beta decay branches of the fission products, coupled for the first time with a complete core model of the OSIRIS reactor core. The MCNP Utility for Reactor Evolution code was used, allowing to take into account the contributions of all beta decayers in-core. This calculation is representative of the isotopic contributions to the antineutrino flux which can be found at research reactors with a standard 19.75\\% enrichment in $^{235}$U. In addition, the required off-equilibrium correction...

  7. 31st International Colloquium in Group Theoretical Methods in Physics

    CERN Document Server

    Gazeau, Jean-Pierre; Faci, Sofiane; Micklitz, Tobias; Scherer, Ricardo; Toppan, Francesco

    2017-01-01

    This proceedings records the 31st International Colloquium on Group Theoretical Methods in Physics (“Group 31”). Plenary-invited articles propose new approaches to the moduli spaces in gauge theories (V. Pestun, 2016 Weyl Prize Awardee), the phenomenology of neutrinos in non-commutative space-time, the use of Hardy spaces in quantum physics, contradictions in the use of statistical methods on complex systems, and alternative models of supersymmetry. This volume’s survey articles broaden the colloquia’s scope out into Majorana neutrino behavior, the dynamics of radiating charges, statistical pattern recognition of amino acids, and a variety of applications of gauge theory, among others. This year’s proceedings further honors Bertram Kostant (2016 Wigner Medalist), as well as S.T. Ali and L. Boyle, for their life-long contributions to the math and physics communities. The aim of the ICGTMP is to provide a forum for physicists, mathematicians, and scientists of related disciplines who develop or apply ...

  8. Control rod calibration methods for fast breeder reactors applied to Phenix; Les methodes d'etalonnage des barres de commande des reacteurs a neutrons rapides application a Phenix

    Energy Technology Data Exchange (ETDEWEB)

    Lecourt, G

    1998-06-18

    The control and the emergency shutdown of a fast breeder reactor depends essentially on control rods. For this reason, it is imperative to know exactly how much anti reactivity is introduced with the rods in the reactor core. Different methods have been compared in order to see if they are compatible with Phenix reactor. Their limits have been studied. The shadow and anti shadow effects that can the rods make one to the other and then their effective weight of the rods screen have been clarified. (N.C.)

  9. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  10. Development and validation of coupled dynamics code 'TRIKIN' for VVER reactors

    Energy Technology Data Exchange (ETDEWEB)

    Obaidurrahman, K; Doshi, J. B.; Jain, R. P. [IIT Bombay, Mumbai (India); Jagannathan, V. [Bhabha Atomic Research Centre, Mumbai (India)

    2010-06-15

    New generation nuclear reactors are designed using advanced safety analysis methods. A thorough understanding of different interacting physical phenomena is necessary to avoid underestimation and overestimation of consequences of off normal transients in the reactor safety analysis results. This feature requires a multiphysics reactor simulation model. In this context, a coupled dynamics model based on a multiphysics formulation is developed indigenously for the transient analysis of large pressurized VVER reactors. Major simplifications are employed in the model by making several assumptions based on the physics of individual phenomenon. Space and time grids are optimized to minimize the computational bulk. The capability of the model is demonstrated by solving a series of international (AER) benchmark problems for VVER reactors. The developed model was used to analyze a number of reactivity transients that are likely to occur in VVER reactors

  11. Study on the Neutrino Oscillation with a Next Generation Medium-Baseline Reactor Experiment

    Directory of Open Access Journals (Sweden)

    Chang Dong Shin

    2014-01-01

    Full Text Available For over fifty years, reactor experiments have played an important role in neutrino physics, in both discoveries and precision measurements. One of the methods to verify the existence of neutrino is the observation of neutrino oscillation phenomena. Electron antineutrinos emitted from a reactor provide the measurement of the small mixing angle θ13, providing rich programs of neutrino properties, detector development, nuclear monitoring, and application. Using reactor neutrinos, future reactor neutrino experiments, more precise measurements of θ12,  Δm122, and mass hierarchy will be explored. The precise measurement of θ13 would be crucial for measuring the CP violation parameters at accelerators. Therefore, reactor neutrino physics will assist in the complete understanding of the fundamental nature and implications of neutrino masses and mixing. In this paper, we investigated several characteristics of RENO-50, which is a future medium-baseline reactor neutrino oscillation experiment, by using the GloBES simulation package.

  12. Statistical physics and computational methods for evolutionary game theory

    CERN Document Server

    Javarone, Marco Alberto

    2018-01-01

    This book presents an introduction to Evolutionary Game Theory (EGT) which is an emerging field in the area of complex systems attracting the attention of researchers from disparate scientific communities. EGT allows one to represent and study several complex phenomena, such as the emergence of cooperation in social systems, the role of conformity in shaping the equilibrium of a population, and the dynamics in biological and ecological systems. Since EGT models belong to the area of complex systems, statistical physics constitutes a fundamental ingredient for investigating their behavior. At the same time, the complexity of some EGT models, such as those realized by means of agent-based methods, often require the implementation of numerical simulations. Therefore, beyond providing an introduction to EGT, this book gives a brief overview of the main statistical physics tools (such as phase transitions and the Ising model) and computational strategies for simulating evolutionary games (such as Monte Carlo algor...

  13. Physical Activity Recognition with Mobile Phones: Challenges, Methods, and Applications

    Science.gov (United States)

    Yang, Jun; Lu, Hong; Liu, Zhigang; Boda, Péter Pál

    In this book chapter, we present a novel system that recognizes and records the physical activity of a person using a mobile phone. The sensor data is collected by built-in accelerometer sensor that measures the motion intensity of the device. The system recognizes five everyday activities in real-time, i.e., stationary, walking, running, bicycling, and in vehicle. We first introduce the sensor's data format, sensor calibration, signal projection, feature extraction, and selection methods. Then we have a detailed discussion and comparison of different choices of feature sets and classifiers. The design and implementation of one prototype system is presented along with resource and performance benchmark on Nokia N95 platform. Results show high recognition accuracies for distinguishing the five activities. The last part of the chapter introduces one demo application built on top of our system, physical activity diary, and a selection of potential applications in mobile wellness, mobile social sharing and contextual user interface domains.

  14. Development of a hybrid deterministic/stochastic method for 1D nuclear reactor kinetics

    Energy Technology Data Exchange (ETDEWEB)

    Terlizzi, Stefano; Dulla, Sandra; Ravetto, Piero [Politecnico di Torino, Corso Duca degli Abruzzi, 24 10129, Torino (Italy); Rahnema, Farzad, E-mail: farzad@gatech.edu [Nuclear & Radiological Engineering and Medical Physics Programs, Georgia Institute of Technology, 770 State Street NW, Atlanta, Ga, 30332-0745 (United States); Nuclear & Radiological Engineering and Medical Physics Programs, Georgia Institute of Technology, 770 State Street NW, Atlanta, Ga, 30332-0745 (United States); Zhang, Dingkang [Nuclear & Radiological Engineering and Medical Physics Programs, Georgia Institute of Technology, 770 State Street NW, Atlanta, Ga, 30332-0745 (United States)

    2015-12-31

    A new method has been implemented for solving the time-dependent neutron transport equation efficiently and accurately. This is accomplished by coupling the hybrid stochastic-deterministic steady-state coarse-mesh radiation transport (COMET) method [1,2] with the new predictor-corrector quasi-static method (PCQM) developed at Politecnico di Torino [3]. In this paper, the coupled method is implemented and tested in 1D slab geometry.

  15. A quasi-static polynomial nodal method for nuclear reactor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Gehin, Jess C. [Massachusetts Inst. of Tech., Cambridge, MA (United States)

    1992-09-01

    Modern nodal methods are currently available which can accurately and efficiently solve the static and transient neutron diffusion equations. Most of the methods, however, are limited to two energy groups for practical application. The objective of this research is the development of a static and transient, multidimensional nodal method which allows more than two energy groups and uses a non-linear iterative method for efficient solution of the nodal equations. For both the static and transient methods, finite-difference equations which are corrected by the use of discontinuity factors are derived. The discontinuity factors are computed from a polynomial nodal method using a non-linear iteration technique. The polynomial nodal method is based upon a quartic approximation and utilizes a quadratic transverse-leakage approximation. The solution of the time-dependent equations is performed by the use of a quasi-static method in which the node-averaged fluxes are factored into shape and amplitude functions. The application of the quasi-static polynomial method to several benchmark problems demonstrates that the accuracy is consistent with that of other nodal methods. The use of the quasi-static method is shown to substantially reduce the computation time over the traditional fully-implicit time-integration method. Problems involving thermal-hydraulic feedback are accurately, and efficiently, solved by performing several reactivity/thermal-hydraulic updates per shape calculation.

  16. A method for studying stability domains in physical models

    Science.gov (United States)

    Gallas, Jason A. C.

    1994-10-01

    We present a method for investigating the simultaneous movement of all zeros of equations of motions defined by discrete mappings. The method is used to show that knowledge of the interplay of all zeros is of fundamental importance for establishing periodicities and relative stability properties of the various possible physical solutions. The method is also used (i) to show that the Frontière set of Fatou is defined primarily by zeros of functions leading to an entire invariant limiting function which underlies every dynamical system, (ii) to identify cyclotomic polynomials as components of the limiting function obtained for a parameter value supporting a particular superstable orbit of the quadratic map, (iii) to describe highly symmetric periodic cycles embedded in these components, and (iv) to provide an unified picture about which mathematical objects form basin boundaries of dynamical systems in general: the closure of all zeros not belonging to “stable” orbits.

  17. Multi-scale, coupled Reactor Physics / Thermal-Hydraulics system and applications to the HPLWR 3 Pass Core

    OpenAIRE

    Monti, Lanfranco

    2009-01-01

    The HPLWR is an innovative reactor concept cooled with water at supercritical pressure. The pronounced changes of water properties with the heat-up demands advanced analyses tools which have been developed and successfully applied. Coupled neutronic/thermal-hydraulic analyses have been performed for the whole core and the coupled solution has been successively investigated at sub-channel resolution evaluating local quantities. The obtained results represent a new quality in core analyses.

  18. Dosing method of physical activity in aerobics classes for students

    Directory of Open Access Journals (Sweden)

    Yu.I. Beliak

    2014-10-01

    Full Text Available Purpose : reasons for the method of dosing of physical activity in aerobics classes for students. The basis of the method is the evaluation of the metabolic cost of funds used in them. Material : experiment involved the assessment of the pulse response of students to load complexes classical and step aerobics (n = 47, age 20-23 years. In complexes used various factors regulating the intensity: perform combinations of basic steps, involvement of movements with his hands, holding in hands dumbbells weighing 1kg increase in the rate of musical accompaniment, varying heights step platform. Results . on the basis of the relationship between heart rate and oxygen consumption was determined by the energy cost of each admission control load intensity. This indicator has been used to justify the intensity and duration of multiplicity aerobics. Figure correspond to the level of physical condition and motor activity deficits students. Conclusions : the estimated component of this method of dosing load makes it convenient for use in automated computer programs. Also it can be easily modified to dispense load other types of recreational fitness.

  19. Pacific Northwest Laboratory Monthly Activities Report for June 1966 AEC Division of Reactor Development and Technology Programs

    Energy Technology Data Exchange (ETDEWEB)

    SL Fawcett

    1966-06-01

    This report has the following sections: Summary; Civilian Power Reactors; Applied and Reactor Physics; Reactor Fuels and Materials; Engineering Development; Plutonium Recycle Program; Advanced Systems; and Nuclear Safety.

  20. Validation of finite difference core diffusion calculation methods with FEM and NEM for VVER-1000 MWe reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jagannathan, V. [Light Water Reactor Physics Section, Reactor Physics Design Div., Bhabha Atomic Research Centre, Mumbai - 400 085 (India); RPDD, Central Complex, BARC, Mumbai - 400085 (India); Singh, T. [Reactor Physics and Nuclear Engineering Section, Reactor Group, BARC, Mumbai (India); Pal, U.; Karthikeyan, R. [Light Water Reactor Physics Section, Reactor Physics Design Div., Bhabha Atomic Research Centre, Mumbai - 400 085 (India); Sundaram, G. [Nuclear Safety Group, KK-NPC, Mumbai (India)

    2006-07-01

    India is developing several in-house fuel management codes for the design evaluation of WER-1000 M We reactors, being built at Kudankulam, Tamil Nadu in collaboration with Russian Federation. A lattice burnup code EXCEL provides the few group lattice parameters of various fuel assembly types constituting the core. The core diffusion analyses have been performed by two methods. In the first method the entire fuel assembly is treated as a single homogenized cell. Each fuel assembly cell is divided into 6n{sup 2} triangles, where 'n' is the number of uniform divisions on a side of the hexagon. Regular triangular meshes are used in the active core as well as in surrounding reflector regions. This method is incorporated in the code TRIHEXFA. In the second method a pin by pin description of the core is accomplished by considering the few group lattice parameters generated by EXCEL code for various fuel and non-fuel cells in each fuel assembly. Regular hexagonal cells of one pin pitch are considered in the core and reflector regions. This method is incorporated in HEXPIN code. Both these codes use centre mesh finite difference method (FDM) for regular triangular or hexagonal meshes. It is well known that the large size of the WER fuel assembly, the zigzag structure of the core-baffle zone, the distribution of water tubes of different diameter in this baffle zone and the surrounding steel and water layers of different thickness, all lead to a very complex description of the core-reflector interface. We are analyzing the WER core in fresh state by two other approaches to obtain independent benchmark reference solutions. They are finite element method (FEM) and nodal expansion method (NEM). The few group cross sections of EXCEL are used in the FEM and NEM analyses. The paper would present the comparison of the results of core followup simulations of FD codes with those of FEM and NEM analyses. (authors)

  1. Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Martens, Frederick H. [Argonne National Laboratory; Jacobson, Norman H.

    1968-09-01

    This booklet discusses research reactors - reactors designed to provide a source of neutrons and/or gamma radiation for research, or to aid in the investigation of the effects of radiation on any type of material.

  2. Characterisation of well-adhered ZrO2 layers produced on structured reactors using the sonochemical sol-gel method

    Science.gov (United States)

    Jodłowski, Przemysław J.; Chlebda, Damian K.; Jędrzejczyk, Roman J.; Dziedzicka, Anna; Kuterasiński, Łukasz; Sitarz, Maciej

    2018-01-01

    The aim of this study was to obtain thin zirconium dioxide coatings on structured reactors using the sonochemical sol-gel method. The preparation method of metal oxide layers on metallic structures was based on the synergistic combination of three approaches: the application of ultrasonic irradiation during the synthesis of Zr sol-gel based on a precursor solution containing zirconium(IV) n-propoxide, the addition of stabilszing agents, and the deposition of ZrO2 on the metallic structures using the dip-coating method. As a result, dense, uniform zirconium dioxide films were obtained on the FeCrAlloy supports. The structured reactors were characterised by various physicochemical methods, such as BET, AFM, EDX, XRF, XRD, XPS and in situ Raman spectroscopy. The results of the structural analysis by Raman and XPS spectroscopy confirmed that the metallic surface was covered by a ZrO2 layer without any impurities. SEM/EDX mapping revealed that the deposited ZrO2 covered the metallic support uniformly. The mechanical and high temperature tests showed that the developed ultrasound assisted sol-gel method is an efficient way to obtain thin, well-adhered zirconium dioxide layers on the structured reactors. The prepared metallic supports covered with thin ZrO2 layers may be a good alternative to layered structured reactors in several dynamics flow processes, for example for gas exhaust abatement.

  3. REACTOR FUEL ELEMENTS TESTING CONTAINER

    Science.gov (United States)

    Whitham, G.K.; Smith, R.R.

    1963-01-15

    This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

  4. NATO Advanced Study Institute on Methods in Computational Molecular Physics

    CERN Document Server

    Diercksen, Geerd

    1992-01-01

    This volume records the lectures given at a NATO Advanced Study Institute on Methods in Computational Molecular Physics held in Bad Windsheim, Germany, from 22nd July until 2nd. August, 1991. This NATO Advanced Study Institute sought to bridge the quite considerable gap which exist between the presentation of molecular electronic structure theory found in contemporary monographs such as, for example, McWeeny's Methods 0/ Molecular Quantum Mechanics (Academic Press, London, 1989) or Wilson's Electron correlation in moleeules (Clarendon Press, Oxford, 1984) and the realization of the sophisticated computational algorithms required for their practical application. It sought to underline the relation between the electronic structure problem and the study of nuc1ear motion. Software for performing molecular electronic structure calculations is now being applied in an increasingly wide range of fields in both the academic and the commercial sectors. Numerous applications are reported in areas as diverse as catalysi...

  5. Physical Fault Injection and Monitoring Methods for Programmable Devices

    CERN Document Server

    AUTHOR|(INSPIRE)INSPIRE-00510096; Ferencei, Jozef

    A method of detecting faults for evaluating the fault cross section of any field programmable gate array (FPGA) was developed and is described in the thesis. The incidence of single event effects in FPGAs was studied for different probe particles (proton, neutron, gamma) using this method. The existing accelerator infrastructure of the Nuclear Physics Institute in Rez was supplemented by more sensitive beam monitoring system to ensure that the tests are done under well defined beam conditions. The bit cross section of single event effects was measured for different types of configuration memories, clock signal phase and beam energies and intensities. The extended infrastructure served also for radiation testing of components which are planned to be used in the new Inner Tracking System (ITS) detector of the ALICE experiment and for selecting optimal fault mitigation techniques used for securing the design of the FPGA-based ITS readout unit against faults induced by ionizing radiation.

  6. Ensemble Methods for Classification of Physical Activities from Wrist Accelerometry.

    Science.gov (United States)

    Chowdhury, Alok Kumar; Tjondronegoro, Dian; Chandran, Vinod; Trost, Stewart G

    2017-09-01

    To investigate whether the use of ensemble learning algorithms improve physical activity recognition accuracy compared to the single classifier algorithms, and to compare the classification accuracy achieved by three conventional ensemble machine learning methods (bagging, boosting, random forest) and a custom ensemble model comprising four algorithms commonly used for activity recognition (binary decision tree, k nearest neighbor, support vector machine, and neural network). The study used three independent data sets that included wrist-worn accelerometer data. For each data set, a four-step classification framework consisting of data preprocessing, feature extraction, normalization and feature selection, and classifier training and testing was implemented. For the custom ensemble, decisions from the single classifiers were aggregated using three decision fusion methods: weighted majority vote, naïve Bayes combination, and behavior knowledge space combination. Classifiers were cross-validated using leave-one subject out cross-validation and compared on the basis of average F1 scores. In all three data sets, ensemble learning methods consistently outperformed the individual classifiers. Among the conventional ensemble methods, random forest models provided consistently high activity recognition; however, the custom ensemble model using weighted majority voting demonstrated the highest classification accuracy in two of the three data sets. Combining multiple individual classifiers using conventional or custom ensemble learning methods can improve activity recognition accuracy from wrist-worn accelerometer data.

  7. Measures of the physical activity: revision of methods

    Directory of Open Access Journals (Sweden)

    Adair da Silva Lopes

    2000-12-01

    Full Text Available RESUMO Physical activity has been linked with several health benefi ts. Despite the evidence from population surveys, the measurement of physical activity has presented a challenge for researches in the area. The present study intends to show the main instruments for measuring physical activity, as well as their limitations and advantages. The choice of the most adequate instrument should observe criteria such as quality, simplicity or others. Currently, the lack of a “gold standard” instrument, suggests that a combination of methods would be the means of providing better data. The development of new instruments, improvement of current ones and the combination of methods are possible choices presented in this study for the use of better quality instruments. A atividade física tem sido relacionada a diversos benefícios para a saúde. Apesar de evidências levantadas em estudos populacionais, a mensuração da atividade física tem representado um desafio para pesquisadores da área. O presente estudo pretende apresentar os principais instrumentos de medida da atividade física, suas vantagens e limitações. A escolha do instrumento mais adequado deve atender a alguns critérios como a qualidade, a praticidade do instrumento entre outros. A falta de um instrumento considerado “padrão” sugere atualmente a utilização de uma combinação de métodos de maneira a fornecer dados mais confi áveis e precisos. A construção de novos instrumentos, o refi namento dos instrumentos existentes e a combinação de métodos são possibilidades consideradas para o uso de instrumentos de melhor qualidade.

  8. Literature in Focus: Statistical Methods in Experimental Physics

    CERN Multimedia

    2007-01-01

    Frederick James was a high-energy physicist who became the CERN "expert" on statistics and is now well-known around the world, in part for this famous text. The first edition of Statistical Methods in Experimental Physics was originally co-written with four other authors and was published in 1971 by North Holland (now an imprint of Elsevier). It became such an important text that demand for it has continued for more than 30 years. Fred has updated it and it was released in a second edition by World Scientific in 2006. It is still a top seller and there is no exaggeration in calling it «the» reference on the subject. A full review of the title appeared in the October CERN Courier.Come and meet the author to hear more about how this book has flourished during its 35-year lifetime. Frederick James Statistical Methods in Experimental Physics Monday, 26th of November, 4 p.m. Council Chamber (Bldg. 503-1-001) The author will be introduced...

  9. Methods for Analyzing Pathways through a Physics Major

    CERN Document Server

    Aiken, John M

    2016-01-01

    Physics Education Research frequently investigates what students studying physics do on small time scales (e.g. single courses, observations within single courses), or post-education time scales (e.g., what jobs do physics majors get?) but there is little research into how students get from the beginning to the end of a physics degree. Our work attempts to visualize students paths through the physics major, and quantitatively describe the students who take physics courses, receive physics degrees, and change degree paths into and out of the physics program at Michigan State University.

  10. CONVECTION REACTOR

    Science.gov (United States)

    Hammond, R.P.; King, L.D.P.

    1960-03-22

    An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

  11. Engineering Evaluation of Proposed Alternative Salt Transfer Method for the Molten Salt Reactor Experiement for the Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Carlberg, Jon A.; Roberts, Kenneth T.; Kollie, Thomas G.; Little, Leslie E.; Brady, Sherman D.

    2009-09-30

    This evaluation was performed by Pro2Serve in accordance with the Technical Specification for an Engineering Evaluation of the Proposed Alternative Salt Transfer Method for the Molten Salt Reactor Experiment at the Oak Ridge National Laboratory (BJC 2009b). The evaluators reviewed the Engineering Evaluation Work Plan for Molten Salt Reactor Experiment Residual Salt Removal, Oak Ridge National Laboratory, Oak Ridge, Tennessee (DOE 2008). The Work Plan (DOE 2008) involves installing a salt transfer probe and new drain line into the Fuel Drain Tanks and Fuel Flush Tank and connecting them to the new salt transfer line at the drain tank cell shield. The probe is to be inserted through the tank ball valve and the molten salt to the bottom of the tank. The tank would then be pressurized through the Reactive Gas Removal System to force the salt into the salt canisters. The Evaluation Team reviewed the work plan, interviewed site personnel, reviewed numerous documents on the Molten Salt Reactor (Sects. 7 and 8), and inspected the probes planned to be used for the transfer. Based on several concerns identified during this review, the team recommends not proceeding with the salt transfer via the proposed alternate salt transfer method. The major concerns identified during this evaluation are: (1) Structural integrity of the tanks - The main concern is with the corrosion that occurred during the fluorination phase of the uranium removal process. This may also apply to the salt transfer line for the Fuel Flush Tank. Corrosion Associated with Fluorination in the Oak Ridge National Laboratory Fluoride Volatility Process (Litman 1961) shows that this problem is significant. (2) Continued generation of Fluorine - Although the generation of Fluorine will be at a lower rate than experienced before the uranium removal, it will continue to be generated. This needs to be taken into consideration regardless of what actions are taken with the salt. (3) More than one phase of material

  12. Advances in Spectral Nodal Methods applied to S{sub N} Nuclear Reactor Global calculations in Cartesian Geometry

    Energy Technology Data Exchange (ETDEWEB)

    Barros, R.C.; Filho, H.A.; Oliveira, F.B.S. [Departamento de Modelagem Computacional, Instituto Politecnico, Universidade do Estado do Rio de Janeiro- UERJ, Rua Alberto Rangel s/n, 28630-050 Nova Friburgo, RJ (Brazil); Silva, F.C. da [Programa de Engenharia Nuclear, COPPE, Universidade Federal do Rio de Janeiro - UFRJ, Caixa Postal 68509, 21945-970 Rio de Janeiro, RJ (Brazil)]. e-mail: dickbarros@uol.com.br

    2004-07-01

    Presented here are the advances in spectral nodal methods for discrete ordinates (SN) eigenvalue problems in Cartesian geometry. These coarse-mesh methods are based on three ingredients: (i) the use of the standard discretized spatial balance SN equations; (ii) the use of the non-standard spectral diamond (SD) auxiliary equations in the multiplying regions of the domain, e.g. fuel assemblies; and (iii) the use of the non-standard spectral Green's function (SGF) auxiliary equations in the non-multiplying regions of the domain, e.g., the reflector. In slab-geometry the hybrid SD-SGF method generates numerical results that are completely free of spatial truncation errors. In X,Y-geometry, we obtain a system of two 'slab-geometry' SN equations for the node-edge average angular fluxes by transverse-integrating the X,Y-geometry SN equations separately in the y- and then in the x-directions within an arbitrary node of the spatial grid set up on the domain. In this paper, we approximate the transverse leakage terms by constants. These are the only approximations considered in the SD-SGF-constant nodal method, as the source terms, that include scattering and eventually fission events, are treated exactly. Moreover, we describe in this paper the progress of the approximate SN albedo boundary conditions for substituting the non-multiplying regions around the nuclear reactor core. We show numerical results to typical model problems to illustrate the accuracy of spectral nodal methods for coarse-mesh SN criticality calculations. (Author)

  13. Mirror reactor surface study

    Energy Technology Data Exchange (ETDEWEB)

    Hunt, A. L.; Damm, C. C.; Futch, A. H.; Hiskes, J. R.; Meisenheimer, R. G.; Moir, R. W.; Simonen, T. C.; Stallard, B. W.; Taylor, C. E.

    1976-09-01

    A general survey is presented of surface-related phenomena associated with the following mirror reactor elements: plasma first wall, ion sources, neutral beams, director converters, vacuum systems, and plasma diagnostics. A discussion of surface phenomena in possible abnormal reactor operation is included. Several studies which appear to merit immediate attention and which are essential to the development of mirror reactors are abstracted from the list of recommended areas for surface work. The appendix contains a discussion of the fundamentals of particle/surface interactions. The interactions surveyed are backscattering, thermal desorption, sputtering, diffusion, particle ranges in solids, and surface spectroscopic methods. A bibliography lists references in a number of categories pertinent to mirror reactors. Several complete published and unpublished reports on surface aspects of current mirror plasma experiments and reactor developments are also included.

  14. Fast method for reactor and feature scale coupling in ALD and CVD

    Science.gov (United States)

    Yanguas-Gil, Angel; Elam, Jeffrey W.

    2017-08-08

    Transport and surface chemistry of certain deposition techniques is modeled. Methods provide a model of the transport inside nanostructures as a single-particle discrete Markov chain process. This approach decouples the complexity of the surface chemistry from the transport model, thus allowing its application under general surface chemistry conditions, including atomic layer deposition (ALD) and chemical vapor deposition (CVD). Methods provide for determination of determine statistical information of the trajectory of individual molecules, such as the average interaction time or the number of wall collisions for molecules entering the nanostructures as well as to track the relative contributions to thin-film growth of different independent reaction pathways at each point of the feature.

  15. Ageing investigation and upgrading of components/systems of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Syarip; Widi Setiawan [Yogyakarta Nuclear Research Centre, Yogyakarta (Indonesia)

    1998-10-01

    Kartini research reactor has been operated in good condition and has demonstrated successful operation for the past 18 years, utilized for: reactor kinetic and control studies, instrumentation tests, neutronic and thermohydraulic studies, routine neutron activation analysis, reactor safety studies, training for research reactor operators and supervisors, and reactor physics experiments. Several components of Kartini reactor use components from the abandoned IRT-2000 Project at Serpong and from Bandung Reactor Centre such as: reactor tank, reactor core, heat exchanger, motor blower for ventilation system, fuel elements, etc. To maintain a good operating performance and also for aging investigation purposes, the component failure data collection has been done. The method used is based on the Manual on Reliability Data Collection For Research Reactor PSAs, IAEA TECDOC 636, and analyzed by using Data Entry System (DES) computer code. Analysis result shows that the components/systems failure rate of Kartini reactor is around 1,5.10{sup -4} up to 2,8.10{sup -4} per hour, these values are within the ranges of the values indicated in IAEA TECDOC 478. Whereas from the analysis of irradiation history shows that the neutron fluence of fuel element with highest burn-up (2,05 gram U-235 in average) is around 1.04.10{sup 16} n Cm{sup -2} and this value is still far below its limiting value. Some reactor components/systems have been replaced and upgraded such as heat exchanger, instrumentation and control system (ICS), etc. The new reactor ICS was installed in 1994 which is designed as a distributed structure by using microprocessor based systems and bus system technology. The characteristic and operating performance of the new reactor ICS, as well as the operation history and improvement of the Kartini research reactor is presented. (J.P.N.)

  16. Diagnostics of Oil Pollution Zones by Electro-Physical Method

    Science.gov (United States)

    Prostov, Sergey; Shabanov, Evgeniy

    2017-11-01

    The article presents the rationale and development of electro physical methods of control of the degree of soils pollution with petroleum products. The ranges of variation for the parameters of the pore space structure and wettability for the Kuzbass conditions are determined by inverse calculations on the basis of experimental data. The theoretical dependences are shown which allow us to assess the degree of soil pollution with oil products according to the results of longitudinal and areal electrical sounding. The study confirms the possibility of rapid prediction of the degree of soil contamination by measuring its electrical resistance. The database has been compiled for interpretation of the results of experimental sounding when monitoring the changes in the degree of contamination. The use of the express forecast of numerical values of the contamination factor was studied.

  17. Diagnostics of Oil Pollution Zones by Electro-Physical Method

    Directory of Open Access Journals (Sweden)

    Prostov Sergey

    2017-01-01

    Full Text Available The article presents the rationale and development of electro physical methods of control of the degree of soils pollution with petroleum products. The ranges of variation for the parameters of the pore space structure and wettability for the Kuzbass conditions are determined by inverse calculations on the basis of experimental data. The theoretical dependences are shown which allow us to assess the degree of soil pollution with oil products according to the results of longitudinal and areal electrical sounding. The study confirms the possibility of rapid prediction of the degree of soil contamination by measuring its electrical resistance. The database has been compiled for interpretation of the results of experimental sounding when monitoring the changes in the degree of contamination. The use of the express forecast of numerical values of the contamination factor was studied.

  18. Diagnostics and correction of disregulation states by physical methods

    OpenAIRE

    Gorsha, O. V.; Gorsha, V. I.

    2017-01-01

    Nicolaus Copernicus University, Toruń, Poland Ukrainian Research Institute for Medicine of Transport, Odesa, Ukraine Gorsha O. V., Gorsha V. I. Diagnostics and correction of disregulation states by physical methods Горша О. В., Горша В. И. Диагностика и коррекция физическими методами дизрегуляторных состояний Toruń, Odesa 2017 Nicolaus Copernicus University, To...

  19. Quantification method of N2O emission from full-scale biological nutrient removal wastewater treatment plant by laboratory batch reactor analysis.

    Science.gov (United States)

    Lim, Yesul; Kim, Dong-Jin

    2014-08-01

    This study proposes a simplified method for the quantification of N2O emission from a biological nutrient removal wastewater treatment plant (WWTP). The method incorporates a laboratory-scale batch reactor which had almost the same operational (wastewater and sludge flow rates) condition of a unit operation/process of the WWTP. Cumulative N2O emissions from the batch reactor at the corresponding hydraulic retention times of the full-scale units (primary and secondary clarifiers, pre-anoxic, anaerobic, anoxic and aerobic basins) were used for the quantification of N2O emission. The analysis showed that the aerobic basin emitted 95% of the total emission and the emission factor (yield) reached 0.8% based on the influent nitrogen load. The method successfully estimated N2O emission from the WWTP and it has shown advantages in measurement time and cost over the direct field measurement (floating chamber) method. Copyright © 2014 Elsevier Ltd. All rights reserved.

  20. Study of evaluation methods for in-vessel corium retention through external vessel cooling and safety of reactor cavity

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hoon; Chang, Soon Heung; Choi, Tae Hoon; Kim, Hyun Sop; Yang, Soo Hyung; Kim, Soo Hyung; Kim, Seung Hop; An Hyung Taek; Jeong, Yong Hoon; Huh, Gyun Young [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-03-15

    Cooling methodologies for the molten corium resulted from the severe accident of the nuclear power plant is suggested as one of most important items for the safety of the NPP. In this regard, considerable experimental and analytical works have been devoted. In the 1st phase of this project, present status related to the external reactor vessel cooling for the retention of the corium in the reactor vessel and corium at the reactor cavity have been investigated and preliminary studies have been accomplished for the detail evaluation of the each cooling methodology. The preliminary studies include the analysis and detail investigation of the possible phenomena, investigation of the heat transfer correlations and preliminary evaluation of the external reactor vessel cooling using the developed computer code.

  1. Comparison of Kinetic-based and Artificial Neural Network Modeling Methods for a Pilot Scale Vacuum Gas Oil Hydrocracking Reactor

    Directory of Open Access Journals (Sweden)

    Sepehr Sadighi

    2013-12-01

    Full Text Available An artificial neural network (ANN and kinetic-based models for a pilot scale vacuum gas oil (VGO hydrocracking plant are presented in this paper. Reported experimental data in the literature were used to develop, train, and check these models. The proposed models are capable of predicting the yield of all main hydrocracking products including dry gas, light naphtha, heavy naphtha, kerosene, diesel, and unconverted VGO (residue. Results showed that kinetic-based and artificial neural models have specific capabilities to predict yield of hydrocracking products. The former is able to accurately predict the yield of lighter products, i.e. light naphtha, heavy naphtha and kerosene. However, ANN model is capable of predicting yields of diesel and residue with higher precision. The comparison shows that the ANN model is superior to the kinetic-base models.  © 2013 BCREC UNDIP. All rights reservedReceived: 9th April 2013; Revised: 13rd August 2013; Accepted: 18th August 2013[How to Cite: Sadighi, S., Zahedi, G.R. (2013. Comparison of Kinetic-based and Artificial Neural Network Modeling Methods for a Pilot Scale Vacuum Gas Oil Hydrocracking Reactor. Bulletin of Chemical Reaction Engineering & Catalysis, 8 (2: 125-136. (doi:10.9767/bcrec.8.2.4722.125-136][Permalink/DOI: http://dx.doi.org/10.9767/bcrec.8.2.4722.125-136

  2. The New Water Moderator of the IBR-2 Reactor with a Canyon on the Lateral Surface. Design and Physical Parameters

    CERN Document Server

    Korneev, D A; Bodnarchuk, V I; Peresedov, V F; Rogov, A D; Shabalin, E P; Yaradaikin, S P

    2003-01-01

    An element of the new cold methane moderator of the reactor IBR-2, the water premoderator, serves as a thermal moderator for the 9th and 1st channels. Neutron radiation in the direction of the 9th channel comes from the lateral surface of the moderator. A specific feature of the reflectometer REFLEX located on the 9th channel is that it only "sees" neutrons emitted from a limited region of the moderator surface. This region is a rectangular extended along a vertical with a horizontal dimension of about 7 mm. To increase the flux on the sample, a groove-like pocket (canyon) with a depth of 80 mm by the width 15 mm and height 200 mm was cut in the premoderator on its lateral surface. The design of the moderator and the results of measurements of the neutron flux distribution on the lateral surface of the moderator are presented.

  3. Methods and systems for fuel production in electrochemical cells and reactors

    Energy Technology Data Exchange (ETDEWEB)

    Marina, Olga A.; Pederson, Larry R.

    2018-01-30

    Methods and systems for fuel, chemical, and/or electricity production from electrochemical cells are disclosed. A voltage is applied between an anode and a cathode of an electrochemical cell. The anode includes a metal or metal oxide electrocatalyst. Oxygen is supplied to the cathode, producing oxygen ions. The anode electrocatalyst is at least partially oxidized by the oxygen ions transported through an electrolyte from the cathode to the anode. A feed gas stream is supplied to the anode electrocatalyst, which is converted to a liquid fuel. The anode electrocatalyst is re-oxidized to higher valency oxides, or a mixture of oxide phases, by supplying the oxygen ions to the anode. The re-oxidation by the ions is controlled or regulated by the amount of voltage applied.

  4. Management of Spent Nuclear Fuel of Nuclear Research Reactor VVR-S at the National Institute of Physics and Nuclear Engineering, Bucharest, Romania

    Science.gov (United States)

    Biro, Lucian

    2009-05-01

    The Nuclear Research Reactor VVR-S (RR-VVR-S) located in Magurele-Bucharest, Romania, was designed for research and radioisotope production. It was commissioned in 1957 and operated without any event or accident for forty years until shut down in 1997. In 2002, by government decree, it was permanently shutdown for decommissioning. The National Institute of Physics and Nuclear Engineering (IFIN-HH) is responsible for decommissioning the RR-VVR-S, the first nuclear decommissioning project in Romania. In this context, IFIN-HH prepared and obtained approval from the Romanian Nuclear Regulatory Body for the Decommissioning Plan. One of the most important aspects for decommissioning the RR-VVR-S is solving the issue of the fresh and spent nuclear fuel (SNF) stored on site in wet storage pools. In the framework of the Russian Research Reactor Fuel Return Program (RRRFR), managed by the U.S. Department of Energy and in cooperation with the International Atomic Energy Agency and the Rosatom State Corporation, Romania repatriated all fresh HEU fuel to the Russian Federation in 2003 and the HEU SNF will be repatriated to Russia in 2009. With the experience and lessons learned from this action and with the financial support of the Romanian Government it will be possible for Romania to also repatriate the LEU SNF to the Russian Federation before starting the dismantling and decontamination of the nuclear facility. [4pt] In collaboration with K. Allen, Idaho National Laboratory, USA; L. Biro, National Commission for Nuclear Activities Control, Romania; and M. Dragusin, National Institute of Physics and Nuclear Engineering, Bucharest-Magurele, Romania.

  5. 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.

  6. POLARIZATION REMOTE SENSING PHYSICAL MECHANISM, KEY METHODS AND APPLICATION

    Directory of Open Access Journals (Sweden)

    B. Yang

    2017-09-01

    Full Text Available China's long-term planning major projects "high-resolution earth observation system" has been invested nearly 100 billion and the satellites will reach 100 to 2020. As to 2/3 of China's area covered by mountains,it has a higher demand for remote sensing. In addition to light intensity, frequency, phase, polarization is also the main physical characteristics of remote sensing electromagnetic waves. Polarization is an important component of the reflected information from the surface and the atmospheric information, and the polarization effect of the ground object reflection is the basis of the observation of polarization remote sensing. Therefore, the effect of eliminating the polarization effect is very important for remote sensing applications. The main innovations of this paper is as follows: (1 Remote sensing observation method. It is theoretically deduced and verified that the polarization can weaken the light in the strong light region, and then provide the polarization effective information. In turn, the polarization in the low light region can strengthen the weak light, the same can be obtained polarization effective information. (2 Polarization effect of vegetation. By analyzing the structure characteristics of vegetation, polarization information is obtained, then the vegetation structure information directly affects the absorption of biochemical components of leaves. (3 Atmospheric polarization neutral point observation method. It is proved to be effective to achieve the ground-gas separation, which can achieve the effect of eliminating the atmospheric polarization effect and enhancing the polarization effect of the object.

  7. Mathematical and computational methods in nuclear physics. Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Dehesa, J.S.; Gomez, J.M.G.; Polls, A.

    1984-01-01

    The present proceedings contain the talks given at the Sixth International Granada Workshop on ''Mathematical and Computational Methods in Nuclear Physics'', held in Granada (Spain), October 3rd-8th, 1983. The lectures covering various aspects of the many-body problem in nuclei, review present knowledge and include some unpublished material as well. Bohigas and Giannoni discuss the fluctuation properties of spectra of many-body systems by means of random matrix theories, and the attempts to search for quantum mechanical manifestations of classical chaotic motion. The role of spectral distributions (expressed as explicit functions of the microscopic matrix elements of the Hamiltonian) in the statistical spectroscopy of nuclear systems is analyzed by French. Zuker, after a brief review of the theoretical basis of the shell model, discusses a reformulation of the theory of effective interactions and gives a survey of the linked cluster theory. Goeke's lectures center on the mean-field methods, particularly TDHF, used in the investigation of the large-amplitude nuclear collective motion, pointing out both the successes and failures of the theory. In addition the present volume also contains the seminars on related topics.

  8. Polarization Remote Sensing Physical Mechanism, Key Methods and Application

    Science.gov (United States)

    Yang, B.; Wu, T.; Chen, W.; Li, Y.; Knjazihhin, J.; Asundi, A.; Yan, L.

    2017-09-01

    China's long-term planning major projects "high-resolution earth observation system" has been invested nearly 100 billion and the satellites will reach 100 to 2020. As to 2/3 of China's area covered by mountains it has a higher demand for remote sensing. In addition to light intensity, frequency, phase, polarization is also the main physical characteristics of remote sensing electromagnetic waves. Polarization is an important component of the reflected information from the surface and the atmospheric information, and the polarization effect of the ground object reflection is the basis of the observation of polarization remote sensing. Therefore, the effect of eliminating the polarization effect is very important for remote sensing applications. The main innovations of this paper is as follows: (1) Remote sensing observation method. It is theoretically deduced and verified that the polarization can weaken the light in the strong light region, and then provide the polarization effective information. In turn, the polarization in the low light region can strengthen the weak light, the same can be obtained polarization effective information. (2) Polarization effect of vegetation. By analyzing the structure characteristics of vegetation, polarization information is obtained, then the vegetation structure information directly affects the absorption of biochemical components of leaves. (3) Atmospheric polarization neutral point observation method. It is proved to be effective to achieve the ground-gas separation, which can achieve the effect of eliminating the atmospheric polarization effect and enhancing the polarization effect of the object.

  9. Interactive Rapid Dose Assessment Model (IRDAM): reactor-accident assessment methods. Vol. 2

    Energy Technology Data Exchange (ETDEWEB)

    Poeton, R.W.; Moeller, M.P.; Laughlin, G.J.; Desrosiers, A.E.

    1983-05-01

    As part of the continuing emphasis on emergency preparedness, the US Nuclear Regulatory Commission (NRC) sponsored the development of a rapid dose assessment system by Pacific Northwest Laboratory (PNL). This system, the Interactive Rapid Dose Assessment Model (IRDAM) is a micro-computer based program for rapidly assessing the radiological impact of accidents at nuclear power plants. This document describes the technical bases for IRDAM including methods, models and assumptions used in calculations. IRDAM calculates whole body (5-cm depth) and infant thyroid doses at six fixed downwind distances between 500 and 20,000 meters. Radionuclides considered primarily consist of noble gases and radioiodines. In order to provide a rapid assessment capability consistent with the capacity of the Osborne-1 computer, certain simplifying approximations and assumptions are made. These are described, along with default values (assumptions used in the absence of specific input) in the text of this document. Two companion volumes to this one provide additional information on IRDAM. The user's Guide (NUREG/CR-3012, Volume 1) describes the setup and operation of equipment necessary to run IRDAM. Scenarios for Comparing Dose Assessment Models (NUREG/CR-3012, Volume 3) provides the results of calculations made by IRDAM and other models for specific accident scenarios.

  10. Thermodynamics, Gibbs Method and Statistical Physics of Electron Gases Gibbs Method and Statistical Physics of Electron Gases

    CERN Document Server

    Askerov, Bahram M

    2010-01-01

    This book deals with theoretical thermodynamics and the statistical physics of electron and particle gases. While treating the laws of thermodynamics from both classical and quantum theoretical viewpoints, it posits that the basis of the statistical theory of macroscopic properties of a system is the microcanonical distribution of isolated systems, from which all canonical distributions stem. To calculate the free energy, the Gibbs method is applied to ideal and non-ideal gases, and also to a crystalline solid. Considerable attention is paid to the Fermi-Dirac and Bose-Einstein quantum statistics and its application to different quantum gases, and electron gas in both metals and semiconductors is considered in a nonequilibrium state. A separate chapter treats the statistical theory of thermodynamic properties of an electron gas in a quantizing magnetic field.

  11. REACTOR COOLING

    Science.gov (United States)

    Quackenbush, C.F.

    1959-09-29

    A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

  12. An evaluation of teaching methods in the introductory physics classroom

    Science.gov (United States)

    Savage, Lauren Michelle Williams

    The introductory physics mechanics course at the University of North Carolina at Charlotte has a history of relatively high DFW rates. In 2011, the course was redesigned from the traditional lecture format to the inverted classroom format (flipped). This format inverts the classroom by introducing material in a video assigned as homework while the instructor conducts problem solving activities and guides discussions during the regular meetings. This format focuses on student-centered learning and is more interactive and engaging. To evaluate the effectiveness of the new method, final exam data over the past 10 years was mined and the pass rates examined. A normalization condition was developed to evaluate semesters equally. The two teaching methods were compared using a grade distribution across multiple semesters. Students in the inverted class outperformed those in the traditional class: "A"s increased by 22% and "B"s increased by 38%. The final exam pass rate increased by 12% under the inverted classroom approach. The same analysis was used to compare the written and online final exam formats. Surprisingly, no students scored "A"s on the online final. However, the percent of "B"s increased by 136%. Combining documented best practices from a literature review with personal observations of student performance and attitudes from first hand classroom experience as a teaching assistant in both teaching methods, reasons are given to support the continued use of the inverted classroom approach as well as the online final. Finally, specific recommendations are given to improve the course structure where weaknesses have been identified.

  13. Reactor flux calculations

    Energy Technology Data Exchange (ETDEWEB)

    Lhuillier, D. [Commissariat à l' Énergie Atomique et aux Énergies Alternatives, Centre de Saclay, IRFU/SPhN, 91191 Gif-sur-Yvette (France)

    2013-02-15

    The status of the prediction of reactor anti-neutrino spectra is presented. The most accurate method is still the conversion of total β spectra of fissionning isotopes as measured at research reactors. Recent re-evaluations of the conversion process led to an increased predicted flux by few percent and were at the origin of the so-called reactor anomaly. The up to date predictions are presented with their main sources of error. Perspectives are given on the complementary ab-initio predictions and upcoming experimental cross-checks of the predicted spectrum shape.

  14. NEUTRONIC REACTOR CONTROL

    Science.gov (United States)

    Metcalf, H.E.

    1958-10-14

    Methods of controlling reactors are presented. Specifically, a plurality of neutron absorber members are adjustably disposed in the reactor core at different distances from the center thereof. The absorber members extend into the core from opposite faces thereof and are operated by motive means coupled in a manner to simultaneously withdraw at least one of the absorber members while inserting one of the other absorber members. This feature effects fine control of the neutron reproduction ratio by varying the total volume of the reactor effective in developing the neutronic reaction.

  15. Wave propagation simulation in the upper core of sodium-cooled fast reactors using a spectral-element method for heterogeneous media

    Science.gov (United States)

    Nagaso, Masaru; Komatitsch, Dimitri; Moysan, Joseph; Lhuillier, Christian

    2018-01-01

    ASTRID project, French sodium cooled nuclear reactor of 4th generation, is under development at the moment by Alternative Energies and Atomic Energy Commission (CEA). In this project, development of monitoring techniques for a nuclear reactor during operation are identified as a measure issue for enlarging the plant safety. Use of ultrasonic measurement techniques (e.g. thermometry, visualization of internal objects) are regarded as powerful inspection tools of sodium cooled fast reactors (SFR) including ASTRID due to opacity of liquid sodium. In side of a sodium cooling circuit, heterogeneity of medium occurs because of complex flow state especially in its operation and then the effects of this heterogeneity on an acoustic propagation is not negligible. Thus, it is necessary to carry out verification experiments for developments of component technologies, while such kind of experiments using liquid sodium may be relatively large-scale experiments. This is why numerical simulation methods are essential for preceding real experiments or filling up the limited number of experimental results. Though various numerical methods have been applied for a wave propagation in liquid sodium, we still do not have a method for verifying on three-dimensional heterogeneity. Moreover, in side of a reactor core being a complex acousto-elastic coupled region, it has also been difficult to simulate such problems with conventional methods. The objective of this study is to solve these 2 points by applying three-dimensional spectral element method. In this paper, our initial results on three-dimensional simulation study on heterogeneous medium (the first point) are shown. For heterogeneity of liquid sodium to be considered, four-dimensional temperature field (three spatial and one temporal dimension) calculated by computational fluid dynamics (CFD) with Large-Eddy Simulation was applied instead of using conventional method (i.e. Gaussian Random field). This three-dimensional numerical

  16. Participation in the U.S. Department of Energy Reactor Sharing Program. Progress report

    Energy Technology Data Exchange (ETDEWEB)

    Mulder, R.U.; Benneche, P.E.; Hosticka, B.

    1997-03-01

    The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would not be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA for use at their institutions. These areas are discussed here.

  17. Gold nanoparticles production using reactor and cyclotron based methods in assessment of (196,198)Au production yields by (197)Au neutron absorption for therapeutic purposes.

    Science.gov (United States)

    Khorshidi, Abdollah

    2016-11-01

    Medical nano-gold radioisotopes is produced regularly using high-flux nuclear reactors, and an accelerator-driven neutron activator can turn out higher yield of (197)Au(n,γ)(196,198)Au reactions. Here, nano-gold production via radiative/neutron capture was investigated using irradiated Tehran Research Reactor flux and also simulated proton beam of Karaj cyclotron in Iran. (197)Au nano-solution, including 20nm shaped spherical gold and water, was irradiated under Tehran reactor flux at 2.5E+13n/cm(2)/s for (196,198)Au activity and production yield estimations. Meanwhile, the yield was examined using 30MeV proton beam of Karaj cyclotron via simulated new neutron activator containing beryllium target, bismuth moderator around the target, and also PbF2 reflector enclosed the moderator region. Transmutation in (197)Au nano-solution samples were explored at 15 and 25cm distances from the target. The neutron flux behavior inside the water and bismuth moderators was investigated for nano-gold particles transmutation. The transport of fast neutrons inside bismuth material as heavy nuclei with a lesser lethargy can be contributed in enhanced nano-gold transmutation with long duration time than the water moderator in reactor-based method. Cyclotron-driven production of βeta-emitting radioisotopes for brachytherapy applications can complete the nano-gold production technology as a safer approach as compared to the reactor-based method. Copyright © 2016 Elsevier B.V. All rights reserved.

  18. Methods for determining thermal stresses values. Some examples relating to nuclear reactors; Methodes de determination des contraintes thermiques. Quelques exemples d'application aux reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, J.; Gautier, A. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Peres, A. [Israel Institute of Technology, Dept. of Nuclear Science Technion (Israel)

    1958-07-01

    As modern techniques develop more elaborate machines, and make their way towards higher and higher temperatures and pressures, the thermal stresses become a matter of major importance in the design of mechanical structures. In the first part of this paper, the authors examine the problem from a theoretical standpoint, and try to evaluate the aptitude and limitation of mathematical techniques to attain the quantitative values of thermal stresses. This paper deals mainly with the experimental methods to measure thermal stresses. The authors show some examples relating to nuclear reactors. (author)Fren. [French] Au fur et a mesure que la technique moderne developpe des machines plus poussees et s'oriente vers des temperatures et des pressions toujours plus elevees, les contraintes thermiques deviennent un facteur d'importance capitale dans le calcul des structures mecaniques. Les auteurs examinent d'abord l'aspect theorique du probleme, ainsi que l'aptitude et les limites du calcul pour exprimer quantitativement la valeur des contraintes thermiques. Les auteurs exposent principalement, ensuite, les methodes experimentales qui permettent de mesurer ces contraintes, et illustrent cet expose de quelques exemples relatifs aux installations nucleaires. (auteur)

  19. A Takagi-Sugeno fuzzy power-distribution method for a prototypical advanced reactor considering pump degradation

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Yue [Institute of Nuclear and New Energy Technology, Tsinghua University, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Beijing (China); Coble, Jamie [Dept. of Nuclear Engineering, University of Tennessee, Knoxville (United States)

    2017-08-15

    Advanced reactor designs often feature longer operating cycles between refueling and new concepts of operation beyond traditional baseload electricity production. Owing to this increased complexity, traditional proportional–integral control may not be sufficient across all potential operating regimes. The prototypical advanced reactor (PAR) design features two independent reactor modules, each connected to a single dedicated steam generator that feeds a common balance of plant for electricity generation and process heat applications. In the current research, the PAR is expected to operate in a load-following manner to produce electricity to meet grid demand over a 24-hour period. Over the operational lifetime of the PAR system, primary and intermediate sodium pumps are expected to degrade in performance. The independent operation of the two reactor modules in the PAR may allow the system to continue operating under degraded pump performance by shifting the power production between reactor modules in order to meet overall load demands. This paper proposes a Takagi–Sugeno (T–S) fuzzy logic-based power distribution system. Two T–S fuzzy power distribution controllers have been designed and tested. Simulation shows that the devised T–S fuzzy controllers provide improved performance over traditional controls during daily load-following operation under different levels of pump degradation.

  20. A Takagi–Sugeno fuzzy power-distribution method for a prototypical advanced reactor considering pump degradation

    Directory of Open Access Journals (Sweden)

    Yue Yuan

    2017-08-01

    Full Text Available Advanced reactor designs often feature longer operating cycles between refueling and new concepts of operation beyond traditional baseload electricity production. Owing to this increased complexity, traditional proportional–integral control may not be sufficient across all potential operating regimes. The prototypical advanced reactor (PAR design features two independent reactor modules, each connected to a single dedicated steam generator that feeds a common balance of plant for electricity generation and process heat applications. In the current research, the PAR is expected to operate in a load-following manner to produce electricity to meet grid demand over a 24-hour period. Over the operational lifetime of the PAR system, primary and intermediate sodium pumps are expected to degrade in performance. The independent operation of the two reactor modules in the PAR may allow the system to continue operating under degraded pump performance by shifting the power production between reactor modules in order to meet overall load demands. This paper proposes a Takagi–Sugeno (T–S fuzzy logic-based power distribution system. Two T–S fuzzy power distribution controllers have been designed and tested. Simulation shows that the devised T–S fuzzy controllers provide improved performance over traditional controls during daily load-following operation under different levels of pump degradation.

  1. Innovative and Advanced Coupled Neutron Transport and Thermal Hydraulic Method (Tool) for the Design, Analysis and Optimization of VHTR/NGNP Prismatic Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rahnema, Farzad; Garimeela, Srinivas; Ougouag, Abderrafi; Zhang, Dingkang

    2013-11-29

    This project will develop a 3D, advanced coarse mesh transport method (COMET-Hex) for steady- state and transient analyses in advanced very high-temperature reactors (VHTRs). The project will lead to a coupled neutronics and thermal hydraulic (T/H) core simulation tool with fuel depletion capability. The computational tool will be developed in hexagonal geometry, based solely on transport theory without (spatial) homogenization in complicated 3D geometries. In addition to the hexagonal geometry extension, collaborators will concurrently develop three additional capabilities to increase the code’s versatility as an advanced and robust core simulator for VHTRs. First, the project team will develop and implement a depletion method within the core simulator. Second, the team will develop an elementary (proof-of-concept) 1D time-dependent transport method for efficient transient analyses. The third capability will be a thermal hydraulic method coupled to the neutronics transport module for VHTRs. Current advancements in reactor core design are pushing VHTRs toward greater core and fuel heterogeneity to pursue higher burn-ups, efficiently transmute used fuel, maximize energy production, and improve plant economics and safety. As a result, an accurate and efficient neutron transport, with capabilities to treat heterogeneous burnable poison effects, is highly desirable for predicting VHTR neutronics performance. This research project’s primary objective is to advance the state of the art for reactor analysis.

  2. Power reactor noise studies and applications

    Energy Technology Data Exchange (ETDEWEB)

    Arzhanov, V

    2002-03-01

    The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

  3. A physics-motivated Centroidal Voronoi Particle domain decomposition method

    Energy Technology Data Exchange (ETDEWEB)

    Fu, Lin, E-mail: lin.fu@tum.de; Hu, Xiangyu Y., E-mail: xiangyu.hu@tum.de; Adams, Nikolaus A., E-mail: nikolaus.adams@tum.de

    2017-04-15

    In this paper, we propose a novel domain decomposition method for large-scale simulations in continuum mechanics by merging the concepts of Centroidal Voronoi Tessellation (CVT) and Voronoi Particle dynamics (VP). The CVT is introduced to achieve a high-level compactness of the partitioning subdomains by the Lloyd algorithm which monotonically decreases the CVT energy. The number of computational elements between neighboring partitioning subdomains, which scales the communication effort for parallel simulations, is optimized implicitly as the generated partitioning subdomains are convex and simply connected with small aspect-ratios. Moreover, Voronoi Particle dynamics employing physical analogy with a tailored equation of state is developed, which relaxes the particle system towards the target partition with good load balance. Since the equilibrium is computed by an iterative approach, the partitioning subdomains exhibit locality and the incremental property. Numerical experiments reveal that the proposed Centroidal Voronoi Particle (CVP) based algorithm produces high-quality partitioning with high efficiency, independently of computational-element types. Thus it can be used for a wide range of applications in computational science and engineering.

  4. A physics-motivated Centroidal Voronoi Particle domain decomposition method

    Science.gov (United States)

    Fu, Lin; Hu, Xiangyu Y.; Adams, Nikolaus A.

    2017-04-01

    In this paper, we propose a novel domain decomposition method for large-scale simulations in continuum mechanics by merging the concepts of Centroidal Voronoi Tessellation (CVT) and Voronoi Particle dynamics (VP). The CVT is introduced to achieve a high-level compactness of the partitioning subdomains by the Lloyd algorithm which monotonically decreases the CVT energy. The number of computational elements between neighboring partitioning subdomains, which scales the communication effort for parallel simulations, is optimized implicitly as the generated partitioning subdomains are convex and simply connected with small aspect-ratios. Moreover, Voronoi Particle dynamics employing physical analogy with a tailored equation of state is developed, which relaxes the particle system towards the target partition with good load balance. Since the equilibrium is computed by an iterative approach, the partitioning subdomains exhibit locality and the incremental property. Numerical experiments reveal that the proposed Centroidal Voronoi Particle (CVP) based algorithm produces high-quality partitioning with high efficiency, independently of computational-element types. Thus it can be used for a wide range of applications in computational science and engineering.

  5. Evolution of accelerometer methods for physical activity research.

    Science.gov (United States)

    Troiano, Richard P; McClain, James J; Brychta, Robert J; Chen, Kong Y

    2014-07-01

    The technology and application of current accelerometer-based devices in physical activity (PA) research allow the capture and storage or transmission of large volumes of raw acceleration signal data. These rich data not only provide opportunities to improve PA characterisation, but also bring logistical and analytic challenges. We discuss how researchers and developers from multiple disciplines are responding to the analytic challenges and how advances in data storage, transmission and big data computing will minimise logistical challenges. These new approaches also bring the need for several paradigm shifts for PA researchers, including a shift from count-based approaches and regression calibrations for PA energy expenditure (PAEE) estimation to activity characterisation and EE estimation based on features extracted from raw acceleration signals. Furthermore, a collaborative approach towards analytic methods is proposed to facilitate PA research, which requires a shift away from multiple independent calibration studies. Finally, we make the case for a distinction between PA represented by accelerometer-based devices and PA assessed by self-report. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://group.bmj.com/group/rights-licensing/permissions.

  6. The Fe removal in pyrophyllite by physical method

    Science.gov (United States)

    Cho, Kanghee; Jo, Jiyu; Bak, GeonYeong; Choi, NagChoul; Park*, CheonYoung

    2015-04-01

    The presence of Fe in ingredient material such as limestone, borax and pyrophyllite can prevent their use mainly in the glass fiber manufacturing industry. The red to yellow pigmentation in pyrophyllite is mainly due to the associated oxides and sulfides of Fe such as hematite, pyrite, etc. The removal of Fe in the pyrophyllite was investigated using high frequency treatment and magnetic separation under various alumina grades in pyrophyllite. The hematite and pyrite were observed in the pyrophyllite from photomicrograph and XRD analysis results. On the decrease of Al2O3 content in pyrophyllite was showed that SiO2, Fe2O3 and TiO2 content were increased by XRF analysis. The high frequency treatment experiment for the pyrophyllite showed that the (1) pyrite phase was transformed hematite and magnetite, (2) mass loss of the sample by volatilization of included sulfur(S) in pyrite. The results of magnetic separation for treated sample by high frequency were identified that Fe removal percent were in the range of 97.6~98.8%. This study demonstrated that physical method (high frequency treatment and magnetic separation) was effective for the removal of Fe in pyrophyllite. This subject is supported by Korea Ministry of Environment(MOE) as "Advanced Technology Program for Environmental Industry".

  7. An innovative way of thinking nuclear waste management - Neutron physics of a reactor directly operating on SNF.

    Science.gov (United States)

    Merk, Bruno; Litskevich, Dzianis; Bankhead, Mark; Taylor, Richard J

    2017-01-01

    A solution for the nuclear waste problem is the key challenge for an extensive use of nuclear reactors as a major carbon free, sustainable, and applied highly reliable energy source. Partitioning and Transmutation (P&T) promises a solution for improved waste management. Current strategies rely on systems designed in the 60's for the massive production of plutonium. We propose an innovative strategic development plan based on invention and innovation described with the concept of developments in s-curves identifying the current boundary conditions, and the evolvable objectives. This leads to the ultimate, universal vision for energy production characterized by minimal use of resources and production of waste, while being economically affordable and safe, secure and reliable in operation. This vision is transformed into a mission for a disruptive development of the future nuclear energy system operated by burning of existing spent nuclear fuel (SNF) without prior reprocessing. This highly innovative approach fulfils the sustainability goals and creates new options for P&T. A proof on the feasibility from neutronic point of view is given demonstrating sufficient breeding of fissile material from the inserted SNF. The system does neither require new resources nor produce additional waste, thus it provides a highly sustainable option for a future nuclear system fulfilling the requests of P&T as side effect. In addition, this nuclear system provides enhanced resistance against misuse of Pu and a significantly reduced fuel cycle. However, the new system requires a demand driven rethinking of the separation process to be efficient.

  8. A facile and efficient method of enzyme immobilization on silica particles via Michael acceptor film coatings: immobilized catalase in a plug flow reactor.

    Science.gov (United States)

    Bayramoglu, Gulay; Arica, M Yakup; Genc, Aysenur; Ozalp, V Cengiz; Ince, Ahmet; Bicak, Niyazi

    2016-06-01

    A novel method was developed for facile immobilization of enzymes on silica surfaces. Herein, we describe a single-step strategy for generating of reactive double bonds capable of Michael addition on the surfaces of silica particles. This method was based on reactive thin film generation on the surfaces by heating of impregnated self-curable polymer, alpha-morpholine substituted poly(vinyl methyl ketone) p(VMK). The generated double bonds were demonstrated to be an efficient way for rapid incorporation of enzymes via Michael addition. Catalase was used as model enzyme in order to test the effect of immobilization methodology by the reactive film surface through Michael addition reaction. Finally, a plug flow type immobilized enzyme reactor was employed to estimate decomposition rate of hydrogen peroxide. The highly stable enzyme reactor could operate continuously for 120 h at 30 °C with only a loss of about 36 % of its initial activity.

  9. The long-term future for civilian nuclear power generation in France: The case for breeder reactors. Breeder reactors: The physical and physical chemistry parameters, associate material thermodynamics and mechanical engineering: Novelties and issues; La production d'electricite d'origine nucleaire en France, dans le futur a long terme: Le cas des surgenerateurs: Les reacteurs nucleaires surgenerateurs: Les parametres physique et physico-chimiques, la thermodynamique associee des materiaux et de l'ingenierie mecanique: Nouveautes et options

    Energy Technology Data Exchange (ETDEWEB)

    Dautray, R. [Academie des sciences, 23, quai de Conti, 75270 Paris cedex 06 (France)

    2011-06-15

    The author gives a summary overview of the knowledge base acquired since the first breeder reactors became operational in the fifties. Neutron transport theory, thermal phenomena, reactor core cooling, various coolants used and envisioned for this function, fuel fabrication from separated materials, main equipment (pumps, valves, heat exchanges...) have now attained maturity, sufficient to implement sodium cooling circuits. However, the use of metallic sodium still raises certain severe questions in terms of safe handling and security considerations. The structural components, both inside the reactor core and outside (i.e. heat exchangers) are undergoing in-depth research so as to last longer. The fuel cycle, notably the re-fabrication of fuel elements and fertile elements, the case of transuranic elements, etc., call for studies into radiation induced phenomena, chemistry separation, separate or otherwise treatments for materials that have different radioactive, physical, thermodynamical, chemical and biological properties. The concerns that surround the definitive disposal of certain radioactive wastes could be qualitatively improved with respect to the pressurized water reactors (PWRs) in service today. Lastly, the author notes that breeder reactors eliminate the need for an isotope separation facility, and this constitutes a significant contribution to contain nuclear proliferation. France was in the forefront of nuclear breeder power generation science, technological research and also in the knowledge base related to breeder reactors. It is in the country's interest to pursue these efforts. (author)

  10. Multifunctional reactors

    NARCIS (Netherlands)

    Westerterp, K.R.

    1992-01-01

    Multifunctional reactors are single pieces of equipment in which, besides the reaction, other functions are carried out simultaneously. The other functions can be a heat, mass or momentum transfer operation and even another reaction. Multifunctional reactors are not new, but they have received much

  11. The cell method a purely algebraic computational method in physics and engineering

    CERN Document Server

    Ferretti, Elena

    2014-01-01

    The Cell Method (CM) is a computational tool that maintains critical multidimensional attributes of physical phenomena in analysis. This information is neglected in the differential formulations of the classical approaches of finite element, boundary element, finite volume, and finite difference analysis, often leading to numerical instabilities and spurious results. This book highlights the central theoretical concepts of the CM that preserve a more accurate and precise representation of the geometric and topological features of variables for practical problem solving. Important applications occur in fields such as electromagnetics, electrodynamics, solid mechanics and fluids. CM addresses non-locality in continuum mechanics, an especially important circumstance in modeling heterogeneous materials. Professional engineers and scientists, as well as graduate students, are offered: A general overview of physics and its mathematical descriptions; Guidance on how to build direct, discrete formulations; Coverag...

  12. Introduction to Methods of Approximation in Physics and Astronomy

    Science.gov (United States)

    van Putten, Maurice H. P. M.

    2017-04-01

    Modern astronomy reveals an evolving Universe rife with transient sources, mostly discovered - few predicted - in multi-wavelength observations. Our window of observations now includes electromagnetic radiation, gravitational waves and neutrinos. For the practicing astronomer, these are highly interdisciplinary developments that pose a novel challenge to be well-versed in astroparticle physics and data analysis. In realizing the full discovery potential of these multimessenger approaches, the latter increasingly involves high-performance supercomputing. These lecture notes developed out of lectures on mathematical-physics in astronomy to advanced undergraduate and beginning graduate students. They are organised to be largely self-contained, starting from basic concepts and techniques in the formulation of problems and methods of approximation commonly used in computation and numerical analysis. This includes root finding, integration, signal detection algorithms involving the Fourier transform and examples of numerical integration of ordinary differential equations and some illustrative aspects of modern computational implementation. In the applications, considerable emphasis is put on fluid dynamical problems associated with accretion flows, as these are responsible for a wealth of high energy emission phenomena in astronomy. The topics chosen are largely aimed at phenomenological approaches, to capture main features of interest by effective methods of approximation at a desired level of accuracy and resolution. Formulated in terms of a system of algebraic, ordinary or partial differential equations, this may be pursued by perturbation theory through expansions in a small parameter or by direct numerical computation. Successful application of these methods requires a robust understanding of asymptotic behavior, errors and convergence. In some cases, the number of degrees of freedom may be reduced, e.g., for the purpose of (numerical) continuation or to identify

  13. Characterisation of radiation field for irradiation of biological samples at nuclear reactor-comparison of twin detector and recombination methods.

    Science.gov (United States)

    Golnik, N; Gryziński, M A; Kowalska, M; Meronka, K; Tulik, P

    2014-10-01

    Central Laboratory for Radiological Protection is involved in achieving scientific project on biological dosimetry. The project includes irradiation of blood samples in radiation fields of nuclear reactor. A simple facility for irradiation of biological samples has been prepared at horizontal channel of the nuclear reactor MARIA in NCBJ in Poland. The radiation field, composed mainly of gamma radiation and thermal neutrons, has been characterised in terms of tissue kerma using twin-detector technique and recombination chambers. © The Author 2013. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  14. Study of evaluation methods for in-vessel corium retention through external vessel cooling and safety of reactor cavity

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hoon; Chang, Soon Heung; Kim, Soo Hyung; Kim, Kee Poong; Lee, Hyoung Wook; Jang, Kwang Keol; Jeong, Yong Hoon; Kim, Sang Jin; Lee, Seong Jin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Park, Jae Hong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    2001-03-15

    In this work, assessment system for methodology for reactor pressure vessel integrity is developed. Assessment system is make up of severe accident assessment code which can calculate the conditions of plant and structural analysis code which can assess the integrity of reactor vessel using given plant conditions. An assessment of cavity flooding using containment spray system has been done. As a result, by the containment spray, cavity can be flooded successfully and CCI can be reduced. The technical backgrounds for external vessel cooling and corium cooling on the cavity are summarized and provided in this report.

  15. Methods to Measure Physical Activity Behaviors in Health Education Research

    Science.gov (United States)

    Fitzhugh, Eugene C.

    2015-01-01

    Regular physical activity (PA) is an important concept to measure in health education research. The health education researcher might need to measure physical activity because it is the primary measure of interest, or PA might be a confounding measure that needs to be controlled for in statistical analysis. The purpose of this commentary is to…

  16. An innovative way of thinking nuclear waste management – Neutron physics of a reactor directly operating on SNF

    Science.gov (United States)

    Litskevich, Dzianis; Bankhead, Mark; Taylor, Richard J.

    2017-01-01

    A solution for the nuclear waste problem is the key challenge for an extensive use of nuclear reactors as a major carbon free, sustainable, and applied highly reliable energy source. Partitioning and Transmutation (P&T) promises a solution for improved waste management. Current strategies rely on systems designed in the 60’s for the massive production of plutonium. We propose an innovative strategic development plan based on invention and innovation described with the concept of developments in s-curves identifying the current boundary conditions, and the evolvable objectives. This leads to the ultimate, universal vision for energy production characterized by minimal use of resources and production of waste, while being economically affordable and safe, secure and reliable in operation. This vision is transformed into a mission for a disruptive development of the future nuclear energy system operated by burning of existing spent nuclear fuel (SNF) without prior reprocessing. This highly innovative approach fulfils the sustainability goals and creates new options for P&T. A proof on the feasibility from neutronic point of view is given demonstrating sufficient breeding of fissile material from the inserted SNF. The system does neither require new resources nor produce additional waste, thus it provides a highly sustainable option for a future nuclear system fulfilling the requests of P&T as side effect. In addition, this nuclear system provides enhanced resistance against misuse of Pu and a significantly reduced fuel cycle. However, the new system requires a demand driven rethinking of the separation process to be efficient. PMID:28749952

  17. An innovative way of thinking nuclear waste management - Neutron physics of a reactor directly operating on SNF.

    Directory of Open Access Journals (Sweden)

    Bruno Merk

    Full Text Available A solution for the nuclear waste problem is the key challenge for an extensive use of nuclear reactors as a major carbon free, sustainable, and applied highly reliable energy source. Partitioning and Transmutation (P&T promises a solution for improved waste management. Current strategies rely on systems designed in the 60's for the massive production of plutonium. We propose an innovative strategic development plan based on invention and innovation described with the concept of developments in s-curves identifying the current boundary conditions, and the evolvable objectives. This leads to the ultimate, universal vision for energy production characterized by minimal use of resources and production of waste, while being economically affordable and safe, secure and reliable in operation. This vision is transformed into a mission for a disruptive development of the future nuclear energy system operated by burning of existing spent nuclear fuel (SNF without prior reprocessing. This highly innovative approach fulfils the sustainability goals and creates new options for P&T. A proof on the feasibility from neutronic point of view is given demonstrating sufficient breeding of fissile material from the inserted SNF. The system does neither require new resources nor produce additional waste, thus it provides a highly sustainable option for a future nuclear system fulfilling the requests of P&T as side effect. In addition, this nuclear system provides enhanced resistance against misuse of Pu and a significantly reduced fuel cycle. However, the new system requires a demand driven rethinking of the separation process to be efficient.

  18. General meeting. Technical reunion: the numerical and experimental simulation applied to the Reactor Physics; Assemblee generale. Reunion technique: la simulation numerique et experimentale appliquee a la physique des reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-10-01

    The SFEN (French Society on Nuclear Energy), organized the 18 october 2001 at Paris, a technical day on the numerical and experimental simulation, applied to the reactor Physics. Nine aspects were discussed, giving a state of the art in the domain:the french nuclear park; the future technology; the controlled thermonuclear fusion; the new organizations and their implications on the research and development programs; Framatome-ANP markets and industrial code packages; reactor core simulation at high temperature; software architecture; SALOME; DESCARTES. (A.L.B.)

  19. Physics and technology of nuclear materials

    CERN Document Server

    Ursu, Ioan

    2015-01-01

    Physics and Technology of Nuclear Materials presents basic information regarding the structure, properties, processing methods, and response to irradiation of the key materials that fission and fusion nuclear reactors have to rely upon. Organized into 12 chapters, this book begins with selectively several fundamentals of nuclear physics. Subsequent chapters focus on the nuclear materials science; nuclear fuel; structural materials; moderator materials employed to """"slow down"""" fission neutrons; and neutron highly absorbent materials that serve in reactor's power control. Other chapters exp

  20. A Synthetic Approach to the Transfer Matrix Method in Classical and Quantum Physics

    Science.gov (United States)

    Pujol, O.; Perez, J. P.

    2007-01-01

    The aim of this paper is to propose a synthetic approach to the transfer matrix method in classical and quantum physics. This method is an efficient tool to deal with complicated physical systems of practical importance in geometrical light or charged particle optics, classical electronics, mechanics, electromagnetics and quantum physics. Teaching…

  1. Physical developer method for detection of latent fingerprints: A review

    Directory of Open Access Journals (Sweden)

    G.S. Sodhi

    2016-06-01

    Full Text Available The physical developer technique is a means to detect fingerprints on dry and wet, porous items, including paper articles, clay-based products and adhesive tapes. The process involves an oxidation–reduction couple whereby a solution of an iron salt reduces aqueous silver nitrate to finely divided metallic silver. The technique derives its name from the photographic physical developer which, during processing of film rolls, undergoes a similar redox reaction. The physical developer reveals the fingerprints as dark gray or black images due to the adsorption of metallic silver particles on the fatty acid and lipid components of sweat residue.

  2. Engineering and Physics Optimization of Breed and Burn Fast Reactor Systems; NUCLEAR ENERGY RESEARCH INITIATIVE (NERI) QUARTERLY PROGRESS REPORT

    Energy Technology Data Exchange (ETDEWEB)

    ERROR, [value too long for type character varying(50); Hejzlar, Pavel; Yarsky, Peter; Driscoll, Mike; Wachs, Dan; Weaver, Kevan; Czerwinski, Ken; Pope, Mike; Parry, James; Marshall, Theron D.; Davis, Cliff B.; Crawford, Dustin; Hartmann, Thomas; Saha, Pradip

    2005-01-31

    This project is organized under four major tasks (each of which has two or more subtasks) with contributions among the three collaborating organizations (MIT, INEEL and ANL-West): Task A: Core Physics and Fuel Cycle; Task B: Core Thermal Hydraulics; Task C: Plant Design; Task D: Fuel Design The lead PI, Michael J. Driscoll, has consolidated and summarized the technical progress submissions provided by the contributing investigators from all sites, under the above principal task headings.

  3. Mathematical Methods For Students of Physics and Related Fields

    CERN Document Server

    Hassani, Sadri

    2009-01-01

    Intended to follow the usual introductory physics courses, this book has the unique feature of addressing the mathematical needs of sophomores and juniors in physics, engineering and other related fields. Many original, lucid, and relevant examples from the physical sciences, problems at the ends of chapters, and boxes to emphasize important concepts help guide the student through the material. Beginning with reviews of vector algebra and differential and integral calculus, the book continues with infinite series, vector analysis, complex algebra and analysis, ordinary and partial differential equations. Discussions of numerical analysis, nonlinear dynamics and chaos, and the Dirac delta function provide an introduction to modern topics in mathematical physics. This new edition has been made more user-friendly through organization into convenient, shorter chapters. Also, it includes an entirely new section on Probability and plenty of new material on tensors and integral transforms. Some praise for the previo...

  4. Mathematical methods for students of physics and related fields

    CERN Document Server

    Hassani, Sadri

    2000-01-01

    Intended to follow the usual introductory physics courses, this book has the unique feature of addressing the mathematical needs of sophomores and juniors in physics, engineering and other related fields Many original, lucid, and relevant examples from the physical sciences, problems at the ends of chapters, and boxes to emphasize important concepts help guide the student through the material Beginning with reviews of vector algebra and differential and integral calculus, the book continues with infinite series, vector analysis, complex algebra and analysis, ordinary and partial differential equations Discussions of numerical analysis, nonlinear dynamics and chaos, and the Dirac delta function provide an introduction to modern topics in mathematical physics This new edition has been made more user-friendly through organization into convenient, shorter chapters Also, it includes an entirely new section on Probability and plenty of new material on tensors and integral transforms Some praise for the previous edi...

  5. Status of Physics and Safety Analyses for the Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Ingersoll, DT

    2005-12-15

    A study has been completed to develop a new baseline core design for the liquid-salt-cooled very high-temperature reactor (LS-VHTR) that is better optimized for liquid coolant and that satisfies the top-level operational and safety targets, including strong passive safety performance, acceptable fuel cycle parameters, and favorable core reactivity response to coolant voiding. Three organizations participated in the study: Oak Ridge National Laboratory (ORNL), Idaho National Laboratory (INL), and Argonne National Laboratory (ANL). Although the intent was to generate a new reference LS-VHTR core design, the emphasis was on performing parametric studies of the many variables that constitute a design. The results of the parametric studies not only provide the basis for choosing the optimum balance of design options, they also provide a valuable understanding of the fundamental behavior of the core, which will be the basis of future design trade-off studies. A new 2400-MW(t) baseline design was established that consists of a cylindrical, nonannular core cooled by liquid {sup 7}Li{sub 2}BeF{sub 4} (Flibe) salt. The inlet and outlet coolant temperatures were decreased by 50 C, and the coolant channel diameter was increased to help lower the maximum fuel and vessel temperatures. An 18-month fuel cycle length with 156 GWD/t burnup was achieved with a two-batch shuffling scheme, while maintaining a core power density of 10 MW/m{sup 3} using graphite-coated uranium oxicarbide particle fuel enriched to 15% {sup 235}U and assuming a 25 vol-% packing of the coated particles in the fuel compacts. The revised design appears to have excellent steady-state and transient performance. The previous concern regarding the core's response to coolant voiding has been resolved for the case of Flibe coolant by increasing the coolant channel diameter and the fuel loading. Also, the LSVHTR has a strong decay heat removal performance and appears capable of surviving a loss of forced

  6. The evaluation method of some physical qualities of boxers

    Directory of Open Access Journals (Sweden)

    Ivanov V.I.

    2010-12-01

    Full Text Available Possibilities of determination of the special physical qualities of boxers are considered by a trainer. In an experiment 40 sportsmen took part in the of age 17 - 19 years. It is rotined that application of the offered trainer will be given by possibility to the trainer constantly to watch change in the special physical preparedness of sportsmen and to bring in urgent corrective in a training process. Resulted recommendation as evaluated by the special capacity of sportsmen.

  7. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  8. Methods to Implement Innovation and Entrepreneurship in Physics

    Science.gov (United States)

    Arion, Douglas

    2015-03-01

    The physics community is beginning to become aware of the benefits of entrepreneurship and innovation education: greater enrollments, improved students satisfaction, a wider range of interesting research problems, and the potential for greater return from more successful alumni. This talk will suggest a variety of mechanisms by which physics departments can include entrepreneurship and innovation content within their programs - without necessarily requiring earth-shattering changes to the curriculum. These approaches will thus make it possible for departments to get involved with entrepreneurship and innovation, and grow those components into vibrant activities for students and faculty.

  9. NUCLEAR REACTOR

    Science.gov (United States)

    Grebe, J.J.

    1959-07-14

    High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

  10. Results of Koo measurements of HTGR lattice by oscillated zero reactivity technique using the AGIP-NUCLEARE RB-2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Benedetti, F; Brighenti, G.; Chiodi, P.L.; Ghilardotti, G.; Giuliani, C.

    1974-10-15

    This paper describes k-infinity measurements conducted using an assembly of loose HTGR coated particles in the BR-2 reactor by means of null reactivity oscillating method comparing the effect of poisoned and unpoisoned lattices like tests performed in the Physical Constants Test Reactor (PCTR) at Hanford. The RB-2 reactor was the property of the Italian firm AGIP NUCLEARE and operated at the Montecuccolino Center in Bologna.

  11. Integration of educational methods and physical settings: Design ...

    African Journals Online (AJOL)

    Quality design and appropriate space organization in preschool settings can support preschool children's educational activities. Although the relationship between the well-being and development of children and physical settings has been emphasized by many early childhood researchers, there is still a need for theoretical ...

  12. Harmonisation of physical and chemical methods for soil ...

    African Journals Online (AJOL)

    As part of a collaborative project to investigate human impacts on Quercus suber L. (cork oak) forests, five research groups from countries in Europe and North Africa undertook a survey of soil quality (physical properties, potentially toxic elements) at sites in NW Tunisia and NW Sardinia. All groups performed the analysis of ...

  13. Continuous-Flow Biochips: Technology, Physical Design Methods and Testing

    DEFF Research Database (Denmark)

    Pop, Paul; Araci, Ismail Emre; Chakrabarty, Krishnendu

    2015-01-01

    This article is a tutorial on continuous-flow biochips where the basic building blocks are microchannels, and microvalves, and by combining them, more complex units such as mixers, switches, and multiplexers can be built. It also presents the state of the art in flow-based biochip technology and ...... and emerging research challenges in the areas of physical design and testing techniques....

  14. Physical-chemical and operational performance of an anaerobic baffled reactor (ABR treating swine wastewater - 10.4025/actascitechnol.v32i4.7203

    Directory of Open Access Journals (Sweden)

    Erlon Lopes Pereira

    2010-12-01

    Full Text Available Since hog raising concentrates a huge amount of swine manure in small areas, it is considered by the environmental government organizations to be one of the most potentially pollutant activities. Therefore the main objective of this research was to evaluate by operational criteria and removal efficiency, the performance of a Anaerobic Baffled Reactor (ABR, working as a biological pre-treatment of swine culture effluents. The physical-chemical analyses carried out were: total COD, BOD5, total solids (TS, fix (TFS and volatiles (TVS, temperature, pH, total Kjeldahl nitrogen, phosphorus, total acidity and alkalinity. The ABR unit worked with an average efficiency of 65.2 and 76.2%, respectively, concerning total COD and BOD5, with a hydraulic retention time (HRT about 15 hours. The results for volumetric organic loading rate (VOLR, organic loading rate (OLR and hydraulic loading rate (HLR were: 4.46 kg BOD m-3 day-1; 1.81 kg BOD5 kg TVS-1 day-1 and 1.57 m3 m-3 day-1, respectively. The average efficiency of the whole treatment system for total COD and BOD5 removal were 66.5 and 77.8%, showing an adequate performance in removing the organic matter from swine wastewater.

  15. Armouring facility? Nuclear-weapon and reactor reseach at the Kaiser-Wilhelm Institute for Physics; Eine Waffenschmiede? Kernwaffen- und Reaktorforschung am Kaiser-Wilhelm-Institut fuer Physik

    Energy Technology Data Exchange (ETDEWEB)

    Hachtmann, R. (ed.); Walker, M.

    2005-07-01

    The Kaiser Wilhelm Institute for Physics is best known as the place where Werner Heisenberg worked on nuclear weapons for Hitler. Although this is essentially true, there is more to the story. At the start of World War II this institute was taken over by the German Army Ordnance to be the central, but not exclusive site for a research project into the economic and military applications of nuclear fission. The Army physicist Kurt Diebner was installed in the institute as its commissarial director. Heisenberg was affiliated with the institute as an advisor at first, and became the director in 1942. Heisenberg and his colleagues, including in particular Karl-Heinz Hoecker, Carl Friedrich von Weizsaecker, and Karl Wirtz, worked on nuclear reactors and isotope separation with the clear knowledge that these were two different paths to atomic bombs [Atombomben]. However, they were clearly ambivalent about what they were doing. New documents recently returned from Russian archives shed new light on this work and the scientists' motivations. (orig.)

  16. Studies on solid-state physics carried out with the Saclay reactor (1962); Etudes de physique du solide realisees a la pile de Saclay (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Herpin, A. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1962-07-01

    This paper deals only with solid-state physics experiments carried out on outgoing beams: rather than giving a general review of the work performed, if refers to only a few of the most important studies or those nearest completion. These are being made with the experimental beams of the two Saclay reactors EL-2, with a central flux of 10{sup 13} n/cm{sup 2}, and - since 1958 - EL-3, whose central flux is equal ta 10{sup 14} n/cm{sup 2}. The experiments are being carried out by two separate groups of physicists, employing different techniques, namely neutron diffraction using a crystal spectrometer, and inelastic scattering using a time-of-flight spectrometer. (author) [French] Cet expose ne relate que des experiences de physique du solide faites sur des faisceaux sortis; plutot que de donner une revue de l'ensemble des travaux effectues, on ne cite que quelques etudes que l'on peut considerer comme plus essentielles ou mieux achevees. On utilise les faisceaux experimentaux des deux piles de Saclay, EL-2 dont le flux au centre est de 10{sup 13}n/cm{sup 2} et, depuis 1958, EL-3 pour laquelle il est egal a 10{sup 14} n/cm{sup 2}. Les experiences sont realisees par deux groupes de physiciens distincts, employant des techniques differentes, la diffraction des neutrons qui utilise un spectrometre a cristal, et la diffusion inelastique avec un spectrometre a temps de vol. (auteur)

  17. Analysis of dpa Rates in the HFIR Reactor Vessel using a Hybrid Monte Carlo/Deterministic Method*

    Directory of Open Access Journals (Sweden)

    Risner J.M.

    2016-01-01

    Full Text Available The Oak Ridge High Flux Isotope Reactor (HFIR, which began full-power operation in 1966, provides one of the highest steady-state neutron flux levels of any research reactor in the world. An ongoing vessel integrity analysis program to assess radiation-induced embrittlement of the HFIR reactor vessel requires the calculation of neutron and gamma displacements per atom (dpa, particularly at locations near the beam tube nozzles, where radiation streaming effects are most pronounced. In this study we apply the Forward-Weighted Consistent Adjoint Driven Importance Sampling (FW-CADIS technique in the ADVANTG code to develop variance reduction parameters for use in the MCNP radiation transport code. We initially evaluated dpa rates for dosimetry capsule locations, regions in the vicinity of the HB-2 beamline, and the vessel beltline region. We then extended the study to provide dpa rate maps using three-dimensional cylindrical mesh tallies that extend from approximately 12 in. below to approximately 12 in. above the height of the core. The mesh tally structures contain over 15,000 mesh cells, providing a detailed spatial map of neutron and photon dpa rates at all locations of interest. Relative errors in the mesh tally cells are typically less than 1%.

  18. Analysis of dpa rates in the HFIR reactor vessel using a hybrid Monte Carlo/deterministic method

    Energy Technology Data Exchange (ETDEWEB)

    Blakeman, Edward [Retired

    2016-01-01

    The Oak Ridge High Flux Isotope Reactor (HFIR), which began full-power operation in 1966, provides one of the highest steady-state neutron flux levels of any research reactor in the world. An ongoing vessel integrity analysis program to assess radiation-induced embrittlement of the HFIR reactor vessel requires the calculation of neutron and gamma displacements per atom (dpa), particularly at locations near the beam tube nozzles, where radiation streaming effects are most pronounced. In this study we apply the Forward-Weighted Consistent Adjoint Driven Importance Sampling (FW-CADIS) technique in the ADVANTG code to develop variance reduction parameters for use in the MCNP radiation transport code. We initially evaluated dpa rates for dosimetry capsule locations, regions in the vicinity of the HB-2 beamline, and the vessel beltline region. We then extended the study to provide dpa rate maps using three-dimensional cylindrical mesh tallies that extend from approximately 12 below to approximately 12 above the axial extent of the core. The mesh tally structures contain over 15,000 mesh cells, providing a detailed spatial map of neutron and photon dpa rates at all locations of interest. Relative errors in the mesh tally cells are typically less than 1%.

  19. Request for Naval Reactors Comment on Proposed Prometheus Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to JPL

    Energy Technology Data Exchange (ETDEWEB)

    D. Kokkinos

    2005-04-28

    The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.

  20. Reactor engineering department annual report. April 1, 1995 - March 31, 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1995 (April 1, 1995 - March 31, 1996). The major Department`s programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermalhydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermalhydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  1. Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-10-01

    This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department`s programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC`s fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)

  2. Reactor Engineering Department annual report. April 1, 1997 - March 31, 1998

    Energy Technology Data Exchange (ETDEWEB)

    Ochiai, Masaaki; Ohnuki, Akira; Ono, Toshihiko [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    1998-11-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1997 (April 1, 1997 - March 31, 1998). The major Department`s programs promoted in the year are the achievement of the world-strongest lasing of Free Electron Laser and the verification of the core thermal integrity during design basis events in PWRs. Other Major tasks of the Department are various basic researches on the advanced reactor system design studies, the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  3. Reactor engineering department annual report. April 1, 1994 - March 31, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1994 (April 1, 1994 - March 31, 1995). The major Department`s programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author).

  4. Texting to increase physical activity among teenagers (TXT Me!): Rationale, design, and methods proposal

    Science.gov (United States)

    Physical activity decreases from childhood through adulthood. Among youth, teenagers (teens) achieve the lowest levels of physical activity, and high school age youth are particularly at risk of inactivity. Effective methods are needed to increase youth physical activity in a way that can be maintai...

  5. Two modelling approaches to water-quality simulation in a flooded iron-ore mine (Saizerais, Lorraine, France): a semi-distributed chemical reactor model and a physically based distributed reactive transport pipe network model.

    Science.gov (United States)

    Hamm, V; Collon-Drouaillet, P; Fabriol, R

    2008-02-19

    The flooding of abandoned mines in the Lorraine Iron Basin (LIB) over the past 25 years has degraded the quality of the groundwater tapped for drinking water. High concentrations of dissolved sulphate have made the water unsuitable for human consumption. This problematic issue has led to the development of numerical tools to support water-resource management in mining contexts. Here we examine two modelling approaches using different numerical tools that we tested on the Saizerais flooded iron-ore mine (Lorraine, France). A first approach considers the Saizerais Mine as a network of two chemical reactors (NCR). The second approach is based on a physically distributed pipe network model (PNM) built with EPANET 2 software. This approach considers the mine as a network of pipes defined by their geometric and chemical parameters. Each reactor in the NCR model includes a detailed chemical model built to simulate quality evolution in the flooded mine water. However, in order to obtain a robust PNM, we simplified the detailed chemical model into a specific sulphate dissolution-precipitation model that is included as sulphate source/sink in both a NCR model and a pipe network model. Both the NCR model and the PNM, based on different numerical techniques, give good post-calibration agreement between the simulated and measured sulphate concentrations in the drinking-water well and overflow drift. The NCR model incorporating the detailed chemical model is useful when a detailed chemical behaviour at the overflow is needed. The PNM incorporating the simplified sulphate dissolution-precipitation model provides better information of the physics controlling the effect of flow and low flow zones, and the time of solid sulphate removal whereas the NCR model will underestimate clean-up time due to the complete mixing assumption. In conclusion, the detailed NCR model will give a first assessment of chemical processes at overflow, and in a second time, the PNM model will provide more

  6. Participation in the US Department of Energy reactor sharing program. Final progress report, October 1996--September 1997

    Energy Technology Data Exchange (ETDEWEB)

    Mulder, R.U.; Benneche, P.E.; Hosticka, B.

    1998-04-01

    The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would not be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA from use at their institutions. These areas are discussed in this report.

  7. Participation in the US Department of Energy Reactor Sharing Program. Annual report, September 30, 1993--September 29, 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-03-01

    The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would not be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA for use at their institutions. These areas are discussed further in the report.

  8. A digital method for period measurements in a nuclear reactor; Um metodo digital para medidas de periodo em um reator nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Mundim, Sergio Gorretta

    1971-02-15

    The present paper begins by giving a theoretical treatment for the nuclear reactor period. The conventional method of measuring the period is analysed and some previously developed digital methods are described. The paper criticises the latter, pointing out some deficiencies which the proposed process is able to eliminate. All errors connected with this process are also analysed. The paper presents suitable solutions to reduce them to a minimum. The total error is found to he less than the error presented by the other methods described. A digital period meter is designed with memory resources and an automatic scaler changer. Integrated circuits specifications are used in it. Real time experiments with nuclear reactors were made in order to check te validity of the method. The data acquired were applied to a simulated digital period meter implemented in a general purpose computer. The nuclear part of the work was developed at the 'Comissao Nacional de Energia Nuclear' and the simulation work was dane at the 'Departamento de Calculo Cientifico' of COPPE, which also advised the author in the completion of this thesis. (author)

  9. Tandem Mirror Reactor Systems Code (Version I)

    Energy Technology Data Exchange (ETDEWEB)

    Reid, R.L.; Finn, P.A.; Gohar, M.Y.; Barrett, R.J.; Gorker, G.E.; Spampinaton, P.T.; Bulmer, R.H.; Dorn, D.W.; Perkins, L.J.; Ghose, S.

    1985-09-01

    A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost.

  10. Reactor monitoring and safeguards using antineutrino detectors

    CERN Document Server

    Bowden, N S

    2008-01-01

    Nuclear reactors have served as the antineutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these very weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Measurements made with antineutrino detectors could therefore offer an alternative means for verifying the power history and fissile inventory of a reactors, as part of International Atomic Energy Agency (IAEA) and other reactor safeguards regimes. Several efforts to develop this monitoring technique are underway across the globe.

  11. A Structured Peer-Mentoring Method for Physical Activity Behavior Change Among Adolescents

    OpenAIRE

    Smith, Laureen H.; Petosa, Rick L.

    2016-01-01

    Despite national guidelines for regular physical activity, most adolescents are not physically active. Schools serve an estimated 60 million youth and provide an educational environment to meet the current physical activity guidelines. The obesity epidemic and chronic disease comorbidities associated with physical inactivity are not likely to be reversed without a strong contribution from local schools. This article describes how a structured peer-mentoring method provides a feasible, flexibl...

  12. Review of Physical and Chemical Methods for Characterization of Fuels

    Science.gov (United States)

    1981-12-01

    Reversed-phase chromatography REF Refractometry CAL Calorimetry POT Potentiometry CC Coordination chromatography MSB Mossbauer spectroscopy FS Flame...2896 REFERENCE: P-95 DESCRIPTION: Total base number of petroleum products by potentiometric perchloric acid titration TEST METHOD: Potentiometry ...DESCRIPTION: Determination of chloride in methyl fuel TEST METHOD: Potentiometry APPLICATION: Methyl fuel SCOPE: Determination of chloride ion concentration

  13. Mathematical methods for mathematicians, physical scientists and engineers

    CERN Document Server

    Dunning-Davies, J

    2003-01-01

    This practical introduction encapsulates the entire content of teaching material for UK honours degree courses in mathematics, physics, chemistry and engineering, and is also appropriate for post-graduate study. It imparts the necessary mathematics for use of the techniques, with subject-related worked examples throughout. The text is supported by challenging problem exercises (and answers) to test student comprehension. Index notation used in the text simplifies manipulations in the sections on vectors and tensors. Partial differential equations are discussed, and special functions introduced

  14. Results of a Neutronic Simulation of HTR-Proteus Core 4.2 using PEBBED and other INL Reactor Physics Tools: FY-09 Report

    Energy Technology Data Exchange (ETDEWEB)

    Hans D. Gougar

    2009-08-01

    The Idaho National Laboratory’s deterministic neutronics analysis codes and methods were applied to the computation of the core multiplication factor of the HTR-Proteus pebble bed reactor critical facility. A combination of unit cell calculations (COMBINE-PEBDAN), 1-D discrete ordinates transport (SCAMP), and nodal diffusion calculations (PEBBED) were employed to yield keff and flux profiles. Preliminary results indicate that these tools, as currently configured and used, do not yield satisfactory estimates of keff. If control rods are not modeled, these methods can deliver much better agreement with experimental core eigenvalues which suggests that development efforts should focus on modeling control rod and other absorber regions. Under some assumptions and in 1D subcore analyses, diffusion theory agrees well with transport. This suggests that developments in specific areas can produce a viable core simulation approach. Some corrections have been identified and can be further developed, specifically: treatment of the upper void region, treatment of inter-pebble streaming, and explicit (multiscale) transport modeling of TRISO fuel particles as a first step in cross section generation. Until corrections are made that yield better agreement with experiment, conclusions from core design and burnup analyses should be regarded as qualitative and not benchmark quality.

  15. Formation, physical characteristics and microbial community structure of aerobic granules in a pilot-scale sequencing batch reactor for real wastewater treatment.

    Science.gov (United States)

    Liu, Yong-Qiang; Moy, Benjamin; Kong, Yun-Hua; Tay, Joo-Hwa

    2010-05-05

    In this study, aerobic granular sludge was successfully developed in a pilot-scale sequencing batch reactor (SBR) installed on site to treat real wastewater using traditional activated sludge as inoculum. Compared with 1 or 2 months required by lab-scale reactor for aerobic granulation, it took about 400 days for activated sludge to transform into granule-dominant sludge in the pilot-scale SBR on site. Although the sludge in the reactor after 400-day operation was a mixture of flocs and granules with floc ratio ranged from 5 to 30%, sludge volume index with 5min settling (SVI5) always maintained at around 30mL/g. The similar microbial community structures represented by denaturing gradient gel electrophoresis (DGGE) between coexisted flocs and granules in the reactor indicated no strong microbial selection after the granules were dominant in the reactor. Chemical oxygen demand (COD) and NH4(+)-N removal efficiencies were above 80 and 98%, respectively, after 50-day operation, and the total inorganic N removal efficiency was about 50%. The results in this study demonstrate that it is feasible to form aerobic granules in pilot-scale SBR reactor and maintain the long-term stability of granular sludge with a high influent quality fluctuation. Meanwhile, stable COD and NH4(+)-N removal efficiencies can be obtained in the reactor. Copyright © 2010 Elsevier Inc. All rights reserved.

  16. Five Lectures on Nuclear Reactors Presented at Cal Tech

    Science.gov (United States)

    Weinberg, Alvin M.

    1956-02-10

    The basic issues involved in the physics and engineering of nuclear reactors are summarized. Topics discussed include theory of reactor design, technical problems in power reactors, physical problems in nuclear power production, and future developments in nuclear power. (C.H.)

  17. Coupling of partitioned physics codes with quasi-Newton methods

    CSIR Research Space (South Africa)

    Haelterman, R

    2017-03-01

    Full Text Available –51 (1997) [10] M.A. Gomez-Ruggiero, J.M. Martinez, The Column-Updating Method for solving nonlinear equations in Hilbert space. RAIRO Mathematical Modelling and Numerical Analysis 26, pp. 309–330 (1992) [11] R. Haelterman, Analytical Study of the Least...). [14] V.L.R. Lopes, J.M. Martinez, Convergence properties of the Inverse Column-Updating Method. Optim. Methods Softw. 6, pp. 127–144 (1995) [15] J.M. Martinez, L.S. Ochi, Sobre Dois Metodos de Broyden. Mat. Apl. Comput. 1/2, pp. 135–143 (1982) [16] J...

  18. Pacific Northwest Laboratory Monthly Activities Report APRIL 1966 on AEC Division of Reactor Development and Technology

    Energy Technology Data Exchange (ETDEWEB)

    S. L. Fawcett

    1966-05-01

    This report has the following sections: Summary of Activities; Civilian Power Reactors; Applied and Reactor Physics; Reactor Fuels and Materials; Engineering Development; Plutonium Recycle Program; Advanced Systems; and Nuclear Safety.

  19. Pacific Northwest Laboratory Monthly Activities Report March 1966 On AEC Division of Reactor Development and Technology

    Energy Technology Data Exchange (ETDEWEB)

    S. L. Fawcett

    1966-04-01

    This report has the following sections: Summary of Activities; Civilian Power Reactors; Applied and Reactor Physics; Reactor Fuels and Materials; Engineering Development; Plutonium Recycle Program; Advanced Systems; and Nuclear Safety.

  20. Research and development program in reactor diagnostics and monitoring with neutron noise methods. Stage 7. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Pazsit, I.; Demaziere, C.; Arzhanov, V. [Chalmers Univ. of Technology, Goeteborg (Sweden). Department of Reactor Physics; Garis, N.S. [Swedish Nuclear Power Inspectorate, Stockholm (Sweden)

    2001-08-01

    This report constitutes stage 7 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. A proposal for the continuation of this program in stage 8 is also given at the end of the report. In stage 6, the basic principles of a 3-D fully coupled neutronic/thermal-hydraulic simulator in the frequency domain were presented. The neutronic model relied on the two-group diffusion approximation, whereas the thermal-hydraulic algorithms relied on the so called 'lumped' model. The key element of this simulator was that only the static data were required which could be obtained from the Studsvik Scandpower CASMO-4/TABLES-3/ SIMULATE-3 code package. The simulator was developed with this underlying idea, which means that the calculation of the static fluxes and the eigenvalue were avoided. Depending on what kind of spatial discretization scheme which is used in the noise simulator to calculate the 'leakage' noise, it is not granted that the system remains critical by using the group constants supplied by SIMULATE. Nevertheless, when the system is critical, the balance equations should be fulfilled in all nodes with respect to the discretization scheme used. In concrete terms, the calculation of the static fluxes and eigenvalue can be avoided if the system is brought back to criticality by modifying the cross-sections so that the balance equations are always fulfilled with the chosen spatial discretization scheme. This approach was used in this study with the finite difference scheme. As pointed out in stage 6, the finite difference scheme is relatively inefficient compared to finite elements or nodal methods, but on the other hand it is rather easy to implement. These two more sophisticated schemes are planned to be investigated at a later stage, but for the time being the simulator relying on the finite difference scheme was improved as much as possible so that a 2-D entirely