Uddin, M.N. [Department of Physics, Jahangirnagar University, Dhaka (Bangladesh); Sarker, M.M. [Reactor Physics and Engineering Division, Institute of Nuclear Science and Technology, Atomic Energy Research Establishment, Savar, GPO Box 3787, Dhaka 1000 (Bangladesh); Khan, M.J.H. [Reactor Physics and Engineering Division, Institute of Nuclear Science and Technology, Atomic Energy Research Establishment, Savar, GPO Box 3787, Dhaka 1000 (Bangladesh)], E-mail: jahirulkhan@yahoo.com; Islam, S.M.A. [Department of Physics, Jahangirnagar University, Dhaka (Bangladesh)
2009-10-15
The aim of this paper is to present the validation of evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 through the analysis of the integral parameters of TRX and BAPL benchmark lattices of thermal reactors for neutronics analysis of TRIGA Mark-II Research Reactor at AERE, Bangladesh. In this process, the 69-group cross-section library for lattice code WIMS was generated using the basic evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 with the help of nuclear data processing code NJOY99.0. Integral measurements on the thermal reactor lattices TRX-1, TRX-2, BAPL-UO{sub 2}-1, BAPL-UO{sub 2}-2 and BAPL-UO{sub 2}-3 served as standard benchmarks for testing nuclear data files and have also been selected for this analysis. The integral parameters of the said lattices were calculated using the lattice transport code WIMSD-5B based on the generated 69-group cross-section library. The calculated integral parameters were compared to the measured values as well as the results of Monte Carlo Code MCNP. It was found that in most cases, the values of integral parameters show a good agreement with the experiment and MCNP results. Besides, the group constants in WIMS format for the isotopes U-235 and U-238 between two data files have been compared using WIMS library utility code WILLIE and it was found that the group constants are identical with very insignificant difference. Therefore, this analysis reflects the validation of evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 through benchmarking the integral parameters of TRX and BAPL lattices and can also be essential to implement further neutronic analysis of TRIGA Mark-II research reactor at AERE, Dhaka, Bangladesh.
McEwan, C.; Ball, M.; Novog, D., E-mail: mcewac2@mcmaster.ca [McMaster Univ., Hamilton, Ontario (Canada)
2013-07-01
Simulation results are of little use if nothing is known about the uncertainty in the results. In order to assess the uncertainty in a set of output parameters due to uncertainty in a set of input parameters, knowledge of the covariance between input parameters is required. Current practice is to apply the covariance between multigroup cross sections at infinite dilution to all cross sections including those at non-infinite dilutions. In this work, the effect of dilution on multigroup cross section covariance is investigated as well as the effect on the covariance between the few group homogenized cross sections produced by lattice code DRAGON. (author)
Mohammadnia Meysam
2013-01-01
Full Text Available The flux expansion nodal method is a suitable method for considering nodalization effects in node corners. In this paper we used this method to solve the intra-nodal flux analytically. Then, a computer code, named MA.CODE, was developed using the C# programming language. The code is capable of reactor core calculations for hexagonal geometries in two energy groups and three dimensions. The MA.CODE imports two group constants from the WIMS code and calculates the effective multiplication factor, thermal and fast neutron flux in three dimensions, power density, reactivity, and the power peaking factor of each fuel assembly. Some of the code's merits are low calculation time and a user friendly interface. MA.CODE results showed good agreement with IAEA benchmarks, i. e. AER-FCM-101 and AER-FCM-001.
Dissecting Reactor Antineutrino Flux Calculations
Sonzogni, A. A.; McCutchan, E. A.; Hayes, A. C.
2017-09-01
Current predictions for the antineutrino yield and spectra from a nuclear reactor rely on the experimental electron spectra from 235U, 239Pu, 241Pu and a numerical method to convert these aggregate electron spectra into their corresponding antineutrino ones. In the present work we investigate quantitatively some of the basic assumptions and approximations used in the conversion method, studying first the compatibility between two recent approaches for calculating electron and antineutrino spectra. We then explore different possibilities for the disagreement between the measured Daya Bay and the Huber-Mueller antineutrino spectra, including the 238U contribution as well as the effective charge and the allowed shape assumption used in the conversion method. We observe that including a shape correction of about +6 % MeV-1 in conversion calculations can better describe the Daya Bay spectrum. Because of a lack of experimental data, this correction cannot be ruled out, concluding that in order to confirm the existence of the reactor neutrino anomaly, or even quantify it, precisely measured electron spectra for about 50 relevant fission products are needed. With the advent of new rare ion facilities, the measurement of shape factors for these nuclides, for many of which precise beta intensity data from TAGS experiments already exist, would be highly desirable.
Lattice calculation of nonleptonic charm decays
Simone, J.N.
1991-11-01
The decays of charmed mesons into two body nonleptonic final states are investigated. Weak interaction amplitudes of interest in these decays are extracted from lattice four-point correlation functions using a effective weak Hamiltonian including effects to order G{sub f} in the weak interactions yet containing effects to all orders in the strong interactions. The lattice calculation allows a quantitative examination of non-spectator processes in charm decays helping to elucidate the role of effects such as color coherence, final state interactions and the importance of the so called weak annihilation process. For D {yields} K{pi}, we find that the non-spectator weak annihilation diagram is not small, and we interpret this as evidence for large final state interactions. Moreover, there is indications of a resonance in the isospin {1/2} channel to which the weak annihilation process contributes exclusively. Findings from the lattice calculation are compared to results from the continuum vacuum saturation approximation and amplitudes are examined within the framework of the 1/N expansion. Factorization and the vacuum saturation approximation are tested for lattice amplitudes by comparing amplitudes extracted from lattice four-point functions with the same amplitude extracted from products of two-point and three-point lattice correlation functions arising out of factorization and vacuum saturation.
Lattice calculation of nonleptonic charm decays
Simone, James Nicholas [Univ. of California, Los Angeles, CA (United States)
1991-11-01
The decays of charmed mesons into two body nonleptonic final states are investigated. Weak interaction amplitudes of interest in these decays are extracted from lattice four-point correlation functions using a effective weak Hamiltonian including effects to order G_{f } in the weak interactions yet containing effects to all orders in the strong interactions. The lattice calculation allows a quantitative examination of non-spectator processes in charm decays helping to elucidate the role of effects such as color coherence, final state interactions and the importance of the so called weak annihilation process. For D → Kπ, we find that the non-spectator weak annihilation diagram is not small, and we interpret this as evidence for large final state interactions. Moreover, there is indications of a resonance in the isospin 1/2 channel to which the weak annihilation process contributes exclusively. Findings from the lattice calculation are compared to results from the continuum vacuum saturation approximation and amplitudes are examined within the framework of the 1/N expansion. Factorization and the vacuum saturation approximation are tested for lattice amplitudes by comparing amplitudes extracted from lattice four-point functions with the same amplitude extracted from products of two-point and three-point lattice correlation functions arising out of factorization and vacuum saturation.
Calculation of reactor antineutrino spectra in TEXONO
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
A Lattice Calculation of Parton Distributions
Alexandrou, Constantia [Cyprus Univ. Nicosia (Cyprus). Dept. of Physics; The Cyprus Institute, Nicosia (Cyprus); Cichy, Krzysztof [Deutsches Elektronen-Synchrotron (DESY), Zeuthen (Germany). John von Neumann-Inst. fuer Computing NIC; Poznan Univ. (Poland). Faculty of Physics; Drach, Vincent [Univ. of Southern Denmark, Odense (Denmark). CP3-Origins; Univ. of Southern Denmark, Odense (Denmark). Danish IAS; Garcia-Ramos, Elena [Deutsches Elektronen-Synchrotron (DESY), Zeuthen (Germany). John von Neumann-Inst. fuer Computing NIC; Humboldt-Universitaet, Berlin (Germany). Inst. fuer Physik; Hadjiyiannakou, Kyriakos [Cyprus Univ. Nicosia (Cyprus). Dept. of Physics; Jansen, Karl; Steffens, Fernanda; Wiese, Christian [Deutsches Elektronen-Synchrotron (DESY), Zeuthen (Germany). John von Neumann-Inst. fuer Computing NIC
2015-04-15
We report on our exploratory study for the direct evaluation of the parton distribution functions from lattice QCD, based on a recently proposed new approach. We present encouraging results using N{sub f}=2+1+1 twisted mass fermions with a pion mass of about 370 MeV. The focus of this work is a detailed description of the computation, including the lattice calculation, the matching to an infinite momentum and the nucleon mass correction. In addition, we test the effect of gauge link smearing in the operator to estimate the influence of the Wilson line renormalization, which is yet to be done.
A Lattice Calculation of Parton Distributions
Alexandrou, Constantia; Drach, Vincent; Garcia-Ramos, Elena; Hadjiyiannakou, Kyriakos; Jansen, Karl; Steffens, Fernanda; Wiese, Christian
2015-01-01
We report on our exploratory study for the direct evaluation of the parton distribution functions from lattice QCD, based on a recently proposed new approach. We present encouraging results using Nf = 2 + 1 + 1 twisted mass fermions with a pion mass of about 370 MeV. The focus of this work is a detailed description of the computation, including the lattice calculation, the matching to an infinite momentum and the nucleon mass correction. In addition, we test the effect of gauge link smearing in the operator to estimate the influence of the Wilson line renormalization, which is yet to be done.
Equation of State from Lattice QCD Calculations
Gupta, Rajan [Los Alamos National Laboratory
2011-01-01
We provide a status report on the calculation of the Equation of State (EoS) of QCD at finite temperature using lattice QCD. Most of the discussion will focus on comparison of recent results obtained by the HotQCD and Wuppertal-Budapest collaborations. We will show that very significant progress has been made towards obtaining high precision results over the temperature range of T = 150-700 MeV. The various sources of systematic uncertainties will be discussed and the differences between the two calculations highlighted. Our final conclusion is that these lattice results of EoS are precise enough to be used in the phenomenological analysis of heavy ion experiments at RHIC and LHC.
Lattice QCD Calculation of Nucleon Structure
Liu, Keh-Fei [University of Kentucky, Lexington, KY (United States). Dept. of Physics and Astronomy; Draper, Terrence [University of Kentucky, Lexington, KY (United States). Dept. of Physics and Astronomy
2016-08-30
It is emphasized in the 2015 NSAC Long Range Plan that "understanding the structure of hadrons in terms of QCD's quarks and gluons is one of the central goals of modern nuclear physics." Over the last three decades, lattice QCD has developed into a powerful tool for ab initio calculations of strong-interaction physics. Up until now, it is the only theoretical approach to solving QCD with controlled statistical and systematic errors. Since 1985, we have proposed and carried out first-principles calculations of nucleon structure and hadron spectroscopy using lattice QCD which entails both algorithmic development and large-scale computer simulation. We started out by calculating the nucleon form factors -- electromagnetic, axial-vector, πNN, and scalar form factors, the quark spin contribution to the proton spin, the strangeness magnetic moment, the quark orbital angular momentum, the quark momentum fraction, and the quark and glue decomposition of the proton momentum and angular momentum. The first round of calculations were done with Wilson fermions in the `quenched' approximation where the dynamical effects of the quarks in the sea are not taken into account in the Monte Carlo simulation to generate the background gauge configurations. Beginning in 2000, we have started implementing the overlap fermion formulation into the spectroscopy and structure calculations. This is mainly because the overlap fermion honors chiral symmetry as in the continuum. It is going to be more and more important to take the symmetry into account as the simulations move closer to the physical point where the u and d quark masses are as light as a few MeV only. We began with lattices which have quark masses in the sea corresponding to a pion mass at ~ 300 MeV and obtained the strange form factors, charm and strange quark masses, the charmonium spectrum and the D_{s} meson decay constant f_{Ds}, the strangeness and charmness, the meson mass
Lattice QCD Calculation of Nucleon Structure
Liu, Keh-Fei; Draper, Terrence
2016-08-30
It is emphasized in the 2015 NSAC Long Range Plan [1] that \\understanding the structure of hadrons in terms of QCD's quarks and gluons is one of the central goals of modern nuclear physics." Over the last three decades, lattice QCD has developed into a powerful tool for ab initio calculations of strong-interaction physics. Up until now, it is the only theoretical approach to solving QCD with controlled statistical and systematic errors. Since 1985, we have proposed and carried out rst-principles calculations of nucleon structure and hadron spectroscopy using lattice QCD which entails both algorithmic development and large scale computer simulation. We started out by calculating the nucleon form factors { electromagnetic [2], axial-vector [3], NN [4], and scalar [5] form factors, the quark spin contribution [6] to the proton spin, the strangeness magnetic moment [7], the quark orbital angular momentum [8], the quark momentum fraction [9], and the quark and glue decomposition of the proton momentum and angular momentum [10]. These rst round of calculations were done with Wilson fermions in the `quenched' approximation where the dynamical e ects of the quarks in the sea are not taken into account in the Monte Carlo simulation to generate the background gauge con gurations. Beginning in 2000, we have started implementing the overlap fermion formulation into the spectroscopy and structure calculations [11, 12]. This is mainly because the overlap fermion honors chiral symmetry as in the continuum. It is going to be more and more important to take the symmetry into account as the simulations move closer to the physical point where the u and d quark masses are as light as a few MeV only. We began with lattices which have quark masses in the sea corresponding to a pion mass at 300 MeV and obtained the strange form factors [13], charm and strange quark masses, the charmonium spectrum and the Ds meson decay constant fDs [14], the strangeness and charmness [15], the
Detailed Burnup Calculations for Research Reactors
Leszczynski, F. [Centro Atomico Bariloche (CNEA), 8400 S. C. de Bariloche (Argentina)
2011-07-01
tasks for each burn up step: 1) Monte Carlo criticality calculation of the full system tallying spatial power distribution for each spatial region of interest. 2) Preparation of depletion code input and cross- section libraries from Monte Carlo calculation output and other auxiliary code, including normalized power density of each spatial zone with an auxiliary program. The 1 group cross section library needed for depletion calculations can be obtained with a cell code such as DRAGON4 vs. burn up. 3) Depletion calculations of isotope concentrations on the input burn up time-step. 4) Preparation of Monte Carlo calculation input with the new isotope concentrations output of depletion calculation with other auxiliary program. This sequence is implemented in an automatic way. On the first stages of RRMCQ development, a simplified version has been tested with a set of dependent numerical and experimental benchmarks using standard nuclear data libraries at lattice cell level. Then a full core model has been developed and it is to day used on RA6 reactor of Bariloche Atomic Centre. (author)
A Lattice Calculation of Parton Distributions
Alexandrou, Constantia; Hadjiyiannakou, Kyriakos; Jansen, Karl; Steffens, Fernanda; Wiese, Christian
2016-01-01
We present results for the $x$ dependence of the unpolarized, helicity, and transversity isovector quark distributions in the proton using lattice QCD, employing the method of quasi-distributions proposed by Ji in 2013. Compared to a previous calculation by us, the errors are reduced by a factor of about 2.5. Moreover, we present our first results for the polarized sector of the proton, which indicate an asymmetry in the proton sea in favor of the $u$ antiquarks for the case of helicity distributions, and an asymmetry in favor of the $d$ antiquarks for the case of transversity distributions.
A lattice calculation of parton distributions
Alexandrou, Constantia [Cyprus Univ., Nicosia (Cyprus). Dept. of Physics; The Cyprus Institute, Nicosia (Cyprus); Cichy, Krzysztof [Frankfurt Univ. (Germany). Inst. fuer Theoretische Physik; Poznann Univ. (Poland). Faculty of Physics; Hadjiyiannakou, Kyriakos [George Washington Univ., Washington, DC (United States). Dept. of Physics; Jansen, Karl; Steffens, Fernanda; Wiese, Christian [Deutsches Elektronen-Synchrotron (DESY), Zeuthen (Germany). John von Neumann-Inst. fuer Computing NIC
2016-09-15
We present results for the x dependence of the unpolarized, helicity, and transversity isovector quark distributions in the proton using lattice QCD, employing the method of quasi-distributions proposed by Ji in 2013. Compared to a previous calculation by us, the errors are reduced by a factor of about 2.5. Moreover, we present our first results for the polarized sector of the proton, which indicate an asymmetry in the proton sea in favor of the u antiquarks for the case of helicity distributions, and an asymmetry in favor of the d antiquarks for the case of transversity distributions.
Neutronic evaluation of two proposed fuel lattice pitches for ET-RR-1 reactor
Ashoub, N.; Saleh, H.G
2000-04-01
The present fuel element of the ET-RR1 research reactor has a 1.75 cm lattice pitch. The neutronic studies were proved that, this lattice pitch is over moderated and not the suitable one from the fuel economic point of view. Two fuel lattice pitches are proposed, one has 1.4 cm lattice pitch with 10% U{sup 235} enrichment and the other has 1.75 cm lattice pitch with 15% U{sup 235} enrichment. The comparative neutronic study was done between these two proposed fuel lattice pitches against the present one in two cases, one for the complete core configuration of the ET-RR-1 which includes 52 fuel elements and the other for one of the actual core configuration load contains 47 fuel elements. This study is included the calculations of different neutronic parameters as the infinite and effective multiplication factor, the multi-group neutron flux along the reactor core, and the power peaking factor. The above factors were calculated by using the WIMSD4 code for lattice cell calculation, and the DIXY2 code for diffusion calculations. The results are represented in some tables and figures.
Saphyr: a code system from reactor design to reference calculations
Akherraz, B.; Baudron, A.M.; Buiron, L.; Coste-Delclaux, M.; Fedon-Magnaud, C.; Lautard, J.J.; Moreau, F.; Nicolas, A.; Sanchez, R.; Zmijarevic, I. [CEA Saclay, Direction de l' Energie Nucleaire, Departement de Modelisation des Systemes et Structures, Service d' Etudes des Reacteurs et de Modelisation Avancee (DENDMSS/SERMA), 91 - Gif sur Yvette (France); Bergeron, A.; Caruge, D.; Fillion, P.; Gallo, D.; Royer, E. [CEA Saclay, Direction de l' Energie Nucleaire, Departement de Modelisation des Systemes et Structures, Service Fluides numeriques, Modelisations et Etudes (DEN/DMSS/SFNME), 91 - Gif sur Yvette (France); Loubiere, S. [CEA Saclay, Direction de l' Energie Nucleaire, Direction de la Simulation et des Outils Experimentaux, 91- Gif sur Yvette (France)
2003-07-01
In this paper we briefly present the package SAPHYR (in French Advanced System for Reactor Physics) which is devoted to reactor calculations, safety analysis and design. This package is composed of three main codes: APOLLO2 for lattice calculations, CRONOS2 for whole core neutronic calculations and FLICA4 for thermohydraulics. Thanks to a continuous development effort, the SAPHYR system is an outstanding tool covering a large domain of applications, from sophisticated 'research and development' studies that need state-of-the-art methodology to routine industrial calculations for reactor and criticality analysis. SAPHYR is powerful enough to carry out calculations for all types of reactors and is invaluable to understand complex phenomena. SAPHYR components are in use in various nuclear companies such as 'Electricite de France', Framatome-ANP, Cogema, SGN, Transnucleaire and Technicatome. Waiting for the next generation tools (DESCARTES for neutronics and NEPTUNE for thermohydraulics) to be available for such a variety of use, with a better level of flexibility and at least equivalent validation and qualification level, the improvement of SAPHYR is going on, to acquire new functions constantly required by users and to improve current performance levels.
Pressure vessel calculations for VVER-440 reactors.
Hordósy, G; Hegyi, Gy; Keresztúri, A; Maráczy, Cs; Temesvári, E; Vértes, P; Zsolnay, E
2005-01-01
For the determination of the fast neutron load of the reactor pressure vessel a mixed calculational procedure was developed. The procedure was applied to the Unit II of Paks NPP, Hungary. The neutron source on the outer surfaces of the reactor was determined by a core design code, and the neutron transport calculations outside the core were performed by the Monte Carlo code MCNP. The reaction rate in the activation detectors at surveillance positions and at the cavity were calculated and compared with measurements. In most cases, fairly good agreement was found.
Status of lattice field theory calculations
Sharpe, S.R.
1990-01-01
This report briefly discusses the following topics: overview of all present calculation; reliability criteria for quenched calculation; quenched versus full QCD, and difficulties facing full QCD; results for the quenched pion wavefunction''; results for the quenched hadron spectrum; results for quenched B{sub K}; A new method for calculating the surface tension; the non-pertubative upper bound on the Higgs mass; and toward the TERAFLOP machine.
A novel lattice energy calculation technique for simple inorganic crystals
Kaya, Cemal [Department of Chemistry, Faculty of Science, Cumhuriyet University, 58140 Sivas (Turkey); Kaya, Savaş, E-mail: savaskaya@cumhuriyet.edu.tr [Department of Chemistry, Faculty of Science, Cumhuriyet University, 58140 Sivas (Turkey); Banerjee, Priyabrata [Surface Engineering and Tribology Group, CSIR-Central Mechanical Engineering Research Institute, Mahatma Gandhi Avenue, Durgapur 713209 (India)
2017-01-01
In this pure theoretical study, a hitherto unexplored equation based on Shannon radii of the ions forming that crystal and chemical hardness of any crystal to calculate the lattice energies of simple inorganic ionic crystals has been presented. To prove the credibility of this equation, the results of the equation have been compared with experimental outcome obtained from Born-Fajans-Haber- cycle which is fundamentally enthalpy-based thermochemical cycle and prevalent theoretical approaches proposed for the calculation of lattice energies of ionic compounds. The results obtained and the comparisons made have demonstrated that the new equation is more useful compared to other theoretical approaches and allows to exceptionally accurate calculation of lattice energies of inorganic ionic crystals without doing any complex calculations.
A novel lattice energy calculation technique for simple inorganic crystals
Kaya, Cemal; Kaya, Savaş; Banerjee, Priyabrata
2017-01-01
In this pure theoretical study, a hitherto unexplored equation based on Shannon radii of the ions forming that crystal and chemical hardness of any crystal to calculate the lattice energies of simple inorganic ionic crystals has been presented. To prove the credibility of this equation, the results of the equation have been compared with experimental outcome obtained from Born-Fajans-Haber- cycle which is fundamentally enthalpy-based thermochemical cycle and prevalent theoretical approaches proposed for the calculation of lattice energies of ionic compounds. The results obtained and the comparisons made have demonstrated that the new equation is more useful compared to other theoretical approaches and allows to exceptionally accurate calculation of lattice energies of inorganic ionic crystals without doing any complex calculations.
QCD Factorization and PDFs from Lattice QCD Calculation
Ma, Yan-Qing
2014-01-01
In this talk, we review a QCD factorization based approach to extract parton distribution and correlation functions from lattice QCD calculation of single hadron matrix elements of quark-gluon operators. We argue that although the lattice QCD calculations are done in the Euclidean space, the nonperturbative collinear behavior of the matrix elements are the same as that in the Minkowski space, and could be systematically factorized into parton distribution functions with infrared safe matching coefficients. The matching coefficients can be calculated perturbatively by applying the factorization formalism on to asymptotic partonic states.
High Precision Calculations of the Lennard-Jones Lattice Constants for Five Lattices
Stein, Matthew
2017-01-01
The total potential energy of a crystal as described by the Lennard-Jones (L-J) potential depends in part upon the calculation of lattice constants. Knowing these constants to high precision is useful for prediction of the lattice type and simulation of crystals such as rare-gas solids or germanium detectors, but reaching higher precision is computationally costly and challenging. Presented here is the extension of the precision of the lattice constants, Lp, up to 32 decimal digits, and in some cases corrections from previous publication. The Lp terms are given for 4 cubic, face-centered cubic, body-centered cubic, hexagonal-close-pack, and diamond lattices. This precision was obtained through the use of careful parallelization technique, exploitation of the symmetries of each lattice, and the ``onionization'' of the simulated crystal. The results of this computation, along with the tools and algorithm strategies to make this computation possible, are explained in detail graphically.
Vectorized Monte Carlo methods for reactor lattice analysis
Brown, F. B.
1984-01-01
Some of the new computational methods and equivalent mathematical representations of physics models used in the MCV code, a vectorized continuous-enery Monte Carlo code for use on the CYBER-205 computer are discussed. While the principal application of MCV is the neutronics analysis of repeating reactor lattices, the new methods used in MCV should be generally useful for vectorizing Monte Carlo for other applications. For background, a brief overview of the vector processing features of the CYBER-205 is included, followed by a discussion of the fundamentals of Monte Carlo vectorization. The physics models used in the MCV vectorized Monte Carlo code are then summarized. The new methods used in scattering analysis are presented along with details of several key, highly specialized computational routines. Finally, speedups relative to CDC-7600 scalar Monte Carlo are discussed.
Modified lattice-statics approach to dislocation calculations. I - Formalism
Esterling, D. M.
1978-01-01
A modified lattice-statics method to calculate the atomic displacements associated with a screw dislocation is outlined. The model incorporates an anharmonic region wherein the forces are derived from a pair potential. Appropriate energy and force expressions are derived. The modifications necessary for the implementation of the conjugate-gradient function minimization method are also derived.
Gravitation Field Calculations on a Dynamic Lattice by Distributed Computing
Mähönen, Petri; Punkka, Veikko
A new method of calculating numerically time evolution of a gravitational field in General Relatity is introduced. Vierbein (tetrad) formalism, dynamic lattice and massively parallelized computation are suggested as they are expected to speed up the calculations considerably and facilitate the solution of problems previously considered too hard to be solved, such as the time evolution of a system consisting of two or more black holes or the structure of worm holes.
Gravitational field calculations on a dynamic lattice by distributed computing.
Mähönen, P.; Punkka, V.
A new method of calculating numerically time evolution of a gravitational field in general relativity is introduced. Vierbein (tetrad) formalism, dynamic lattice and massively parallelized computation are suggested as they are expected to speed up the calculations considerably and facilitate the solution of problems previously considered too hard to be solved, such as the time evolution of a system consisting of two or more black holes or the structure of worm holes.
Improvement of Neutronics Calculation Methods for Fast Reactors
Takeda, Toshikazu
2011-01-01
To accurately estimate neutronics properties of fast reactors, particularly Japan Sodium-cooled Fast Reactor of1,500 MW electric, calculational methods are being improved in Japan.This paper describes the planning and the ongoing development of the neutronics calculation methods in the fieldof 1) assembly calculations including the calculations of effective cross sections, 2) core calculations and 3) uncertaintyevaluation and uncertainty reduction.
Uncertainty quantification in lattice QCD calculations for nuclear physics
Beane, Silas R. [Univ. of Washington, Seattle, WA (United States); Detmold, William [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Orginos, Kostas [College of William and Mary, Williamsburg, VA (United States); Thomas Jefferson National Accelerator Facility (TJNAF), Newport News, VA (United States); Savage, Martin J. [Institute for Nuclear Theory, Seattle, WA (United States)
2015-02-05
The numerical technique of Lattice QCD holds the promise of connecting the nuclear forces, nuclei, the spectrum and structure of hadrons, and the properties of matter under extreme conditions with the underlying theory of the strong interactions, quantum chromodynamics. A distinguishing, and thus far unique, feature of this formulation is that all of the associated uncertainties, both statistical and systematic can, in principle, be systematically reduced to any desired precision with sufficient computational and human resources. As a result, we review the sources of uncertainty inherent in Lattice QCD calculations for nuclear physics, and discuss how each is quantified in current efforts.
Binding Energy Calculations for Novel Ternary Ionic Lattices
Rodríguez-Mijangos, Ricardo; Vazquez-Polo, Gustavo
2002-03-01
Theoretical calculations for the binding energy between metalic ions and negative ions on a novel ternary ionic lattice is carried out for several solid solutions prepared with different concentrations and characterized recently (1). The ternary lattices that reach a good miscibility are: KCl(x)KBr(y)RbCl(z) in three different concentrations: (x=y=z=0.33), (x=0.5, y=0.25, z=0.25) and (x=0.33, y=0.07, z=0.60). The binding energy for these novel structures is calculated from the lattice constants obtained by X ray diffractometry analysis performed on the samples and the Vegard law (2). For the repulsive force exponent m, an average of the m values was considered. The energy values obtained by the Born´expression are compared with corresponding energy values from the lattice with more complex expressions, such as the Born Mayer, Born-Van der Walls. There is a good aggreement between all these calculations. (1)R. R. Mijangos, A. Cordero-Borboa, E. Alvarez, M. Cervantes, Physics Letters A 282 (2001) 195-200. (2) G. Vazquez-Polo, R. R. Mijangos et al. Revista Mexicana de Fisica, 47, Diciembre 2001. In Press.
Reactor physics analysis for the design of nuclear fuel lattices with burnable poisons
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
On the Calculation of Reactor Time Constants Using the Monte Carlo Method
Leppaenen, Jaakko [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT (Finland)
2008-07-01
Full-core reactor dynamics calculation involves the coupled modelling of thermal hydraulics and the time-dependent behaviour of core neutronics. The reactor time constants include prompt neutron lifetimes, neutron reproduction times, effective delayed neutron fractions and the corresponding decay constants, typically divided into six or eight precursor groups. The calculation of these parameters is traditionally carried out using deterministic lattice transport codes, which also produce the homogenised few-group constants needed for resolving the spatial dependence of neutron flux. In recent years, there has been a growing interest in the production of simulator input parameters using the stochastic Monte Carlo method, which has several advantages over deterministic transport calculation. This paper reviews the methodology used for the calculation of reactor time constants. The calculation techniques are put to practice using two codes, the PSG continuous-energy Monte Carlo reactor physics code and MORA, a new full-core Monte Carlo neutron transport code entirely based on homogenisation. Both codes are being developed at the VTT Technical Research Centre of Finland. The results are compared to other codes and experimental reference data in the CROCUS reactor kinetics benchmark calculation. (author)
Validation of WIMS-AECL reactivity device calculations for CANDU reactor
Min, Byung Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Donnelly, J. V. [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)
1997-06-01
An important component of the overall program to validate WIMS-AECL for use with RFSP in the analysis of CANDU-6 reactors for design and safety analysis calculations is the validation of calculations of incremental cross sections used to represent reactivity devices. A method has been developed for the calculation of the three-dimensional neutron flux distribution in and around CANDU reactor fuel channels and reactivity control devices. The methods is based on one- and two dimensional transport calculations with the WIMS-AECL lattice cell code, SPH homogenization, and three-dimensional flux calculations with finite-difference diffusion theory using the MULTICELL code. Simulations of Wolsung 1 Phase-B commissioning measurements and Point Lepreau restart tests have been performed, as a part of the program to validate WIMS-AECL lattice cell calculations for application to CANDU reactor simulations in RFSP. The incremental cross section properties of the Wolsung 1 and Point Lepreau adjusters, Mechanical Control Absorbers(MCA) and liquid Zone Control Unit (ZCU) is based on the WIMS-AECL/MULTICELL modelling methods and the results are compared with those of WIMS-AECL/DRAGON-2 modelling methods. (author). 13 tabs., 4 figs., 11 refs.
Standard Guide for Benchmark Testing of Light Water Reactor Calculations
American Society for Testing and Materials. Philadelphia
2010-01-01
1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactor fluences with ...
Molecular modeling study of chiral drug crystals: lattice energy calculations.
Li, Z J; Ojala, W H; Grant, D J
2001-10-01
The lattice energies of a number of chiral drugs with known crystal structures were calculated using Dreiding II force field. The lattice energies, including van der Waals, Coulombic, and hydrogen-bonding energies, of homochiral and racemic crystals of some ephedrine derivatives and of several other chiral drugs, are compared. The calculated energies are correlated with experimental data to probe the underlying intermolecular forces responsible for the formation of racemic species, racemic conglomerates, or racemic compounds, termed chiral discrimination. Comparison of the calculated energies among ephedrine derivatives reveals that a greater Coulombic energy corresponds to a higher melting temperature, while a greater van der Waals energy corresponds to a larger enthalpy of fusion. For seven pairs of homochiral and racemic compounds, correlation of the differences between the two forms in the calculated energies and experimental enthalpy of fusion suggests that the van der Waals interactions play a key role in the chiral discrimination in the crystalline state. For salts of the chiral drugs, the counter ions diminish chiral discrimination by increasing the Coulombic interactions. This result may explain why salt forms favor the formation of racemic conglomerates, thereby facilitating the resolution of racemates.
The use of Schoonschip and form in perturbative lattice calculations
Capitani, S; Capitani, Stefano; Rossi, Giancarlo
1995-01-01
Using the formal languages Schoonschip and Form, we have developed general codes that are able to carry out all the algebraic manipulations needed to perform analytic lattice calculations, starting from the elementary building blocks (propagators and vertices) of each Feynman diagram. The main difficulty resides in the fact that, although there are many built in instructions to deal with Dirac gamma-matrices, Schoonschip and Form have been conceived having in mind a continuum theory, which is invariant with respect to the Lorentz group. On the lattice, on the contrary, a field theory is only invariant with respect to the hypercubic group, contained in the (euclidean) Lorentz group and not every pair of equal indices should be summed over. Being impossible to directly use the `gammatrics' of Schoonschip and Form as they are, special routines have been developed to correctly treat gamma matrices on the lattice, while using as much as possible of the built in Schoonschip and Form commands. We have used our codes...
A floating point engine for lattice gauge calculations
Husby, D.; Atac, R.; Cook, A.; Deppe, J.; Fischler, M.; Gaines, I.; Wash, T.; Pham, T.; Zmuda, T.
1989-02-01
The latest in low cost computing solutions from the Fermilab Advanced Computer Program is targeted at Lattice Gauge theory calculations and delivers supercomputer performance at a fraction of the cost. A typical system with 256 processors, 2.5 Gigabytes of memory, and 64 Gigabytes of on-line tape storage, delivers a peak performance of 5 billion floating point operations per second. The programming environment, Canopy, provides a comprehensive, hardware independent, distributed processing platform from within the more familiar environments of FORTRAN, C, and UNIX. This paper describes the individual processing elements of the system and gives a brief description of the Canopy software.
A floating point engine for lattice gauge calculations
Husby, D.; Atac, R.; Cook, A.; Deppe, J.; Fischler, M.; Gaines, I.; Nash, T.; Pham, T.; Zmuda, T.; Eichten, E.
1988-11-01
The latest in low cost computing solutions from the Fermilab Advanced Computer Program is targeted at Lattice Gauge theory calculations and delivers supercomputer performance at a fraction of the cost. A typical system with 256 processors, 2.5 Gigabytes of memory, and 64 Gigabytes of on-line tape storage, delivers a peak performance of 5 billion floating point operations per second. The programming environment, Canopy, provides a comprehensive, hardware independent, distributed processing platform from within the more familiar environments of FORTRAN, C, and UNIX. This paper describes the individual processing elements of the system and gives a brief description of the Canopy software. 8 refs., 3 figs.
Variational Calculation in SU(3) Lattice Gauge Theory
YANG Chun; ZHANG Qi-Ren; GAO Chun-Yuan
2001-01-01
Using the Hamiltonian lattice gauge theory, we perform some variational calculations to obtain the ground-state energy of SU(3) gauge field and scalar (0++) glueball mass. The agreement of our data with the strong and weak expansion results in the corresponding limits indicates that this method can provide us with reliable information in the most interesting medium region. The trial wavefunction used in our variational method is also proven to be a good first approximation of the ground-state of the SU(3) gauge field. Upgrading this function according to correlations of adjacent plaquettes may mean better results.
Pressure Vessel Calculations for VVER-440 Reactors
Hordósy, G.; Hegyi, Gy.; Keresztúri, A.; Maráczy, Cs.; Temesvári, E.; Vértes, P.; Zsolnay, É.
2003-06-01
Monte Carlo calculations were performed for a selected cycle of the Paks NPP Unit II to test a computational model. In the model the source term was calculated by the core design code KARATE and the neutron transport calculations were performed by the MCNP. Different forms of the source specification were examined. The calculated results were compared with measurements and in most cases fairly good agreement was found.
Uncertainty Analysis of Light Water Reactor Fuel Lattices
C. Arenas
2013-01-01
Full Text Available The study explored the calculation of uncertainty based on available cross-section covariance data and computational tool on fuel lattice levels, which included pin cell and the fuel assembly models. Uncertainty variations due to temperatures changes and different fuel compositions are the main focus of this analysis. Selected assemblies and unit pin cells were analyzed according to the OECD LWR UAM benchmark specifications. Criticality and uncertainty analysis were performed using TSUNAMI-2D sequence in SCALE 6.1. It was found that uncertainties increase with increasing temperature, while kinf decreases. This increase in the uncertainty is due to the increase in sensitivity of the largest contributing reaction of uncertainty, namely, the neutron capture reaction 238U(n, γ due to the Doppler broadening. In addition, three types (UOX, MOX, and UOX-Gd2O3 of fuel material compositions were analyzed. A remarkable increase in uncertainty in kinf was observed for the case of MOX fuel. The increase in uncertainty of kinf in MOX fuel was nearly twice the corresponding value in UOX fuel. The neutron-nuclide reaction of 238U, mainly inelastic scattering (n, n′, contributed the most to the uncertainties in the MOX fuel, shifting the neutron spectrum to higher energy compared to the UOX fuel.
Parameter analysis calculation on characteristics of portable FAST reactor
Otsubo, Akira; Kowata, Yasuki [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center
1998-06-01
In this report, we performed a parameter survey analysis by using the analysis program code STEDFAST (Space, TErrestrial and Deep sea FAST reactor-gas turbine system). Concerning the deep sea fast reactor-gas turbine system, calculations with many variable parameters were performed on the base case of a NaK cooled reactor of 40 kWe. We aimed at total equipment weight and surface area necessary to remove heat from the system as important values of the characteristics of the system. Electric generation power and the material of a pressure hull were specially influential for the weight. The electric generation power, reactor outlet/inlet temperatures, a natural convection heat transfer coefficient of sea water were specially influential for the area. Concerning the space reactor-gas turbine system, the calculations with the variable parameters of compressor inlet temperature, reactor outlet/inlet temperatures and turbine inlet pressure were performed on the base case of a Na cooled reactor of 40 kWe. The first and the second variable parameters were influential for the total equipment weight of the important characteristic of the system. Concerning the terrestrial fast reactor-gas turbine system, the calculations with the variable parameters of heat transferred pipe number in a heat exchanger to produce hot water of 100degC for cogeneration, compressor stage number and the kind of primary coolant material were performed on the base case of a Pb cooled reactor of 100 MWt. In the comparison of calculational results for Pb and Na of primary coolant material, the primary coolant weight flow rate was naturally large for the former case compared with for the latter case because density is very different between them. (J.P.N.)
Lattice QCD Calculations in Nuclear Physics towards the Exascale
Joo, Balint
2017-01-01
The combination of algorithmic advances and new highly parallel computing architectures are enabling lattice QCD calculations to tackle ever more complex problems in nuclear physics. In this talk I will review some computational challenges that are encountered in large scale cold nuclear physics campaigns such as those in hadron spectroscopy calculations. I will discuss progress in addressing these with algorithmic improvements such as multi-grid solvers and software for recent hardware architectures such as GPUs and Intel Xeon Phi, Knights Landing. Finally, I will highlight some current topics for research and development as we head towards the Exascale era This material is funded by the U.S. Department of Energy, Office Of Science, Offices of Nuclear Physics, High Energy Physics and Advanced Scientific Computing Research, as well as the Office of Nuclear Physics under contract DE-AC05-06OR23177.
Lattice calculation of composite dark matter form factors
Appelquist, T; Buchoff, M I; Cheng, M; Cohen, S D; Fleming, G T; Kiskis, J; Lin, M F; Neil, E T; Osborn, J C; Rebbi, C; Schaich, D; Schroeder, C; Syritsyn, S N; Voronov, G; Vranas, P; Wasem, J
2013-01-01
Composite dark matter candidates, which can arise from new strongly-coupled sectors, are well-motivated and phenomenologically interesting, particularly in the context of asymmetric generation of the relic density. In this work, we employ lattice calculations to study the electromagnetic form factors of electroweak-neutral dark-matter baryons for a three-color, QCD-like theory with Nf = 2 and 6 degenerate fermions in the fundamental representation. We calculate the (connected) charge radius and anomalous magnetic moment, both of which can play a significant role for direct detection of composite dark matter. We find minimal Nf dependence in these quantities. We generate mass-dependent cross-sections for dark matter-nucleon interactions and use them in conjunction with experimental results from XENON100, excluding dark matter candidates of this type with masses below 10 TeV.
Improved resonance reaction rate calculation for lattice physics subsystem
Finch, D.R.
1974-02-08
The resonance capture calculations of the HAMMER System and HAMBUR System are derived from a consistent statement of the integral slowing down equation and definitions of the resonance integral. The assumptions made in these treatments are explicitly stated, and and an attempt is made to estimate the possible error in the resonance integral arising from these assumptions. This analysis is made to pin-point those parts of the calculation that can be improved and updated. Based on the analysis of existing calculations a method of calculation is derived which avoids most of the problems encountered in HAMMER and HAMBUR. The chief improvements that result are as follows: Careful attention is paid to calculation of the resonance flux as most errors in existing calculations result from consistently overpredicting fluxes in all regions of a lattice cell. The calculation can be modified to produce as crude or detailed a resonance calculation, at the expense of computer time, as required by the user. Resonances that overlap group boundaries contribute the correct contribution to each group's reaction rates. Overlap between resonances of different isotopes is correctly accounted for. Up-to-date resonance formalisms are used including the Adler-Adler multi-level formulations. Provision is made to easily add new formalisms when required. Streaming effects from neutron leaking into a cell may optionally be included in the calculation of resonance reaction rates. A complete description of the physics contained in this new computational module is provided along with additional information on the numerical techniques employed in the module.
A Framework for Lattice QCD Calculations on GPUs
Winter, Frank; Clark, M A; Edwards, Robert G; Joo, Balint
2014-08-01
Computing platforms equipped with accelerators like GPUs have proven to provide great computational power. However, exploiting such platforms for existing scientific applications is not a trivial task. Current GPU programming frameworks such as CUDA C/C++ require low-level programming from the developer in order to achieve high performance code. As a result porting of applications to GPUs is typically limited to time-dominant algorithms and routines, leaving the remainder not accelerated which can open a serious Amdahl's law issue. The lattice QCD application Chroma allows to explore a different porting strategy. The layered structure of the software architecture logically separates the data-parallel from the application layer. The QCD Data-Parallel software layer provides data types and expressions with stencil-like operations suitable for lattice field theory and Chroma implements algorithms in terms of this high-level interface. Thus by porting the low-level layer one can effectively move the whole application in one swing to a different platform. The QDP-JIT/PTX library, the reimplementation of the low-level layer, provides a framework for lattice QCD calculations for the CUDA architecture. The complete software interface is supported and thus applications can be run unaltered on GPU-based parallel computers. This reimplementation was possible due to the availability of a JIT compiler (part of the NVIDIA Linux kernel driver) which translates an assembly-like language (PTX) to GPU code. The expression template technique is used to build PTX code generators and a software cache manages the GPU memory. This reimplementation allows us to deploy an efficient implementation of the full gauge-generation program with dynamical fermions on large-scale GPU-based machines such as Titan and Blue Waters which accelerates the algorithm by more than an order of magnitude.
Xenon poisoning calculation code for miniature neutron source reactor (MNSR)
无
2001-01-01
In line with the actual requirements and based upon the specific char acteristics of MNSR, a revised point-reactor model was adopted to model MNSR's xenon poisoning. The corresponding calculation code, MNSRXPCC (Xenon Poison ing Calculation Code for MNSR), was developed and tested by the Shanghai MNSR data.
A lattice QCD calculation of the transverse decay constant of the b1(1235) meson
Jansen, K; Michael, C; Urbach, C
2009-01-01
We review various B meson decays that require knowledge of the transverse decay constant of the b1(1235) meson. We report on an exploratory lattice QCD calculation of the transverse decay constant of the b1 meson. The lattice QCD calculations used unquenched gauge configurations, at two lattice spacings, generated with two flavours of sea quarks. The twisted mass formalism is used.
Lattice Calculation of the Strangeness Magnetic Moment of the Nucleon
Dong, S J; Williams, A G
1998-01-01
We report on a lattice QCD calculation of the strangeness magnetic moment of the nucleon. Our result is $G_M^s(0) = - 0.36 \\pm 0.20 $. The sea contributions from the u and d quarks are about 80% larger. However, they cancel to a large extent due to their electric charges, resulting in a smaller net sea contribution of $ - 0.097 \\pm 0.037 \\mu_N$ to the nucleon magnetic moment. As far as the neutron to proton magnetic moment ratio is concerned, this sea contribution tends to cancel out the cloud-quark effect from the Z-graphs and result in a ratio of $ -0.68 \\pm 0.04$ which is close to the SU(6) relation and the experiment. The strangeness Sachs electric mean-square radius $_E$ is found to be small and negative and the total sea contributes substantially to the neutron electric form factor.
Final safeguards analysis, high temperature lattice test reactor. Revision 1
Hanthorn, H.E.; Brown, W.W.; Clark, R.G.; Heineman, R.E.; Humes, R.M.
1966-01-01
The PMACS `reactor-normal` signal signifies that important process variables do not exceed their set points, that various interlocks are properly set, that functional tests of the computer operation are satisfactory, and that the reactor flux level and period derived from two additional, independent, and dissimilar channels are within set limits. This safety circuit combines the features of redundancy, dissimilar components, and frequent testing which are required for best reliability. The experimental equipment auxiliary to the reactor includes two oscillator mechanisms, one to move the test cell or the adjoining cell into and out of position, the other to move small specimens in the test cell or adjoining cells. They have cooling chambers for the removal of specimens from the test cell without the necessity of cooling the reactor. A neutron chopper and time-of-flight spectrometer are provided; the neutron detectors, at the end of a 25-meter flight tube, are in an adjoining small building. Test cores may be assembled on a core dolly have a load capacity of 14,000 lb. Two wire traverse mechanisms are provided for measurements of flux distribution.
Assessment of uncertainty in full core reactor physics calculations using statistical methods
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)
Lattice thermal conductivity of borophene from first principle calculation
Xiao, Huaping; Cao, Wei; Ouyang, Tao; Guo, Sumei; He, Chaoyu; Zhong, Jianxin
2017-04-01
The phonon transport property is a foundation of understanding a material and predicting the potential application in mirco/nano devices. In this paper, the thermal transport property of borophene is investigated by combining first-principle calculations and phonon Boltzmann transport equation. At room temperature, the lattice thermal conductivity of borophene is found to be about 14.34 W/mK (error is about 3%), which is much smaller than that of graphene (about 3500 W/mK). The contributions from different phonon modes are qualified, and some phonon modes with high frequency abnormally play critical role on the thermal transport of borophene. This is quite different from the traditional understanding that thermal transport is usually largely contributed by the low frequency acoustic phonon modes for most of suspended 2D materials. Detailed analysis further reveals that the scattering between the out-of-plane flexural acoustic mode (FA) and other modes likes FA + FA/TA/LA/OP ↔ TA/LA/OP is the predominant phonon process channel. Finally the vibrational characteristic of some typical phonon modes and mean free path distribution of different phonon modes are also presented in this work. Our results shed light on the fundamental phonon transport properties of borophene, and foreshow the potential application for thermal management community.
Exposure calculation code module for reactor core analysis: BURNER
Vondy, D.R.; Cunningham, G.W.
1979-02-01
The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.
Exposure calculation code module for reactor core analysis: BURNER
Vondy, D.R.; Cunningham, G.W.
1979-02-01
The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.
New burnup calculation of TRIGA IPR-R1 reactor
Meireles, Sincler P. de; Campolina, Daniel de A.M.; Santos, Andre A. Campagnole dos; Menezes, Maria A.B.C.; Mesquita, Amir Z., E-mail: sinclercdtn@hotmail.com.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)
2015-07-01
The IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, operates since 1960.The reactor is operating for more than fifty years and has a long history of operation. Determining the current composition of the fuel is very important to calculate various parameters. The reactor burnup calculation has been performed before, however, new techniques, methods, software and increase of the processing capacity of the new computers motivates new investigations to be performed. This work presents the evolution of effective multiplication constant and the results of burnup. This new model has a more detailed geometry with the introduction of the new devices, like the control rods and the samarium discs. This increase of materials in the simulation in burnup calculation was very important for results. For these series of simulations a more recently cross section library, ENDF/B-VII, was used. To perform the calculations two Monte Carlo particle transport code were used: Serpent and MCNPX. The results obtained from two codes are presented and compared with previous studies in the literature. (author)
Calculation of Conduction Spectra in Quantum Dot Composed of Penrose Lattice
Tomita, Ryutaro; Fujimoto, Shigeo; Natsume, Yuhei [Graduate School of Science and Technology, Chiba University, 1-33 Yayoi-cho, Inage-ku, Chiba 263-8522 (Japan)
2007-04-15
Conductance through finite two-dimensional Penrose lattices (PLs) are calculated as a function of gate voltage. To investigate effects of lattice defects and lattice vibrations, three types of PLs are taken into account; (A) PL without lattice defects, (B) PL with lattice defects, and (C) PL with lattice vibrations, which is applied the electric field across the conductor. When energy levels of PL coincides with the chemical potential of the electrodes with increasing gate voltage, electron transfer through PL takes place. However, electron transfer is forbidden at certain states; These states have confined states, which have multiple-degeneracy in their electronic states. In addition, the states are localized due to interference of their wave-functions though the wave-function of each state is extended. We would like to point out that rigidity of confined states with respect to lattice defects and lattice vibrations.
Fuel depletion calculation in MTR-LEU NUR reactor
Zeggar Foudil
2008-01-01
Full Text Available In this article, we present the results of a few energy groups calculations for the NUR reactor fuel depletion analysis up to 45 000 MWd/tU taken as the maximum fuel burn up. The WIMSD-4 cell code has been employed as a calculation tool. In this study, we are interested in actinides such as the uranium and plutonium isotopes, as well as fission products Xe-135, Sm-149, Sm-151, Eu-155, and Gd-157. Calculation results regarding the five energy groups are in a good agreement with those obtained with only two energy groups which can, therefore, be used in all subsequent calculations. Calculation results presented in this article can be used as a microscopic data base for estimating the amount of radioactive sources randomly dispersed in the environment. They can also be used to monitor the fuel assemblies inventory at the core level.
Fuel burnup calculation of a research reactor plate element
Santos, Nadia Rodrigues dos; Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: nadiasam@gmail.com, E-mail: zrlima@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)
2013-07-01
This work consists in simulating the burnup of two different plate type fuel elements, where one is the benchmark MTR of the IAEA, which is made of an alloy of uranium and aluminum, while the other belonging to a typical multipurpose reactor is composed of an alloy of uranium and silicon. The simulation is performed using the WIMSD-5B computer code, which makes use of deterministic methods for solving neutron transport. In developing this task, fuel element equivalent cells were calculated representing each of the reactors to obtain the initial concentrations of each isotope constituent element of the fuel cell and the thicknesses corresponding to each region of the cell, since this information is part of the input data. The compared values of the k∞ showed a similar behavior for the case of the MTR calculated with the WIMSD-5B and EPRI-CELL codes. Relating the graphs of the concentrations in the burnup of both reactors, there are aspects very similar to each isotope selected. The application WIMSD-5B code to calculate isotopic concentrations and burnup of the fuel element, proved to be satisfactory for the fulfillment of the objective of this work. (author)
One-loop calculations in Supersymmetric Lattice QCD
Costa M.
2017-01-01
We present here results from dimensional regularization, relegating to a forthcoming publication [1] our results along with a more complete list of references. Part of the lattice study regards also the renormalization of quark bilinear operators which, unlike the nonsupersymmetric case, exhibit a rich pattern of operator mixing at the quantum level.
Large-scale calculation of ferromagnetic spin systems on the pyrochlore lattice
Soldatov, Konstantin; Nefedev, Konstantin; Komura, Yukihiro; Okabe, Yutaka
2017-02-01
We perform the high-performance computation of the ferromagnetic Ising model on the pyrochlore lattice. We determine the critical temperature accurately based on the finite-size scaling of the Binder ratio. Comparing with the data on the simple cubic lattice, we argue the universal finite-size scaling. We also calculate the classical XY model and the classical Heisenberg model on the pyrochlore lattice.
Selective Oxidation of Propane by Lattice Oxygen of Vanadium-Phosphorous Oxide in a Pulse Reactor
Rusong Zhao; Jian Wang; Qun Dong; Jianhong Liu
2005-01-01
Selective oxidation of propane by lattice oxygen of vanadium-phosphorus oxide (VPO) catalysts was investigated with a pulse reactor in which the oxidation of propane and the re-oxidation of catalyst were implemented alternately in the presence of water vapor. The principal products are acrylic acid (AA),acetic acid (HAc), and carbon oxides. In addition, small amounts of C1 and C2 hydrocarbons were also found, molar ratio of AA to HAc is 1.4-2.2. The active oxygen species are those adsorbed on catalyst surface firmly and/or bound to catalyst lattice, i.e. lattice oxygen; the selective oxidation of propane on VPO catalysts can be carried out in a circulating fluidized bed (CFB) riser reactor. For propane oxidation over VPO catalysts, the effects of reaction temperature in a pulse reactor were found almost the same as in a steady-state flow reactor. That is, as the reaction temperature increases, propane conversion and the amount of C1+C2 hydrocarbons in the product increase steadily, while selectivity to acrylic acid and to acetic acid increase prior to 350 ℃ then begin to drop at temperatures higher than 350 ℃, and yields of acrylic acid and of acetic acid attained maximum at about 400 ℃. The maximum yields of acrylic acid and of acetic acid for a single-pass are 7.50% and 4.59% respectively, with 38.2% propane conversion. When the amount of propane pulsed decreases or the amount of catalyst loaded increases, the conversion increases but the selectivity decreases. Increasing the flow rate of carrier gases causes the conversion pass through a minimum but selectivity and yields pass through a maximum. In a fixed bed reactor, it is hard to obtain high selectivity at a high reaction conversion due to the further degradation of acrylic acid and acetic acid even though propane was oxidized by the lattice oxygen. The catalytic performance can be improved in the presence of excess propane. Propylene can be oxidized by lattice oxygen of VPO catalyst at 250
A Framework for Lattice QCD Calculations on GPUs
Winter, F T; Edwards, R G; Joó, B
2014-01-01
Computing platforms equipped with accelerators like GPUs have proven to provide great computational power. However, exploiting such platforms for existing scientific applications is not a trivial task. Current GPU programming frameworks such as CUDA C/C++ require low-level programming from the developer in order to achieve high performance code. As a result porting of applications to GPUs is typically limited to time-dominant algorithms and routines, leaving the remainder not accelerated which can open a serious Amdahl's law issue. The lattice QCD application Chroma allows to explore a different porting strategy. The layered structure of the software architecture logically separates the data-parallel from the application layer. The QCD Data-Parallel software layer provides data types and expressions with stencil-like operations suitable for lattice field theory and Chroma implements algorithms in terms of this high-level interface. Thus by porting the low-level layer one can effectively move the whole applica...
FORMATION (DECOMPOSITION) ENTHALPY CALCULATIONS FOR CRYSTAL LATTICES OF ALKALINE-EARTH FLUORIDES
Gruba, O.; Germanyuk, N.; Ryabukhin, A.
2015-01-01
A series of calculations of structural and thermochemical properties has been carried out for the alkaline-earth fluorides. The calculations have been carried out using the modified model of effective ionic radii and the model of enthalpy calculation for the crystal lattice. The results of the calculations are in accordance with the known experimental data within confidence intervals.
Framework for improved lattice calculations of epsilion'/epsilon
Laiho, Jack
In this thesis we show that it is possible to construct epsilon '/epsilon to NLO using both full and partially quenched chiral perturbation theory (PQChPT) from amplitudes that are computable using numerical lattice gauge theory. We find that the electro-weak penguin (Delta I = 3/2 and 1/2) contributions to epsilon'/epsilon in PQChPT can be determined to NLO using only degenerate (mK = mpi) K → pi computations without momentum insertion. All one-loop formulas needed to extract the necessary NLO constants from the lattice are presented in this work. Issues pertaining to power divergent contributions, originating from mixing with lower dimensional operators in a lattice regularization, are addressed. In embedding the QCD penguin left-right operator onto PQChPT an ambiguity arises when the number of light sea quarks is not the physical value of three, as first emphasized by Golterman and Pallante. In the quenched theory they have pointed out that there are additional effective operators that appear in the quenched chiral perturbation theory needed to make contact with K → pipi amplitudes at physical kinematics. They have also proposed a method for determining the leading order low-energy constant, aNSq , associated with the new operators. We show that their method has difficulties due to power divergent contributions and propose a new method to obtain this constant from the lattice which does not suffer from this problem. Using this alternative method, we obtain aNSq , and show that our value implies a large ambiguity in the quenched contribution of Q6 to epsilon'/epsilon.
Framework For Improved Lattice Calculations Of Epsilion'/epsilon
Laiho, J
2004-01-01
In this thesis we show that it is possible to construct ε ′/ε to NLO using both full and partially quenched chiral perturbation theory (PQChPT) from amplitudes that are computable using numerical lattice gauge theory. We find that the electro- weak penguin (Δ I = 3/2 and 1/2) contributions to ε′/ε in PQChPT can be determined to NLO using only degenerate (mK = mπ) K → π computations without momentum insertion. All one-loop formulas needed to extract the necessary NLO constants from the lattice are presented in this work. Issues pertaining to power divergent contributions, originating from mixing with lower dimensional operators in a lattice regularization, are addressed. In embedding the QCD penguin left-right operator onto PQChPT an ambiguity arises when the number of light sea quarks is not the physical value of three, as first emphasized by Golterman and Pallante. In the quenched theory they have pointed out that there...
A Simple Spreadsheet Program for the Calculation of Lattice-Site Distributions
McCaffrey, John G.
2009-01-01
A simple spreadsheet program is presented that can be used by undergraduate students to calculate the lattice-site distributions in solids. A major strength of the method is the natural way in which the correct number of ions or atoms are present, or absent, at specific lattice distances. The expanding-cube method utilized is straightforward to…
Lattice Calculation of the Decay of Primordial Higgs Condensate
Enqvist, Kari; Rusak, Stanislav; Weir, David
2015-01-01
We study the resonant decay of the primordial Standard Model Higgs condensate after inflation into $SU(2)$ gauge bosons on the lattice. We find that the non-Abelian interactions between the gauge bosons quickly extend the momentum distribution towards high values, efficiently destroying the condensate after the onset of backreaction. For the inflationary scale $H = 10^8$ GeV, we find that 90% of the Higgs condensate has decayed after $n \\sim 10$ oscillation cycles. This differs significantly from the Abelian case where, given the same couplings strengths, most of the condensate would persist after the resonance.
Lattice calculation of the decay of primordial Higgs condensate
Enqvist, Kari; Nurmi, Sami; Rusak, Stanislav; Weir, David J.
2016-02-01
We study the resonant decay of the primordial Standard Model Higgs condensate after inflation into SU(2) gauge bosons on the lattice. We find that the non-Abelian interactions between the gauge bosons quickly extend the momentum distribution towards high values, efficiently destroying the condensate after the onset of backreaction. For the inflationary scale H = 108 GeV, we find that 90% of the Higgs condensate has decayed after n~ 10 oscillation cycles. This differs significantly from the Abelian case where, given the same coupling strengths, most of the condensate would persist after the resonance.
On calculating disconnected-type hadronic light-by-light scattering diagrams from lattice QCD
Hayakawa, M; Christ, N H; Izubuchi, T; Jin, L C; Lehner, C
2015-01-01
For reliable comparison of the standard model prediction to the muon g-2 with its experimental value, the hadronic light-by-light scattering (HLbL) contribution must be calculated by lattice QCD simulation. HLbL contribution has many types of disconnected-type diagrams. Here, we start with recalling the point that must be taken care of in every method to calculate them by lattice QCD, and present one concrete method called nonperturbative QED method.
Lattice location of dopant atoms: An -body model calculation
N K Deepak
2010-03-01
The channelling and scattering yields of 1 MeV -particles in the $\\langle 1 0 0 \\rangle$, $\\langle 1 1 0 \\rangle and $\\langle 1 1 1 \\rangle$ directions of silicon implanted with bismuth and ytterbium have been simulated using -body model. The close encounter yield from dopant atoms in silicon is determined from the flux density, using the Bontemps and Fontenille method. All previous works reported in literature so far have been done with computer programmes using a statistical analytical expression or by a binary collision model or a continuum model. These results at the best gave only the transverse displacement of the lattice site from the concerned channelling direction. Here we applied the superior -body model to study the yield from bismuth in silicon. The finding that bismuth atom occupies a position close to the silicon substitutional site is new. The transverse displacement of the suggested lattice site from the channelling direction is consistent with the experimental results. The above model is also applied to determine the location of ytterbium in silicon. The present values show good agreement with the experimental results.
Direct calculation of the lattice Green function with arbitrary interactions for general crystals.
Yasi, Joseph A; Trinkle, Dallas R
2012-06-01
Efficient computation of lattice defect geometries such as point defects, dislocations, disconnections, grain boundaries, interfaces, and free surfaces requires accurate coupling of displacements near the defect to the long-range elastic strain. Flexible boundary condition methods embed a defect in infinite harmonic bulk through the lattice Green function. We demonstrate an efficient and accurate calculation of the lattice Green function from the force-constant matrix for general crystals with an arbitrary basis by extending a method for Bravais lattices. New terms appear due to the presence of optical modes and the possible loss of inversion symmetry. By separately treating poles and discontinuities in reciprocal space, numerical accuracy is controlled at all distances. We compute the lattice Green function for a two-dimensional model with broken symmetry to elucidate the role of different coupling terms. The algorithm is generally applicable in two and three dimensions to crystals with arbitrary number of atoms in the unit cell, symmetry, and interactions.
Fuel lattice design in a boiling water reactor using an ant-colony-based system
Montes, Jose Luis, E-mail: joseluis.montes@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, La Marquesa, Ocoyoacac, Estado de Mexico, CP 52750 (Mexico); Facultad de Ciencias, Universidad Autonoma del Estado de Mexico (Mexico); Francois, Juan-Luis, E-mail: juan.luis.francois@gmail.com [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, Jiutepec, Mor., CP 62550 (Mexico); Ortiz, Juan Jose, E-mail: juanjose.ortiz@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, La Marquesa, Ocoyoacac, Estado de Mexico, CP 52750 (Mexico); Martin-del-Campo, Cecilia, E-mail: cecilia.martin.del.campo@gmail.com [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, Jiutepec, Mor., CP 62550 (Mexico); Perusquia, Raul, E-mail: raul.perusquia@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, La Marquesa, Ocoyoacac, Estado de Mexico, CP 52750 (Mexico)
2011-06-15
Research highlights: > We present an ant-colony-based system for BWR fuel lattice design and optimization. > Assessment of candidate solutions at 0.0 MWd/kg {sup 235}U seems to have a limited scope. > Suitable heuristic rules enable more realistic fuel lattice designs. > The election of the objective has a large impact in CPU time. > ACS enables an important decrease of the initial average U-235 enrichment. - Abstract: This paper presents a new approach to deal with the boiling water reactor radial fuel lattice design. The goal is to optimize the distribution of both, the fissionable material, and the reactivity control poison material inside the fuel lattice at the beginning of its life. An ant-colony-based system was used to search for either: the optimum location of the poisoned pin inside the lattice, or the U{sup 235} enrichment and Gd{sub 2}O{sub 3} concentrations. In the optimization process, in order to know the parameters of the candidate solutions, the neutronic simulator CASMO-4 transport code was used. A typical 10 x 10 BWR fuel lattice with an initial average U{sup 235} enrichment of 4.1%, used in the current operation of Laguna Verde Nuclear Power Plant was taken as a reference. With respect to that reference lattice, it was possible to decrease the average U{sup 235} enrichment up to 3.949%, this obtained value represents a decrease of 3.84% with respect to the reference U{sup 235} enrichment; whereas, the k-infinity was inside the {+-}100 pcm's range, and there was a difference of 0.94% between the local power peaking factor and the lattice reference value. Particular emphasis was made on defining the objective function which is used for making the assessment of candidate solutions. In a typical desktop personal computer, about four hours of CPU time were necessary for the algorithm to fulfill the goals of the optimization process. The results obtained with the application of the implemented system showed that the proposed approach represents a
Lattice QCD calculation of $K^+ K^-$ scattering length
Fu, Ziwen
2012-01-01
We deliver ab initio calculation of s-wave $K^+K^-$ scattering length ($a_0^{K^+K^-}$) by L\\"uscher's formula. In the "Asqtad" improved staggered dynamical fermion formulation, we measure $K^+K^-$ four-point correlation function by moving wall sources without gauge fixing, and find $a_0^{K^+K^-} = 0.456 \\pm 0.272$ fm, which is in reasonable agreement with tree-level prediction and comparable with experimental result. An essential ingredient in our calculation is to explicitly include the disconnected diagram.
Casimir energy calculations within the formalism of the noncompact lattice QED
Pavlovsky, Oleg
2009-01-01
A new method based on the Monte-Carlo calculation on the lattice is proposed to study the Casimir effect in the noncompact lattice QED. We have studied the standard Casimir problem with two parallel plane surfaces (mirrors) and oblique boundary conditions on those as a test of our method. Physically, this boundary conditions may appear in the problem of modelling of the thin material films interaction and are generated by additional Chern-Simons boundary term. This approach for the boundary condition generation is very suitable for the lattice formulation of the Casimir problem due to gauge invariance.
Recent progress in lattice calculations of properties of open-charm mesons
Mohler, Daniel
2015-01-01
Recent progress in lattice calculations of properties of open-charm mesons, both regular and exotic, is reviewed, with an emphasis on spectroscopy. After reviewing recent calculations of excited state energy levels I will discuss progress in extracting hadronic masses and widths of charmed states from Lattice QCD simulations including low-lying scattering channels directly, to determine phase shift data and bound state/ resonance properties. With regard to other properties results from recent calculations of the $DD^*\\pi$ and $DD\\rho$, $D^*D^*\\rho$ couplings are presented. Beyond regular mesons, searches for explicitly exotic (tetraquark) states are also reviewed.
Calculation of body-centered-cubic lattice sums with an application to ferromagnetism.
Wintucky, E. G.
1972-01-01
The lattice sums for the bcc lattice are recalculated using the method of Flax and Raich to obtain more general expressions, valid for all temperatures, in terms of a Langevin function and its derivatives. Formulas are presented which enable easy numerical evaluation. A comparison with well-known low-temperature expansions and with the results of direct numerical integration demonstrates the validity at low temperatures of the more general expressions calculated here.
Application of perturbation theory to lattice calculations based on method of cyclic characteristics
Assawaroongruengchot, Monchai
computing time when both direct and adjoint solutions are required. A problem that arises for the generalized adjoint problem is that the direct use of the negative external generalized adjoint sources in the adjoint solution algorithm results in negative generalized adjoint functions. A coupled flux biasing/decontamination scheme is applied to make the generalized adjoint functions positive using the adjoint functions in such a way that it can be used for the multigroup rebalance technique. Next we consider the application of the perturbation theory to the reactor problems. Since the coolant void reactivity (CVR) is a important factor in reactor safety analysis, we have decided to select this parameter for optimization studies. We consider the optimization and adjoint sensitivity techniques for the adjustments of CVR at beginning of burnup cycle (BOC) and k eff at end of burnup cycle (EOC) for a 2D Advanced CANDU Reactor (ACR) lattice. The sensitivity coefficients are evaluated using the perturbation theory based on the integral transport equations. Three sets of parameters for CVR-BOC and keff-EOC adjustments are studied: (1) Dysprosium density in the central pin with Uranium enrichment in the outer fuel rings, (2) Dysprosium density and Uranium enrichment both in the central pin, and (3) the same parameters as in the first case but the objective is to obtain a negative checkerboard CVR at beginning of cycle (CBCVR-BOC). To approximate the sensitivity coefficient at EOC, we perform constant-power burnup/depletion calculations for 600 full power days (FPD) using a slightly perturbed nuclear library and the unperturbed neutron fluxes to estimate the variation of nuclide densities at EOC. Sensitivity analyses of CVR and eigenvalue are included in the study. In addition the optimization and adjoint sensitivity techniques are applied to the CBCVR-BOC and keff-EOC adjustment of the ACR lattices with Gadolinium in the central pin. Finally we apply these techniques to the CVR
Zahra Nasrazadani
2017-02-01
Full Text Available The heavy water zero power reactor (HWZPR, which is a critical assembly with a maximum power of 100 W, can be used in different lattice pitches. The last change of core configuration was from a lattice pitch of 18–20 cm. Based on regulations, prior to the first operation of the reactor, a new core was simulated with MCNP (Monte Carlo N-Particle-4C and WIMS (Winfrith Improved Multigroup Scheme–CITATON codes. To investigate the criticality of this core, the effective multiplication factor (Keff versus heavy water level, and the critical water level were calculated. Then, for safety considerations, the reactivity worth of D2O, the reactivity worth of safety and control rods, and temperature reactivity coefficients for the fuel and the moderator, were calculated. The results show that the relevant criteria in the safety analysis report were satisfied in the new core. Therefore, with the permission of the reactor safety committee, the first criticality operation was conducted, and important physical parameters were measured experimentally. The results were compared with the corresponding values in the original core.
Lattice calculations for A=3,4,6,12 nuclei using chiral effective field theory
Epelbaum, Evgeny; Lee, Dean; Meißner, Ulf-G
2010-01-01
We present lattice calculations for the ground state energies of tritium, helium-3, helium-4, lithium-6, and carbon-12 nuclei. Our results were previously summarized in a letter publication. This paper provides full details of the calculations. We include isospin-breaking, Coulomb effects, and interactions up to next-to-next-to-leading order in chiral effective field theory.
Palau, J.M. [CEA Cadarache, Service de Physique des Reacteurs et du Cycle, Lab. de Projets Nucleaires, 13 - Saint-Paul-lez-Durance (France)
2005-07-01
This paper presents how Monte-Carlo calculations (French TRIPOLI4 poly-kinetic code with an appropriate pre-processing and post-processing software called OVNI) are used in the case of 3-dimensional heterogeneous benchmarks (slab reactor cores) to reduce model biases and enable a thorough and detailed analysis of the performances of deterministic methods and their associated data libraries with respect to key neutron parameters (reactivity, local power). Outstanding examples of application of these tools are presented regarding the new numerical methods implemented in the French lattice code APOLLO2 (advanced self-shielding models, new IDT characteristics method implemented within the discrete-ordinates flux solver model) and the JEFF3.1 nuclear data library (checked against JEF2.2 previous file). In particular we have pointed out, by performing multigroup/point-wise TRIPOLI4 (assembly and core) calculations, the efficiency (in terms of accuracy and computation time) of the new IDT method developed in APOLLO2. In addition, by performing 3-dimensional TRIPOLI4 calculations of the whole slab core (few millions of elementary volumes), the high quality of the new JEFF3.1 nuclear data files and revised evaluations (U{sup 235}, U{sup 238}, Hf) for reactivity prediction of slab cores critical experiments has been stressed. As a feedback of the whole validation process, improvements in terms of nuclear data (mainly Hf capture cross-sections) and numerical methods (advanced quadrature formulas accounting validation results, validation of new self-shielding models, parallelization) are suggested to improve even more the APOLLO2-CRONOS2 standard calculation route. (author)
S. Mattedi
2000-12-01
Full Text Available A modified form of the Hicks and Young algorithm was used with the Mattedi-Tavares-Castier lattice equation of state (MTC lattice EOS to calculate critical points of binary mixtures that exhibit several types of critical behavior. Several qualitative aspects of the critical curves, such as maxima and minima in critical pressure, and minima in critical temperature, could be predicted using the MTC lattice EOS. These results were in agreement with experimental information available in the literature, illustrating the flexibility of the functional form of the MTC lattice EOS. We observed however that the MTC lattice EOS failed to predict maxima in pressure for two of the studied systems: ethane + ethanol and methane + n-hexane. We also observed that the agreement between the calculated and experimental critical properties was at most semi-quantitative in some examples. Despite these limitations, in many ways similar to those of other EOS in common use when applied to critical point calculations, we can conclude that the MTC lattice EOS has the ability to predict several types of critical curves of complex shape.
Application of MCNP for neutronic calculations at VR-1 training reactor
Huml, Ondřej; Rataj, Jan; Bílý, Tomáš
2014-06-01
The paper presents utilization of Monte Carlo MCNP transport code for neutronic calculations of training reactor VR-1. Results of calculations are compared with results of measurements realized during last few critical experiments with various reactor core configurations. Very good agreement between calculations and measurements is observed.
On the effect of excited states in lattice calculations of the nucleon axial charge
Hansen, Maxwell T
2016-01-01
Excited-state contamination is one of the dominant uncertainties in lattice calculations of the nucleon axial-charge, $g_A$. Recently published results in leading-order chiral perturbation theory (ChPT) predict the excited-state contamination to be independent of the nucleon interpolator and positive. However, empirical results from numerical lattice calculations show negative contamination (downward curvature), indicating that present-day calculations are not in the regime where the leading-order ChPT predictions apply. In this paper we show that, under plausible assumptions, one can reproduce the behavior of lattice correlators by taking into account final-state $N \\pi$ interactions, in particular the effect of the Roper resonance, and by postulating a sign change in the infinite-volume $N \\to N \\pi$ axial-vector transition amplitude.
NONE
1963-07-01
This note constitutes the first edition of a Handbook for the calculation of reactor protections. This handbook makes it possible to calculate simply the different neutron and gamma fluxes and consequently, to fix the minimum quantities of materials necessary under general safety conditions both for the personnel and for the installations. It contains a certain amount of nuclear data, calculation methods, and constants corresponding to the present state of our knowledge. (authors) [French] Cette note constitue la premiere edition du 'Formulaire sur le calcul de la protection des reacteurs'. Ce formulaire permet de calculer de facon simple les difterents flux de neutrons et de gamma et, par suite, de fixer les quantites minima de materiaux a utiliser pour que les conditions generales de securite soient respectees, tant pour le personnel que pour les installations. Il contient un certain nombre de donnees nucleaires, de methodes de calcul et de constantes correspondant a l'etat actuel de nos connaissances. (auteurs)
SPECIFICATIONS FOR HIGH TEMPERATURE LATTICE TEST REACTOR BUILDING 318 PROJECT CAH-100
Vitro Engineering Company
1964-07-15
This is the specifications for the High Temperature Lattice Test Reactor Building 318 and it is divided into the following 21 divisions or chapters: (1) Excavating, Backfill & Grading; (2) Reinforced Concrete; (3) Masonry; (4) Structural Steel & Miscellaneous Metal Items, Contents - Division V; (5) Plumbing, Process & Service Piping; (6) Welding; (7) Insulated Metal Siding; (8) Roof Decks & Roofing; (9) Plaster Partitions & Ceiling; (10) Standard Doors, Windows & Hardware; (11) Shielding Doors; (12) Sprinkler System & Fire Extinguishers, Contents - Division XIII; (13) Heating, Ventilating & Air Conditioning; (14) Painting, Protective Coating & Floor Covering, Contents - Division XV; (15) Electrical; (16) Communications & Alarm Systems; (17) Special Equipment & Furnishings; (18) Overhead Bridge Crane; (19) Prefabricated Steel Building; (20) Paved Drive; and (21) Landscaping & Irrigation Sprinklers.
Lattice calculation of the pion transition form factor $\\pi^0 \\to \\gamma^* \\gamma^*$
Antoine, Gérardin; Nyffeler, Andreas
2016-01-01
We calculate the pion transition form factor ${\\cal F}_{\\pi^0\\gamma^*\\gamma^*}(q_1^2,q_2^2)$, which describe the interaction of an on-shell pion with two off-shell photons, using lattice QCD simulations with two degenerate flavors of dynamical quarks. This form factor is the main ingredient in the calculation of the pion-pole contribution to hadronic light-by-light scattering in the muon $g-2$, $a_\\mu^{\\mathrm{HLbL}; \\pi^0}$. We focus our study on the spacelike region with photon virtualities up to $1.5~\\mathrm{GeV}^2$, not yet measured experimentally. Several lattice spacings and pion masses are used to extrapolate the results to the physical point and a comparison with different phenomenological models is performed. Finally, we use our extrapolated form factor to provide a lattice determinaiton of $a_\\mu^{\\mathrm{HLbL}; \\pi^0}$.
KOELINK, MH; DEMUL, FFM; GREVE, J; GRAAFF, R; DASSEL, ACM; AARNOUDSE, JG
1992-01-01
In addition to the static cubic lattice model for photon migration in turbid biological media by Bonner et al. [J. Opt. Soc. Am. A 4, 423-432 (1987)], a dynamic method is presented to calculate the average absolute Doppler shift as a function of the distance between the point of injection of photons
Numerical calculations for Heisenberg ferromagnet on honeycomb lattice using Oguchi’s method
Mert, Gülistan; Mert, H. Şevki [Department of Physics, Selcuk University, 42075, Konya (Turkey)
2015-03-10
Magnetic properties such as the magnetization, internal energy and specific heat for Heisenberg ferromagnet with spin - 1/2 on honeycomb lattice are have been calculated using Oguchi’s method. We have found that the magnetic specific heat exhibits two peaks.
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)
The abnormal lattice contraction of plutonium hydrides studied by first-principles calculations
Ao Bing-Yun; Shi Peng; Guo Yong; Gao Tao
2013-01-01
Pu can be loaded with H forming complicated continuous solid solutions and compounds,and causing remarkable electronic and structural changes.Full potential linearized augmented plane wave methods combined with Hubbard parameter U and the spin-orbit effects are employed to investigate the electronic and structural properties of stoichiometric and non-stoichiometric face-centered cubic Pu hydrides (PuHx,x =2,2.25,2.5,2.75,3).The decreasing trend with increasing x of the calculated lattice parameters is in reasonable agreement with the experimental findings.A comparative analysis of the electronic-structure results for a series of PuHx compositions reveals that the lattice contraction results from the associated effects of the enhanced chemical bonding and the size effects involving the interstitial atoms.We find that the size effects are the driving force for the abnormal lattice contraction.
Shao, Yongliang; Zhang, Lei; Hao, Xiaopeng; Wu, Yongzhong; Dai, Yuanbin; Tian, Yuan; Huo, Qin
2014-08-05
We report a method to obtain the stress of crystalline materials directly from lattice deformation by Hooke's law. The lattice deformation was calculated using the crystallographic orientations obtained from electron backscatter diffraction (EBSD) technology. The stress distribution over a large area was obtained efficiently and accurately using this method. Wurtzite structure gallium nitride (GaN) crystal was used as the example of a hexagonal crystal system. With this method, the stress distribution of a GaN crystal was obtained. Raman spectroscopy was used to verify the stress distribution. The cause of the stress distribution found in the GaN crystal was discussed from theoretical analysis and EBSD data. Other properties related to lattice deformation, such as piezoelectricity, can also be analyzed by this novel approach based on EBSD data.
CCSD(T)/CBS fragment-based calculations of lattice energy of molecular crystals
Červinka, Ctirad; Fulem, Michal; Růžička, Květoslav
2016-02-01
A comparative study of the lattice energy calculations for a data set of 25 molecular crystals is performed using an additive scheme based on the individual energies of up to four-body interactions calculated using the coupled clusters with iterative treatment of single and double excitations and perturbative triples correction (CCSD(T)) with an estimated complete basis set (CBS) description. The CCSD(T)/CBS values on lattice energies are used to estimate sublimation enthalpies which are compared with critically assessed and thermodynamically consistent experimental values. The average absolute percentage deviation of calculated sublimation enthalpies from experimental values amounts to 13% (corresponding to 4.8 kJ mol-1 on absolute scale) with unbiased distribution of positive to negative deviations. As pair interaction energies present a dominant contribution to the lattice energy and CCSD(T)/CBS calculations still remain computationally costly, benchmark calculations of pair interaction energies defined by crystal parameters involving 17 levels of theory, including recently developed methods with local and explicit treatment of electronic correlation, such as LCC and LCC-F12, are also presented. Locally and explicitly correlated methods are found to be computationally effective and reliable methods enabling the application of fragment-based methods for larger systems.
Tang, Xiaoli [Physics Department, Auburn University, Auburn, Alabama (United States); Dong, Jianjun [Physics Department, Auburn University, Auburn, Alabama (United States)
2009-06-01
We report a recent first-principles calculation of harmonic and anharmonic lattice dynamics of MgO. The 2nd order harmonic and 3rd order anharmonic interatomic interaction terms are computed explicitly, and their pressure dependences are discussed. The phonon mode Grueneisen parameters derived based on our calculated 3rd order lattice anharmonicity are in good agreement with those estimated using the finite difference method. The implications for lattice thermal conductivity at high pressure are discussed based on a simple kinetic transport theory.
Calculation of the Nucleon Axial Form Factor Using Staggered Lattice QCD
Meyer, Aaron S; Kronfeld, Andreas S; Li, Ruizi; Simone, James N
2016-01-01
The nucleon axial form factor is a dominant contribution to errors in neutrino oscillation studies. Lattice QCD calculations can help control theory errors by providing first-principles information on nucleon form factors. In these proceedings, we present preliminary results on a blinded calculation of $g_A$ and the axial form factor using HISQ staggered baryons with 2+1+1 flavors of sea quarks. Calculations are done using physical light quark masses and are absolutely normalized. We discuss fitting form factor data with the model-independent $z$ expansion parametrization.
Anisotropic intrinsic lattice thermal conductivity of borophane from first-principles calculations.
Liu, Gang; Wang, Haifeng; Gao, Yan; Zhou, Jian; Wang, Hui
2017-01-25
Borophene (boron sheet) as a new type of two-dimensional (2D) material was grown successfully recently. Unfortunately, the structural stability of freestanding borophene is still an open issue. Theoretical research has found that full hydrogenation can remove such instability, and the product is called borophane. In this paper, using first-principles calculations we investigate the lattice dynamics and thermal transport properties of borophane. The intrinsic lattice thermal conductivity and the relaxation time of borophane are investigated by solving the phonon Boltzmann transport equation (BTE) based on first-principles calculations. We find that the intrinsic lattice thermal conductivity of borophane is anisotropic, as the higher value (along the zigzag direction) is about two times of the lower one (along the armchair direction). The contributions of phonon branches to the lattice thermal conductivities along different directions are evaluated. It is found that both the anisotropy of thermal conductivity and the different phonon branches which dominate the thermal transport along different directions are decided by the group velocity and the relaxation time of phonons with very low frequency. In addition, the size dependence of thermal conductivity is investigated using cumulative thermal conductivity. The underlying physical mechanisms of these unique properties are also discussed in this paper.
Decay heat measurement on fusion reactor materials and validation of calculation code system
Maekawa, Fujio; Ikeda, Yujiro; Wada, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1998-03-01
Decay heat rates for 32 fusion reactor relevant materials irradiated with 14-MeV neutrons were measured for the cooling time period between 1 minute and 400 days. With using the experimental data base, validity of decay heat calculation systems for fusion reactors were investigated. (author)
System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors
Moiseyev, A.V. [SSC RF - IPPE, 1 Bondarenko Square, Obninsk, Kaluga Region 249033 (Russian Federation)
2008-07-01
There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k{sub eff}, control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)
Entanglement entropy for a Maxwell field: Numerical calculation on a two dimensional lattice
Casini, Horacio
2014-01-01
We study entanglement entropy (EE) for a Maxwell field in 2+1 dimensions. We do numerical calculations in two dimensional lattices. This gives a concrete example of the general results of our recent work on entropy for lattice gauge fields using an algebraic approach. To evaluate the entropies we extend the standard calculation methods for the entropy of Gaussian states in canonical commutation algebras to the more general case of algebras with center and arbitrary numerical commutators. We find that while the entropy depends on the details of the algebra choice, mutual information has a well defined continuum limit. We study several universal terms for the entropy of the Maxwell field and compare with the case of a massless scalar field. We find some interesting new phenomena: An "evanescent" logarithmically divergent term in the entropy with topological coefficient which does not have any correspondence with ultraviolet entanglement in the universal quantities, and a non standard way in which strong subaddi...
$\\pi_0$ pole mass calculation in a strong magnetic field and lattice constraints
Avancini, Sidney S; Pinto, Marcus Benghi; Tavares, William R; Timóteo, Varese S
2016-01-01
The $\\pi_0$ neutral meson pole mass is calculated in a strongly magnetized medium using the SU(2) Nambu-Jona-Lasinio model within the random phase approximation (RPA) at zero temperature and zero baryonic density. We employ a magnetic field dependent coupling $G(eB)$ fitted to reproduce lattice QCD results for the quark condensates. Divergent quantities are handled with a magnetic field independent regularization scheme in order to avoid unphysical oscillations. A comparison between the running and the fixed couplings reveals that the former produces results much closer to the predictions from recent lattice calculations. In particular, we find that the $\\pi_0$ meson mass systematically decreases when the magnetic field increases while the scalar mass remains almost constant. We also investigate how the magnetic background influences other mesonic properties such as $f_{{\\pi}_0}$ and $g_{\\pi_0 q q}$.
Michel, K. H.; ćakır, D.; Sevik, C.; Peeters, F. M.
2017-03-01
The elastic constant C11 and piezoelectric stress constant e1 ,11 of two-dimensional (2D) dielectric materials comprising h-BN, 2 H -MoS2 , and other transition-metal dichalcogenides and dioxides are calculated using lattice dynamical theory. The results are compared with corresponding quantities obtained with ab initio calculations. We identify the difference between clamped-ion and relaxed-ion contributions with the dependence on inner strains which are due to the relative displacements of the ions in the unit cell. Lattice dynamics allows us to express the inner-strain contributions in terms of microscopic quantities such as effective ionic charges and optoacoustical couplings, which allows us to clarify differences in the piezoelectric behavior between h-BN and MoS2. Trends in the different microscopic quantities as functions of atomic composition are discussed.
Lattice dynamics and spin-phonon interactions in multiferroic RMn2O5: Shell model calculations
Litvinchuk, A. P.
2009-08-01
The results of the shell model lattice dynamics calculations of multiferroic RMn2O5 materials (space group Pbam) are reported. Theoretical even-parity eigenmode frequencies are compared with those obtained experimentally in polarized Raman scattering experiments for R=Ho,Dy. Analysis of displacement patterns allows to identify vibrational modes which facilitate spin-phonon coupling by modulating the Mn-Mn exchange interaction and provides explanation of the observed anomalous temperature behavior of phonons.
Calculation of the Spin-Dependent Optical Lattice in Rubidium Bose-Einstein Condensation
CAO Ming-Tao; HAN Liang; QI Yue-Rong; ZHANG Shou-Gang; GAO Hong; LI Fu-Li
2012-01-01
We provide a theoretical study to calculate the spin-dependent optical lattice with rubidium Bose-Einstein condensation (BEC) in a steady magnetic field.The optical dipole potential variation at different Zeeman levels are obtained.We also show that atoms can be transported in three dimensions by changing the polarization of the trapping field.An explanation of this transportation process in an atomic coordinate is presented.
Lattice dynamics of wurtzite CdS: Neutron scattering and ab-initio calculations
Debernardi, A.; Pyka, N. M.; Göbel, A.; Ruf, T.; Lauck, R.; Kramp, S.; Cardona, M.
1997-08-01
We have measured the phonon dispersion of wurtzite CdS by inelastic neutron scattering in a single crystal made from the nonabsorbing isotope 114Cd. One of the two silent B 1-modes occurs at 3.96 THz ( k = 0 ). It is significantly lower and less dispersive than so far assumed. Previous semiempirical lattice dynamical models need to be reanalyzed. However, the observed dispersion branches compare favorably with an ab-initio calculation.
Moriarty, K.J.M. (Royal Holloway Coll., Englefield Green (UK). Dept. of Mathematics); Blackshaw, J.E. (Floating Point Systems UK Ltd., Bracknell)
1983-04-01
The computer program calculates the average action per plaquette for SU(6)/Z/sub 6/ lattice gauge theory. By considering quantum field theory on a space-time lattice, the ultraviolet divergences of the theory are regulated through the finite lattice spacing. The continuum theory results can be obtained by a renormalization group procedure. Making use of the FPS Mathematics Library (MATHLIB), we are able to generate an efficient code for the Monte Carlo algorithm for lattice gauge theory calculations which compares favourably with the performance of the CDC 7600.
Gabriel, T.A.; Bishop, B.L.; Wiffen, F.W.
1979-08-01
In order to plan radiation damage experiments in fission reactors keyed toward fusion reactor applications, it is necessary to have available for these facilities displacement per atom (dpa) and gas production rates for many potential materials. This report supplies such data for the elemental constituents of alloys of interest to the United States fusion reactor alloy development program. The calculations are presented for positions of interest in the HFIR, ORR, and EBR-II reactors. DPA and gas production rates in alloys of interest can be synthesized from these results.
Calculation to experiment comparison of SPND signals in various nuclear reactor environments
Barbot, Loic; Radulovic, Vladimir; Fourmentel, Damien [CEA, DEN, DER, Instrumentation, Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance, (France); Snoj, Luka [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana, (Slovenia); Tarchalski, Mikolaj [National Centre for Nuclear Research, ulica Andrzeja Soltana 7, 05-400 Otwock (Swierk), (Poland); Dewynter-Marty, Veronique [CEA, DEN, DANS, DRSN, SIREN, LESCI, Saclay, F-91191 Gif sur Yvette, (France); Malouch, Fadhel [CEA, DEN, DANS, DM2S, SERMA, Saclay, F-91191 Gif sur Yvette, (France)
2015-07-01
In the perspective of irradiation experiments in the future Jules Horowitz Reactor (JHR), the Instrumentation Sensors and Dosimetry Laboratory of CEA Cadarache (France) is developing a numerical tool for SPND design, simulation and operation. In the frame of the SPND numerical tool qualification, dedicated experiments have been performed both in the Slovenian TRIGA Mark II reactor (JSI) and very recently in the French CEA Saclay OSIRIS reactor, as well as a test of two detectors in the core of the Polish MARIA reactor (NCBJ). A full description of experimental set-ups and neutron-gamma calculations schemes are provided in the first part of the paper. Calculation to experiment comparison of the various SPNDs in the different reactors is thoroughly described and discussed in the second part. Presented comparisons show promising final results. (authors)
OPAL REACTOR: Calculation/Experiment comparison of Neutron Flux Mapping in Flux Coolant Channels
Barbot, L.; Domergue, C.; Villard, J. F.; Destouches, C. [CEA, Paris (France); Braoudakis, G.; Wassink, D.; Sinclair, B.; Osborn, J. C.; Huayou, Wu [ANSTO, Syeney (Australia)
2013-07-01
The measurement and calculation of the neutron flux mapping of the OPAL research reactor are presented. Following an investigation of fuel coolant channels using sub-miniature fission chambers to measure thermal neutron flux profiles, neutronic calculations were performed. Comparison between calculation and measurement shows very good agreement.
International Electrotechnical Commission. Geneva
1988-01-01
This standard applies to the following types of reactors: shunt reactors, current-limiting reactors including neutral-earthing reactors, damping reactors, tuning (filter) reactors, earthing transformers (neutral couplers), arc-suppression reactors, smoothing reactors, with the exception of the following reactors: small reactors with a rating generally less than 2 kvar single-phase and 10 kvar three-phase, reactors for special purposes such as high-frequency line traps or reactors mounted on rolling stock.
Lattice calculation of electric dipole moments and form factors of the nucleon
Abramczyk, M.; Aoki, S.; Blum, T.; Izubuchi, T.; Ohki, H.; Syritsyn, S.
2017-07-01
We analyze commonly used expressions for computing the nucleon electric dipole form factors (EDFF) F3 and moments (EDM) on a lattice and find that they lead to spurious contributions from the Pauli form factor F2 due to inadequate definition of these form factors when parity mixing of lattice nucleon fields is involved. Using chirally symmetric domain wall fermions, we calculate the proton and the neutron EDFF induced by the C P -violating quark chromo-EDM interaction using the corrected expression. In addition, we calculate the electric dipole moment of the neutron using a background electric field that respects time translation invariance and boundary conditions, and we find that it decidedly agrees with the new formula but not the old formula for F3. Finally, we analyze some selected lattice results for the nucleon EDM and observe that after the correction is applied, they either agree with zero or are substantially reduced in magnitude, thus reconciling their difference from phenomenological estimates of the nucleon EDM.
Lattice calculation of electric dipole moments and form factors of the nucleon
Abramczyk, M.; Aoki, S.; Blum, T.; Izubuchi, T.; Ohki, H.; Syritsyn, S.
2017-07-01
We analyze commonly used expressions for computing the nucleon electric dipole form factors (EDFF) $F_3$ and moments (EDM) on a lattice and find that they lead to spurious contributions from the Pauli form factor $F_2$ due to inadequate definition of these form factors when parity mixing of lattice nucleon fields is involved. Using chirally symmetric domain wall fermions, we calculate the proton and the neutron EDFF induced by the CP-violating quark chromo-EDM interaction using the corrected expression. In addition, we calculate the electric dipole moment of the neutron using background electric field that respects time translation invariance and boundary conditions, and find that it decidedly agrees with the new formula but not the old formula for $F_3$. Finally, we analyze some selected lattice results for the nucleon EDM and observe that after the correction is applied, they either agree with zero or are substantially reduced in magnitude, thus reconciling their difference from phenomenological estimates of the nucleon EDM.
Validation of the CAREM calculation line on WWER-type lattices
Lecot, Carlos A. [Investigacion Aplicada SE (INVAP), San Carlos de Bariloche (Argentina)
1997-12-01
The INVAP calculation line, based on the CONDOR and CITVAP codes, was validated against the WWER-type lattices. Experimental results were extracted from the volumes I, II and III of the `Final Report of TIC: Experimental Investigations of the Physical Properties of WWER-type Uranium-Water Lattices`. The results are presented in two stages: a brief summary about the validation of the cell code CONDOR v 1.3 and a full explanation related with the validation of the CITVAP core code. The critical codes evaluated included homogeneous and regularly perturbed configurations, with a high variety of enrichments, boric acid concentration in moderator, lattice pitches and perturbation such as water holes, Zr B{sub 2} and B{sub 4} C rods, Gd{sub 2} O{sub 3} rods, among others. The conclusions obtained in these evaluations are considered as part of the CAREM calculation line evaluation to be carried out in the CONCRIT R A-8 critical facility. (author). 10 refs., 2 figs., 13 tabs.
Reactor Physics Analysis Models for a CANDU Reactor
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.
Reactor Physics Analysis Models for a CANDU Reactor
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.
Model calculations of edge dislocation defects and vacancies in {alpha}-Iron lattice
Petrov, L; Troev, T; Nankov, N; Popov, E, E-mail: lpetrov@inrne.bas.b [Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences, 72 Tzarigradsko Chaussee Blvd., 1784 Sofia (Bulgaria)
2010-01-01
Two models of defects in perfect {alpha}-iron lattice were discussed. In the perfect bcc iron lattice 42x42x42 a{sub o} (a{sub o} = 2,87 A) an edge dislocation was created, moving the second half of the bulk on one a{sub o} distance. This action generates a little volume in the middle of the bulk witch increases of the positron lifetime (PLT) calculated using the superimposed-atom method of Puska and Nieminen [1]. The result of 118 ps PLT in simple edge dislocation's model is in a good concurrence with earlier publications and experimental data [2]. Through the dislocation line one, two and three vacancies were localized. These models give the results for PLT of 146, 157 and 167 ps respectively. The computer simulations were performed using Finnis-Sinclair (FS) N-body potential.
Kaya, Savaş; Kaya, Cemal
2015-09-08
This paper presents a new technique for estimation of lattice energies of inorganic ionic compounds using a simple formula. This new method demonstrates the relationship between chemical hardness and lattice energies of ionic compounds. Here chemical hardness values of ionic compounds are calculated via our molecular hardness equation. The results obtained using the present method and comparisons made by considering experimental data and the results from other theoretical methods in the literature showed that the new method allows easy evaluation of lattice energies of inorganic ionic crystals without the need for ab initio calculations and complex calculations.
The concerted calculation of the BN-600 reactor for the deterministic and stochastic codes
Bogdanova, E. V.; Kuznetsov, A. N.
2017-01-01
The solution of the problem of increasing the safety of nuclear power plants implies the existence of complete and reliable information about the processes occurring in the core of a working reactor. Nowadays the Monte-Carlo method is the most general-purpose method used to calculate the neutron-physical characteristic of the reactor. But it is characterized by large time of calculation. Therefore, it may be useful to carry out coupled calculations with stochastic and deterministic codes. This article presents the results of research for possibility of combining stochastic and deterministic algorithms in calculation the reactor BN-600. This is only one part of the work, which was carried out in the framework of the graduation project at the NRC “Kurchatov Institute” in cooperation with S. S. Gorodkov and M. A. Kalugin. It is considering the 2-D layer of the BN-600 reactor core from the international benchmark test, published in the report IAEA-TECDOC-1623. Calculations of the reactor were performed with MCU code and then with a standard operative diffusion algorithm with constants taken from the Monte - Carlo computation. Macro cross-section, diffusion coefficients, the effective multiplication factor and the distribution of neutron flux and power were obtained in 15 energy groups. The reasonable agreement between stochastic and deterministic calculations of the BN-600 is observed.
Structure and lattice dynamics of rare-earth ferroborate crystals: Ab initio calculation
Chernyshev, V. A.; Nikiforov, A. E.; Petrov, V. P.; Serdtsev, A. V.; Kashchenko, M. A.; Klimin, S. A.
2016-08-01
The ab initio calculation of the crystal structure and the phonon spectrum of crystals RFe3(BO3)4 ( R = Pr, Nd, Sm) has been performed in the framework of the density functional theory. The ion coordinates in the unit cell, the lattice parameters, the frequencies and the types of fundamental vibrations, and also the intensities of lines in the Raman spectrum and infrared reflection spectra have been found. The elastic constants of the crystals have been calculated. For low-frequency A 2 mode in PrFe3(BO3)4, a "seed" vibration frequency that strongly interacts with the electronic excitation on a praseodymium ion was found. The calculation results satisfactory agree with the experimental data.
Parallelization of heterogeneous reactor calculations on a graphics processing unit
Malofeev, V. M., E-mail: vm-malofeev@mail.ru; Pal’shin, V. A. [National Research Center Kurchatov Institute (Russian Federation)
2016-12-15
Parallelization is applied to the neutron calculations performed by the heterogeneous method on a graphics processing unit. The parallel algorithm of the modified TREC code is described. The efficiency of the parallel algorithm is evaluated.
Some methods for calculation of perturbations in nuclear reactors
Abramov, B. D., E-mail: abramov@ippe.ru [Leypunsky Institute of Physics and Power Engineering (Russian Federation)
2015-12-15
Some methods for calculation of local perturbations of neutron fields and reactivity effects accompanying them are considered. Existence, uniqueness, properties and methods for finding solutions to the considered problems are discussed.
Lattice calculation of the pion transition form factor $\\pi^0 \\to \\gamma^* \\gamma^*$
Gérardin, Antoine; Nyffeler, Andreas
2016-01-01
We calculate the $\\pi^0\\to \\gamma^*\\gamma^*$ transition form factor ${\\cal F}_{\\pi^0\\gamma^*\\gamma^*}(q_1^2,q_2^2)$ in lattice QCD with two flavors of quarks. Our main motivation is to provide the input to calculate the $\\pi^0$-pole contribution to hadronic light-by-light scattering in the muon $(g-2)$, $a_\\mu^{\\rm HLbL;\\pi^0}$. We therefore focus on the region where both photons are spacelike up to virtualities of about $1.5~$GeV$^2$, which has so far not been experimentally accessible. Results are obtained in the continuum at the physical pion mass by a combined extrapolation. We reproduce the prediction of the chiral anomaly for real photons with an accuracy of about $8-9\\%$. We also compare to various recently proposed models and find reasonable agreement for the parameters of some of these models with their phenomenological values. Finally, we use the parametrization of our lattice data by these models to calculate $a_\\mu^{\\rm HLbL;\\pi^0}$.
Peng-Jen Chen; Horng-Tay Jeng
2016-01-01
A new semiconducting phase of two-dimensional phosphorous in the Kagome lattice is proposed from first-principles calculations. The band gaps of the monolayer (ML) and bulk Kagome phosphorous (Kagome-P) are 2.00 and 1.11 eV, respectively. The magnitude of the band gap is tunable by applying the in-plane strain and/or changing the number of stacking layers. High optical absorption coefficients at the visible light region are predicted for multilayer Kagome-P, indicating potential applications ...
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2002-01-01
The paper specifies an unambiguous basic relationship between the published results of ab initio calculations of lattice energies,EL,and heats of sublimation,ΔHs,of individual energetic materials. In this relationship,the ΔHs value has been replaced by heats of fusion,ΔHm,tr. Thereby its unambiguity has been lost,and the similarity of details of molecular structure begins to be of decisive importance. The resulting partial relationships,together with the basic relationship,have been used for prediction of ΔHs,and ΔHm,tr values of technically attractive polynitro compounds.
Jordan, K. A.; Schubring, D. [Univ. of Florida, Florida (United States); Girardin, G.; Pautz, A. [Swiss Federal Institute of Technology, Zuerich (Switzerland)
2013-07-01
domains will be expanded and the validation base of commonlyused calculation methods will be expanded to cover a new range of research reactor types. From a practical perspective, CROCUS and the UFTR will have fully validated reactor dynamic and transient models for dynamic and accident analysis. With these validated models, both facilities will have improved capabilities and flexibility for extended operations in the future. CROCUS and the UFTR will be able to make future reactor modifications with reduced regulatory resistance. A feasibility analysis of future power uprates at these facilities will also result.
Core calculations for the upgrading of the IEA-R1 research reactor
Santos, Adimir dos; Perrotta, Jose A.; Bastos, Jose Luis F.; Yamaguchi, Mitsuo; Umbehaun, Pedro E. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)]. E-mail: asantos@net.ipen.br; perrotta@net.ipen.br; mitsuo@net.ipen.br
1998-07-01
The IEA-R1 Research Reactor is a multipurpose reactor. It has been used for basic and applied research in the nuclear area, training and radioisotopes production since 1957. In 1995, the Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) took the decision to modernize and upgrade the power from 2 to 5 MW and increase the operational cycle. This work presents the design requirements and the calculations effectuated to reach this goal. (author)
Blanc-Tranchant, P
2001-07-01
The general field in which this work takes place is the field of the accuracy improvement of neutronic calculations, required to operate Pressurized Water Reactors (PWR) with a better precision and a lower cost. More specifically, this thesis deals with the calculation of the absorber clusters used to control these reactors. The first aim of that work was to define and validate a reference calculation route of such an absorber cluster, based on the deterministic code APOLLO2. This calculation scheme was then to be checked against experimental data. This study of the complex situation of absorber clusters required several intermediate studies, of simpler problems, such as the study of fuel rods lattices and the study of single absorber rods (B4C, AIC, Hafnium) isolated in such lattices. Each one of these different studies led to a particular reference calculation route. All these calculation routes were developed against reference continuous energy Monte-Carlo calculations, carried out with the stochastic code TRIPOLI4. They were then checked against experimental data measured during French experimental programs, undertaken within the EOLE experimental reactor, at the Nuclear Research Center of Cadarache: the MISTRAL experiments for the study of isolated absorber rods and the EPICURE experiments for the study of absorber clusters. This work led to important improvements in the calculation of isolated absorbers and absorber clusters. The reactivity worth of these clusters in particular, can now be obtained with a great accuracy: the discrepancy observed between the calculated and the experimental values is less than 2.5 %, and then slightly lower than the experimental uncertainty. (author)
Blanc-Tranchant, P
1999-11-01
The general field in which this work takes place is the field of the accuracy improvement of neutronic calculations, required to operate Pressurized Water Reactors (PWR) with a better precision and a lower cost. More specifically, this thesis deals with the calculation of the absorber clusters used to control these reactors. The first aim of that work was to define and validate a reference calculation route of such an absorber cluster, based on the deterministic code Apollo 2. This calculation scheme was then to be checked against experimental data. This study of the complex situation of absorber clusters required several intermediate studies, of simpler problems, such as the study of fuel rods lattices and the study of single absorber rods (B{sub 4}C, AIC, Hafnium) isolated in such lattices. Each one of these different studies led to a particular reference calculation route. All these calculation routes were developed against reference continuous energy Monte-Carlo calculations, carried out with the stochastic code TRIPOLI14. They were then checked against experimental data measured during french experimental programs, undertaken within the EOLE experimental reactor, at the Nuclear Research Center of Cadarache: the MISTRAL experiments for the study of isolated absorber rods and the EPICURE experiments for the study of absorber clusters. This work led to important improvements in the calculation of isolated absorbers and absorber clusters. The reactivity worth of these clusters in particular, can now be obtained with a great accuracy: the discrepancy observed between the calculated and the experimental values is less than 2.5 %, and then slightly lower than the experimental uncertainty. (author)
Criticality safety calculations of the Soreq research reactor storage pool
Caner, M.; Hirshfeld, H.; Nagler, A.; Silverman, I.; Bettan, M. [Soreq Nuclear Research Center, Yavne 81800 (Israel); Levine, S.H. [Penn State University, University Park 16802 (United States)
2001-07-01
The IRR-l spent fuel is to be relocated in a storage pool. The present paper describes the actual facility and summarizes the Monte Carlo criticality safety calculations. The fuel elements are to be placed inside cadmium boxes to reduce their reactivity. The fuel element is 7.6 cm by 8.0 cm in the horizontal plane. The cadmium box is effectively 9.7 cm by 9.7 cm, providing significant water between the cadmium and the fuel element. The present calculations show that the spent fuel storage pool is criticality safe even for fresh fuel elements. (author)
Laureau, A., E-mail: laureau.axel@gmail.com; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.
2017-05-15
Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.
Clerc, T., E-mail: thomas.clerc2@gmail.com [Institut de Génie Nucléaire, P.O. Box 6079, Station “Centre-Ville”, Montréal, Qc., Canada H3C 3A7 (Canada); Hébert, A., E-mail: alain.hebert@polymtl.ca [Institut de Génie Nucléaire, P.O. Box 6079, Station “Centre-Ville”, Montréal, Qc., Canada H3C 3A7 (Canada); Leroyer, H.; Argaud, J.P.; Bouriquet, B.; Ponçot, A. [Électricité de France, R and D, SINETICS, 1 Av. du Général de Gaulle, 92141 Clamart (France)
2014-07-01
Highlights: • We present a computational scheme for the determination of reflector properties in a PWR. • The approach is based on the minimization of a functional. • We use a data assimilation method or a parametric complementarity principle. • The reference target is a solution obtained with the method of characteristics. • The simplified flux solution is based on diffusion theory or on the simplified Pn method. - Abstract: This paper presents a computational scheme for the determination of equivalent 2D multi-group spatially dependant reflector parameters in a Pressurized Water Reactor (PWR). The proposed strategy is to define a full-core calculation consistent with a reference lattice code calculation such as the Method Of Characteristics (MOC) as implemented in APOLLO2 lattice code. The computational scheme presented here relies on the data assimilation module known as “Assimilation de données et Aide à l’Optimisation (ADAO)” of the SALOME platform developed at Électricité De France (EDF), coupled with the full-core code COCAGNE and with the lattice code APOLLO2. A first code-to-code verification of the computational scheme is made using the OPTEX reflector model developed at École Polytechnique de Montréal (EPM). As a result, we obtain 2D multi-group, spatially dependant reflector parameters, using both diffusion or SP{sub N} operators. We observe important improvements of the power discrepancies distribution over the core when using reflector parameters computed with the proposed computational scheme, and the SP{sub N} operator enables additional improvements.
The spectral code Apollo2: from lattice to 2D core calculations
Coste-Delclaux, M.; Santandrea, S.; Damian, F.; Blanc-Tranchant, P.; Zmijarevic, I. [CEA Saclay (DEN/DANS/SERMA), 91 - Gif-sur-Yvette (France); Santamarina, A. [CEA Cadarache (CEA/DEN/DER/SPRC), 13 - Saint Paul lez Durance (France)
2005-07-01
Apollo2 is a powerful code dedicated to neutron transport, it is a highly qualified tool for a wide range of applications from research and development studies to industrial applications. Today Apollo2 is part of several advanced 3-dimensional nuclear code packages dedicated to reactor physics, fuel cycle, criticality and safety analysis. The presentations have been organized into 7 topics: -) an introduction to Apollo2, -) cross-sections, -) flux calculation, -) advanced applications, -) Apollo2 users, specialized packages, -) qualification program, and -) the future of Apollo2. This document gathers only the slides of the presentations.
Parallelizing the QUDA Library for Multi-GPU Calculations in Lattice Quantum Chromodynamics
Ronald Babich, Michael Clark, Balint Joo
2010-11-01
Graphics Processing Units (GPUs) are having a transformational effect on numerical lattice quantum chromodynamics (LQCD) calculations of importance in nuclear and particle physics. The QUDA library provides a package of mixed precision sparse matrix linear solvers for LQCD applications, supporting single GPUs based on NVIDIA's Compute Unified Device Architecture (CUDA). This library, interfaced to the QDP++/Chroma framework for LQCD calculations, is currently in production use on the "9g" cluster at the Jefferson Laboratory, enabling unprecedented price/performance for a range of problems in LQCD. Nevertheless, memory constraints on current GPU devices limit the problem sizes that can be tackled. In this contribution we describe the parallelization of the QUDA library onto multiple GPUs using MPI, including strategies for the overlapping of communication and computation. We report on both weak and strong scaling for up to 32 GPUs interconnected by InfiniBand, on which we sustain in excess of 4 Tflops.
Fuel lattice design in a boiling water reactor using a knowledge-based automation system
Tung, Wu-Hsiung, E-mail: wstong@iner.gov.tw; Lee, Tien-Tso; Kuo, Weng-Sheng; Yaur, Shung-Jung
2015-11-15
Highlights: • An automation system was developed for the fuel lattice radial design of BWRs. • An enrichment group peaking equalizing method is applied to optimize the design. • Several heuristic rules and restrictions are incorporated to facilitate the design. • The CPU time for the system to design a 10x10 lattice was less than 1.2 h. • The beginning-of-life LPF was improved from 1.319 to 1.272 for one of the cases. - Abstract: A knowledge-based fuel lattice design automation system for BWRs is developed and applied to the design of 10 × 10 fuel lattices. The knowledge implemented in this fuel lattice design automation system includes the determination of gadolinium fuel pin location, the determination of fuel pin enrichment and enrichment distribution. The optimization process starts by determining the gadolinium distribution based on the pin power distribution of a flat enrichment lattice and some heuristic rules. Next, a pin power distribution flattening and an enrichment grouping process are introduced to determine the enrichment of each fuel pin enrichment type and the initial enrichment distribution of a fuel lattice design. Finally, enrichment group peaking equalizing processes are performed to achieve lower lattice peaking. Several fuel lattice design constraints are also incorporated in the automation system such that the system can accomplish a design which meets the requirements of practical use. Depending on the axial position of the lattice, a different method is applied in the design of the fuel lattice. Two typical fuel lattices with U{sup 235} enrichment of 4.471% and 4.386% were taken as references. Application of the method demonstrates that improved lattice designs can be achieved through the enrichment grouping and the enrichment group peaking equalizing method. It takes about 11 min and 1 h 11 min of CPU time for the automation system to accomplish two design cases on an HP-8000 workstation, including the execution of CASMO-4
Core Physics and Kinetics Calculations for the Fissioning Plasma Core Reactor
Butler, C.; Albright, D.
2007-01-01
Highly efficient, compact nuclear reactors would provide high specific impulse spacecraft propulsion. This analysis and numerical simulation effort has focused on the technical feasibility issues related to the nuclear design characteristics of a novel reactor design. The Fissioning Plasma Core Reactor (FPCR) is a shockwave-driven gaseous-core nuclear reactor, which uses Magneto Hydrodynamic effects to generate electric power to be used for propulsion. The nuclear design of the system depends on two major calculations: core physics calculations and kinetics calculations. Presently, core physics calculations have concentrated on the use of the MCNP4C code. However, initial results from other codes such as COMBINE/VENTURE and SCALE4a. are also shown. Several significant modifications were made to the ISR-developed QCALC1 kinetics analysis code. These modifications include testing the state of the core materials, an improvement to the calculation of the material properties of the core, the addition of an adiabatic core temperature model and improvement of the first order reactivity correction model. The accuracy of these modifications has been verified, and the accuracy of the point-core kinetics model used by the QCALC1 code has also been validated. Previously calculated kinetics results for the FPCR were described in the ISR report, "QCALC1: A code for FPCR Kinetics Model Feasibility Analysis" dated June 1, 2002.
Calculation of the axion mass based on high-temperature lattice quantum chromodynamics
Borsanyi, S.; Fodor, Z.; Guenther, J.; Kampert, K.-H.; Katz, S. D.; Kawanai, T.; Kovacs, T. G.; Mages, S. W.; Pasztor, A.; Pittler, F.; Redondo, J.; Ringwald, A.; Szabo, K. K.
2016-11-01
Unlike the electroweak sector of the standard model of particle physics, quantum chromodynamics (QCD) is surprisingly symmetric under time reversal. As there is no obvious reason for QCD being so symmetric, this phenomenon poses a theoretical problem, often referred to as the strong CP problem. The most attractive solution for this requires the existence of a new particle, the axion—a promising dark-matter candidate. Here we determine the axion mass using lattice QCD, assuming that these particles are the dominant component of dark matter. The key quantities of the calculation are the equation of state of the Universe and the temperature dependence of the topological susceptibility of QCD, a quantity that is notoriously difficult to calculate, especially in the most relevant high-temperature region (up to several gigaelectronvolts). But by splitting the vacuum into different sectors and re-defining the fermionic determinants, its controlled calculation becomes feasible. Thus, our twofold prediction helps most cosmological calculations to describe the evolution of the early Universe by using the equation of state, and may be decisive for guiding experiments looking for dark-matter axions. In the next couple of years, it should be possible to confirm or rule out post-inflation axions experimentally, depending on whether the axion mass is found to be as predicted here. Alternatively, in a pre-inflation scenario, our calculation determines the universal axionic angle that corresponds to the initial condition of our Universe.
DEVELOPMENT OF CALCULATION METHOD OF SENSITIVITIES FOR LIGHT WATER REACTORS
TOSHIKAZU TAKEDA
2013-11-01
Full Text Available A new method of calculating sensitivity coefficients of core characteristics relative to infinite-dilution cross sections has been developed. Conventional sensitivity coefficients are evaluated for the changes of effective cross sections which are dependent on individual models of core and cell. Therefore a correction has been derived to the conventional sensitivity coefficients based on the perturbation theory. The accuracy of the present method has been verified by comparing numerical results of sensitivity coefficients with a reference Monte-Carlo method.
Porta A.
2016-01-01
Full Text Available Beta decay of fission products is at the origin of decay heat and antineutrino emission in nuclear reactors. Decay heat represents about 7% of the reactor power during operation and strongly impacts reactor safety. Reactor antineutrino detection is used in several fundamental neutrino physics experiments and it can also be used for reactor monitoring and non-proliferation purposes. 92,93Rb are two fission products of importance in reactor antineutrino spectra and decay heat, but their β-decay properties are not well known. New measurements of 92,93Rb β-decay properties have been performed at the IGISOL facility (Jyväskylä, Finland using Total Absorption Spectroscopy (TAS. TAS is complementary to techniques based on Germanium detectors. It implies the use of a calorimeter to measure the total gamma intensity de-exciting each level in the daughter nucleus providing a direct measurement of the beta feeding. In these proceedings we present preliminary results for 93Rb, our measured beta feedings for 92Rb and we show the impact of these results on reactor antineutrino spectra and decay heat calculations.
Porta, A.; Zakari-Issoufou, A.-A.; Fallot, M.; Algora, A.; Tain, J. L.; Valencia, E.; Rice, S.; Bui, V. M.; Cormon, S.; Estienne, M.; Agramunt, J.; Äystö, J.; Bowry, M.; Briz, J. A.; Caballero-Folch, R.; Cano-Ott, D.; Cucouanes, A.; Elomaa, V.-V.; Eronen, T.; Estévez, E.; Farrelly, G. F.; Garcia, A. R.; Gelletly, W.; Gomez-Hornillos, M. B.; Gorlychev, V.; Hakala, J.; Jokinen, A.; Jordan, M. D.; Kankainen, A.; Karvonen, P.; Kolhinen, V. S.; Kondev, F. G.; Martinez, T.; Mendoza, E.; Molina, F.; Moore, I.; Perez-Cerdán, A. B.; Podolyák, Zs.; Penttilä, H.; Regan, P. H.; Reponen, M.; Rissanen, J.; Rubio, B.; Shiba, T.; Sonzogni, A. A.; Weber, C.
2016-03-01
Beta decay of fission products is at the origin of decay heat and antineutrino emission in nuclear reactors. Decay heat represents about 7% of the reactor power during operation and strongly impacts reactor safety. Reactor antineutrino detection is used in several fundamental neutrino physics experiments and it can also be used for reactor monitoring and non-proliferation purposes. 92,93Rb are two fission products of importance in reactor antineutrino spectra and decay heat, but their β-decay properties are not well known. New measurements of 92,93Rb β-decay properties have been performed at the IGISOL facility (Jyväskylä, Finland) using Total Absorption Spectroscopy (TAS). TAS is complementary to techniques based on Germanium detectors. It implies the use of a calorimeter to measure the total gamma intensity de-exciting each level in the daughter nucleus providing a direct measurement of the beta feeding. In these proceedings we present preliminary results for 93Rb, our measured beta feedings for 92Rb and we show the impact of these results on reactor antineutrino spectra and decay heat calculations.
Gonzalez, Alejandro; Milian, Daniel [Instituto Superior de Ciencias y Tecnologias Nucleares (ISCTN), La Habana (Cuba). E-mail: agg@ctn.isctn.edu.cu
2000-07-01
The task of the physical calculation of the reactor demand of the management of a great volume of information and inclose the stages for processing of data, calculations and analysis of their results. These stages are highly sensible to human mistakes, that's why is imprescindible that them undergo automatization, doing tracked all the process against mistake or unexpected result. The user-interface SYSMOD was developed over the platform IDE Delphi 3.0, visual language driven to events. It to consist in of the principal menu, which inclose between its options the preparation of the input data (File and Edit) to the pre-processors for the calculation codes of reactors. The output information may be showed in graphic and/or alphanumeric format (Data-Process). SYSMOD endures two applications for the management of the data base for the data during the preparation of the input for the pre-processors of the spectral calculation, so as for the organization, conservation and presentation for the obtained results. The carried out of the lattices and global codes, takes place from this application, over the platform MS-DOS (Run). SYSMOD regards the possibility for the debugging of the codes (Debugging), so as the benchmarks qualified to so effect (Benchmark). SYSMOD has been applied for the analysis of te WWER-440 of the first unity of Juragua Nuclear Power Plant. (author)
无
2001-01-01
To improve nuclear fuel utilization efficiency and prolong fuel cycle burn-up, a tight pitch lattice pressured heavy water reactor was investigated as an alternative of next generation of power reactors. It is shown that the high conversion ratio and negative coolant void reactivity coefficient are challenges in the reactor core physics designs. Various techniques were proposed to solve these problems. In this work, a tight pitch lattice and mixed fuel assemblies pressured heavy water reactor concept was investigated. By utilizing numerical simulation technique, it is demonstrated that reactor core mixed with Pu/U and Th/U assemblies can achieve high conversion ratio (0.98), long burn-up (60 GWD/t) and negative void reactivity coefficients.
Jingyu Zhang
2016-01-01
Full Text Available In water-cooled reactor, the dominant radioactive source term under normal operation is activated corrosion products (ACPs, which have an important impact on reactor inspection and maintenance. A three-node transport model of ACPs was introduced into the new version of ACPs source term code CATE in this paper, which makes CATE capable of theoretically simulating the variation and the distribution of ACPs in a water-cooled reactor and suitable for more operating conditions. For code testing, MIT PWR coolant chemistry loop was simulated, and the calculation results from CATE are close to the experimental results from MIT, which means CATE is available and credible on ACPs analysis of water-cooled reactor. Then ACPs in the blanket cooling loop of water-cooled fusion reactor ITER under construction were analyzed using CATE and the results showed that the major contributors are the short-life nuclides, especially Mn-56. At last a point kernel integration code ARShield was coupled with CATE, and the dose rate around ITER blanket cooling loop was calculated. Results showed that after shutting down the reactor only for 8 days, the dose rate decreased nearly one order of magnitude, which was caused by the rapid decay of the short-life ACPs.
Measured and calculated effective delayed neutron fraction of the IPR-R1 Triga reactor
Souza, Rose Mary G.P.; Dalle, Hugo M.; Campolina, Daniel A.M., E-mail: souzarm@cdtn.b, E-mail: dallehm@cdtn.b, E-mail: campolina@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)
2011-07-01
The effective delayed neutron fraction, {beta}{sub eff}, one of the most important parameter in reactor kinetics, was measured for the 100 kW IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil. The current reactor core has 63 fuel elements, containing about 8.5% and 8% by weight of uranium enriched to 20% in U{sup 235}. The core has cylindrical configuration with an annular graphite reflector. Since the first criticality of the reactor in November 1960, the core configuration and the number of fuel elements have been changed several times. At that time, the reactor power was 30 kW, there were 56 fuel elements in the core, and the {beta}{sub eff} value for the reactor recommended by General Atomic (manufacturer of TRIGA) was 790 pcm. The current {beta}{sub eff} parameter was determined from experimental methods based on inhour equation and on the control rod drops. The estimated values obtained were (774 {+-} 38) pcm and (744 {+-} 20) pcm, respectively. The {beta}{sub eff} was calculated by Monte Carlo transport code MCNP5 and it was obtained 747 pcm. The calculated and measured values are in good agreement, and the relative percentage error is -3.6% for the first case, and 0.4% for the second one. (author)
Quantitative calculation of reaction performance in sonochemical reactor by bubble dynamics
Xu, Zheng; Yasuda, Keiji; Liu, Xiao-Jun
2015-10-01
In order to design a sonochemical reactor with high reaction efficiency, it is important to clarify the size and intensity of the sonochemical reaction field. In this study, the reaction field in a sonochemical reactor is estimated from the distribution of pressure above the threshold for cavitation. The quantitation of hydroxide radical in a sonochemical reactor is obtained from the calculation of bubble dynamics and reaction equations. The distribution of the reaction field of the numerical simulation is consistent with that of the sonochemical luminescence. The sound absorption coefficient of liquid in the sonochemical reactor is much larger than that attributed to classical contributions which are heat conduction and shear viscosity. Under the dual irradiation, the reaction field becomes extensive and intensive because the acoustic pressure amplitude is intensified by the interference of two ultrasonic waves. Project supported by the National Natural Science Foundation of China (Grant Nos. 11404245, 11204129, and 11211140039).
A Mechanistic Source Term Calculation for a Metal Fuel Sodium Fast Reactor
Grabaskas, David; Bucknor, Matthew; Jerden, James
2017-06-26
A mechanistic source term (MST) calculation attempts to realistically assess the transport and release of radionuclides from a reactor system to the environment during a specific accident sequence. The U.S. Nuclear Regulatory Commission (NRC) has repeatedly stated its expectation that advanced reactor vendors will utilize an MST during the U.S. reactor licensing process. As part of a project to examine possible impediments to sodium fast reactor (SFR) licensing in the U.S., an analysis was conducted regarding the current capabilities to perform an MST for a metal fuel SFR. The purpose of the project was to identify and prioritize any gaps in current computational tools, and the associated database, for the accurate assessment of an MST. The results of the study demonstrate that an SFR MST is possible with current tools and data, but several gaps exist that may lead to possibly unacceptable levels of uncertainty, depending on the goals of the MST analysis.
Quantitative calculation of reaction performance in sonochemical reactor by bubble dynamics
徐峥; 安田启司; 刘晓峻
2015-01-01
In order to design a sonochemical reactor with high reaction efficiency, it is important to clarify the size and intensity of the sonochemical reaction field. In this study, the reaction field in a sonochemical reactor is estimated from the distribution of pressure above the threshold for cavitation. The quantitation of hydroxide radical in a sonochemical reactor is obtained from the calculation of bubble dynamics and reaction equations. The distribution of the reaction field of the numerical simulation is consistent with that of the sonochemical luminescence. The sound absorption coefficient of liquid in the sonochemical reactor is much larger than that attributed to classical contributions which are heat conduction and shear viscosity. Under the dual irradiation, the reaction field becomes extensive and intensive because the acoustic pressure amplitude is intensified by the interference of two ultrasonic waves.
Zagar, Tomaz [Reactor Physics Department, Jozef Stefan Institute, Jamova 39, 1000 Ljubljana (Slovenia)]. E-mail: tomaz.zagar@ijs.si; Bozic, Matjaz [Nuklearna elektrarna Krsko, Vrbina 12, 8270 Krsko (Slovenia); Ravnik, Matjaz [Reactor Physics Department, Jozef Stefan Institute, Jamova 39, 1000 Ljubljana (Slovenia)
2004-12-01
In this paper, a process of long-lived activity determination in research reactor concrete shielding is presented. The described process is a combination of experiment and calculations. Samples of original heavy reactor concrete containing mineral barite were irradiated inside the reactor shielding to measure its long-lived induced radioactivity. The most active long-lived ({gamma} emitting) radioactive nuclides in the concrete were found to be {sup 133}Ba, {sup 60}Co and {sup 152}Eu. Neutron flux, activation rates and concrete activity were calculated for actual shield geometry for different irradiation and cooling times using TORT and ORIGEN codes. Experimental results of flux and activity measurements showed good agreement with the results of calculations. Volume of activated concrete waste after reactor decommissioning was estimated for particular case of Jozef Stefan Institute TRIGA reactor. It was observed that the clearance levels of some important long-lived isotopes typical for barite concrete (e.g. {sup 133}Ba, {sup 41}Ca) are not included in the IAEA and EU basic safety standards.
Neutronics calculations for the Oak Ridge National Laboratory Tokamak Reactor Studies
Santoro, R.T.; Baker, V.C.; Barnes, J.M.
1976-01-01
Neutronics calculations have been carried out to analyze the nuclear performance of conceptual blanket and shield designs for the Tokamak Experimental Power Reactor (EPR) and the Tokamak Demonstration Reactor Plant (DRP) being considered at the Oak Ridge National Laboratory. These reactor designs represent a sequence in the commercialization of fusion-generated electrical power. All of the calculations were carried out using the one-dimensional discrete ordinates code ANISN and the latest available ENDF/B-IV coupled neutron-gamma-ray transport cross-section data, fluence-to-kerma conversion factors, and radiation damage cross-section data. The calculations include spatial and integral heating-rate estimates in the reactor with emphasis on the recovery of fusion neutron energy in the blanket and limiting the heat-deposition rate in the superconducting toroidal field coils. Radiation damage due to atomic displacements and gas production produced in the reactor structural material and in the toroidal field coil windings were also estimated. The tritium-breeding ratio when natural lithium is used as the fertile material in the DRP blanket and in the experimental breeding modules in the EPR is also given.
Yoon, B; Engelhardt, M; Green, J; Gupta, R; Hägler, P; Musch, B; Negele, J; Pochinsky, A; Syritsyn, S
2016-01-01
We present a lattice QCD calculation of transverse momentum dependent parton distribution functions (TMDs) of protons using staple-shaped Wilson lines. For time-reversal odd observables, we calculate the generalized Sivers and Boer-Mulders transverse momentum shifts in SIDIS and DY cases, and for T-even observables we calculate the transversity related to the tensor charge and the generalized worm-gear shift. The calculation is done on two different n_f=2+1 ensembles: domain-wall fermion (DWF) with lattice spacing 0.084 fm and pion mass of 297 MeV, and clover fermion with lattice spacing 0.114 fm and pion mass of 317 MeV. The results from those two different discretizations are consistent with each other.
A direct hybrid S{sub N} method for slab-geometry lattice calculations
Silva, Davi J.M.; Barros, Ricardo C., E-mail: rcbarros@pq.cnpq.b [Universidade do Estado do Rio de Janeiro (IPRJ/UERJ), Nova Friburgo, RJ (Brazil). Programa de Pos-graduacao em Modelagem Computacional; Zani, Jose H. [Fundacao Educacional Serra dos Orgaos, Teresopolis, RJ (Brazil). Ciencia da Computacao
2011-07-01
In this work we describe a hybrid direct method for calculating the thermal disadvantage factor and the neutron flux distribution in fuel-moderator lattices. For the mathematical model, we use the one-speed slab-geometry discrete ordinates (S{sub N}) transport equation with linearly anisotropic scattering. The basic idea is to use higher order angular quadrature set in the highly absorbing fuel region (S{sub NF}) and lower order angular quadrature set in the diffusive moderator region (S{sub NM}) , i.e., N{sub F} > N{sub M}. We apply special continuity conditions based on the equivalence of the S{sub N} and P{sub N-1} equations, which characterize the hybrid model. Numerical results to a typical model problem are given to illustrate the accuracy and the efficiency of the offered hybrid method. (author)
New real-space renormalization-group calculation for the critical properties of lattice spin systems
Hecht, Charles E.; Kikuchi, Ryoichi
1982-05-01
In evaluating the critical properties of lattice spin systems in the real-space renormalization-group theory we use the cluster variation method. A configuration in the transformed system is constrained and the probability of occurrence of this configuration is calculated both in the transformed system and in the original system. By equating the two probabilities and forming ratios of two such equalities (for two or more constrained configurations) the fixed point of the renormalization transformation is evaluated. The method can avoid the trouble due to different singularities in the original and transformed systems, and hence can obviate the possible development of spurious singularities in the transformation at low temperatures. The two-dimensional triangular Ising model is treated with numerical results comparable with those obtained by the cluster treatment of Niemeijer and van Leeuwen who used more and larger cluster types than those we introduce.
Lattice dynamics calculations for ferropericlase with internally consistent LDA+U method
Fukui, Hiroshi; Tsuchiya, Taku; Baron, Alfred Q. R.
2012-12-01
Vibrational densities of states and phonon dispersion relations for Mg0.875Fe0.125O ferropericlase in the high- and low-spin (HS and LS) states were calculated from first principles lattice dynamics using the internally consistent LDA+Utechnique. Finite-temperature thermodynamic properties were determined based on the quasi-harmonic approximation including the HS and LS mixing entropy and the magnetic entropy effects, which gave pressure and temperature variations of the low-spin fraction. Our results suggest that for thermodynamic modeling of the earth's interior, the effect of the mixed spin state cannot be ignored in the lower mantle, especially the lowermost part. The anomaly in the seismic wave velocity due to the spin crossover transition of ferropericlase, if it exists, is difficult to detect because of the wide pressure range of the transition, which is broadened by the temperature effect and the damping of the amplitude of the slow seismic wave.
Neutronic calculation of fast reactors by the EUCLID/V1 integrated code
Koltashev, D. A.; Stakhanova, A. A.
2017-01-01
This article considers neutronic calculation of a fast-neutron lead-cooled reactor BREST-OD-300 by the EUCLID/V1 integrated code. The main goal of development and application of integrated codes is a nuclear power plant safety justification. EUCLID/V1 is integrated code designed for coupled neutronics, thermomechanical and thermohydraulic fast reactor calculations under normal and abnormal operating conditions. EUCLID/V1 code is being developed in the Nuclear Safety Institute of the Russian Academy of Sciences. The integrated code has a modular structure and consists of three main modules: thermohydraulic module HYDRA-IBRAE/LM/V1, thermomechanical module BERKUT and neutronic module DN3D. In addition, the integrated code includes databases with fuel, coolant and structural materials properties. Neutronic module DN3D provides full-scale simulation of neutronic processes in fast reactors. Heat sources distribution, control rods movement, reactivity level changes and other processes can be simulated. Neutron transport equation in multigroup diffusion approximation is solved. This paper contains some calculations implemented as a part of EUCLID/V1 code validation. A fast-neutron lead-cooled reactor BREST-OD-300 transient simulation (fuel assembly floating, decompression of passive feedback system channel) and cross-validation with MCU-FR code results are presented in this paper. The calculations demonstrate EUCLID/V1 code application for BREST-OD-300 simulating and safety justification.
Mohammadi, A; Hassanzadeh, M; Gharib, M
2016-02-01
In this study, shielding calculation and criticality safety analysis were carried out for general material testing reactor (MTR) research reactors interim storage and relevant transportation cask. During these processes, three major terms were considered: source term, shielding, and criticality calculations. The Monte Carlo transport code MCNP5 was used for shielding calculation and criticality safety analysis and ORIGEN2.1 code for source term calculation. According to the results obtained, a cylindrical cask with body, top, and bottom thicknesses of 18, 13, and 13 cm, respectively, was accepted as the dual-purpose cask. Furthermore, it is shown that the total dose rates are below the normal transport criteria that meet the standards specified.
Parametric HECTR calculations of hydrogen transport and combustion at N Reactor
Payne, A.C. Jr.; Camp, A.L.
1987-06-01
This report describes a limited number of parametric calculations of hydrogen transport and combustion in the N Reactor confinement for selected accident sequences. The calculations are performed using the HECTR computer code, which is a lumped-parameter code developed specifically for evaluating hydrogen behavior in reactor containments. A number of parameters are evaluated in this study, including hydrogen source rate, spray effects, and source location. The calculations indicate that mixing within major compartments tends to occur fairly rapidly, but that mixing between compartments can be inhibited in certain situations, resulting in the formation of flammable mixtures. These results are being compared to calculations performed with other computer codes, including a code that uses finite-difference models. United Nuclear Corporation will present the results of these code comparisons in future reports.
Post test calculations of a severe accident experiment for VVER-440 reactors by the ATHLET code
Gyoergy, Hunor [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques (BME NTI); Trosztel, Istvan [Hungarian Academy of Sciences, Budapest (Hungary). Centre for Energy Research (MTA EK)
2013-09-15
Severe accident - if no mitigation action is taken - leads to core melt. An effective severe accident management strategy can be the external reactor pressure vessel cooling for corium localization and stabilization. For some time discussion was going on, whether the in-vessel retention can be applied for the VVER-440 type reactors. It had to be demonstrated that the available space between the reactor vessel and biological protection allows sufficient cooling to keep the melted core in the vessel, without the reactor pressure vessel losing its integrity. In order to demonstrate the feasibility of the concept an experimental facility was realized in Hungary. The facility called Cooling Effectiveness on the Reactor External Surface (CERES) is modeling the vessel external surface and the biological protection of Paks NPP. A model of the CERES facility for the ATHLET TH system code was developed. The results of the ATHLET calculation agree well with the measurements showing that the vessel cooling can be insured for a long time in a VVER-440 reactor. (orig.)
Sensitivity of Candu-Scwr Reactor Physics Calculations to Nuclear Data Files
Kozier, K. S.; Dyck, G. R.
2006-04-01
A simplified MCNP model of a CANDU-SCWR lattice was used to test the sensitivity of the calculated reactivity to various nuclear data files involving issues of current interest. These tests were performed for cooled and voided conditions, with and without axial neutron leakage, for a uniform lattice of mid-life fuel and a mixed lattice with high-burnup and low-burnup fuel in alternate channels. Results were compared using different room-temperature data files for deuterium, various thermal-scattering-law data files for hydrogen bound in light water and deuterium bound in heavy water, and for pre-ENDF/B-VII and ENDF/B-VI.8 data for uranium. The reactivity differences observed were small (typically <1 mk) and increased with axial neutron leakage.
Patel, Kinjal D.; Patel, Urmila H.
2017-01-01
Sulfamonomethoxine, 4-Amino-N-(6-methoxy-4-pyrimidinyl) benzenesulfonamide (C11H12N4O3S), is investigated by single crystal X-ray diffraction technique. Pair of N-H⋯N and C-H⋯O intermolecular interactions along with π···π interaction are responsible for the stability of the molecular packing of the structure. In order to understand the nature of the interactions and their quantitative contributions towards the crystal packing, the 3D Hirshfeld surface and 2D fingerprint plot analysis are carried out. PIXEL calculations are performed to determine the lattice energies correspond to intermolecular interactions in the crystal structure. Ab initio quantum chemical calculations of sulfamonomethoxine (SMM) have been performed by B3LYP method, using 6-31G** basis set with the help of Schrodinger software. The computed geometrical parameters are in good agreement with the experimental data. The Mulliken charge distribution, calculated using B3LYP method to confirm the presence of electron acceptor and electron donor atoms, responsible for intermolecular hydrogen bond interactions hence the molecular stability.
Jacobsen, C.J.H.; Dahl, Søren; Boisen, A.
2002-01-01
For ammonia synthesis catalysts a volcano-type relationship has been found experimentally. We demonstrate that by combining density functional theory calculations with a microkinetic model the position of the maximum of the volcano curve is sensitive to the reaction conditions. The catalytic ammo......-principle quantum mechanical calculations of gas-surface interactions, reactor design, and catalyst selection has been established for the first time....
Calculated power distribution of a thermionic, beryllium oxide reflected, fast-spectrum reactor
Mayo, W.; Lantz, E.
1973-01-01
A procedure is developed and used to calculate the detailed power distribution in the fuel elements next to a beryllium oxide reflector of a fast-spectrum, thermionic reactor. The results of the calculations show that, although the average power density in these outer fuel elements is not far from the core average, the power density at the very edge of the fuel closest to the beryllium oxide is about 1.8 times the core avearge.
Adjoint-based uncertainty quantification and sensitivity analysis for reactor depletion calculations
Stripling, Hayes Franklin
Depletion calculations for nuclear reactors model the dynamic coupling between the material composition and neutron flux and help predict reactor performance and safety characteristics. In order to be trusted as reliable predictive tools and inputs to licensing and operational decisions, the simulations must include an accurate and holistic quantification of errors and uncertainties in its outputs. Uncertainty quantification is a formidable challenge in large, realistic reactor models because of the large number of unknowns and myriad sources of uncertainty and error. We present a framework for performing efficient uncertainty quantification in depletion problems using an adjoint approach, with emphasis on high-fidelity calculations using advanced massively parallel computing architectures. This approach calls for a solution to two systems of equations: (a) the forward, engineering system that models the reactor, and (b) the adjoint system, which is mathematically related to but different from the forward system. We use the solutions of these systems to produce sensitivity and error estimates at a cost that does not grow rapidly with the number of uncertain inputs. We present the framework in a general fashion and apply it to both the source-driven and k-eigenvalue forms of the depletion equations. We describe the implementation and verification of solvers for the forward and ad- joint equations in the PDT code, and we test the algorithms on realistic reactor analysis problems. We demonstrate a new approach for reducing the memory and I/O demands on the host machine, which can be overwhelming for typical adjoint algorithms. Our conclusion is that adjoint depletion calculations using full transport solutions are not only computationally tractable, they are the most attractive option for performing uncertainty quantification on high-fidelity reactor analysis problems.
1975-10-01
Information is presented concerning the radioactive releases from the containment following accidents; radioactive inventory of the reactor core; atmospheric dispersion; reactor sites and meteorological data; radioactive decay and deposition from plumes; finite distance of plume travel; dosimetric models; health effects; demographic data; mitigation of radiation exposure; economic model; and calculated results with consequence model.
Improving the Volume Dependence of Two-Body Binding Energies Calculated with Lattice QCD
Davoudi, Zohreh
2011-01-01
Volume modifications to the binding of two-body systems in large cubic volumes of extent L depend upon the total momentum and exponentially upon the ratio of L to the size of the boosted system. Recent work by Bour et al determined the momentum dependence of the leading volume modifications to nonrelativistic systems with periodic boundary conditions imposed on the single-particle wavefunctions, enabling them to numerically determine the scattering of such bound states using a low-energy effective field theory and Luschers finite-volume method. The calculation of bound nuclear systems directly from QCD using Lattice QCD has begun, and it is important to reduce the systematic uncertainty introduced into such calculations by the finite spatial extent of the gauge-field configurations. We extend the work of Bour et al from nonrelativistic quantum mechanics to quantum field theory by generalizing the work of Luscher and of Gottlieb and Rummukainen to boosted two-body bound states. The volume modifications to bind...
Preliminary safety calculations to improve the design of Molten Salt Fast Reactor
Brovchenko, M.; Heuer, D.; Merle-Lucotte, E.; Allibert, M.; Capellan, N.; Ghetta, V.; Laureau, A. [LPSC, CNRS/IN2P3, Grenoble INP, 53,rue des Martyrs, 38026 Grenoble Cedex (France)
2012-07-01
Molten salt reactors are liquid fuel reactors so that they are flexible in operation but very different in the safety approach from solid fuel reactors. This study bears on the specific concept named Molten Salt Fast Reactor (MSFR). Since this new nuclear technology is in development, safety is an essential point to be considered all along the R and D studies. This paper presents the first step of the safety approach: the systematic description of the MSFR, limited here to the main systems surrounding the core. This systematic description is the basis on which we will be able to devise accidental scenarios. Thanks to the negative reactivity feedback coefficient, most accidental scenarios lead to reactor shut down. Because of the decay heat generated in the fuel salt, it must be cooled. After the description of the tools developed to calculate the residual heat, the different contributions are discussed in this study. The decay heat of fission products in the MSFR is evaluated to be low (3% of nominal power), mainly due to the reprocessing that transfers the fission products to the gas reprocessing unit. As a result, the contribution of the actinides is significant (0.5% of nominal power). The unprotected loss of heat sink transients are studied in this paper. It appears that slow transients are favorable (> 1 min) to minimize the temperature increase of the fuel salt. This work will be the basis of further safety studies as well as an essential parameter for the design of the draining system. (authors)
Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR
Brovchenko Mariya
2017-01-01
Full Text Available The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR. The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.
Reference neutron transport calculation note for Korea nuclear power plants with 3-loop PWR reactors
Kim, Byung Cheol; Chang, Ki Oak
1997-05-01
Reactor pressure vessel (RPV) steels are subjected to neutron irradiation at a temperature of about 290 deg C. This radiation exposure alters the mechanical properties, leading to a shift of the brittle-to-ductile transition temperature toward higher temperatures and to a diminution of the rupture energy as determined by Charpy V-notch tests. This radiation embrittlement is one of the important aging factors of nuclear power plants. U.S. NRC recommended the basic requirements for the determination of the pressure vessel fluence by regulatory guide DG-1025 in order to reduce the uncertainty in the determination of neutron fluence calculation and measurements. The determination of the pressure vessel fluence is based on both calculations and measurements. The fluence prediction is made with a calculation and the measurements are used to qualify the calculational methodology. Because of the importance and the difficulty of these calculations, the method`s qualification by comparison to measurement must be made to ensure a reliable and accurate vessel fluence determination. This reference calculation note is to provide a series of forward and adjoint neutron transport calculations for use in the evaluation of neutron dosimetry from surveillance capsule irradiations at 3-loop PWR reactor as well as for use in the determination of the neutron exposure of the reactor vessel wall in accordance with U.S Regulatory Guide DG-1025 requirements. The calculations of the pressure vessel fluence consist of the following steps; (1) Determination of the geometrical and material input data, (2) Determination of the core neutron source, and (3) Propagation of the neutron fluence from the core to the vessel and into the cavity. (author). 12 tabs., 3 figs., 7 refs.
Calculation of the reactor neutron time of flight spectrum by convolution technique
Cheng Jin-Xing; Ouyang Xiao-Ping; Zheng Yi; Zhang An-Hui; Ouyang Mao-Jie
2008-01-01
It is a very complex and tlme-consuming process to simulate the nuclear reactor neutron spectrum from the reactor core to the export channel by applying a Monte Carlo program. This paper presents a new method to calculate the neutron spectrum by using the convolution technique which considers the channel transportation as a linear system and the transportation scattering as the response function. It also applies Monte Carlo Neutron and Photon Transport Code (MCNP) to simulate the response function numerically. With the application of convolution technique to calculate thespectrum distribution from the core to the channel, the process is then much more convenient only with the simple numerical integral numeration. This saves computer time and reduces some trouble in re-writing of the MCNP program.
Heat Transfer Calculation on Plate-Type Fuel Assembly of High Flux Research Reactor
Daxin Gong
2015-01-01
Full Text Available Heat transfer characteristics of fuel assemblies for a high flux research reactor with a neutron trap are numerically investigated in this study. Single-phase turbulence flow is calculated by a commercial code, FLUENT, where the computational objective covers standard and control fuel assemblies. The simulation is carried out with an inlet coolant velocity varying from 4.5 m/s to 7.5 m/s in hot assemblies. The results indicate that the cladding temperature is always lower than the saturation temperature in the calculated ranges. The temperature rise in the control fuel assembly is smaller than that of the standard fuel assembly. Additionally, the assembly with a hot spot is specially studied, and the safety of the research reactor is also approved.
Tarchalski, M.; Pytel, K.; Wroblewska, M.; Marcinkowska, Z.; Boettcher, A.; Prokopowicz, R. [NCBJ Institute, MARIA Reactor, ul.Andrzeja Soltana 7, 05-400 Swierk (Poland); Sireta, P.; Gonnier, C.; Bignan, G. [CEA, DEN, Reactor Studies Department, Cadarache, F-13108 St-Paul-Lez-Durance (France); Lyoussi, A.; Fourmentel, D.; Barbot, L.; Villard, J.F.; Destouches, C. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France); Reynard-Carette, C.; Brun, J. [Aix Marseille Universite, CNRS, Universite de Toulon, IM2NP UMR 7334, 13397, Marseille (France); Jagielski, J. [NCBJ Institute, MARIA Reactor, ul.Andrzeja Soltana 7, 05-400 Swierk (Poland); Institute of Electronic Materials Technolgy, Wolczynska 133, 01-919 Warszawa (Poland); Luks, A. [Institute of Heat Engineering, Nowowiejska 21/25, 00-665 Warsaw (Poland)
2015-07-01
Precise computational determination of nuclear heating which consists predominantly of gamma heating (more than 80 %) is one of the challenges in material testing reactor exploitation. Due to sophisticated construction and conditions of experimental programs planned in JHR it became essential to use most accurate and precise gamma heating model. Before the JHR starts to operate, gamma heating evaluation methods need to be developed and qualified in other experimental reactor facilities. This is done inter alia using OSIRIS, MINERVE or EOLE research reactors in France. Furthermore, MARIA - Polish material testing reactor - has been chosen to contribute to the qualification of gamma heating calculation schemes/tools. This reactor has some characteristics close to those of JHR (beryllium usage, fuel element geometry). To evaluate gamma heating in JHR and MARIA reactors, both simulation tools and experimental program have been developed and performed. For gamma heating simulation, new calculation scheme and gamma heating model of MARIA have been carried out using TRIPOLI4 and APOLLO2 codes. Calculation outcome has been verified by comparison to experimental measurements in MARIA reactor. To have more precise calculation results, model of MARIA in TRIPOLI4 has been made using the whole geometry of the core. This has been done for the first time in the history of MARIA reactor and was complex due to cut cone shape of all its elements. Material composition of burnt fuel elements has been implemented from APOLLO2 calculations. An experiment for nuclear heating measurements and calculation verification has been done in September 2014. This involved neutron, photon and nuclear heating measurements at selected locations in MARIA reactor using in particular Rh SPND, Ag SPND, Ionization Chamber (all three from CEA), KAROLINA calorimeter (NCBJ) and Gamma Thermometer (CEA/SCK CEN). Measurements were done in forty points using four channels. Maximal nuclear heating evaluated from
Chen, Peng-Jen; Jeng, Horng-Tay
2016-03-16
A new semiconducting phase of two-dimensional phosphorous in the Kagome lattice is proposed from first-principles calculations. The band gaps of the monolayer (ML) and bulk Kagome phosphorous (Kagome-P) are 2.00 and 1.11 eV, respectively. The magnitude of the band gap is tunable by applying the in-plane strain and/or changing the number of stacking layers. High optical absorption coefficients at the visible light region are predicted for multilayer Kagome-P, indicating potential applications for solar cell devices. The nearly dispersionless top valence band of the ML Kagome-P with high density of states at the Fermi level leads to superconductivity with Tc of ~9 K under the optimal hole doping concentration. We also propose that the Kagome-P can be fabricated through the manipulation of the substrate-induced strain during the process of the sample growth. Our work demonstrates the high applicability of the Kagome-P in the fields of electronics, photovoltaics, and superconductivity.
Lattice QCD calculation of form factors for $\\Lambda_b \\to \\Lambda(1520) \\ell^+ \\ell^-$ decays
Meinel, Stefan
2016-01-01
Experimental results for mesonic $b \\to s \\mu^+ \\mu^-$ decays show a pattern of deviations from Standard-Model predictions, which could be due to new fundamental physics or due to an insufficient understanding of hadronic effects. Additional information on the $b \\to s \\mu^+ \\mu^-$ transition can be obtained from $\\Lambda_b$ decays. This was recently done using the process $\\Lambda_b \\to \\Lambda \\mu^+ \\mu^-$, where the $\\Lambda$ is the lightest strange baryon. A further interesting channel is $\\Lambda_b \\to p^+ K^- \\mu^+ \\mu^-$, where the $p^+ K^-$ final state receives contributions from multiple higher-mass $\\Lambda$ resonances. The narrowest and most prominent of these is the $\\Lambda(1520)$, which has $J^P=\\frac32^-$. Here we present an ongoing lattice QCD calculation of the relevant $\\Lambda_b \\to \\Lambda(1520)$ form factors. We discuss the choice of interpolating field for the $\\Lambda(1520)$, and explain our method for extracting the fourteen $\\Lambda_b \\to \\Lambda(1520)$ helicity form factors from corr...
Osuský, F.; Bahdanovich, R.; Farkas, G.; Haščík, J.; Tikhomirov, G. V.
2017-01-01
The paper is focused on development of the coupled neutronics-thermal hydraulics model for the Gas-cooled Fast Reactor. It is necessary to carefully investigate coupled calculations of new concepts to avoid recriticality scenarios, as it is not possible to ensure sub-critical state for a fast reactor core under core disruptive accident conditions. Above mentioned calculations are also very suitable for development of new passive or inherent safety systems that can mitigate the occurrence of the recriticality scenarios. In the paper, the most promising fuel material compositions together with a geometry model are described for the Gas-cooled fast reactor. Seven fuel pin and fuel assembly geometry is proposed as a test case for coupled calculation with three different enrichments of fissile material in the form of Pu-UC. The reflective boundary condition is used in radial directions of the test case and vacuum boundary condition is used in axial directions. During these condition, the nuclear system is in super-critical state and to achieve a stable state (which is numerical representation of operational conditions) it is necessary to decrease the reactivity of the system. The iteration scheme is proposed, where SCALE code system is used for collapsing of a macroscopic cross-section into few group representation as input for coupled code NESTLE.
Allès, B; Di Giacomo, Adriano; Pica, C
2006-01-01
A Ginsparg-Wilson based calibration of the topological charge is used to calculate the renormalization constants which appear in the field-theoretical determination of the topological susceptibility on the lattice. A systematic comparison is made with calculations based on cooling. The two methods agree within present statistical errors (3%-4%). We also discuss the independence of the multiplicative renormalization constant Z from the background topological charge used to determine it.
Methods for calculating group cross sections for doubly heterogeneous thermal reactor systems. [HTGR
Stamatelatos, M G; LaBauve, R J
1977-02-01
The report discusses methods used at LASL for calculating group cross sections for doubly heterogeneous HTGR systems of the General Atomic design. These cross sections have been used for the neutronic safety analysis calculations of such HTGR systems at various points in reactor lifetime (e.g., beginning-of-life, end-of-equilibrium cycle). They were also compared with supplied General Atomic cross sections generated with General Atomic codes. The overall agreement between the LASL and the GA cross sections has been satisfactory.
Axial power distribution calculation using a neural network in the nuclear reactor core
Kim, Y. H.; Cha, K. H.; Lee, S. H. [Korea Electric Power Research Institute, Taejon (Korea, Republic of)
1997-12-31
This paper is concerned with an algorithm based on neural networks to calculate the axial power distribution using excore detector signals in the nuclear reactor core. The fundamental basis of the algorithm is that the detector response can be fairly accurately estimated using computational codes. In other words, the training set, which represents relationship between detector signals and axial power distributions, for the neural network can be obtained through calculations instead of measurements. Application of the new method to the Yonggwang nuclear power plant unit 3 (YGN-3) shows that it is superior to the current algorithm in place. 7 refs., 4 figs. (Author)
Nevinitsa, V. A.; Dudnikov, A. A.; Blandinskiy, V. Yu.; Balanin, A. L.; Alekseev, P. N.; Titarenko, Yu. E.; Batyaev, V. F.; Pavlov, K. V.; Titarenko, A. Yu.
2015-12-01
A subcritical molten salt reactor with an external neutron source is studied computationally as a facility for incineration and transmutation of minor actinides from spent nuclear fuel of reactors of VVER-1000 type and for producing 233U from 232Th. The reactor configuration is chosen, the requirements to be imposed on the external neutron source are formulated, and the equilibrium isotopic composition of heavy nuclides and the key parameters of the fuel cycle are calculated.
Nevinitsa, V. A., E-mail: Neviniza-VA@nrcki.ru; Dudnikov, A. A.; Blandinskiy, V. Yu.; Balanin, A. L.; Alekseev, P. N. [National Research Centre Kurchatov Institute (Russian Federation); Titarenko, Yu. E.; Batyaev, V. F.; Pavlov, K. V.; Titarenko, A. Yu., E-mail: yuri.titarenko@itep.ru [Institute for Theoretical and Experimental Physics (Russian Federation)
2015-12-15
A subcritical molten salt reactor with an external neutron source is studied computationally as a facility for incineration and transmutation of minor actinides from spent nuclear fuel of reactors of VVER-1000 type and for producing {sup 233}U from {sup 232}Th. The reactor configuration is chosen, the requirements to be imposed on the external neutron source are formulated, and the equilibrium isotopic composition of heavy nuclides and the key parameters of the fuel cycle are calculated.
Nuclide Inventory Calculation Using MCNPX for Wolsung Unit 1 Reactor Decommissioning
Rabir, Mohamad Hairie; Noh, Kyoung Ho; Hah, Chang Joo [KEPCO International Nuclear Graduate School, Daejeon (Korea, Republic of)
2014-05-15
The CINDER90 computation process involves utilizing linear Markovian chains to determine the time dependent nuclide densities. The CINDER90 depletion algorithm is implemented the MCNPX code package. The coupled depletion process involves a Monte-Carlo steady-state reaction rate calculation linked to a deterministic depletion calculation. The process is shown in Fig.1. MCNPX runs a steady state calculation to determine the system eigenvalue collision densities, recoverable energies from fission and neutrons per fission events. In order to generate number densities for the next time step, the CINDER90 code takes the MCNPX generated values and performs a depletion calculation. MCNPX then takes the new number densities and caries out a new steady-stated calculation. The process repeats itself until the final time step. This paper describe the preliminary source term and nuclide inventory calculation of Candu single fuel channel using MCNPX, as a part of the activities to support the equilibrium core model development and decommissioning evaluation process of a Candu reactor. The aim of this study was to apply the MCNPX code for source term and nuclide inventory calculation of Candu single fuel channel. Nuclide inventories as a function of burnup will be used to model an equilibrium core for Candu reactor. The core lifetime neutron fluence obtained from the model is used to estimate radioactivity at the stage of decommisioning. In general, as expected, the actinides and fission products build up increase with increasing burnup. Despite the fact that the MCNPX code is still in development we can conclude that the code is capable of obtaining relevant results in burnup and source term calculation. It is recommended that in the future work, the calculation has to be verified on the basis of experimental data or comparison with other codes.
Ao, B.Y., E-mail: aobingyun24@yahoo.com.cn [Science and Technology on Surface Physics and Chemistry Laboratory, P.O. Box 718-35, Mianyang 621907 (China); Wang, X.L.; Shi, P.; Chen, P.H.; Ye, X.Q.; Lai, X.C. [Science and Technology on Surface Physics and Chemistry Laboratory, P.O. Box 718-35, Mianyang 621907 (China); Gao, T., E-mail: gaotao@scu.edu.cn [Institute of Atomic and Molecular Physics, Sichuan University, Chengdu 610065 (China)
2012-05-15
Plutonium metal can be loaded with hydrogen, which forms complicated solid solutions and compounds, and leads to significant changes in electronic structure. A first-principles pseudopotential plane wave method with added Hubbard parameter U was employed to investigate the electronic and structural properties of face-centered cubic Pu hydrides (PuH{sub x}, x = 2, 2.25, and 3). The decrease in calculated lattice parameters with increasing x is in reasonable agreement with experimental findings. Comparative analysis of the electronic-structure results for a series of PuH{sub x} compositions reveals that lattice contraction occurs due to enhanced chemical bonding and the size effects involving interstitial atoms. We find that the size effects are the driving force for the abnormal lattice contraction.
Kim, Changhoan
We report the results of a calculation of the K → pipi matrix elements of the DeltaI = 3/2 operators. Relying on the 3-flavor effective Hamiltonian, we calculate the low energy contribution to the matrix elements in quenched lattice QCD with the DBW2 action using domain wall fermions, while the high energy contribution is included in the Wilson coefficients. In order to generate interacting pipi states with non-zero relative momentum in lattice, we apply anti-periodic boundary conditions on pions. Since only the magnitude of the overlap of our interpolating operators with the initial and final state is determined, we can calculate only the magnitude of the matrix elements. From the comparison with the experimental result, however, we find some degree of discrepancy. This discrepancy might be ascribed to the unphysical kinematics we choose in this report.
Calculations of the Spin-Lattice Coupling Coefficients Fij and Zij for MgO:Co2+Crystal
ZHENG Wen-Chen; WU Shao-Yi
2001-01-01
According to a uniform and simple method of calculating spin-lattice coupling coefficients and the pert1rbation formulas of gi factors and hyperfine structure constants Ai based on the cluster approach for 3d7 ions in cubic,tetragonal and trigonal octahedral crystal fields, the spin-lattice coupling coefficients Fij (F11, Fl2, F44), Zij (Z11, Z12,Z44) and also g factor and hyperfine constant A for MgO:Co2+ are calculated by using the parameters obtained from the optical spectra without adjustable parameters. The calculated results show good agreement with the observed values.The difiiculty in explaining the coeficients Fij and Zij is therefore removed.``
Chizhikova, Z.N.; Kalashnikov, A.G.; Kapranova, E.N.; Korobitsyn, V.E.; Manturov, G.N.; Tsiboulia, A.A.
1998-12-01
One of the important problems for ensuring the VVER type reactor safety when the reactor is partially loaded with MOX fuel is the choice of appropriate physical zoning to achieve the maximum flattening of pin-by-pin power distribution. When uranium fuel is replaced by MOX one provided that the reactivity due to fuel assemblies is kept constant, the fuel enrichment slightly decreases. However, the average neutron spectrum fission microscopic cross-section for {sup 239}Pu is approximately twice that for {sup 235}U. Therefore power peaks occur in the peripheral fuel assemblies containing MOX fuel which are aggravated by the interassembly water. Physical zoning has to be applied to flatten the power peaks in fuel assemblies containing MOX fuel. Moreover, physical zoning cannot be confined to one row of fuel elements as is the case with a uniform lattice of uranium fuel assemblies. Both the water gap and the jump in neutron absorption macroscopic cross-sections which occurs at the interface of fuel assemblies with different fuels make the problem of calculating space-energy neutron flux distribution more complicated since it increases nondiffusibility effects. To solve this problem it is necessary to update the current codes, to develop new codes and to verify all the codes including nuclear-physical constants libraries employed. In so doing it is important to develop and validate codes of different levels--from design codes to benchmark ones. This paper presents the results of the burnup calculation for a multiassembly structure, consisting of MOX fuel assemblies surrounded by uranium dioxide fuel assemblies. The structure concerned can be assumed to model a fuel assembly lattice symmetry element of the VVER-1000 type reactor in which 1/4 of all fuel assemblies contains MOX fuel.
Kozier, K.S
1999-05-01
This paper compares the results of coolant-void reactivity (CVR) reactor-physics calculations performed using the Monte Carlo N-particle transport code, MCNP version 4B, with those obtained using Atomic Energy of Canada Limited's (AECL's) latest version of the Winfrith improved multigroup scheme (WIMS) code, WIMS-AECL version 2-5c. Cross sections derived from the evaluated nuclear data file version B-VI (ENDF/B-VI) are used for both the WIMS-AECL and MCNP4B calculations. The comparison is made for uniform lattices at room temperature containing either fresh natural uranium or mixed oxide (MOX) 37-element CANDU fuel. The MOX fuel composition corresponds roughly to that of irradiated CANDU fuel at a burnup of about 4500 MWd/tU. The level of agreement between the CVR predictions of WIMS-AECL and MCNP4B is studied as a function of lattice buckling (a measure of the curvature of the neutron-flux distribution) over the range from 0.0 to 4.1 m{sup -2} . For the cases studied, it is found that the absolute k values calculated by WIMS-AECL are higher than those of MCNP4B by several mk (1 mk is a change of 0.001 in k), amounts that depend on the fuel type being modelled and the particular cross-section data used. However, the agreement between WIMS-AECL and MCNP4B is much better for the CVR (i.e., the {delta}k on coolant voiding), and is relatively insensitive to the fuel type. (author)
Kozier, K.S. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)
1999-07-01
This paper compares the results of coolant-void reactivity (CVR) reactor-physics calculations performed using the Monte Carlo N-particle transport code, MCNP version 4B, with those obtained using Atomic Energy of Canada Limited's (AECL's) latest version of the Winfrith improved multigroup scheme (WIMS) code, WIMS-AECL version 2-5c. Cross sections derived from the evaluated nuclear data file version B-VT (ENDF/B-VI) are used for both the WIMS-AECL and MCNP4B calculations. The comparison is made for uniform lattices at room temperature containing either fresh natural uranium or mixed oxide (MOX) 37-element CANDU fuel. The MOX fuel composition corresponds roughly to that of irradiated CANDU fuel at a burnup of about 4500 MWd/tU. The level of agreement between the CVR predictions of WIMS-AECL and MCNP4B is studied as a function of lattice buckling (a measure of the curvature of the neutron-flux distribution) over the range from 0.0 to 4.1 m{sup -2}. For the cases studied, it is found that the absolute keff values calculated by WIMS-AECL are higher than those of MCNP4B by several mk (1 mk is a change of 0.001 in keff), amounts that depend on the fuel type being modelled and the particular cross-section data used. However, the agreement between WIMS-AECL and MCNP4B is much better for the CVR (i.e., the {delta}keff on coolant voiding), and is relatively insensitive to the fuel type. (author)
Algora, A.; Tain, J.L.; Perez-Cerdan, A.B.; Rubio, B.; Agramunt, J.; Caballero, L.; Nacher, E.; Jordan, D.; Molina, F. [Valencia Univ., IFIC (Spain); Algora, A.; Krasznahorkay, A.; Hunyadi, M.D.; Gulyas, J.; Vitez, A.; Csatlos, M.; Csige, L. [Institute of Nuclear Research, Debrecen (Hungary); Aysto, J.; Penttila, H.; Rinta-Antila, S.; Moore, I.; Eronen, T.; Jokinen, A.; Nieminen, A.; Hakala, J.; Karvonen, P.; Kankainen, A.; Hager, U.; Sonoda, T.; Saastamoinen, A.; Rissanen, J.; Kessler, T.; Weber, C.; Ronkainen, J.; Rahaman, S.; Elomaa, V. [Jyvaskyla Univ. (Finland); Burkard, K.; Huller, W. [GSI, Darmstadt (Germany); Batist, L. [PNPI, Gatchina (Russian Federation); Gelletly, W. [Surrey Univ., Guildford (United Kingdom); Yoshida, T. [Mushashi Institute of Technology (Japan); Nichols, A.L. [IAEA Nuclear Data Section, Vienna (Austria); Sonzogni, A. [National Nuclear Data Center, Brookhaven National Laboratory, Upton, NY (United States); Perajarvi, K. [STUK, Helsinki (Finland)
2008-07-01
The decay heat of fission products plays an important role in predictions of the heat up of nuclear fuel in reactors. The released energy is calculated as the summation of the activities of all fission products P(t) equals {sigma}(E{sub i}*{lambda}{sub i}*N{sub i}(t)), where E{sub i} is the decay energy of nuclide i (gamma and beta component), {lambda}{sub i} is the decay constant of nuclide i and N{sub i}(t) is the number of nuclide i at cooling time t. Even though the reproduction of the measured decay heat has improved in recent years, there is still a long standing discrepancy in the t about 1000 s cooling time for some fuels. A possible explanation to this improper description has been found in the work of Yoshida et al., where it has been shown that the incomplete knowledge of the {beta}-decay of some T{sub c} isotopes can be the source of the systematic discrepancy. We have recently measured the {beta}-decay process of some Tc isotopes using a total absorption spectrometer at the IGISOL facility in Jyvaskyla (Finland). The results of the measurements as well as the their consequences on summation calculations are discussed. (authors)
Lahti, G. P.; Mueller, R. A.
1973-01-01
Measurements of MeV neutron were made at the surface of a lithium hydride and depleted uranium shielded reactor. Four shield configurations were considered: these were assembled progressively with cylindrical shells of 5-centimeter-thick depleted uranium, 13-centimeter-thick lithium hydride, 5-centimeter-thick depleted uranium, 13-centimeter-thick lithium hydride, 5-centimeter-thick depleted uranium, and 3-centimeter-thick depleted uranium. Measurements were made with a NE-218 scintillation spectrometer; proton pulse height distributions were differentiated to obtain neutron spectra. Calculations were made using the two-dimensional discrete ordinates code DOT and ENDF/B (version 3) cross sections. Good agreement between measured and calculated spectral shape was observed. Absolute measured and calculated fluxes were within 50 percent of one another; observed discrepancies in absolute flux may be due to cross section errors.
Preliminary Development of Thermal Power Calculation Code H-Power for a Supercritical Water Reactor
Fan Zhang
2014-01-01
Full Text Available SCWR (Supercritical Water Reactor is one of the promising Generation IV nuclear systems, which has higher thermal power efficiency than current pressurized water reactor. It is necessary to perform the thermal equilibrium and thermal power calculation for the conceptual design and further monitoring and calibration of the SCWR. One visual software named H-Power was developed to calculate thermal power and its uncertainty of SCWR, in which the advanced IAPWS-IF97 industrial formulation was used to calculate the thermodynamic properties of water and steam. The ISO-5167-4: 2003 standard was incorporated in the code as the basis of orifice plate to compute the flow rate. New heat balance model and uncertainty estimate have also been included in the code. In order to validate H-Power, an assessment was carried out by using data published by US and Qinshan Phase II. The results showed that H-Power was able to estimate the thermal power of SCWR.
Casoli, P.; Authier, N.; Baud, J. [Commissariat a l' energie Atomique, Centre de Valduc, 21120 Is-sur-Tille (France)
2009-07-01
Several experimental devices are operated by the Criticality and Neutron Science Research Department of the CEA Valduc Laboratory. One of these is the metallic core reactor Caliban. The knowledge of the fundamental kinetic parameters of the reactor is very useful, indeed necessary, to the operator. The purpose of this study was to develop and perform experiments allowing to determinate some of these parameters. The prompt neutron decay constant and particularly its value at criticality can be measured with reactor noise techniques such as the interval-distribution, the Feynman variance-to-mean, and the Rossi-{alpha} methods. By introducing the Nelson number, the effective delayed neutron fraction and the average neutron lifetime can also be calculated with the Rossi-{alpha} method. Subcritical, critical, and even supercritical experiments were performed. With the Rossi-{alpha} technique, it was found that the prompt neutron decay constant at criticality was (6.02*10{sup 5} {+-} 9%). Experiments also brought out the limitations of the used experimental parameters. (authors)
Zhou, Fei [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Nielson, Weston [Univ. of California, Los Angeles, CA (United States); Xia, Yi [Univ. of California, Los Angeles, CA (United States); Ozoliņš, Vidvuds [Univ. of California, Los Angeles, CA (United States)
2014-10-01
First-principles prediction of lattice thermal conductivity κ_{L} of strongly anharmonic crystals is a long-standing challenge in solid-state physics. Making use of recent advances in information science, we propose a systematic and rigorous approach to this problem, compressive sensing lattice dynamics. Compressive sensing is used to select the physically important terms in the lattice dynamics model and determine their values in one shot. Nonintuitively, high accuracy is achieved when the model is trained on first-principles forces in quasirandom atomic configurations. The method is demonstrated for Si, NaCl, and Cu_{12}Sb_{4}S_{13}, an earth-abundant thermoelectric with strong phonon-phonon interactions that limit the room-temperature κ_{L} to values near the amorphous limit.
Zhou, Fei; Nielson, Weston; Xia, Yi; Ozoliņš, Vidvuds
2014-10-31
First-principles prediction of lattice thermal conductivity κ(L) of strongly anharmonic crystals is a long-standing challenge in solid-state physics. Making use of recent advances in information science, we propose a systematic and rigorous approach to this problem, compressive sensing lattice dynamics. Compressive sensing is used to select the physically important terms in the lattice dynamics model and determine their values in one shot. Nonintuitively, high accuracy is achieved when the model is trained on first-principles forces in quasirandom atomic configurations. The method is demonstrated for Si, NaCl, and Cu(12)Sb(4)S(13), an earth-abundant thermoelectric with strong phonon-phonon interactions that limit the room-temperature κ(L) to values near the amorphous limit.
Lattice calculation of $1^{-+}$ hybrid mesons with improved Kogut-Susskind fermions
Bernard, C; DeTar, C E; Gottlieb, S; Gregory, E B; Heller, U M; Osborn, J; Sugar, R; Toussaint, D; Gottlieb, Steven
2003-01-01
We report on a lattice determination of the mass of the exotic $1^{-+}$ hybrid meson using an improved Kogut-Susskind action. Results from both quenched and dynamical quark simulations are presented. We also compare with earlier results using Wilson quarks at heavier quark masses. The results on lattices with three flavors of dynamical quarks show effects of sea quarks on the hybrid propagators which probably result from coupling to two meson states. We extrapolate the quenched results to the physical light quark mass to allow comparison with experimental candidates for the $1^{-+}$ hybrid meson. The lattice result remains somewhat heavier than the experimental result, although it may be consistent with the $\\pi_1(1600)$.
Shchurovskaya, M. V.; Alferov, V. P.; Geraskin, N. I.; Radaev, A. I.
2017-01-01
The results of the validation of a research reactor calculation using Monte Carlo and deterministic codes against experimental data and based on code-to-code comparison are presented. The continuous energy Monte Carlo code MCU-PTR and the nodal diffusion-based deterministic code TIGRIS were used for full 3-D calculation of the IRT MEPhI research reactor. The validation included the investigations for the reactor with existing high enriched uranium (HEU, 90 w/o) fuel and low enriched uranium (LEU, 19.7 w/o, U-9%Mo) fuel.
Thermal-hydraulic Fortran program for steady-state calculations of plate-type fuel research reactors
Khedr Ahmed
2008-01-01
Full Text Available The safety assessment of research and power reactors is a continuous process covering their lifespan and requiring verified and validated codes. Power reactor codes all over the world are well established and qualified against real measuring data and qualified experimental facilities. These codes are usually sophisticated, require special skills and consume a lot of running time. On the other hand, most research reactor codes still require much more data for validation and qualification. It is, therefore, of benefit to any regulatory body to develop its own codes for the review and assessment of research reactors. The present paper introduces a simple, one-dimensional Fortran program called THDSN for steady-state thermal-hydraulic calculations of plate-type fuel research reactors. Besides calculating the fuel and coolant temperature distributions and pressure gradients in an average and hot channel, the program calculates the safety limits and margins against the critical phenomena encountered in research reactors, such as the onset of nucleate boiling, critical heat flux and flow instability. Well known thermal-hydraulic correlations for calculating the safety parameters and several formulas for the heat transfer coefficient have been used. The THDSN program was verified by comparing its results for 2 and 10 MW benchmark reactors with those published in IAEA publications and a good agreement was found. Also, the results of the program are compared with those published for other programs, such as the PARET and TERMIC.
A first look at maximally twisted mass lattice QCD calculations at the physical point
Abdel-Rehim, A. [The Cyprus Institute, Nicosia (Cyprus). CaSToRC; Boucaud, P. [Paris XI Univ., Orsay (France). Laboratoire de Physique Theorique; Carrasco, N. [Valencia-CSIC Univ. (Spain). Dept. de Fisica Teorica; IFIC, Valencia (Spain); and others
2013-11-15
In this contribution, a first look at simulations using maximally twisted mass Wilson fermions at the physical point is presented. A lattice action including clover and twisted mass terms is presented and the Monte Carlo histories of one run with two mass-degenerate flavours at a single lattice spacing are shown. Measurements from the light and heavy-light pseudoscalar sectors are compared to previous N{sub f}=2 results and their phenomenological values. Finally, the strategy for extending simulations to N{sub f}=2+1+1 is outlined.
A first look at maximally twisted mass lattice QCD calculations at the physical point
Abdel-Rehim, A; Carrasco, N; Deuzeman, A; Dimopoulos, P; Frezzotti, R; Herdoiza, G; Jansen, K; Kostrzewa, B; Mangin-Brinet, M; Montvay, I; Palao, D; Rossi, G C; Sanfilippo, F; Scorzato, L; Shindler, A; Urbach, C; Wenger, U
2013-01-01
In this contribution, a first look at simulations using maximally twisted mass Wilson fermions at the physical point is presented. A lattice action including clover and twisted mass terms is presented and the Monte Carlo histories of one run with two mass-degenerate flavours at a single lattice spacing are shown. Measurements from the light and heavy-light pseudoscalar sectors are compared to previous $N_f = 2$ results and their phenomenological values. Finally, the strategy for extending simulations to $N_f = 2 + 1 + 1$ is outlined.
Lattice Boltzmann equation calculation of internal, pressure-driven turbulent flow
Hammond, L A; Care, C M; Stevens, A
2002-01-01
We describe a mixing-length extension of the lattice Boltzmann approach to the simulation of an incompressible liquid in turbulent flow. The method uses a simple, adaptable, closure algorithm to bound the lattice Boltzmann fluid incorporating a law-of-the-wall. The test application, of an internal, pressure-driven and smooth duct flow, recovers correct velocity profiles for Reynolds number to 1.25 x 10 sup 5. In addition, the Reynolds number dependence of the friction factor in the smooth-wall branch of the Moody chart is correctly recovered. The method promises a straightforward extension to other curves of the Moody chart and to cylindrical pipe flow.
Lattice effective field theory calculations for A = 3,4,6,12 nuclei
Epelbaum, Evgeny; Lee, Dean; Meißner, Ulf-G
2009-01-01
We present lattice results for the ground state energies of tritium, helium-3, helium-4, lithium-6, and carbon-12 nuclei. Our analysis includes isospin-breaking, Coulomb effects, and interactions up to next-to-next-to-leading order in chiral effective field theory.
Calculation of the Perturbative Expansion of Wilson Operators on the Lattice
Liu, Da-Qing; Wu, Ji-Min; Chen, Ying
2001-11-01
We introduce an approach to expand gauge-invariant Wilson operators on the lattice. This approach is based on the non-Abelian-Stokes theorem and overcomes the disadvantages of the Luscher-Weisz method. It is also suitable for expanding any Wilson operator.
Calculation of the Perturbative Expansion of Wilson Operators on the Lattice
刘大庆; 吴济民; 陈莹
2001-01-01
We introduce an approach to expand gauge-invariant Wilson operators on the lattice. This approach is based on the non-Abelian-Stokes theorem and overcomes the disadvantages of the Luscher-Weisz method. It is also suitable for expanding any Wilson operator.
Fellinger, Michael R; Hector, Louis G; Trinkle, Dallas R
2017-02-01
We present computed datasets on changes in the lattice parameter and elastic stiffness coefficients of bcc Fe due to substitutional Al, B, Cu, Mn, and Si solutes, and octahedral interstitial C and N solutes. The data is calculated using the methodology based on density functional theory (DFT) presented in Ref. (M.R. Fellinger, L.G. Hector Jr., D.R. Trinkle, 2017) [1]. All the DFT calculations were performed using the Vienna Ab initio Simulations Package (VASP) (G. Kresse, J. Furthmüller, 1996) [2]. The data is stored in the NIST dSpace repository (http://hdl.handle.net/11256/671).
Pham, Binh Thi-Cam [Idaho National Laboratory; Hawkes, Grant Lynn [Idaho National Laboratory; Einerson, Jeffrey James [Idaho National Laboratory
2015-08-01
This paper presents the quantification of uncertainty of the calculated temperature data for the Advanced Gas Reactor (AGR) fuel irradiation experiments conducted in the Advanced Test Reactor at Idaho National Laboratory in support of the Advanced Reactor Technology Research and Development program. Recognizing uncertainties inherent in physics and thermal simulations of the AGR tests, the results of the numerical simulations are used in combination with statistical analysis methods to improve qualification of measured data. The temperature simulation data for AGR tests are also used for validation of the fission product transport and fuel performance simulation models. These crucial roles of the calculated fuel temperatures in ensuring achievement of the AGR experimental program objectives require accurate determination of the model temperature uncertainties. To quantify the uncertainty of AGR calculated temperatures, this study identifies and analyzes ABAQUS model parameters of potential importance to the AGR predicted fuel temperatures. The selection of input parameters for uncertainty quantification of the AGR calculated temperatures is based on the ranking of their influences on variation of temperature predictions. Thus, selected input parameters include those with high sensitivity and those with large uncertainty. Propagation of model parameter uncertainty and sensitivity is then used to quantify the overall uncertainty of AGR calculated temperatures. Expert judgment is used as the basis to specify the uncertainty range for selected input parameters. The input uncertainties are dynamic accounting for the effect of unplanned events and changes in thermal properties of capsule components over extended exposure to high temperature and fast neutron irradiation. The sensitivity analysis performed in this work went beyond the traditional local sensitivity. Using experimental design, analysis of pairwise interactions of model parameters was performed to establish
Hybrid parallel code acceleration methods in full-core reactor physics calculations
Courau, T.; Plagne, L.; Ponicot, A. [EDF R and D, 1, Avenue du General de Gaulle, 92141 Clamart Cedex (France); Sjoden, G. [Nuclear and Radiological Engineering, Georgia Inst. of Technology, Atlanta, GA 30332 (United States)
2012-07-01
When dealing with nuclear reactor calculation schemes, the need for three dimensional (3D) transport-based reference solutions is essential for both validation and optimization purposes. Considering a benchmark problem, this work investigates the potential of discrete ordinates (Sn) transport methods applied to 3D pressurized water reactor (PWR) full-core calculations. First, the benchmark problem is described. It involves a pin-by-pin description of a 3D PWR first core, and uses a 8-group cross-section library prepared with the DRAGON cell code. Then, a convergence analysis is performed using the PENTRAN parallel Sn Cartesian code. It discusses the spatial refinement and the associated angular quadrature required to properly describe the problem physics. It also shows that initializing the Sn solution with the EDF SPN solver COCAGNE reduces the number of iterations required to converge by nearly a factor of 6. Using a best estimate model, PENTRAN results are then compared to multigroup Monte Carlo results obtained with the MCNP5 code. Good consistency is observed between the two methods (Sn and Monte Carlo), with discrepancies that are less than 25 pcm for the k{sub eff}, and less than 2.1% and 1.6% for the flux at the pin-cell level and for the pin-power distribution, respectively. (authors)
White, J.E.; Roussin, R.W.; Gilpin, H.
1988-12-01
A version of the CRAC2 computer code applicable for use in analyses of consequences and risks of reactor accidents in case work for environmental statements has been implemented for use on the Nuclear Regulatory Commission Data General MV/8000 computer system. Input preparation is facilitated through the use of an interactive computer program which operates on an IBM personal computer. The resulting CRAC2 input deck is transmitted to the MV/8000 by using an error-free file transfer mechanism. To facilitate the use of CRAC2 at NRC, relevant background material on input requirements and model descriptions has been extracted from four reports - ''Calculations of Reactor Accident Consequences,'' Version 2, NUREG/CR-2326 (SAND81-1994) and ''CRAC2 Model Descriptions,'' NUREG/CR-2552 (SAND82-0342), ''CRAC Calculations for Accident Sections of Environmental Statements, '' NUREG/CR-2901 (SAND82-1693), and ''Sensitivity and Uncertainty Studies of the CRAC2 Computer Code,'' NUREG/CR-4038 (ORNL-6114). When this background information is combined with instructions on the input processor, this report provides a self-contained guide for preparing CRAC2 input data with a specific orientation toward applications on the MV/8000. 8 refs., 11 figs., 10 tabs.
Impact of nuclear data on sodium-cooled fast reactor calculations
Aures, Alexander; Bostelmann, Friederike; Zwermann, Winfried; Velkov, Kiril
2016-03-01
Neutron transport and depletion calculations are performed in combination with various nuclear data libraries in order to assess the impact of nuclear data on safety-relevant parameters of sodium-cooled fast reactors. These calculations are supplemented by systematic uncertainty analyses with respect to nuclear data. Analysed quantities are the multiplication factor and nuclide densities as a function of burn-up and the Doppler and Na-void reactivity coefficients at begin of cycle. While ENDF/B-VII.0 / -VII.1 yield rather consistent results, larger discrepancies are observed between the JEFF libraries. While the newest evaluation, JEFF-3.2, agrees with the ENDF/B-VII libraries, the JEFF-3.1.2 library yields significant larger multiplication factors.
Impact of nuclear data on sodium-cooled fast reactor calculations
Aures Alexander
2016-01-01
Full Text Available Neutron transport and depletion calculations are performed in combination with various nuclear data libraries in order to assess the impact of nuclear data on safety-relevant parameters of sodium-cooled fast reactors. These calculations are supplemented by systematic uncertainty analyses with respect to nuclear data. Analysed quantities are the multiplication factor and nuclide densities as a function of burn-up and the Doppler and Na-void reactivity coefficients at begin of cycle. While ENDF/B-VII.0 / -VII.1 yield rather consistent results, larger discrepancies are observed between the JEFF libraries. While the newest evaluation, JEFF-3.2, agrees with the ENDF/B-VII libraries, the JEFF-3.1.2 library yields significant larger multiplication factors.
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 10^{3} - 10^{5} 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.
A framework for the calculation of the $\\Delta N\\gamma^*$ transition form factor on the lattice
Agadjanov, Andria; Meißner, Ulf-G; Rusetsky, Akaki
2014-01-01
Using the non-relativistic effective field theory framework in a finite volume, we discuss the extraction of the $\\Delta N\\gamma^*$ transition form factors from lattice data. A counterpart of the L\\"uscher approach for the matrix elements of unstable states is formulated. In particular, we thoroughly discuss various kinematic settings, which are used in the calculation of the above matrix element on the lattice. The emerging L\\"uscher-Lellouch factor and the analytic continuation of the matrix elements into the complex plane are also considered in detail. A full group-theoretical analysis of the problem is made, including the partial-wave mixing and projecting out the invariant form factors from data.
Activation calculation and radiation analysis for China Fusion Engineering Test Reactor
Chen, Zhi, E-mail: zchen@ustc.edu.cn; Qiao, Shiji; Jiang, Shuai; Xu, X. George
2016-11-01
Highlights: • Activation calculation was performed using FLUKA for the main components of CFETR. • Radionuclides and radioactive wastes were assessed for CFETR. • The Waste Disposal Ratings (WDR) were assessed for CFETR. - Abstract: The activation calculation and analysis for the China Fusion Engineering Test Reactor (CFETR) will play an important role in its system design, maintenance, inspection and assessment of nuclear waste. Using the multi-particle transport code FLUKA and its associated data library, we calculated the radioactivity, specific activity, waste disposal rating from activation products, nuclides in the tritium breeding blanket, shielding layer, vacuum vessel and toroidal field coil (TFC) of CFETR. This paper presents the calculation results including neutron flux, activation products and waste disposal rating after one-year full operation of the CFETR. The findings show that, under the assumption of one-year operation at the 200 MW fusion power, the total radioactivity inventory will be 1.05 × 10{sup 19} Bq at shutdown and 1.03 × 10{sup 17} Bq after ten years. The primary residual nuclide is found to be {sup 55}Fe in ten years after the shutdown. The waste disposal rating (WDR) values are very low (<<1), according to Class C limits, CFETR materials are qualified for shallow land burial. It is shown that CFETR has no serious activation safety issue.
Junxiao Zheng
2016-01-01
Full Text Available Maintaining the structural integrity of the reactor pressure vessel (RPV is a critical concern related to the safe operation of nuclear power plants. To estimate the structural integrity over the designed lifetime and to support analyses for a potential plant life extension, an accurate calculation of the fast neutron fluence (E>1.0 MeV or E>0.1 MeV at the RPV is significant. The discrete ordinates method is one of the main methods to solve such problems. During the calculation process, many factors will affect the results. In this paper, the deviations introduced by different differencing schemes and mesh sizes on the AP1000 RPV fast neutron fluence have been studied, which are based on new discrete ordinates code ARES. The analysis shows that the differencing scheme (diamond difference with or without linear zero fix-up, theta weighted, directional theta weighted, and exponential directional weighted introduces a deviation within 4%. The coarse mesh (4 × 4 cm meshes in XY plane leads to approximately 23.7% calculation deviation compared to those of refined mesh (1 × 1 cm meshes in XY plane. Comprehensive study on the deviation introduced by differencing scheme and mesh size has great significance for reasoned evaluation of RPV fast neutron fluence calculation results.
Vdovin, S. A. [JSC ' E and E' (Russian Federation); Shalimov, A. S. [LLC Selekt Co. (Russian Federation)
2013-05-15
The use of the function of effective current braking of the longitudinal differential protection of shunt reactors to offset current surges, which enables the sensitivity of differential protection to be increased when there are short circuits with low damage currents, is considered. It is shown that the use of the calculated braking characteristic enables the reliability of offset protection from transients to be increased when the reactor is connected, which is accompanied by the flow of asymmetric currents containing an aperiodic component.
Development and verification of fuel burn-up calculation model in a reduced reactor geometry
Sembiring, Tagor Malem [Center for Reactor Technology and Nuclear Safety (PTKRN), National Nuclear Energy Agency (BATAN), Kawasan PUSPIPTEK Gd. No. 80, Serpong, Tangerang 15310 (Indonesia)], E-mail: tagorms@batan.go.id; Liem, Peng Hong [Research Laboratory for Nuclear Reactor (RLNR), Tokyo Institute of Technology (Tokyo Tech), O-okayama, Meguro-ku, Tokyo 152-8550 (Japan)
2008-02-15
A fuel burn-up model in a reduced reactor geometry (2-D) is successfully developed and implemented in the Batan in-core fuel management code, Batan-FUEL. Considering the bank mode operation of the control rods, several interpolation functions are investigated which best approximate the 3-D fuel assembly radial power distributions across the core as function of insertion depth of the control rods. Concerning the applicability of the interpolation functions, it can be concluded that the optimal coefficients of the interpolation functions are not very sensitive to the core configuration and core or fuel composition in RSG GAS (MPR-30) reactor. Consequently, once the optimal interpolation function and its coefficients are derived then they can be used for 2-D routine operational in-core fuel management without repeating the expensive 3-D neutron diffusion calculations. At the selected fuel elements (at H-9 and G-6 core grid positions), the discrepancy of the FECFs (fuel element channel power peaking factors) between the 2-D and 3-D models are within the range of 3.637 x 10{sup -4}, 3.241 x 10{sup -4} and 7.556 x 10{sup -4} for the oxide, silicide cores with 250 g {sup 235}U/FE and the silicide core with 300 g {sup 235}U/FE, respectively.
Khare, Ankur; Himmetoglu, Burak; Johnson, Melissa; Norris, David J.; Cococcioni, Matteo; Aydil, Eray S.
2012-04-01
The electronic structure, lattice dynamics, and Raman spectra of the kesterite, stannite, and pre-mixed Cu-Au (PMCA) structures of Cu2ZnSnS4 (CZTS) and Cu2ZnSnSe4 (CZTSe) were calculated using density functional theory (DFT). Differences in longitudinal and transverse optical (LO-TO) splitting in kesterite, stannite, and PMCA structures can be used to differentiate them. The Γ-point phonon frequencies, which give rise to Raman scattering, exhibit small but measurable shifts, for these three structures. Experimentally measured Raman scattering from CZTS and CZTSe thin films were examined in light of DFT calculations and deconvoluted to explain subtle shifts and asymmetric line shapes often observed in CZTS and CZTSe Raman spectra. Raman spectroscopy in conjunction with ab initio calculations can be used to differentiate between kesterite, stannite, and PMCA structures of CZTS and CZTSe.
Lattice calculation of the leading strange quark-connected contribution to the muon g−2
Blum, T. [Physics Department, University of Connecticut,Storrs, CT 06269-3046 (United States); Boyle, P.A.; Debbio, L. Del [School of Physics and Astronomy, University of Edinburgh,Peter Guthrie Tait Road, Edinburgh EH9 3JZ (United Kingdom); Hudspith, R.J. [Department of Physics and Astronomy, York University,4700 Keele Street, Toronto, Ontario, M3J 1P3 (Canada); Izubuchi, T. [Physics Department, Brookhaven National Laboratory,Upton, NY 11973 (United States); RIKEN-BNL Research Center, Brookhaven National Laboratory,Upton, NY 11973 (United States); Jüttner, A. [School of Physics and Astronomy, University of Southampton,Southampton SO17 1BJ (United Kingdom); Lehner, C. [Physics Department, Brookhaven National Laboratory,Upton, NY 11973 (United States); Lewis, R. [Department of Physics and Astronomy, York University,4700 Keele Street, Toronto, Ontario, M3J 1P3 (Canada); Maltman, K. [Department of Mathematics and Statistics, York University,4700 Keele Street, Toronto, Ontario, M3J 1P3 (Canada); CSSM, University of Adelaide,Adelaide, SA 5005 (Australia); Marinković, M. Krstić [School of Physics and Astronomy, University of Southampton,Southampton SO17 1BJ (United Kingdom); CERN, Theoretical Physics Department, CERN,Geneva (Switzerland); Portelli, A. [School of Physics and Astronomy, University of Edinburgh,Peter Guthrie Tait Road, Edinburgh EH9 3JZ (United Kingdom); School of Physics and Astronomy, University of Southampton,Southampton SO17 1BJ (United Kingdom); Spraggs, M. [School of Physics and Astronomy, University of Southampton,Southampton SO17 1BJ (United Kingdom); Collaboration: The RBC/UKQCD collaboration
2016-04-11
We present results for the leading hadronic contribution to the muon anomalous magnetic moment due to strange quark-connected vacuum polarisation effects. Simulations were performed using RBC-UKQCD’s N{sub f}=2+1 domain wall fermion ensembles with physical light sea quark masses at two lattice spacings. We consider a large number of analysis scenarios in order to obtain solid estimates for residual systematic effects. Our final result in the continuum limit is a{sub μ}{sup (2)} {sup had,} {sup s}=53.1(9)({sub −3}{sup +1})×10{sup −10}.
Lattice calculation of the leading strange quark-connected contribution to the muon $g-2$
Blum, T.; Del Debbio, L.; Hudspith, R.J.; Izubuchi, T.; Jüttner, A.; Lehner, C.; Lewis, R.; Maltman, K.; Krstić Marinković, M.; Portelli, A.; Spraggs, M.
2016-01-01
We present results for the leading hadronic contribution to the muon anomalous magnetic moment due to strange quark-connected vacuum polarisation effects. Simulations were performed using RBC--UKQCD's $N_f=2+1$ domain wall fermion ensembles with physical light sea quark masses at two lattice spacings. We consider a large number of analysis scenarios in order to obtain solid estimates for residual systematic effects. Our final result in the continuum limit is $a_\\mu^{(2)\\,{\\rm had},\\,s}=53.1(9)\\left(^{+1}_{-3}\\right)\\times10^{-10}$.
Phenomenology of {Λ}_b\\to {Λ}_cτ {\\overline{ν}}_{τ } using lattice QCD calculations
Datta, Alakabha; Kamali, Saeed; Meinel, Stefan; Rashed, Ahmed
2017-08-01
In a recent paper we studied the effect of new-physics operators with different Lorentz structures on the semileptonic {Λ}_b\\to {Λ}_cτ {\\overline{ν}}_{τ } decay. This decay is of interest in light of the R( D (*)) puzzle in the semileptonic \\overline{B}\\to {D}^{(\\ast )}τ {\\overline{ν}}_{τ } decays. In this work we add tensor operators to extend our previous results and consider both model-independent new physics (NP) and specific classes of models proposed to address the R( D (*)) puzzle. We show that a measurement of R({Λ}_c)=\\mathrmB[{Λ}_b\\to {Λ}_cτ {\\overline{ν}}_{τ}]/\\mathrmB[{Λ}_b\\to {Λ}_cℓ {\\overline{ν}}_{ℓ}] can strongly constrain the NP parameters of models discussed for the R( D (*)) puzzle. We use form factors from lattice QCD to calculate all {Λ}_b\\to {Λ}_cτ {\\overline{ν}}_{τ } observables. The Λ b → Λ c tensor form factors had not previously been determined in lattice QCD, and we present new lattice results for these form factors here.
Two-dimensional TBR calculations for conceptual compact reversed-field pinch reactor blanket
Davidson, J. W.; Battat, M. E.; Dudziak, D. J.
A detailed two-dimensional nucleonic analysis was performed for a conceptual first wall, blanket, and shield design for the Compact Reversed-Field Pinch Reactor. The design includes significant two-dimensional aspects presented by the limiter, vacuum ducts, and coolant manifolds; these aspects seriously degrade the tritium-breeding reaction (TBR) predicted by one-dimensional calculations. A range of design change to increase the TBR were investigated within the two-dimensional analysis. The results of this investigation indicated that an adequate TBR could be achieved with a thinning copper first wall, a (6)Li enrichment near 90%, the proper selection of reflector, and a small addition to the blanket thickness, determined by the one-dimensional analysis.
Girardi, E.; Ruggieri, J.M. [CEA Cadarache, CEA/DEN/CAD/DER/SPRC/LEPH, 13 - Saint-Paul Lez Durance (France)
2003-07-01
The aim of this paper is to present the last developments made on a domain decomposition method applied to reactor core calculations. In this method, two kind of balance equation with two different numerical methods dealing with two different unknowns are coupled. In the first part the two balance transport equations (first order and second order one) are presented with the corresponding following numerical methods: Variational Nodal Method and Discrete Ordinate Nodal Method. In the second part, the Multi-Method/Multi-Domain algorithm is introduced by applying the Schwarz domain decomposition to the multigroup eigenvalue problem of the transport equation. The resulting algorithm is then provided. The projection operators used to coupled the two methods are detailed in the last part of the paper. Finally some preliminary numerical applications on benchmarks are given showing encouraging results. (authors)
Calculation of the heavy-hadron axial couplings g1, g2, and g3 using lattice QCD
Will Detmold, David Lin, Stefan Meinel
2012-06-01
In a recent paper [arXiv:1109.2480] we have reported on a lattice QCD calculation of the heavy-hadron axial couplings g{sub 1}, g{sub 2}, and g{sub 3}. These quantities are low-energy constants of heavy-hadron chiral perturbation theory (HH{chi}PT) and are related to the B*B{pi}, {Sigma}{sub b}*{Sigma}{sub b}{pi}, and {Sigma}{sub b}{sup (*)}{Lambda}{sub b}{pi} couplings. In the following, we discuss important details of the calculation and give further results. To determine the axial couplings, we explicitly match the matrix elements of the axial current in QCD with the corresponding matrix elements in HH{chi}PT. We construct the ratios of correlation functions used to calculate the matrix elements in lattice QCD, and study the contributions from excited states. We present the complete numerical results and discuss the data analysis in depth. In particular, we demonstrate the convergence of SU(4|2) HH{chi}PT for the axial current matrix elements at pion masses up to about 400 MeV and show the impact of the nonanalytic loop contributions. Finally, we present additional predictions for strong and radiative decay widths of charm and bottom baryons.
Teherani, James T
2013-01-01
We have developed a physically-intuitive method to calculate the local lattice constant as a function of position in a high-resolution transmission electron microscopy image by performing a two-dimensional fast Fourier transform. We apply a Gaussian filter with appropriate spatial full-width-half-max (FWHM) bandwidth to the image centered at the desired location to calculate the local lattice constant (as opposed to the average lattice constant). Fourier analysis of the filtered image yields the vertical and horizontal lattice constants at this location. The process is repeated by stepping the Gaussian filter across the image to produce a set of local lattice constants in the vertical and horizontal direction as a function of position in the image. The method has been implemented in a freely available tool on nanoHUB.
Xiaozhi Wu; Shaofeng Wang
2007-01-01
Applying the parametric derivation method, Peierls energy and Peierls stress are calculated with a non-sinusoidal force law in the lattice theory, while the results obtained by the power-series expansion according to sinusoidal law can be deduced as a limiting case of nonsinusoidal law. The simplified expressions of Peierls energy and Peierls stress are obtained for the limit of wide and narrow. Peierls energy and Peierls stress decrease monotonically with the factor of modification of force law. Present results can be used expediently for prediction of the correct order of magnitude of Peierls stress for materials.
Fourmentel, D.; Radulovic, V.; Barbot, L.; Villard, J-F. [Alternative Energies and Atomic Energy Commission, CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, 13108 Saint- Paul-Lez-Durance (France); Zerovnik, G.; Snoj, L. [Reactor Physics Department, Jozef Stefan Institute, SI-1000 Ljubljana (Slovenia); Tarchalski, M.; Pytel, K. [National Centre for Nuclear Research A. Soltana 7, 05-400 Swierk (Poland); Malouch, F. [Alternative Energies and Atomic Energy Commission - CEA, DEN, DM2S, Saclay, 91191, Gif-sur-Yvette (France)
2015-07-01
Neutron and gamma flux levels are key parameters in nuclear research reactors. In Material Testing Reactors, such as the future Jules Horowitz Reactor, under construction at the French Alternative Energies and Atomic Energy Commission (CEA Cadarache, France), the expected gamma flux levels are very high (nuclear heating is of the order of 20 W/g at 100 MWth). As gamma rays deposit their energy in the reactor structures and structural materials it is important to take them into account when designing irradiation devices. There are only a few sensors which allow measurements of the nuclear heating ; a recent development at the CEA Cadarache allows measurements of the gamma flux using a miniature ionization chamber (MIC). The measured MIC response is often compared with calculation using modern Monte Carlo (MC) neutron and photon transport codes, such as TRIPOLI-4 and MCNP6. In these calculations only the production of prompt gamma rays in the reactor is usually modelled thus neglecting the delayed gamma rays. Hence calculations and measurements are usually in better accordance for the neutron flux than for the gamma flux. In this paper we study the contribution of delayed gamma rays to the total MIC signal in order to estimate the systematic error in gamma flux MC calculations. In order to experimentally determine the delayed gamma flux contributions to the MIC response, we performed gamma flux measurements with CEA developed MIC at three different research reactors: the OSIRIS reactor (MTR - 70 MWth at CEA Saclay, France), the TRIGA MARK II reactor (TRIGA - 250 kWth at the Jozef Stefan Institute, Slovenia) and the MARIA reactor (MTR - 30 MWth at the National Center for Nuclear Research, Poland). In order to experimentally assess the delayed gamma flux contribution to the total gamma flux, several reactor shut down (scram) experiments were performed specifically for the purpose of the measurements. Results show that on average about 30 % of the MIC signal is due to
Bangart, R.L.; Bell, L.G.; Boegli, J.S.; Burke, W.C.; Lee, J.Y.; Minns, J.L.; Stoddart, P.G.; Weller, R.A.; Collins, J.T.
1978-12-01
The calculational procedures described in the report reflect current NRC staff practice. The methods described will be used in the evaluation of applications for construction permits and operating licenses docketed after January 1, 1979, until this NUREG is revised as a result of additional staff review. The BWR-GALE (Boiling Water Reactor Gaseous and Liquid Effluents) Code is a computerized mathematical model for calculating the release of radioactive material in gaseous and liquid effluents from boiling water reactors (BWRs). The calculations are based on data generated from operating reactors, field tests, laboratory tests, and plant-specific design considerations incorporated to reduce the quantity of radioactive materials that may be released to the environment.
Methodology comparison for gamma-heating calculations in material-testing reactors
Lemaire, M.; Vaglio-Gaudard, C.; Lyoussi, A. [CEA, DEN, DER, Cadarache F-13108 Saint Paul les Durance (France); Reynard-Carette, C. [Aix Marseille Universite, CNRS, Universite de Toulon, IM2NP UMR 7334, 13397, Marseille (France)
2015-07-01
The Jules Horowitz Reactor (JHR) is a Material-Testing Reactor (MTR) under construction in the south of France at CEA Cadarache (French Alternative Energies and Atomic Energy Commission). It will typically host about 20 simultaneous irradiation experiments in the core and in the beryllium reflector. These experiments will help us better understand the complex phenomena occurring during the accelerated ageing of materials and the irradiation of nuclear fuels. Gamma heating, i.e. photon energy deposition, is mainly responsible for temperature rise in non-fuelled zones of nuclear reactors, including JHR internal structures and irradiation devices. As temperature is a key parameter for physical models describing the behavior of material, accurate control of temperature, and hence gamma heating, is required in irradiation devices and samples in order to perform an advanced suitable analysis of future experimental results. From a broader point of view, JHR global attractiveness as a MTR depends on its ability to monitor experimental parameters with high accuracy, including gamma heating. Strict control of temperature levels is also necessary in terms of safety. As JHR structures are warmed up by gamma heating, they must be appropriately cooled down to prevent creep deformation or melting. Cooling-power sizing is based on calculated levels of gamma heating in the JHR. Due to these safety concerns, accurate calculation of gamma heating with well-controlled bias and associated uncertainty as low as possible is all the more important. There are two main kinds of calculation bias: bias coming from nuclear data on the one hand and bias coming from physical approximations assumed by computer codes and by general calculation route on the other hand. The former must be determined by comparison between calculation and experimental data; the latter by calculation comparisons between codes and between methodologies. In this presentation, we focus on this latter kind of bias. Nuclear
Nuclear Reactions from Lattice QCD
Briceño, Raúl A; Luu, Thomas C
2014-01-01
One of the overarching goals of nuclear physics is to rigorously compute properties of hadronic systems directly from the fundamental theory of strong interactions, Quantum Chromodynamics (QCD). In particular, the hope is to perform reliable calculations of nuclear reactions which will impact our understanding of environments that occur during big bang nucleosynthesis, the evolution of stars and supernovae, and within nuclear reactors and high energy/density facilities. Such calculations, being truly ab initio, would include all two-nucleon and three- nucleon (and higher) interactions in a consistent manner. Currently, lattice QCD provides the only reliable option for performing calculations of some of the low- energy hadronic observables. With the aim of bridging the gap between lattice QCD and nuclear many-body physics, the Institute for Nuclear Theory held a workshop on Nuclear Reactions from Lattice QCD on March 2013. In this review article, we report on the topics discussed in this workshop and the path ...
Nur Krisna, Dwita; Su'ud, Zaki
2017-01-01
Nuclear reactor technology is growing rapidly, especially in developing Nuclear Power Plant (NPP). The utilization of nuclear energy in power generation systems has been progressing phase of the first generation to the fourth generation. This final project paper discusses the analysis neutronic one-cooled fast reactor type Pb-Bi, which is capable of operating up to 20 years without refueling. This reactor uses Thorium Uranium Nitride as fuel and operating on power range 100-500MWtNPPs. The method of calculation used a computer simulation program utilizing the SRAC. SPINNOR reactor is designed with the geometry of hexagonal shaped terrace that radially divided into three regions, namely the outermost regions with highest percentage of fuel, the middle regions with medium percentage of fuel, and most in the area with the lowest percentage. SPINNOR fast reactor operated for 20 years with variations in the percentage of Uranium-233 by 7%, 7.75%, and 8.5%. The neutronic calculation and analysis show that the design can be optimized in a fast reactor for thermal power output SPINNOR 300MWt with a fuel fraction 60% and variations of Uranium-233 enrichment of 7%-8.5%.
First principle calculation of structure and lattice dynamics of Lu2Si2O7
Nazipov, D. V.; Nikiforov, A. E.
2016-12-01
Ab initio calculations of crystal structure and Raman spectra has been performed for single crystal of lutetium pyrosilicate Lu2Si2O7. The types of fundamental vibrations, their frequencies and intensities in the Raman spectrum has been obtained for two polarizations. Calculations were made in the framework of density functional theory (DFT) with hybrid functionals. The isotopic substitution was calculated for all inequivalent ions in cell. The results in a good agreement with experimental data.
First principle calculation of structure and lattice dynamics of Lu2Si2O7
Nazipov D.V.
2017-01-01
Full Text Available Ab initio calculations of crystal structure and Raman spectra has been performed for single crystal of lutetium pyrosilicate Lu2Si2O7. The types of fundamental vibrations, their frequencies and intensities in the Raman spectrum has been obtained for two polarizations. Calculations were made in the framework of density functional theory (DFT with hybrid functionals. The isotopic substitution was calculated for all inequivalent ions in cell. The results in a good agreement with experimental data.
A CG Method for Multiple Right Hand Sides and Multiple Shifts in Lattice QCD Calculations
Birk, Sebastian
2012-01-01
We consider the task of computing solutions of linear systems that only differ by a shift with the identity matrix as well as linear systems with several different right hand sides. In the past Krylov subspace methods have been developed which exploit either the need for solutions to multiple right hand sides (e.g. deflation type methods and block methods) or multiple shifts (e.g. shifted CG) with some success. In this paper we present a block Krylov subspace method which, based on a block Lanczos process, exploits both features - shifts and multiple right hand sides - at once. Such situations arise, for example, in lattice QCD simulations within the Rational Hybrid Monte Carlo algorithm. We give numerical evidence that our method is superior to applying other iterative methods to each of the systems individually as well as, in some cases, to shifted or block Krylov subspace methods.
Yang Ping; Li Pei; Zhang Li-Qiang; Wang Xiao-Liang; Wang Huan; Song Xi-Fu; Xie Fang-Wei
2012-01-01
The lattice,the band gap and the optical properties of n-type ZnO under uniaxial stress are investigated by firstprinciples calculations.The results show that the lattice constants change linearly with stress.Band gaps are broadened linearly as the uniaxial compressive stress increases.The change of band gap for n-type ZnO comes mainly from the contribution of stress in the c-axis direction,and the reason for band gap of n-type ZnO changing with stress is also explained.The calculated results of optical properties reveal that the imaginary part of the dielectric function decreases with the increase of uniaxial compressive stress at low energy.However,when the energy is higher than 4.0 eV,the imaginary part of the dielectric function increases with the increase of stress and a blueshift appears.There are two peaks in the absorption spectrum in an energy range of 4.0-13.0 eV.The stress coefficient of the band gap of n-type ZnO is larger than that of pure ZnO,which supplies the theoretical reference value for the modulation of the band gap of doped ZnO.
Gamma-point lattice free energy estimates from O(1) force calculations
Voss, Johannes; Vegge, Tejs
2008-01-01
We present a new method for estimating the vibrational free energy of crystal (and molecular) structures employing only a single force calculation, for a particularly displaced configuration, in addition to the calculation of the ground state configuration. This displacement vector is the sum...
Gamma-point lattice free energy estimates from O(1) force calculations
Voss, Johannes; Vegge, Tejs
2008-01-01
We present a new method for estimating the vibrational free energy of crystal (and molecular) structures employing only a single force calculation, for a particularly displaced configuration, in addition to the calculation of the ground state configuration. This displacement vector is the sum...
Ab initio calculations of phonon dispersion and lattice dynamics in TlGaTe{sub 2}
Jafarova, Vusala; Orudzhev, Guseyn; Alekperov, Oktay; Mamedov, Nazim; Abdullayev, Nadir; Najafov, Arzu [Institute of Physics (Innovation Sector), 33 H. Javid ave, Baku 1143 (Azerbaijan); Paucar, Raul [Institute of Physics (Innovation Sector), 33 H. Javid ave, Baku 1143 (Azerbaijan); Chiba Institute of Technology, 2-17-1 Tsudanuma, Narashino, Chiba 275-0016 (Japan); Shim, YongGu [Osaka Prefecture University, 1-1 Gakuen-cho, Naka-ku, Sakai, Osaka 599-8531 (Japan); Wakita, Kazuki [Chiba Institute of Technology, 2-17-1 Tsudanuma, Narashino, Chiba 275-0016 (Japan)
2015-06-15
This work reports the results of DFT-based calculations of phonon spectra of TlGaTe{sub 2}. The dispersion of phonon bands was calculated along the directions of Brillouin zone (BZ) that include symmetry points. The calculated phonon frequencies at the centre of BZ were compared with those obtained by Raman spectroscopy with the aid of a confocal laser microscopy system. A fairly good agreement between the calculated and experimental data was found. Complimentary, molar heat capacity at constant volume and Debye temperature were calculated in the range 5/500 K on the base of the obtained phonon density of states. The obtained temperature dependencies were compared with available experimental data.The results of comparison were satisfactory. (copyright 2015 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)
Madni, I.K. [Brookhaven National Lab., Upton, NY (United States); Cazzoli, E.G.; Khatib-Rahbar, M. [Energy Research, Inc., Rockville, MD (United States)
1995-11-01
During certain hypothetical severe accidents in a nuclear power plant, radionuclides could be released to the environment as a plume. Prediction of the atmospheric dispersion and transport of these radionuclides is important for assessment of the risk to the public from such accidents. A simplified PC-based model was developed that predicts time-integrated air concentration of each radionuclide at any location from release as a function of time integrated source strength using the Gaussian plume model. The solution procedure involves direct analytic integration of air concentration equations over time and position, using simplified meteorology. The formulation allows for dry and wet deposition, radioactive decay and daughter buildup, reactor building wake effects, the inversion lid effect, plume rise due to buoyancy or momentum, release duration, and grass height. Based on air and ground concentrations of the radionuclides, the early dose to an individual is calculated via cloudshine, groundshine, and inhalation. The model also calculates early health effects based on the doses. This paper presents aspects of the model that would be of interest to the prediction of environmental flows and their public consequences.
Decay heat experiment and validation of calculation code systems for fusion reactor
Maekawa, Fujio; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Wada, Masayuki
1999-10-01
Although accurate estimation of decay heat value is essential for safety analyses of fusion reactors against loss of coolant accidents and so on, no experimental work has been devoted to validating the estimation. Hence, a decay heat measurement experiment was performed as a task (T-339) of ITER/EDA. A new detector, the Whole Energy Absorption Spectrometer (WEAS), was developed for accurate and efficient measurements of decay heat. Decay heat produced in the thirty-two sample materials which were irradiated by 14-MeV neutrons at FNS/JAERI were measured with WEAS for a wide cooling time period from 1 min to 400 days. The data presently obtained were the first experimental decay heat data in the field of fusion. Validity of decay heat calculation codes of ACT4 and CINAC-V4, activation cross section libraries of FENDL/A-2.0 and JENDL Activation File, and decay data was investigated through analyses of the experiment. As a result, several points that should be modified were found in the codes and data. After solving the problems, it was demonstrated that decay heat valued calculated for most of samples were in good agreement with the experimental data. Especially for stainless steel 316 and copper, which were important materials for ITER, decay heat could be predicted with accuracy of {+-}10%. (author)
Trifonenkov, A. V.; Trifonenkov, V. P.
2017-01-01
This article deals with a feature of problems of calculating time-average characteristics of nuclear reactor optimal control sets. The operation of a nuclear reactor during threatened period is considered. The optimal control search problem is analysed. The xenon poisoning causes limitations on the variety of statements of the problem of calculating time-average characteristics of a set of optimal reactor power off controls. The level of xenon poisoning is limited. There is a problem of choosing an appropriate segment of the time axis to ensure that optimal control problem is consistent. Two procedures of estimation of the duration of this segment are considered. Two estimations as functions of the xenon limitation were plot. Boundaries of the interval of averaging are defined more precisely.
Calculations of lattice vibrational mode lifetimes using Jazz: a Python wrapper for LAMMPS
Gao, Y.; Wang, H.; Daw, M. S.
2015-06-01
Jazz is a new python wrapper for LAMMPS [1], implemented to calculate the lifetimes of vibrational normal modes based on forces as calculated for any interatomic potential available in that package. The anharmonic character of the normal modes is analyzed via the Monte Carlo-based moments approximation as is described in Gao and Daw [2]. It is distributed as open-source software and can be downloaded from the website http://jazz.sourceforge.net/.
Hummel, D.W.; Langton, S.E.; Ball, M.R.; Novog, D.R.; Buijs, A., E-mail: hummeld@mcmaster.ca [McMaster Univ., Hamilton, Ontario (Canada)
2013-07-01
Discrepancies have been observed among a number of recent reactor physics studies in support of the PT-SCWR pre-conceptual design, including differences in lattice-level predictions of infinite neutron multiplication factor, coolant void reactivity, and radial power profile. As a first step to resolving these discrepancies, a lattice-level benchmark problem was designed based on the 78-element plutonium-thorium PT-SCWR fuel design under a set of prescribed local conditions. This benchmark problem was modeled with a suite of both deterministic and Monte Carlo neutron transport codes. The results of these models are presented here as the basis of a code-to-code comparison. (author)
Alcala Ruiz, F.
1973-07-01
By a Point-Reactor model and Perturbation Theory, changes in reactivity during some regular operating periods of JEN-1 MOD Reactor have been calculated and compared with available measured values. they were in good agreement. Also changes in reactivity have been calculated during operations at higher power levels than the present one, concluding some practical consequences for the case of increasing the present power of this reactor. (Author)
Cossu, Guido; Fukaya, Hidenori; Hashimoto, Shoji; Kaneko, Takashi; Noaki, Jun-Ichi
2016-09-01
We compute the chiral condensate in 2 + 1-flavor QCD through the spectrum of low-lying eigenmodes of the Dirac operator. The number of eigenvalues of the Dirac operator is evaluated using a stochastic method with an eigenvalue filtering technique on the background gauge configurations generated by lattice QCD simulations including the effects of dynamical up, down, and strange quarks described by the Möbius domain-wall fermion formulation. The low-lying spectrum is related to the chiral condensate, which is one of the leading-order low-energy constants in chiral effective theory, as dictated by the Banks-Casher relation. The spectrum shape and its dependence on the sea quark masses calculated in numerical simulations are consistent with the expectation from one-loop chiral perturbation theory. After taking the chiral limit as well as the continuum limit using the data at three lattice spacings in the range 0.080-0.045 fm, we obtain Σ(2 GeV) = 270.0(4.9) MeV, with the error combining those from statistical and various sources of systematic error. The finite volume effect is confirmed to be under control by a direct comparison of the results from two different volumes at the lightest available sea quarks corresponding to 230 MeV pions.
Lu, Peng-Xian, E-mail: pengxian_lu@haut.edu.cn; Xia, Yi
2017-05-01
How to further optimize the thermoelectric figure of merit of silicon (Si) nanostructure? Constructing the layered structure composed of two different Si nano morphologies should be viewed an effective approach. The figure of merit of the layered structure could be further optimized by tuning the different contribution from the composed nano morphologies on the electron and phonon transport. In order to reveal the thermoelectric transport mechanism, the electronic structure, the lattice dynamics and the thermoelectric properties of Si nanosphere, Si nanoribbon and the layered structure composed of the two nano morphologies were investigated through first-principles calculation, lattice dynamics simulation and Boltzmann transport theory. The results suggest that the figure of merit of the layered structure is improved significantly in whole although its specific thermoelectric parameters are unsatisfactory as compared to the single nano morphologies. Therefore we provide a complete understanding on the thermoelectric transport of the layered structure and an effective route to further optimize the figure of merit of Si nanostructure.
Determination of $|V_{us}|$ from a lattice-QCD calculation of the $K\\to\\pi\\ell\
Bazavov, A; Bouchard, C; DeTar, C; Du, D; El-Khadra, A X; Foley, J; Freeland, E D; Gámiz, E; Gottlieb, Steven; Heller, U M; Kim, J; Kronfeld, A S; Laiho, J; Levkova, L; Mackenzie, P B; Neil, E T; Oktay, M B; Qiu, Si-Wei; Simone, J N; Sugar, R; Toussaint, D; Van de Water, R S; Zhou, Ran
2013-01-01
We calculate the kaon semileptonic form factor $f_+(0)$ from lattice QCD, working, for the first time, at the physical light-quark masses. We use gauge configurations generated by the MILC collaboration with $N_f=2+1+1$ flavors of sea quarks, which incorporate the effects of dynamical charm quarks as well as those of up, down, and strange. We employ data at three lattice spacings to extrapolate to the continuum limit. Our result, $f_+(0) = 0.9704(32)$, where the error is the total statistical plus systematic uncertainty added in quadrature, is the most precise determination to date. Combining our result with the latest experimental measurements of $K$ semileptonic decays, one obtains the Cabibbo-Kobayashi-Maskawa matrix element $|V_{us}|=0.22290(74)(52)$, where the first error is from $f_+(0)$ and the second one is from experiment. In the first-row test of Cabibbo-Kobayashi-Maskawa unitarity, the error stemming from $|V_{us}|$ is now comparable to that from $|V_{ud}|$.
Paratte, J.M. [Laboratory for Reactor Physics and Systems Behaviour (LRS), Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Frueh, R. [Ecole Polytechnique Federale de Lausanne (EPFL), CH-1015 Lausanne (Switzerland); Kasemeyer, U. [Laboratory for Reactor Physics and Systems Behaviour (LRS), Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Kalugin, M.A. [Kurchatov Institute, 123182 Moscow (Russian Federation); Timm, W. [Framatome-ANP, D-91050 Erlangen (Germany); Chawla, R. [Laboratory for Reactor Physics and Systems Behaviour (LRS), Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne (EPFL), CH-1015 Lausanne (Switzerland)
2006-05-15
Measurements in the CROCUS reactor at EPFL, Lausanne, are reported for the critical water level and the inverse reactor period for several different sets of delayed supercritical conditions. The experimental configurations were also calculated by four different calculation methods. For each of the supercritical configurations, the absolute reactivity value has been determined in two different ways, viz.: (i) through direct comparison of the multiplication factor obtained employing a given calculation method with the corresponding value for the critical case (calculated reactivity: {rho} {sub calc}); (ii) by application of the inhour equation using the kinetic parameters obtained for the critical configuration and the measured inverse reactor period (measured reactivity: {rho} {sub meas}). The calculated multiplication factors for the reference critical configuration, as well as {rho} {sub calc} for the supercritical cases, are found to be in good agreement. However, the values of {rho} {sub meas} produced by two of the applied calculation methods differ appreciably from the corresponding {rho} {sub calc} values, clearly indicating deficiencies in the kinetic parameters obtained from these methods.
On the structure of Lattice code WIMSD-5B
Kim, Won Young; Min, Byung Joo
2004-03-15
The WIMS-D code is a freely available thermal reactor physics lattice code used widely for thermal research and power reactor calculation. Now the code WIMS-AECL, developed on the basis of WIMS-D, has been used as one of lattice codes for the cell calculation in Canada and also, in 1998, the latest version WIMSD-5B is released for OECD/NEA Data Bank. While WIMS-KAERI was developed and has been used, originated from WIMS-D, in Korea, it was adjusted for the cell calculation of research reactor HANARO and so it has no confirmaty to CANDU reactor. Therefore, the code development applicable to cell calculation of CANDU reactor is necessary not only for technological independence and but also for the establishment of CANDU safety analysis system. A lattice code WIMSD-5B was analyzed in order to set the system of reactor physics computer codes, to be used in the assessment of void reactivity effect. In order to improve and validate WIMSD-5B code, the analysis of the structure of WIMSD-5B lattice code was made and so its structure, algorithm and the subroutines of WIMSD-5B were presented for the cluster type and the pij method modelling the CANDU-6 fuel
Calculations of reactor-accident consequences, Version 2. CRAC2: computer code user's guide
Ritchie, L.T.; Johnson, J.D.; Blond, R.M.
1983-02-01
The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems.
Lattice dynamics of diamond-like crystals from a tight-binding calculation of valence bands
Roman, R.; Pascual, J.
1988-11-01
We report on the results of calculations of the TA(X) phonon energy in the series of C, Si, Ge, Sn homopolar crystals. The starting point is the tight-binding model for the electronic Hamiltonian where Es and Ep are taken to be the free atomic energies while the interatomic matrix elements are described by a universal d-2 Harrison's scaling law. The change of the total energy with the atomic distortion is given in terms of changes in the valence band energy and changes in the overlap energy. The numerical calculations for Si gives U1 = -21.77eV and U2 = 60.44eV, close to the values predicted by Harrison U1 = -17.76eV and U2 = 53.28eV. The calculations of the TA(X) phonon energy gives (in the case the interatomic distances are held constant): 26.09 THz (C), 6.46 THz (Si), 3.37THz (Ge) and 1.91 THz (Sn), in reasonably good agreement with the experimental results 24.1 THz (C), 4.49 THz (Si), 2.39 THz (Ge) and 1.26 THz (Sn).
Wang, Yan; Lu, Zexi; Ruan, Xiulin
2016-06-01
The effect of phonon-electron (p-e) scattering on lattice thermal conductivity is investigated for Cu, Ag, Au, Al, Pt, and Ni. We evaluate both phonon-phonon (p-p) and p-e scattering rates from first principles and calculate the lattice thermal conductivity (κL). It is found that p-e scattering plays an important role in determining the κL of Pt and Ni at room temperature, while it has negligible effect on the κL of Cu, Ag, Au, and Al. Specifically, the room temperature κLs of Cu, Ag, Au, and Al predicted from density-functional theory calculations with the local density approximation are 16.9, 5.2, 2.6, and 5.8 W/m K, respectively, when only p-p scattering is considered, while it is almost unchanged when p-e scattering is also taken into account. However, the κL of Pt and Ni is reduced from 7.1 and 33.2 W/m K to 5.8 and 23.2 W/m K by p-e scattering. Even though Al has quite high electron-phonon coupling constant, a quantity that characterizes the rate of heat transfer from hot electrons to cold phonons in the two-temperature model, p-e scattering is not effective in reducing κL owing to the relatively low p-e scattering rates in Al. The difference in the strength of p-e scattering in different metals can be qualitatively understood by checking the amount of electron density of states that is overlapped with the Fermi window. Moreover, κL is found to be comparable to the electronic thermal conductivity in Ni.
无
2010-01-01
Viscosity is an important physical parameter of fluid,and the Eyring viscosity equation is a popular viscosity theory.Based on the Eyring reaction rate equation and Boltzmann statistical theory,and including the probabilities of creating a hole in liquid and the transition to the neighboring hole,a modified Eyring viscosity equation was proposed.According to the structural characteristics of short-range order,liquid is treated as a quasi-lattice structure in a small region.The activation energy,which is the minimum energy needed for the molecule to jump to its neighboring hole because of the restriction of other molecules around it,was analytically calculated from an intermolecular Lennard-Jones potential function and a Stockmayer potential function.The viscosity values of 37 kinds of typical liquids at 25°C and the dependence of viscosity of three kinds of liquids on temperatures were calculated with this modified viscosity equation,and the calculated results agree with the experimental values to some extent.This work not only enriches the understanding of the mechanism of liquid viscosity,but also could provide some theoretical guides for the relevant studies and applications.
Lattice dynamics and electron-phonon coupling calculations using nondiagonal supercells
Lloyd-Williams, Jonathan; Monserrat, Bartomeu
Quantities derived from electron-phonon coupling matrix elements require a fine sampling of the vibrational Brillouin zone. Converged results are typically not obtainable using the direct method, in which a perturbation is frozen into the system and the total energy derivatives are calculated using a finite difference approach, because the size of simulation cell needed is prohibitively large. We show that it is possible to determine the response of a periodic system to a perturbation characterized by a wave vector with reduced fractional coordinates (m1 /n1 ,m2 /n2 ,m3 /n3) using a supercell containing a number of primitive cells equal to the least common multiple of n1, n2, and n3. This is accomplished by utilizing supercell matrices containing nonzero off-diagonal elements. We present the results of electron-phonon coupling calculations using the direct method to sample the vibrational Brillouin zone with grids of unprecedented size for a range of systems, including the canonical example of diamond. We also demonstrate that the use of nondiagonal supercells reduces by over an order of magnitude the computational cost of obtaining converged vibrational densities of states and phonon dispersion curves. J.L.-W. is supported by the Engineering and Physical Sciences Research Council (EPSRC). B.M. is supported by Robinson College, Cambridge, and the Cambridge Philosophical Society. This work was supported by EPSRC Grants EP/J017639/1 and EP/K013564/1.
Supplemental Reactor Physics Calculations and Analysis of ELF Mk 1A Fuel
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.
Shekhanova, M. E.
2017-01-01
In this paper we propose a method of using neutronic calculation code CORNER to the analysis of experiments on the protection of fast neutron reactor and CNFC equipment. An example of Winfrith Graphite Benchmark experiment calculation using this approach is presented. This task can be considered as one step in the general theme of the safety analysis of FR with liquid metal coolant, their fuel cycles and related equipment. CORNER implement a solution of the kinetic equation with a source in the three-dimensional hexagonal geometry based on Sn-method. The purpose of this paper is a demonstration of the application of CORNER’s possibilities for the analysis of the actual reactor problems.
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.
Evans, Robert M.
1976-10-05
1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.
Zeiner, T. [Technical University of Dortmund, Lehrstuhl fuer Fluidverfahrenstechnik, Emil-Figge Str. 70, 44227 Dortmund (Germany); Browarzik, C. [Technical University of Berlin, Institut fuer Prozess- und Verfahrenstechnik (TK7), Fachgebiet Thermodynamik und thermische Verfahrenstechnik, Strasse des 17, Juni 135, 10623 Berlin (Germany); Browarzik, D., E-mail: dieter.browarzik@chemie.uni-halle.de [Martin-Luther University Halle-Wittenberg, Institut fuer Chemie/Physikalische Chemie, 06099 Halle (Germany); Enders, S. [Technical University of Berlin, Institut fuer Prozess- und Verfahrenstechnik (TK7), Fachgebiet Thermodynamik und thermische Verfahrenstechnik, Strasse des 17, Juni 135, 10623 Berlin (Germany)
2011-12-15
Highlights: > The (liquid + liquid) equilibrium of hyperbranched polyester solutions is calculated. > The solvents are n-alkanes, propan-1-ol, and butan-1-ol. > The lattice-cluster theory is combined with a chemical association model. > The solvent molecules are assumed to be linear chains of segments. > The calculations agree reasonably well with the experimental data. - Abstract: The (liquid + liquid) equilibrium of solutions of hyperbranched polyesters is calculated with the lattice-cluster theory (LCT) combined with a chemical association model. The considered solvents are n-alkanes as well as propan-1-ol and butan-1-ol. The structure of the solvents is also considered in the framework of the LCT, assuming the solvent molecules as linear chains of several segments. For polymer solutions with the non-associating n-alkanes only the self association of the hyperbranched polymer molecules has to be considered by the chemical association lattice model (CALM). For the solutions of the type alcohol + hyperbranched polymer additionally the cross association is taken into account by a modified version of the extended chemical association lattice model (ECALM). The association effects are proved to influence strongly the phase equilibrium. Calculating the cloud-point curve and the critical point the polydispersity of the polymer samples is neglected. There is a reasonable agreement of the calculated curves with the experimental data taken from the literature.
Lattice dynamics of lithium oxide
Prabhatasree Goel; N Choudhury; S L Chaplot
2004-08-01
Li2O finds several important technological applications, as it is used in solid-state batteries, can be used as a blanket breeding material in nuclear fusion reactors, etc. Li2O exhibits a fast ion phase, characterized by a thermally induced dynamic disorder in the anionic sub-lattice of Li+, at elevated temperatures around 1200 K. We have carried out lattice-dynamical calculations of Li2O using a shell model in the quasi-harmonic approximation. The calculated phonon frequencies are in excellent agreement with the reported inelastic neutron scattering data. Thermal expansion, specific heat, elastic constants and equation of state have also been calculated which are in good agreement with the available experimental data.
Garcia M, T.; Cruz G, H. S.; Ruiz C, M. A.; Angeles C, A., E-mail: teodoro.garcia@inin.gob.mx [ININ, Carretera Mexico-Toluca sn, 52750 Ocoyoacac, Estado de Mexico (Mexico)
2014-10-15
In the pool of TRIGA Mark III reactor of the Instituto Nacional de Investigaciones Nucleares (ININ), there are control rods that were removed from the core, and which are currently on shelves of decay. These rods were part of the reactor core when only had fuel standard (from 1968-1989). To conduct a proper activation analysis of the rods, is very important to have well-characterized the materials which are built, elemental composition of the same ones, the atomic densities and weight fractions of the elements that constitute them. To determine the neutron activation of the control rods MCNP5 code was used, this code allows us to have well characterized the radionuclides inventory that were formed during irradiation of the control rods. This work is limited to determining the activation of the steel that is part of the shielding of the control rods, the nuclear fuel that is in the fuel follower does not include. The calculation model of the code will be validated with experimental measurements and calculating the activity of fission products of the fuel follower which will take place at the end of 2014. (Author)
Test problem for thermal-hydraulics and neutronic coupled calculation fore ALFREAD reactor core
Filip, A.; Darie, G.; Saldikov, I. S.; Smirnov, A. D.; Tikhomirov, G. V.
2017-01-01
The beginning of a new era of nuclear reactor requires technological advances and also multiples studies. The European Liquid metal cooled Fast breeder Reactor is one of the designs for the generation IV nuclear reactor, selected by ENEA. A pioneer of its time, ELFR needs a demonstrator in order to prove the feasibility of this project and to acquire more data and experience in operating a LFR. For this reason the ALFRED project was started and it is expected to be under operation by the year 2030. This paper has the objective of analyzing the neutronic and thermohydraulics of the ALFRED core by the means of a coupled scheme. The selected code for neutronic simulation is MCNP and the selected code for thermohydraulics is ANSYS.
Gonçalves, I. F.; Ramalho, A. G.; Gonçalves, I. C.; Salgado, J.
The work presented concerns the calculation of the external biological shielding for a neutron beam tube that will be installed at the Portuguese Research Reactor, RPI. This tube will have enough versatility to be used in fields so different as the analysis of the composition of samples or research work in Boron Neutron Capture Therapy, BNCT. The calculation was made by using the MCNP code. This code is a well validated and widely used code, and has therefore become an important tool in the design and optimisation work of experiences related to neutrons and gamma radiation.
Gritzay, Olena; Kalchenko, Oleksandr; Klimova, Nataliya; Razbudey, Volodymyr; Sanzhur, Andriy; Binney, Stephen
2005-05-01
The presented results show our consecutive steps in developing a neutron source with parameters required by Boron Neutron Capture Therapy (BNCT) at the Kyiv Research Reactor (KRR). The main goal of this work was to analyze the influence of installation of different types of uranium converters close to the reactor core on neutron beam characteristics and on level of reactor safety. The general Monte Carlo radiation transport code MCNP, version 4B, has been used for these calculations.
Cai, Xiao-Xiao; Llamas-Jansa, Isabel; Mullet, Steven
2013-01-01
to reactor modelling. Before version 9.5, Geant4 HP thermal scattering model (i.e. the S(α; β) model ) supports only three bounded isotopes, namely, H in water and polyethylene, and C in graphite. Newly supported materials include D in heavy water, O and Be in beryllium oxide, H and Zr in zirconium hydride......, U and O in uranium dioxide, Al metal, Be metal, and Fe metal. The native HP cross section library G4NDL does not include data for elements with atomic number larger than 92. Therefore, transuranic elements, which have impacts for a realistic reactor, can not be simulated by the combination of the HP...
Developments in Sensitivity Methodologies and the Validation of Reactor Physics Calculations
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.
Calculation of $K \\to \\pi\\pi$ decay amplitudes with improved Wilson fermion action in lattice QCD
Ishizuka, N; Ukawa, A; Yoshié, T
2015-01-01
We present our results for the $K\\to\\pi\\pi$ decay amplitudes for both the $\\Delta I=1/2$ and $3/2$ channels. Calculations are carried out with $N_f=2+1$ gauge configurations generated with the Iwasaki gauge action and non-perturbatively $O(a)$-improved Wilson fermion action at $a=0.091\\,{\\rm fm}$, $m_\\pi=280\\,{\\rm MeV}$ and $m_K=580\\,{\\rm MeV}$ on a $32^3\\times 64$ ($La=2.9\\,{\\rm fm}$) lattice. For the quark loops in the penguin and disconnected contributions in the $I=0$ channel, the combined hopping parameter expansion and truncated solver method work very well for variance reduction. We obtain, for the first time with a Wilson-type fermion action, that ${\\rm Re}A_0 = 60(36) \\times10^{ -8}\\,{\\rm GeV}$ and ${\\rm Im}A_0 =-67(56) \\times10^{-12}\\,{\\rm GeV}$ for a matching scale $q^* =1/a$. The dependence on the matching scale $q^*$ for these values is weak.
Sherbini, S; Tamasanis, D; Sykes, J; Porter, S W
1986-12-01
A program was developed to calculate the exposure rate resulting from airborne gases inside a reactor containment building. The calculations were performed at the location of a wall-mounted area radiation monitor. The program uses Monte Carlo techniques and accounts for both the direct and scattered components of the radiation field at the detector. The scattered component was found to contribute about 30% of the total exposure rate at 50 keV and dropped to about 7% at 2000 keV. The results of the calculations were normalized to unit activity per unit volume of air in the containment. This allows the exposure rate readings of the area monitor to be used to estimate the airborne activity in containment in the early phases of an accident. Such estimates, coupled with containment leak rates, provide a method to obtain a release rate for use in offsite dose projection calculations.
Fernandes, A C; Gonçalves, I C; Santos, J; Cardoso, J; Santos, L; Ferro Carvalho, A; Marques, J G; Kling, A; Ramalho, A J G; Osvay, M
2006-01-01
This work presents an extensive study on Monte Carlo radiation transport simulation and thermoluminescent (TL) dosimetry for characterising mixed radiation fields (neutrons and photons) occurring in nuclear reactors. The feasibility of these methods is investigated for radiation fields at various locations of the Portuguese Research Reactor (RPI). The performance of the approaches developed in this work is compared with dosimetric techniques already existing at RPI. The Monte Carlo MCNP-4C code was used for a detailed modelling of the reactor core, the fast neutron beam and the thermal column of RPI. Simulations using these models allow to reproduce the energy and spatial distributions of the neutron field very well (agreement better than 80%). In the case of the photon field, the agreement improves with decreasing intensity of the component related to fission and activation products. (7)LiF:Mg,Ti, (7)LiF:Mg,Cu,P and Al(2)O(3):Mg,Y TL detectors (TLDs) with low neutron sensitivity are able to determine photon dose and dose profiles with high spatial resolution. On the other hand, (nat)LiF:Mg,Ti TLDs with increased neutron sensitivity show a remarkable loss of sensitivity and a high supralinearity in high-intensity fields hampering their application at nuclear reactors.
Calculation of the Thermal State of the Graphite Moderator of the RBMK Reactor
Vorobiev Alexander V.
2017-01-01
Full Text Available This work is devoted to study the temperature field of the graphite stack of the RBMK reactor. In work was analyzed the influence of contact pressure between the components of the masonry on the temperature of the graphite moderator.
Huang, Ran; Gujrati, Purushottam D.
2017-01-01
An inhomogeneous 2-dimensional recursive lattice formed by planar elements has been designed to investigate the thermodynamics of Ising spin system on the surface/thin film. The lattice is constructed as a hybrid of partial Husimi square lattice representing the bulk and 1D single bonds representing the surface. Exact calculations can be achieved with the recursive property of the lattice. The model has an anti-ferromagnetic interaction to give rise to an ordered phase identified as crystal, and a solution with higher energy to represent the amorphous/metastable phase. Free energy and entropy of the ideal crystal and supercooled liquid state of the model on the surface are calculated by the partial partition function. By analyzing the free energies and entropies of the crystal and supercooled liquid state, we are able to identify the melting and ideal glass transition on the surface. The results show that due to the variation of coordination number, the transition temperatures on the surface decrease significantly compared to the bulk system. Our calculation qualitatively agrees with both experimental and simulation works on the thermodynamics of surfaces and thin films conducted by others. Interactions between particles farther than the nearest neighbor distance are taken into consideration, and their effects are investigated. Supported by the National Natural Science Foundation of China under Grant No. 11505110, the Shanghai Pujiang Talent Program under Grant No. 16PJ1431900, and the China Postdoctoral Science Foundation under Grant No. 2016M591666
Pressure tube creep impact on the physics parameters for CANDU-6 reactors
Kim, W. Y.; Min, B. J. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Kam, S. C.; Kim, M. E. [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)
2004-07-01
The lattice cell calculations are performed to assess the sensitivity of the reactor physics parameters to pressure tube creep resulting from radiation aging. The physics parameters of the lattice cell are calculated by using WIMSD-5B code, WIMS- AECL code, and MCNP code. The reference model(normal state) and two perturbed models accounting for the pressure tube creep are developed on the basis of CANDU-6 lattice cell. The 2.5% and 5% values of pressure tube diameter creep are considered. Also, The effects of the analyzed lattice parameters which are the coolant void reactivity, the fuel fission density and the atom density of Pu isotopes on the lattice.
EQ6 Calculations for Chemical Degradation Of N Reactor (U-Metal) Spent Nuclear Fuel Waste Packages
P. Bernot
2001-02-27
The Monitored Geologic Repository (MGR) Waste Package Department of the Civilian Radioactive Waste Management System Management & Operating Contractor (CRWMS M&O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the N Reactor, a graphite moderated reactor at the Department of Energy's (DOE) Hanford Site (ref. 1). The N Reactor core was fueled with slightly enriched (0.947 wt% and 0.947 to 1.25 wt% {sup 235}U in Mark IV and Mark IA fuels, respectively) U-metal clad in Zircaloy-2 (Ref. 1, Sec. 3). Both types of N Reactor SNF have been considered for disposal at the proposed Yucca Mountain site. For some WPs, the outer shell and inner shell may breach (Ref. 3) allowing the influx of water. Water in the WP will moderate neutrons, increasing the likelihood of a criticality event within the WP; and the water may, in time, gradually leach the fissile components from the WP, further affecting the neutronics of the system. This study presents calculations of the long-term geochemical behavior of WPs containing two multi-canister overpacks (MCO) with either six baskets of Mark IA or five baskets of Mark IV intact N Reactor SNF rods (Ref. 1, Sec. 4) and two high-level waste (HLW) glass pour canisters (GPCs) arranged according to the codisposal concept (Ref. 4). The specific study objectives were to determine: (1) The extent to which fissile uranium will remain in the WP after corrosion/dissolution of the initial WP configuration (2) The extent to which fissile uranium will be carried out of the degraded WP by infiltrating water (such that internal criticality is no longer possible, but the possibility of external criticality may be enhanced); and (3) The nominal chemical composition for the criticality evaluations of the WP design, and to suggest the range of parametric variations for additional evaluations. The scope of this calculation, the chemical compositions (and subsequent criticality evaluations) of the simulations, is limited
Reflected kinetics model for nuclear space reactor kinetics and control scoping calculations
Washington, K.E.
1986-05-01
The objective of this research is to develop a model that offers an alternative to the point kinetics (PK) modelling approach in the analysis of space reactor kinetics and control studies. Modelling effort will focus on the explicit treatment of control drums as reactivity input devices so that the transition to automatic control can be smoothly done. The proposed model is developed for the specific integration of automatic control and the solution of the servo mechanism problem. The integration of the kinetics model with an automatic controller will provide a useful tool for performing space reactor scoping studies for different designs and configurations. Such a tool should prove to be invaluable in the design phase of a space nuclear system from the point of view of kinetics and control limitations.
Developments in Sensitivity Methodologies and the Validation of Reactor Physics Calculations
Giuseppe Palmiotti; Massimo Salvatores
2012-01-01
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 metho...
Grabaskas, David [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Bucknor, Matthew [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Jerden, James [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Brunett, Acacia J. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Denman, Matthew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Nuclear Engineering Division; Clark, Andrew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Nuclear Engineering Division; Denning, Richard S. [Consultant, Columbus, OH (United States)
2016-10-01
The potential release of radioactive material during a plant incident, referred to as the source term, is a vital design metric and will be a major focus of advanced reactor licensing. The U.S. Nuclear Regulatory Commission has stated an expectation for advanced reactor vendors to present a mechanistic assessment of the potential source term in their license applications. The mechanistic source term presents an opportunity for vendors to realistically assess the radiological consequences of an incident and may allow reduced emergency planning zones and smaller plant sites. However, the development of a mechanistic source term for advanced reactors is not without challenges, as there are often numerous phenomena impacting the transportation and retention of radionuclides. This project sought to evaluate U.S. capabilities regarding the mechanistic assessment of radionuclide release from core damage incidents at metal fueled, pool-type sodium fast reactors (SFRs). The purpose of the analysis was to identify, and prioritize, any gaps regarding computational tools or data necessary for the modeling of radionuclide transport and retention phenomena. To accomplish this task, a parallel-path analysis approach was utilized, as shown below. One path, led by Argonne and Sandia National Laboratories, sought to perform a mechanistic source term assessment using available codes, data, and models, with the goal to identify gaps in the current knowledge base. The second path, performed by an independent contractor, performed sensitivity analyses to determine the importance of particular radionuclides and transport phenomena in regards to offsite consequences. The results of the two pathways were combined to prioritize gaps in current capabilities.
Kostal, M.; Juricek, V.; Rypar, V.; Svadlenkova, M. [Research Center Rez Ltd., 250 68 Husinec-Rez 130 (Czech Republic); Cvachovec, F. [Univ. of Defence, Kounicova 65, 662 10 Brno (Czech Republic)
2011-07-01
The power density distribution in a reactor has significant influence on core structures and pressure vessel mechanical resistance, as well as on the physical characteristics of nuclear fuel. This quantity also has an effect on the leakage neutron and photon field. This issue has become of increasing importance, as it touches on actual questions of the VVER nuclear power plant life time extension. This paper shows the comparison of calculated and experimentally determined pin by pin power distributions. The calculation has been performed with deterministic and Monte Carlo approaches. This quantity is accompanied by the neutron and photon flux density calculation and measurements at different points of the light water zero-power (LR-0) research reactor mock-up core, reactor built-in component (core barrel), and reactor pressure vessel and model. The effect of the different data libraries used for calculation is discussed. (authors)
Validation of SCALE4.4a for Calculation of Xe-Sm Transients After a Scram of the BR2 Reactor
Kalcheva, S.; Ponsard, B.; Koonen, E.
2007-07-15
The aim of this report is to validate the computational modules system SCALE4.4a for evaluation of reactivity changes, macroscopic absorption cross sections and calculations of the positions of the Control Rods during their motion in Xe-Sm transient after a scram of the BR-2 reactor. The rapid shutting down of the reactor by inserting of negative reactivity by the Control Rods is known as a reactor scram. Following reactor scram, a large xenon and samarium buildup occur in the reactor, which may appreciably affect the multiplication factor of the core due to enormous neutron absorption. The validation of the calculations of Xe-Sm transients by SCALE4.4a has been performed on the measurements of the positions of the Control Rods during their motion in Xe-Sm transients of the BR-2 reactor and on comparison with the calculations by the standard procedure XESM, developed at the BR-2 reactor. A final conclusion is made that the SCALE4.4a modules system can be used for evaluation of Xe-Sm transients of the BR-2 reactor. The utilization of the code is simple, the computational time takes from few seconds.
Schlömer Luc
2017-01-01
Full Text Available The decommissioning of a light water reactor (LWR, which is licensed under § 7 of the German Atomic Energy Act, following the post-operational phase requires a comprehensive licensing procedure including in particular radiation protection aspects and possible impacts to the environment. Decommissioning includes essential changes in requirements for the systems and components and will mainly lead to the direct dismantling. In this context, neutron induced activation calculations for the structural components have to be carried out to predict activities in structures and to estimate future costs for conditioning and packaging. To avoid an overestimation of the radioactive inventory and to calculate the expenses for decommissioning as accurate as possible, modern state-of-the-art Monte-Carlo-Techniques (MCNP™ are applied and coupled with present-day activation and decay codes (ORIGEN-S. In this context ADVANTG is used as weight window generator for MCNP™ i. e. as variance reduction tool to speed up the calculation in deep penetration problems. In this paper the calculation procedure is described and the obtained results are presented with a validation along with measured activities and photon dose rates measured in the post-operational phase. The validation shows that the applied calculation procedure is suitable for the determination of the radioactive inventory of a nuclear power plant. Even the measured gamma dose rates in the post-operational phase at different positions in the reactor building agree within a factor of 2 to 3 with the calculation results. The obtained results are accurate and suitable to support effectively the decommissioning planning process.
Schlömer, Luc; Phlippen, Peter-W.; Lukas, Bernard
2017-09-01
The decommissioning of a light water reactor (LWR), which is licensed under § 7 of the German Atomic Energy Act, following the post-operational phase requires a comprehensive licensing procedure including in particular radiation protection aspects and possible impacts to the environment. Decommissioning includes essential changes in requirements for the systems and components and will mainly lead to the direct dismantling. In this context, neutron induced activation calculations for the structural components have to be carried out to predict activities in structures and to estimate future costs for conditioning and packaging. To avoid an overestimation of the radioactive inventory and to calculate the expenses for decommissioning as accurate as possible, modern state-of-the-art Monte-Carlo-Techniques (MCNP™) are applied and coupled with present-day activation and decay codes (ORIGEN-S). In this context ADVANTG is used as weight window generator for MCNP™ i. e. as variance reduction tool to speed up the calculation in deep penetration problems. In this paper the calculation procedure is described and the obtained results are presented with a validation along with measured activities and photon dose rates measured in the post-operational phase. The validation shows that the applied calculation procedure is suitable for the determination of the radioactive inventory of a nuclear power plant. Even the measured gamma dose rates in the post-operational phase at different positions in the reactor building agree within a factor of 2 to 3 with the calculation results. The obtained results are accurate and suitable to support effectively the decommissioning planning process.
Lattice dynamics of strontium tungstate
Prabhatasree Goel; R Mittal; S L Chaplot; A K Tyagi
2008-11-01
We report here measurements of the phonon density of states and the lattice dynamics calculations of strontium tungstate (SrWO4). At ambient conditions this compound crystallizes to a body-centred tetragonal unit cell (space group I41/a) called scheelite structure. We have developed transferable interatomic potentials to study the lattice dynamics of this class of compounds. The model parameters have been fitted with respect to the experimentally available Raman and infra-red frequencies and the equilibrium unit cell parameters. Inelastic neutron scattering measurements have been carried out in the triple-axis spectrometer at Dhruva reactor. The measured phonon density of states is in good agreement with the theoretical calculations, thus validating the inter-atomic potential developed.
Jacquet, X.; Casoli, P.; Authier, N.; Rousseau, G. [CEA, Centre de Valduc, 21120 Is-sur-Tille (France); Barsu, C. [Pl. de la fontaine, 25410 Corcelles-Ferrieres (France)
2011-07-01
Caliban is a cylindrical metallic core reactor mainly composed of uranium 235. It is operated by the Criticality and Neutron Science Research Laboratory located at the French Atomic Energy Commission research center in Valduc. As with other fast burst reactors, Caliban is used extensively for determining the responses of electronic parts or other objects and materials to neutron-induced displacements. Therefore, Caliban's irradiation characteristics, and especially its central cavity neutron spectrum, have to be very accurately evaluated. The foil activation method has been used in the past by the Criticality and Neutron Science Research Laboratory to evaluate the neutron spectrum of the different facilities it operated, and in particular to characterize the Caliban cavity spectrum. In order to strengthen and to improve our knowledge of the Caliban cavity neutron spectrum and to reduce the uncertainties associated with the available evaluations, new measurements have been performed on the reactor and interpreted by the foil activation method. A sensor set has been selected to sample adequately the studied spectrum. Experimental measured reaction rates have been compared to the results from UMG spectrum unfolding software and to values obtained with the activation code Fispact. Experimental and simulation results are overall in good agreement, although gaps exist for some sensors. UMG software has also been used to rebuild the Caliban cavity neutron spectrum from activation measurements. For this purpose, a default spectrum is needed, and one has been calculated with the Monte-Carlo transport code Tripoli 4 using the benchmarked Caliban description. (authors)
Devan, K.; Gopalakrishnan, V.; Lee, S.M. [Nuclear Data Section Indira Ganhi Centre for Atomic Research, Tamilnadu (India)
1994-12-31
We have created a 25 group neutron cross section set (IGCJENDL) for nuclides of interest to LMFBRs from the Japanese Evaluated Nuclear Data Library - Version 2 (JENDL-2) in the format of French adjusted Cadarache Version 2 set (1969). The integral validation of IGCJENDL set was done by analyzing nine fast critical assemblies proposed by Cross Section Evaluation Working Group (CSEWG). The calculated integral parameters agreed reasonably well with the reported measured values. It is found that this set predicts the integral parameters, k-eff in particular, close to that predicted by adjusted CARNAVAL IV (French) or BNAB-78 (Russian) sets, for a 1200 MWe theoretical benchmark, representing a large power reactor.
Dual Lattice of ℤ-module Lattice
Futa Yuichi
2017-07-01
Full Text Available In this article, we formalize in Mizar [5] the definition of dual lattice and their properties. We formally prove that a set of all dual vectors in a rational lattice has the construction of a lattice. We show that a dual basis can be calculated by elements of an inverse of the Gram Matrix. We also formalize a summation of inner products and their properties. Lattice of ℤ-module is necessary for lattice problems, LLL(Lenstra, Lenstra and Lovász base reduction algorithm and cryptographic systems with lattice [20], [10] and [19].
Calculation with MCNP of capture photon flux in VVER-1000 experimental reactor.
Töre, Candan; Ortego, Pedro
2005-01-01
The aim of this study is to obtain by Monte Carlo method the high energy photon flux due to neutron capture in the internals and vessel layers of the experimental reactor LR-0 located in REZ, Czech Republic, and loaded with VVER-1000 fuel. The calclated neutron, photon and photon to neutron flux ratio are compared with experimental measurements performed with a multi-parameter stilbene detector. The results show clear underestimation of photon flux in downcomer and some overestimation at vessel surface and 1/4 thickness but a good fitting for deeper points in vessel.
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.
Diego Ferraro
2011-01-01
Full Text Available Monte Carlo neutron transport codes are usually used to perform criticality calculations and to solve shielding problems due to their capability to model complex systems without major approximations. However, these codes demand high computational resources. The improvement in computer capabilities leads to several new applications of Monte Carlo neutron transport codes. An interesting one is to use this method to perform cell-level fuel assembly calculations in order to obtain few group constants to be used on core calculations. In the present work the VTT recently developed Serpent v.1.1.7 cell-oriented neutronic calculation code is used to perform cell calculations of a theoretical BWR lattice benchmark with burnable poisons, and the main results are compared to reported ones and with calculations performed with Condor v.2.61, the INVAP's neutronic collision probability cell code.
Braffort, P.; Chaigne, M. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires
1958-07-01
1) Introduction: The difficulties of the formulation of the equations of phenomena occurring during the operation of a fusion reactor are underlined. 2) The possibilities presented by analog computation of the solution of nonlinear differential equations are enumerated. The accuracy and limitations of this method are discussed. 3) The analog solution in the stationary problem of the measurement of the discharge confinement is given and comparison with experimental results. 4) The analog solution of the dynamic problem of the evolution of the discharge current in a simple case is given and it is compared with experimental data. 5) The analog solution of the motion of an isolated ion in the electromagnetic field is given. A spatial field simulator used for this problem (bidimensional problem) is described. 6) The analog solution of the preceding problem for a tridimensional case for particular geometrical configurations using simultaneously 2 field simulators is given. 7) A method of computation derived from Monte Carlo method for the study of dynamic of plasma is described. 8) Conclusion: the essential differences between the analog computation of fission reactors and fusion reactors are analysed. In particular the theory of control of a fusion reactor as described by SCHULTZ is discussed and the results of linearized formulations are compared with those of nonlinear simulation. (author)Fren. [French] 1) Introduction. On souligne les difficultes que presente la mise en equation des phenomenes mis en jeu lors du fonctionnement d'un reacteur a fusion. On selectionne un certain nombre d'equations generalement utilisees et on montre les impossibilites analytiques auxquelles on se heurte alors. 2) On rappelle les possibilites du calcul analogique pour la resolution des systemes differentiels non lineaires et on indique la precision de la methode ainsi que ses limitations. 3) On decrit esolution analogique du probleme statique de la mesure du confinement de la
Measurement and calculation of fast neutron and gamma spectra in well defined cores in LR-0 reactor.
Košťál, Michal; Matěj, Zdeněk; Cvachovec, František; Rypar, Vojtěch; Losa, Evžen; Rejchrt, Jiří; Mravec, Filip; Veškrna, Martin
2017-02-01
A well-defined neutron spectrum is essential for many types of experimental topics and is also important for both calibration and testing of spectrometric and dosimetric detectors. Provided it is well described, such a spectrum can also be employed as a reference neutron field that is suitable for validating selected cross sections. The present paper aims to compare calculations and measurements of such a well-defined spectra in geometrically similar cores of the LR-0 reactor with fuel containing slightly different enrichments (2%, 3.3% and 3.6%). The common feature to all cores is a centrally located dry channel which can be used for the insertion of studied materials. The calculation of neutron and gamma spectra was realized with the MCNP6 code using ENDF/B-VII.0, JEFF-3.1, JENDL-3.3, ROSFOND-2010 and CENDL-3.1 nuclear data libraries. Only minor differences in neutron and gamma spectra were found in the comparison of the presented reactor cores with different fuel enrichments. One exception is the gamma spectrum in the higher energy region (above 8MeV), where more pronounced variations could be observed.
Sanson, Andrea, E-mail: andrea.sanson@unipd.it [Department of Physics and Astronomy, University of Padova, Padova (Italy); Giarola, Marco; Mariotto, Gino [Department of Computer Science, University of Verona, Verona (Italy); Hu, Lei; Chen, Jun; Xing, Xianran [Department of Physical Chemistry, University of Science and Technology Beijing, Beijing (China)
2016-09-01
Very recently it has been found that CaZrF{sub 6} exhibits a very large and isotropic negative thermal expansion (NTE), even greater than the current most popular NTE materials. In this work, the vibrational dynamics of CaZrF{sub 6} has been investigated by temperature-dependent Raman spectroscopy combined with ab initio calculations. As expected on the basis of the group theory for CaZrF{sub 6}, three Raman-active modes were identified: the F{sub 2g} mode peaked at about 236 cm{sup −1}, the E{sub g} mode at around 550–555 cm{sup −1}, and the A{sub g} mode peaked at about 637 cm{sup −1}. The temperature dependence of their frequencies follows an unusual trend: the F{sub 2g} mode, due to bending vibrations of fluorine atoms in the linear Ca-F-Zr chain, is hardened with increasing temperature, while the A{sub g} mode, corresponding to Ca-F-Zr bond stretching vibrations, is softened. We explain this anomalous behavior by separating implicit and explicit anharmonicity for both F{sub 2g} and A{sub g} modes. In fact, cubic anharmonicity (three-phonon processes) is observed to dominate the higher-frequency A{sub g} phonon-mode, quartic anharmonicity (four-phonon processes) is found to dominate the lower-frequency F{sub 2g} phonon-mode. As a result, the large NTE of CaZrF{sub 6} cannot be accurately predicted through the quasi-harmonic approximation. - Highlights: • A Raman and ab initio study of the lattice dynamics of CaZrF{sub 6} was performed. • All the Raman-active modes expected on the basis of the group theory were identified. • The temperature-dependence of the CaZrF{sub 6} Raman frequencies follows an unusual trend. • Explicit anharmonicity dominates for both F{sub 2g} and A{sub g} Raman modes. • The NTE of CaZrF{sub 6} cannot be accurately predicted by the quasi-harmonic approximation.
Physics parameter calculations for a Tandem Mirror Reactor with thermal barriers
Boghosian, B.M.; Lappa, D.A.; Logan, B.G.
1979-11-06
Thermal barriers are localized reductions in potential between the plugs and the central cell, which effectively insulate trapped plug electrons from the central cell electrons. By then applying electron heating in the plug, it is possible to obtain trapped electron temperatures that are much greater than those of the central cell electrons. This, in turn, effects an increase in the plug potential and central cell confinement with a concomitant decrease in plug density and injection power. Ions trapped in the barrier by collisions are removed by the injection of neutral beams directed inside the barrier cell loss cone; these beam neutrals convert trapped barrier ions to neutrals by charge exchange permitting their escape. We describe a zero-dimensional physics model for this type of reactor, and present some preliminary results for Q.
ENFORM II: a calculational system for light water reactor logistics and effluent analysis
Heeb, C.M.; Lewallen, M.A.; Purcell, W.L.; Cole, B.M.
1979-09-01
ENFORM is a computer-based information system that addresses the material logistics, environmental releases and economics of light water reactor (LWR) operation. The most important system inputs consist of electric energy generation requirements, details of plant construction scheduling, unit costs, and environmental release factors. From these inputs the ENFORM system computes the mass balances and generates the environmental release information for noxious chemicals and radionuclides from various fuel cycle facilities (except waste disposal). Fuel cycle costs and electric power costs are also computed. All code development subsequent to 1977 is summarized. Programming instructions are provided for the modules that are comprised in the ENFORM system. ENGEN, a code that uses a generation schedule specified by the user and isotopic data generated by ORIGEN, has been developed to produce a scenario-specific data base. Other codes (ENMAT, ENRAD, etc) have been developed to use data base information to estimate radioactive and nonradioactive release information.
Analysis of offsite dose calculation methodology for a nuclear power reactor
Moser, Donna Smith [Univ. of North Carolina, Chapel Hill, NC (United States)
1995-01-01
This technical study reviews the methodology for calculating offsite dose estimates as described in the offsite dose calculation manual (ODCM) for Pennsylvania Power and Light - Susquehanna Steam Electric Station (SSES). An evaluation of the SSES ODCM dose assessment methodology indicates that it conforms with methodology accepted by the US Nuclear Regulatory Commission (NRC). Using 1993 SSES effluent data, dose estimates are calculated according to SSES ODCM methodology and compared to the dose estimates calculated according to SSES ODCM and the computer model used to produce the reported 1993 dose estimates. The 1993 SSES dose estimates are based on the axioms of Publication 2 of the International Commission of Radiological Protection (ICRP). SSES Dose estimates based on the axioms of ICRP Publication 26 and 30 reveal the total body estimates to be the most affected.
New Decay Data Sub-library for Calculation of Nuclear Reactors Antineutrino Spectra
Sonzogni, Alejandro; McCutchan, Elizabeth; Johnson, Timothy
2015-10-01
The ENDF/B-VII.1 decay data sub-library contains up-to-date decay properties for all known nuclides and can be used in a wide variety of applications such as decay heat, delayed nu-bar and astrophysics. We have recently completed an upgrade to the ENDF/B-VII.1 decay data sub-library in order to better calculate antineutrino spectra from fission of actinide nuclides. This sub-library has been used to identify the main contributors to the antineutrino spectra as well as to derive a systematic behavior of the energy integrated spectra similar to that of the beta-delayed neutron multiplicities. The main improvements have been the use of the TAGS data from Algora et al and Greenwood et al, as well as some of the single beta spectrum data from Rudstam et al to obtain beta minus level feedings. Additionally, we have calculated the antineutrino spectra for neutron energies higher than thermal, needed for highly-enriched uranium cores, such as the HFIR in ORNL that will be used in the PROSPECT experiment. These calculations are relevant since the high precision beta spectra which are used in many antineutrino calculations were measured at thermal energies. The impact of the fission yield data on these calculations will be discussed. This work was sponsored by the Office of Nuclear Physics, Office of Science of the U.S. Department of Energy under Contract No. DE-AC02-98CH10886.
Rousseau, G.; Chambru, L.; Authier, N. [CEA, Centre de Valduc, 21120 Is-sur-Tille, (France)
2015-07-01
In the context of criticality accident alarm system tests, several experiments were carried out in 2013 on the PROSPERO reactor to study the response to neutron and gamma of different devices and dosimeters, particularly on the SNAC2 dosimeter. This article presents the results of this criticality dosimeter in different configurations, and compares the experimental measurements with the results of calculation performed with the TRIPOLI-4 Monte-Carlo Neutral Particles transport code. PROSPERO is a metallic critical assembly managed by the Criticality, Neutron Science and Measurement Department located at the French CEA Research Center of Valduc. The core, surrounded by a reflector of depleted uranium, is composed of 2 horizontal cylindrical blocks made of a highly enriched uranium alloy which can be placed in contact, and of 4 depleted uranium control rods which allow the reactor to be driven. This reactor, placed in a cell 10 m x 8 m x 6 m high, with 1.4-meter-thick concrete walls, is used as a fast neutron spectrum source and is operated at stable power level in delayed critical state, which can vary from 3 mW to 3 kW. PROSPERO is extensively used for electronic hardening or to study the effect of the neutrons on various materials. The SNAC2 criticality dosimeter is a zone dosimeter allowing the off line measurement of criticality accident neutron doses. This dosimeter consists of the pile up of seven activation foils embedded into a 23 mm diameter x 21 mm height cadmium container. The activation measurement of each foil, using a gamma spectroscopy technique, gives information about the neutron reaction rates. The SNAC2 software allows the spectrum unfolding from these values, taking into account the hypothesis of a particular spectrum shape, in three components: a Maxwell spectrum component for the thermal range, a 1/E component for the epithermal range, and a Watt spectrum component for the high energy range. Moreover, from the neutron spectrum, the SNAC
Radiation Transport Calculation of the UGXR Collimators for the Jules Horowitz Reactor (JHR)
Chento, Yelko; Hueso, César; Zamora, Imanol; Fabbri, Marco; Fuente, Cristina De La; Larringan, Asier
2017-09-01
Jules Horowitz Reactor (JHR), a major infrastructure of European interest in the fission domain, will be built and operated in the framework of an international cooperation, including the development and qualification of materials and nuclear fuel used in nuclear industry. For this purpose UGXR Collimators, two multi slit gamma and X-ray collimation mechatronic systems, will be installed at the JHR pool and at the Irradiated Components Storage pool. Expected amounts of radiation produced by the spent fuel and X-ray accelerator implies diverse aspects need to be verified to ensure adequate radiological zoning and personnel radiation protection. A computational methodology was devised to validate the Collimators design by means of coupling different engineering codes. In summary, several assessments were performed by means of MCNP5v1.60 to fulfil all the radiological requirements in Nominal scenario (TEDE < 25µSv/h) and in Maintenance scenario (TEDE < 2mSv/h) among others, detailing the methodology, hypotheses and assumptions employed.
Radiation Transport Calculation of the UGXR Collimators for the Jules Horowitz Reactor (JHR
Chento Yelko
2017-01-01
Full Text Available Jules Horowitz Reactor (JHR, a major infrastructure of European interest in the fission domain, will be built and operated in the framework of an international cooperation, including the development and qualification of materials and nuclear fuel used in nuclear industry. For this purpose UGXR Collimators, two multi slit gamma and X-ray collimation mechatronic systems, will be installed at the JHR pool and at the Irradiated Components Storage pool. Expected amounts of radiation produced by the spent fuel and X-ray accelerator implies diverse aspects need to be verified to ensure adequate radiological zoning and personnel radiation protection. A computational methodology was devised to validate the Collimators design by means of coupling different engineering codes. In summary, several assessments were performed by means of MCNP5v1.60 to fulfil all the radiological requirements in Nominal scenario (TEDE < 25µSv/h and in Maintenance scenario (TEDE < 2mSv/h among others, detailing the methodology, hypotheses and assumptions employed.
Comparison of statistical evaluation of criticality calculations for reactors VENUS-F and ALFRED
Janczyszyn Jerzy
2017-01-01
Full Text Available Limitations of correct evaluation of keff in Monte Carlo calculations, claimed in literature, apart from the nuclear data uncertainty, need to be addressed more thoroughly. Respective doubts concern: the proper number of discarded initial cycles, the sufficient number of neutrons in a cycle and the recognition and dealing with the keff bias. Calculations were performed to provide more information on these points with the use of the MCB code, solely for fast cores. We present applied methods and results, such as: calculation results for stability of variance, relation between standard deviation reported by MCNP and this from the dispersion of multiple independent keff values, second order standard deviations obtained from different numbers of grouped results. All obtained results for numbers of discarded initial cycles from 0 to 3000 were analysed leading for interesting conclusions.
Haerer, Bastian; Prof. Dr. Schmidt, Ruediger; Dr. Holzer, Bernhard
Following the recommendations of the European Strategy Group for High Energy Physics, CERN launched the Future Circular Collider Study (FCC) to investigate the feasibility of large-scale circular colliders for future high energy physics research. This thesis presents the considerations taken into account during the design process of the magnetic lattice in the arc sections of the electron-positron version FCC-ee. The machine is foreseen to operate at four different centre-of-mass energies in the range of 90 to 350 GeV. Different beam parameters need to be achieved for every energy, which requires a flexible lattice design in the arc sections. Therefore methods to tune the horizontal beam emittance without re-positioning machine components are implemented. In combination with damping and excitation wigglers a precise adjustment of the emittance can be achieved. A very first estimation of the vertical emittance arising from lattice imperfections is performed. Special emphasis is put on the optimisation of the ...
Liang, Jingang; Wang, Kan; Qiu, Yishu [Dept. of Engineering Physics, LiuQing Building, Tsinghua University, Beijing (China); Chai, Xiao Ming; Qiang, Sheng Long [Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu (China)
2016-06-15
Because of prohibitive data storage requirements in large-scale simulations, the memory problem is an obstacle for Monte Carlo (MC) codes in accomplishing pin-wise three-dimensional (3D) full-core calculations, particularly for whole-core depletion analyses. Various kinds of data are evaluated and quantificational total memory requirements are analyzed based on the Reactor Monte Carlo (RMC) code, showing that tally data, material data, and isotope densities in depletion are three major parts of memory storage. The domain decomposition method is investigated as a means of saving memory, by dividing spatial geometry into domains that are simulated separately by parallel processors. For the validity of particle tracking during transport simulations, particles need to be communicated between domains. In consideration of efficiency, an asynchronous particle communication algorithm is designed and implemented. Furthermore, we couple the domain decomposition method with MC burnup process, under a strategy of utilizing consistent domain partition in both transport and depletion modules. A numerical test of 3D full-core burnup calculations is carried out, indicating that the RMC code, with the domain decomposition method, is capable of pin-wise full-core burnup calculations with millions of depletion regions.
Grabaskas, David [Argonne National Lab. (ANL), Argonne, IL (United States); Bucknor, Matthew [Argonne National Lab. (ANL), Argonne, IL (United States); Jerden, James [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, Acacia J. [Argonne National Lab. (ANL), Argonne, IL (United States)
2016-02-01
The overall objective of the SFR Regulatory Technology Development Plan (RTDP) effort is to identify and address potential impediments to the SFR regulatory licensing process. In FY14, an analysis by Argonne identified the development of an SFR-specific MST methodology as an existing licensing gap with high regulatory importance and a potentially long lead-time to closure. This work was followed by an initial examination of the current state-of-knowledge regarding SFR source term development (ANLART-3), which reported several potential gaps. Among these were the potential inadequacies of current computational tools to properly model and assess the transport and retention of radionuclides during a metal fuel pool-type SFR core damage incident. The objective of the current work is to determine the adequacy of existing computational tools, and the associated knowledge database, for the calculation of an SFR MST. To accomplish this task, a trial MST calculation will be performed using available computational tools to establish their limitations with regard to relevant radionuclide release/retention/transport phenomena. The application of existing modeling tools will provide a definitive test to assess their suitability for an SFR MST calculation, while also identifying potential gaps in the current knowledge base and providing insight into open issues regarding regulatory criteria/requirements. The findings of this analysis will assist in determining future research and development needs.
Malouch, F. [Saclay Center CEA, DEN/DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette Cedex (France)
2011-07-01
The control of the temperature in material samples irradiated in a material testing reactor requires the knowledge of the nuclear heating caused by the energy deposition by neutrons and photons interacting in the irradiation device structures. Thus, a neutron-photonic three-dimensional calculation scheme has been developed to evaluate the nuclear heating in experimental devices irradiated in the core of the OSIRIS MTR reactor (CEA/Saclay Center). The aim is to obtain a predictive tool for the nuclear heating estimation in irradiation devices. This calculation scheme is mainly based on the TRIPOLI-4 three-dimensional continuous-energy Monte Carlo transport code, developed by CEA (Saclay Center). An experimental validation has been carried out on the basis of nuclear heating measurements performed in the OSIRIS core. After an overview of the experimental devices irradiated in the OSIRIS reactor, we present the calculation scheme and the first results of the experimental validation. (authors)
Atkinson, D; van Steenwijk, F.J.
The resistance between two arbitrary nodes in an infinite square lattice of:identical resistors is calculated, The method is generalized to infinite triangular and hexagonal lattices in two dimensions, and also to infinite cubic and hypercubic lattices in three and more dimensions. (C) 1999 American
Coarse mesh methods for the transport calculation in the CRONOS reactor code
Fedon-Magnaud, C.; Lautard, J.J.; Akherraz, B.; Wu, G.J. [Commissariat a l`Energie Atomique, Gif sur Yvette (France)
1995-12-31
Homogeneous transport methods have been recently implemented in the kinetic code CRONOS dedicated mainly to PWR calculations. Two different methods are presented. The first one is based on the even parity flux formalism and uses finite element spatial discretization and a discrete ordinates angular approximation; the treatment of the anisotropic scattering is described in detail. The second method uses the odd flux as the main unknown, it is closely connected to nodal methods. This method is used to solve two different problems, the simplified PN equations and the exact transport equation using an angular PN expansion. Numerical results are presented for some standard benchmarks and the methods are compared.
Coarse mesh methods for the transport calculation in the Cronos reactor code
Fedon-Magnaud, C.; Lautard, J.J.; Akherraz, B.; Wu, G.J.
1995-12-31
Homogeneous transports methods have been recently implemented in the kinetic code CRONOS dedicated mainly to PWR calculations. Two different methods are presented. The first one is based on the even parity flux formalism and uses finite element spatial discretization and a discrete ordinates angular approximation; the treatment of the anisotropic scattering is described in detail. The second method uses the odd flux as the main unknown, it is closely to nodal methods. This method is used to solve different problems, the simplified PN equations and the exact transport equation using an angular PN expansion. Numerical results are presented for some standard benchmarks and the method are compared. (authors). 18 refs., 3 tabs.
Jin, Luchang; Christ, Norman; Hayakawa, Masashi; Izubuchi, Taku; Lehner, Christoph
2015-01-01
The anomalous magnetic moment of muon, $g-2$, is a very precisely measured quantity. However, the current measurement disagrees with standard model by about 3 standard deviations. Hadronic vacuum polarization and hadronic light by light are the two types of processes that contribute most to the theoretical uncertainty. I will describe how lattice methods are well-suited to provide a first-principle's result for the hadronic light by light contribution, the various numerical strategies that are presently being used to evaluate it, our current results and the important remaining challenges which must be overcome.
Osmera, B; Cvachovec, F; Kyncl, J; Smutný, V
2005-01-01
The fast-neutron and photon space-energy distributions have been measured in an axially (1.25 m active height) and azimuthally (60 degree symmetry sector) shortened model of the WWER-1000 reactor assembled in the LR-0 experimental reactor. The space-energy distributions have been calculated with the stochastic code MCNP and the deterministic three-dimensional code TORT. Selected results are presented and discussed in the paper. This work has been done in the frame of the EU 5th FW project REDOS REDOS, Reactor Dosimetry: Accurate determination and benchmarking of radiation field parameters, relevant for reactor pressure vessel monitoring. EURATOM Programme, Call 2000/C 294/04). All geometry and material composition data of the model as well as the available experimental data were carefully checked and revised.
Nagaya, Yasunobu
2014-06-01
The methods to calculate the kinetics parameters of βeff and Λ with the differential operator sampling have been reviewed. The comparison of the results obtained with the differential operator sampling and iterated fission probability approaches has been performed. It is shown that the differential operator sampling approach gives the same results as the iterated fission probability approach within the statistical uncertainty. In addition, the prediction accuracy of the evaluated nuclear data library JENDL-4.0 for the measured βeff/Λ and βeff values is also examined. It is shown that JENDL-4.0 gives a good prediction except for the uranium-233 systems. The present results imply the need for revisiting the uranium-233 nuclear data evaluation and performing the detailed sensitivity analysis.
Jacobsen, C.J.H.; Dahl, Søren; Boisen, A.
2002-01-01
For ammonia synthesis catalysts a volcano-type relationship has been found experimentally. We demonstrate that by combining density functional theory calculations with a microkinetic model the position of the maximum of the volcano curve is sensitive to the reaction conditions. The catalytic...... ammonia synthesis activity, to a first approximation, is a function only of the binding energy of nitrogen to the catalyst. Therefore, it is possible to evaluate which nitrogen binding energy is optimal under given reaction conditions. This leads to the concept of optimal catalyst curves, which illustrate...... the nitrogen binding energies of the optimal catalysts at different temperatures, pressures, and synthesis gas compositions. Using this concept together with the ability to prepare catalysts with desired binding energies it is possible to optimize the ammonia process. In this way a link between first...
Gauntt, Randall O.; Powers, Dana Auburn; Ashbaugh, Scott G.; Leonard, Mark Thomas; Longmire, Pamela
2010-04-01
In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs2MoO4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU
Zizin, M. N.; Ivanov, L. D.
2013-12-01
In the present paper, an attempt is made to analyze the accuracy of calculating the effectiveness of the VVER-1000 reactor scram system by means of the inverted solution of the kinetics equation (ISKE). In the numerical studies in the intellectual ShIPR software system, the actuation of the reactor scram system with the possible jamming of one of the two most effective rods is simulated. First, the connection of functionals calculated in the space-time computation in different approximations with the kinetics equation is considered on the theoretical level. The formulas are presented in a manner facilitating their coding. Then, the results of processing of several such functions by the ISKE are presented. For estimating the effectiveness of the VVER-1000 reactor scram system, it is proposed to use the measured currents of ionization chambers (IC) jointly with calculated readings of IC imitators. In addition, the integral of the delayed neutron (DN) generation rate multiplied by the adjoint DN source over the volume of the reactor, calculated for the instant of time when insertion of safety rods ends, is used. This integral is necessary for taking into account the spatial reactivity effects. Reasonable agreement was attained for the considered example between the effectiveness of the scram system evaluated by this method and the values obtained by steady-state calculations as the difference of the reciprocal effective multiplication factors with withdrawn and inserted control rods. This agreement was attained with the use of eight-group DN parameters.
Improving accuracy of the calculation of in-core power distributions for light water reactors
Tsuiki, M.; Beere, W.H. (Institute for energy technology, OECD Halden Reactor Project (Norway))
2009-10-15
Comparisons have been made of VNEM prototype system to the measured data obtained from Ringhals unit 3 NPP at its beginning of life, hot-stand-by state. Three cases with difference control rod bank positions and Boron concentrations have been compared: Case 1: nearly all rod banks withdrawn, Boron = 1315 ppm Case 2: bank C = nearly half-inserted, bank D = fully inserted, Boron = 1131 ppm Case 3: banks C and D = fully inserted, Boron = 1060 ppm The results can be summarized as: error: maximum detector reading (%) error: keff (%) Case 1 -2.1 -0.175 Case 2 1.5 -0.022 Case 3 -0.5 -0.044 Excellent agreement was observed in the comparison of the neutron detector readings and the core eigenvalues. The method of core modelling and parameters used in calculation of VNEM is completely the same as the 'PWR standard option' determined from similar comparisons of VNEM and other PWRs. No empirical, or any sort of adjustment was done. (author)
Shahzad, Munir; Sengupta, Pinaki
2017-08-01
We study the Shastry-Sutherland Kondo lattice model with additional Dzyaloshinskii-Moriya (DM) interactions, exploring the possible magnetic phases in its multi-dimensional parameter space. Treating the local moments as classical spins and using a variational ansatz, we identify the parameter ranges over which various common magnetic orderings are potentially stabilized. Our results reveal that the competing interactions result in a heightened susceptibility towards a wide range of spin configurations including longitudinal ferromagnetic and antiferromagnetic order, coplanar flux configurations and most interestingly, multiple non-coplanar configurations including a novel canted-flux state as the different Hamiltonian parameters like electron density, interaction strengths and degree of frustration are varied. The non-coplanar and non-collinear magnetic ordering of localized spins behave like emergent electromagnetic fields and drive unusual transport and electronic phenomena.
Calculation of lattice dynamics, elastic and dielectric properties of γ-BiB3O6 and δ-BiB3O6
Pavlovskii, M. S.; Shinkorenko, A. S.; Zinenko, V. I.
2015-04-01
The crystal lattice vibration frequencies, densities of phonon states, elastic moduli, and high-frequency permittivities have been calculated in terms of the density functional theory method for two polymorphs γ-BiB3O6 and δ-BiB3O6. Based on the calculated densities of phonon states, the temperature dependences of the free energies of two considered bismuth triborate modifications have been constructed, and the temperature of the phase transition between these modifications has been determined (1100 K). The structure of a possible nonpolar praphase of δ-BiB3O6 has been proposed. The polarization of δ-BiB3O6 has been estimated as 131 μC/cm2.
Caldeira, Alexandre D. [Centro Tecnico Aeroespacial (CTA-IEAv), Sao Jose dos Campos, SP (Brazil). Inst. de Estudos Avancados
2000-07-01
This work deals with the cell neutron spectra calculated with the transport equation for an infinite medium applied to the homogenized cell. Considering a radioisotope production reactor fuel cell, as a sample case, the maximum deviation found between the approximated and the S{sub N} methods was 13%. (author)
Francois L, J.L.; Guzman A, J.R. [UNAM-FI, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, Jiutepec, Morelos (Mexico)]. e-mail: jlfl@fi-b.unam.mx
2006-07-01
In this work a comparison of the obtained results with the HELIOS code is made and those obtained by other similar codes, used in the international community, respect to the transmutation of smaller actinides. For this the one it is analyzed the international benchmark: 'Calculations of Different Transmutation Concepts', of the Nuclear Energy Agency. In this benchmark two cell types are analyzed: one small corresponding to a PWR standard, and another big one corresponding to a PWR highly moderated. Its are considered two types of burnt of discharge: 33 GWd/tHM and 50 GWd/tHM. The following types of results are approached: the k{sub eff} like a function of the burnt one, the atomic densities of the main isotopes of the actinides, the radioactivities in the moment in that the reactor it is off and in the times of cooling from 7 up to 50000 years, the reactivity by holes and the Doppler reactivity. The results are compared with those obtained by the following institutions: FZK (Germany), JAERI (Japan), ITEP (Russia) and IPPE (Russian Federation). In the case of the eigenvalue, the obtained results with HELIOS showed a discrepancy around 3% {delta}k/k, which was also among other participants. For the isotopic concentrations: {sup 241}Pu, {sup 242} Pu and {sup 242m} Am the results of all the institutions present a discrepancy bigger every time, as the burnt one increases. Regarding the activities, the discrepancy of results is acceptable, except in the case of the {sup 241} Pu. In the case of the Doppler coefficients the discrepancy of results is acceptable, except for the cells with high moderation; in the case of the holes coefficients, the discrepancy of results increases in agreement with the holes fraction increases, being quite high to 95% of holes. In general, the results are consistent and in good agreement with those obtained by all the participants in the benchmark. The results are inside of the established limits by the work group on Plutonium Fuels
Borodkin Pavel
2016-01-01
Full Text Available Paper describes the new approach of fitting axial fuel burn-up patterns in peripheral fuel assemblies of VVER-1000 type reactors, on the base of ex-core neutron leakage measurements, neutron-physical calculations and in-core SPND measured data. The developed approach uses results of new ex-vessel measurements on different power units through different reactor cycles and their uncertainties to clear the influence of a fitted fuel burn-up profile to the RPV neutron fluence calculations. The new methodology may be recommended to be included in the routine fluence calculations used in RPV lifetime management and may be taken into account during VVER-1000 core burn-up pattern correction.
Borodkin, Pavel; Borodkin, Gennady; Khrennikov, Nikolay
2016-02-01
Paper describes the new approach of fitting axial fuel burn-up patterns in peripheral fuel assemblies of VVER-1000 type reactors, on the base of ex-core neutron leakage measurements, neutron-physical calculations and in-core SPND measured data. The developed approach uses results of new ex-vessel measurements on different power units through different reactor cycles and their uncertainties to clear the influence of a fitted fuel burn-up profile to the RPV neutron fluence calculations. The new methodology may be recommended to be included in the routine fluence calculations used in RPV lifetime management and may be taken into account during VVER-1000 core burn-up pattern correction.
Proskuryakov, K.N. (Moskovskij Ehnergeticheskij Inst. (USSR))
1983-03-01
Mathematical models are proposed for calculating acoustic oscillation resonance frequencies in the coolant in various components of the WWER type primary circuit (core, steam generator, pressurizer, piping). Due to the correspondence between model calculations and experimental results obtained in operating nuclear power plants, the developed models can be used for practical calculations. The possibility of calculating the eigenfrequencies of the coolant oscillation under different operating conditions leads to the interpretation of operational data, to the analysis of operational conditions, to the detection of coolant boiling in the reactor, and to design changes in order to prevent resonance oscillations within the coolant.
Lautard, J.J.; Magnaud, C.; Moreau, F.; Baudron, A.M. [CEA Saclay, Dept. de Mecanique et de Technologie (DMT/SERMA), 91 - Gif-sur-Yvette (France)
1999-07-01
The CRONOS2 software is that part of the SAPHYR code system dedicated to neutronic core calculations. CRONOS2 is a powerful tool for reactor design, fuel management and safety studies. Its modular structure and great flexibility make CRONOS2 an unique simulation tool for research and development for a wide variety of reactor systems. CRONOS2 is a versatile tool that covers a large range of applications from very fast calculations used in training simulators to time and memory consuming reference calculations needed to understand complex physical phenomena. CRONOS2 has a procedure library named CPROC that allows the user to create its own application environment fitted to a specific industrial use. (authors)
Stefani, Giovanni Laranjo de; Conti, Thadeu das Neves; Fedorenko, Giuliana G.; Castro, Vinicius A.; Maio, Mireia F., E-mail: gstefani@ipen.b, E-mail: tnconti@ipen.b, E-mail: g.fedorenko@ipen.b, E-mail: vcastro@ipen.b, E-mail: mfmaio@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Santos, Thiago Augusto dos, E-mail: tsantos@ipen.b [Universidade de Sao Paulo (IFUSP), Sao Paulo, SP (Brazil). Inst. de Fisica
2011-07-01
This work objective was to create a manager program that would automate the programs and computer codes in use for neutronic calculation and thermo-hydraulic in IEA-R1 reactor thus making the process for calculation of safety parameters and for configuration change up to 98% faster than that used in the reactor today. This process was tested in combination with the reactor operators and is being implemented by the quality department. The main codes and programs involved in the calculations of configuration change are Leopard, Hammier-Technion, Twodb, Citation and Cobra. Calculations of delayed neutron and criticality coefficients given in the process of safety parameters calculation are given by the Hammer-Technion and Citation in a process that involves about eleven repetitions so that it meets all the necessary conditions (such different temperatures of the moderator and fuel). The results are entirely consistent with the expected and absolutely the same as those given by manual process. Thus the work shows its reliability as well the advantage of saving time, once a process that could take up to four hours was turned in one that takes around five minutes when done in a home computer. Much of this advantage is due to the fact that were created subprograms to treat the output of each program used and transform them into the input of the other programs, removing from it the intermediate essential data for this to occur, thus avoiding also a possible human error by handling the various data supplied. (author)
Cossu, Guido; Hashimoto, Shoji; Kaneko, Takashi; Noaki, Jun-Ichi
2016-01-01
We compute the chiral condensate in 2+1-flavor QCD through the spectrum of low-lying eigenmodes of Dirac operator. The number of eigenvalues of the Dirac operator is evaluated using a stochastic method with an eigenvalue filtering technique on the background gauge configurations generated by lattice QCD simulations including the effects of dynamical up, down and strange quarks described by the Mobius domain-wall fermion formulation. The low-lying spectrum is related to the chiral condensate, which is one of the leading order low-energy constants in chiral effective theory, as dictated by the Banks-Casher relation. The spectrum shape and its dependence on the sea quark masses calculated in numerical simulations are consistent with the expectation from one-loop chiral perturbation theory. After taking the chiral limit as well as the continuum limit using the data at three lattice spacings ranging 0.080-0.045 fm, we obtain $\\Sigma^{1/3}$(2 GeV) = 270.0(4.9) MeV, with the error combining those from statistical an...
Slater, C.O.
1990-07-01
Results are reported for two-dimensional discrete ordinates, X-Y geometry calculations performed for seven Halden Heavy Boiling Water Reactor core configurations. The calculations were performed in support of an effort to reassess the neutron fluence received by the reactor vessel. Nickel foil measurement data indicated considerable underprediction of fluences by the previously used multigroup removal- diffusion method. Therefore, calculations by a more accurate method were deemed appropriate. For each core configuration, data are presented for (1) integral fluxes in the core and near the vessel wall, (2) neutron spectra at selected locations, (3) isoflux contours superimposed on the geometry models, (4) plots of the geometry models, and (5) input for the calculations. The initial calculations were performed with several mesh sizes. Comparisons of the results from these calculations indicated that the uncertainty in the calculated fluxes should be less than 10%. However, three-dimensional effects (such as axial asymmetry in the fuel loading) could contribute to much greater uncertainty in the calculated neutron fluxes. 7 refs., 22 figs., 11 tabs.
Mazufri, Claudio M. [Investigacion Aplicada SE (INVAP), San Carlos de Bariloche (Argentina)
1995-12-31
According to the nuclear reactors characteristics, can be found different methodologies to appraise the thermal margin available in the core. In the particular case of the CAREM (25 MWe) reactor, where the core is cooled by low mass flux and there are zones with positive steam quality, such evaluation is critical. Due to these characteristics, it was necessary to develop one proper methodology. In the present work, the different steps of that development are described: the election of figures of merit for measure the thermal margin, the hypothesis to use, the election of the critical heat flux prediction model, model qualification and the specification of the core wide procedure. In each step assume criteria are discussed. (author). 9 refs, 1 tab, 1 fig.
Chernyshev, V. A.; Petrov, V. P.; Nikiforov, A. E.; Zakir'yanov, D. O.
2015-06-01
The effect of hydrostatic compression on the lattice structure and dynamics of elpasolites Cs2NaYbF6 and Cs2NaYF6 (sp. gr. 225) has been investigated ab initio. The frequencies and types of fundamental oscillations are determined, and elastic constants are calculated. The computation is performed within the molecular orbitals-linear combinations of atomic orbitals (MO LCAO) approach using the density functional theory (DFT) method with hybrid functionals B3LYP and PBE0 in the CRYSTAL09 program. The rare-earth ion was described by representing the inner (in particular, 4 f) orbitals in the form of a pseudopotential. The outer 5 s and 5 p orbitals, which determine chemical bonding, were described using valence basis sets.
Aleshin, Sergey S.; Gorodkov, Sergey S.; Shcherenko, Anna I. [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation)
2016-09-15
A burn-up calculation of large systems by Monte-Carlo code (MCU) is complex process and it requires large computational costs. Previously prepared isotopic compositions are proposed to be used for the Monte-Carlo code calculations of different system states with burnt fuel. Isotopic compositions are calculated by an approximation method. The approximation method is based on usage of a spectral functionality and reference isotopic compositions, that are calculated by the engineering codes (TVS-M, BIPR-7A and PERMAK-A). The multiplication factors and power distributions of FAs from a 3-D reactor core are calculated in this work by the Monte-Carlo code MCU using earlier prepared isotopic compositions. The separate conditions of the burnt core are observed. The results of MCU calculations were compared with those that were obtained by engineering codes.
Surkov, A. V., E-mail: surkov.andrew@gmail.com; Kochkin, V. N.; Pesnya, Yu. E.; Nasonov, V. A.; Vihrov, V. I.; Erak, D. Yu. [National Research Center Kurchatov Institute (Russian Federation)
2015-12-15
A comparison of measured and calculated neutronic characteristics (fast neutron flux and fission rate of {sup 235}U) in the core and reflector of the IR-8 reactor is presented. The irradiation devices equipped with neutron activation detectors were prepared. The determination of fast neutron flux was performed using the {sup 54}Fe (n, p) and {sup 58}Ni (n, p) reactions. The {sup 235}U fission rate was measured using uranium dioxide with 10% enrichment in {sup 235}U. The determination of specific activities of detectors was carried out by measuring the intensity of characteristic gamma peaks using the ORTEC gamma spectrometer. Neutron fields in the core and reflector of the IR-8 reactor were calculated using the MCU-PTR code.
Medina C, D.; Hernandez A, P. L.; Hernandez D, V. M.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico); Sajo B, L., E-mail: dmedina_c@hotmail.com [Universidad Simon Bolivar, Laboratorio de Fisica Nuclear, Apdo. Postal 89000, Caracas 1080A (Venezuela, Bolivarian Republic of)
2015-10-15
This paper describes a heterogeneous subcritical nuclear reactor with molten salts based on thorium, with graphite moderator and a source of {sup 252}Cf, whose dose levels in the periphery allows its use in teaching and research activities. The design was done by the Monte Carlo method with the code MCNP5 where the geometry, dimensions and fuel was varied in order to obtain the best design. The result is a cubic reactor of 110 cm side with graphite moderator and reflector. In the central part they have 9 ducts that were placed in the direction of axis Y. The central duct contains the source of {sup 252}Cf, of 8 other ducts, are two irradiation ducts and the other six contain a molten salt ({sup 7}LiF - BeF{sub 2} - ThF{sub 4} - UF{sub 4}) as fuel. For design the k{sub eff}, neutron spectra and ambient dose equivalent was calculated. In the first instance the above calculation for a virgin fuel was called case 1, then a percentage of {sup 233}U was used and the percentage of Th was decreased and was called case 2. This with the purpose to compare two different fuels working inside the reactor. In the case 1 a value was obtained for the k{sub eff} of 0.13 and case 2 of 0.28, maintaining the subcriticality in both cases. In the dose levels the higher value is in case 2 in the axis Y with a value of 3.31 e-3 ±1.6% p Sv/Q this value is reported in for one. With this we can calculate the exposure time of personnel working in the reactor. (Author)
Barranco R, F.
2015-07-01
In this thesis criticality and shielding calculations to evaluate the design of a container of dry storage of spent nuclear fuel generated in research reactors were made. The design of such container was originally proposed by Argentina and Brazil, and the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico. Additionally, it is proposed to modify the design of this container to store spent fuel 120 that are currently in the pool of TRIGA Mark III reactor, the Nuclear Center of Mexico and calculations and analyzes are made to verify that the settlement of these fuel elements is subcritical limits and dose rates to workers and the general public are not exceeded. These calculations are part of the design criteria for security protection systems in dry storage system (Dss for its acronym in English) proposed by the Nuclear Regulatory Commission (NRC) of the United States. To carry out these calculations simulation codes of Monte Carlo particle transport as MCNPX and MCNP5 were used. The initial design (design 1) 78 intended to store spent fuel with a maximum of 115. The ININ has 120 fuel elements and spent 3 control rods (currently stored in the reactor pool). This leads to the construction of two containers of the original design, but for economic reasons was decided to modify (design 2) to store in a single container. Criticality calculations are performed to 78, 115 and fresh fuel elements 124 within the container, to the two arrangements described in Chapter 4, modeling the three-dimensional geometry assuming normal operating conditions and accident. These calculations are focused to demonstrate that the container will remain subcritical, that is, that the effective multiplication factor is less than 1, in particular not greater than 0.95 (as per specified by the NRC). Spent fuel 78 and 124 within the container, both gamma radiation to neutron shielding calculations for only two cases were simulated. First actinides and fission products generated
Jing Xingqing E-mail: jingxq@d103.inet.tsinghua.edu.cn; Xu Xiaolin; Yang Yongwei; Qu Ronghong
2002-10-01
The 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) is a pebble bed experimental reactor built by the Institute of Nuclear Energy Technology (INET), Tsinghua University. This paper introduces the first critical prediction calculations and the experiments for the HTR-10. The German VSOP neutronics code is used for the prediction calculations of the first loading. The characteristics of pebble-bed high temperature gas-cooled reactors are taken into account, including the double heterogeneity of the fuel element, the buckling feedback of the spectrum calculation, the effect of the mixture of fuel elements and graphite balls, and the correction of the diffusion coefficients in the upper cavity based on transport theory. Also considered are the effects of impurities in the fuel elements, in the graphite balls and in the reflector graphite on the reactivity. The number of fuel elements and graphite balls in the initial core is predicted to provide reference for the first criticality experiment. The critical experiment adopts a method of extrapolating to approach criticality. The first criticality was attained on December 1, 2000. The first criticality experiment shows that the predicted critical number of the fuel elements and graphite balls is in close agreement with the experimental results. Their relative error is less than 1.0%, implying the physical predictions and the results of the criticality experiment are much beyond expectations.
Mitenkova, E. F.; Novikov, N. V. [Nuclear Safety Inst. of Russian Academy of Sciences, B. Tulskaya 52, Moscow, 115119 (Russian Federation); Blokhin, A. I. [State Scientific Center of Russian Federation, Inst. of Physics and Power Engineering Named after A.I. Leypunsky, Bondarenko Square 1, Obninsk, Kaluga Region, 249030 (Russian Federation)
2012-07-01
The different (U-Pu) fuel compositions are considered for next generation of sodium fast breeder reactors. The considerable discrepancies in axial and radial neutron spectra for hybrid reactor systems compared to the cores with UO{sub 2} fuel cause increasing uncertainty of generating the group nuclear constants in those reactor systems. The calculation results of BFS-62-3A critical assembly which is considered as full-scale model of BN-600 hybrid core with steel reflector specify quite different spectra in local areas. For those systems the MCNP 5 calculations demonstrate significant sensitivity of effective multiplication factor K{sub eff} and spectral indices to nuclear data libraries. For {sup 235}U, {sup 238}U, {sup 239}Pu the results of calculated radial fission rate distributions against the reconstructed ones are analyzed. Comparative analysis of spectral indices, neutron spectra and radial fission rate distributions are performed using the different versions of ENDF/B, JENDL-3.3, JENDL-4, JEFF-3.1.1 libraries and BROND-3 for Fe, Cr isotopes. For analyzing the fission rate sensitivity to the plutonium presence in the fuel {sup 239}Pu is substituted for {sup 235}U (enrichment 90%) in the FA areas containing the plutonium. For {sup 235}U, {sup 238}U, {sup 239}Pu radial fission rate distributions the explanation of pick values discrepancies is based on the group fission constants analyses and possible underestimation of some features at the experimental data recovery method (Westcott factors, temperature dependence). (authors)
Malouch, Fadhel [Alternative Energies and Atomic Energy Commission - CEA, Saclay Center, DEN/DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette Cedex (France)
2015-07-01
Technological irradiations carried out in material testing reactors (MTRs) are used to study the behavior of materials under irradiation conditions required by different types of nuclear power plants (NPPs). For MTRs, specific instrumentation is required for the experiment monitoring and for the characterization of irradiation conditions, in particular the flux of neutrons and photons. To measure neutron and photon flux in experimental locations, different sensors can be used, such as SPNDs (self-powered neutron detectors), SPGDs (self-powered gamma detectors) and ionization chambers. These sensors involve interactions producing ultimately a measurable electric current. Various sensors have been recently tested in the core periphery of the OSIRIS reactor (located at the CEA-Saclay center) in order to qualify their responses to the neutron and the photon flux. One of the key input data for this qualification is to have a relevant evaluation of neutron and gamma fluxes at the irradiation location. The objective of this work is to evaluate the neutron and the gamma flux in the core periphery of the OSIRIS reactor. With this intention, specific neutron-photonic three-dimensional calculations have been performed and are mainly based on the TRIPOLI-4{sup R} three-dimensional continuous-energy Monte Carlo code, developed by CEA (Saclay Center) and extensively validated against reactor dosimetry benchmarks. In the case of the OSIRIS reactor, TRIPOLI-4{sup R} code has been validated against experimental results based on neutron flux and nuclear heating measurements performed in ex-core and in-core experiments. In this work, simultaneous contribution of neutrons and gamma photons in the core periphery is considered using neutron-photon coupled transport calculations. Contributions of prompt and decay photons have been taken into account for the gamma flux calculation. Specific depletion codes are used upstream to provide the decay-gamma sources required by TRIPOLI-4
Feng, Xuan-Kai; Shi, Siqi; Shen, Jian-Yun; Shang, Shun-Li; Yao, Mei-Yi; Liu, Zi-Kui
2016-10-01
Since Zr-Fe-Sn is one of the key ternary systems for cladding and structural materials in nuclear industry, it is of significant importance to understand physicochemical properties related to Zr-Fe-Sn system. In order to design the new Zr alloys with advanced performance by CALPHAD method, the thermodynamic model for the lower order systems is required. In the present work, first-principles calculations are employed to obtain phonon, thermodynamic and elastic properties of Zr6FeSn2 with C22 structure and the end-members (C22-Zr6FeFe2, C22-Zr6SnSn2 and C22-Zr6SnFe2) in the model of (Zr)6(Fe, Sn)2(Fe, Sn)1. It is found that the imaginary phonon modes are absent for C22-Zr6FeSn2 and C22-Zr6SnSn2, indicating they are dynamically stable, while the other two end-members are unstable. Gibbs energies of C22-Zr6FeSn2 and C22-Zr6SnSn2 are obtained from the quasiharmonic phonon approach and can be added in the thermodynamic database: Nuclearbase. The C22-Zr6FeSn2's single-crystal elasticity tensor components along with polycrystalline bulk, shear and Young's moduli are computed with a least-squares approach based upon the stress tensor computed from first-principles method. The results indicate that distortion is more difficult in the directions normal the c-axis than along to it.
Liu, Hong-Xia [Department of Materials Science and Engineering, Lanzhou University of Technology, State Key Laboratory of Advanced Processing and Recycling of Non-ferrous Metals, Lanzhou 730050 (China); Tang, Fu-Ling, E-mail: tfl03@mails.tsinghua.edu.cn [Department of Materials Science and Engineering, Lanzhou University of Technology, State Key Laboratory of Advanced Processing and Recycling of Non-ferrous Metals, Lanzhou 730050 (China); Xue, Hong-Tao; Zhang, Yu [Department of Materials Science and Engineering, Lanzhou University of Technology, State Key Laboratory of Advanced Processing and Recycling of Non-ferrous Metals, Lanzhou 730050 (China); Feng, Yu-Dong [Science and Technology on Surface Engineering Laboratory, Lanzhou Institute of Physics, Lanzhou 730000 (China)
2015-10-01
Graphical abstract: The atomic structure, bonding energy and electronic properties of the perfect WZ-CIS (1 0 0)/MoS{sub 2} (−1 0 0) interface with a lattice mismatch less than 3.5% are studied using the first principles calculation. - Highlights: • The degree of lattice mismatch of WZ-CuInS{sub 2} (1 0 0)/MoS{sub 2} (−1 0 0) is about 3.5%. • The interface bonding energy is −0.65 J/m{sup 2}, the interface has better stability. • On the interface there are some interface states near the Fermi level mainly caused by In-5s and S-3p orbital. • Difference charge density and Bader charges analysis find that the atoms near the interface have strong charge transfer. • A lot of atomic orbital hybridizations appear on the interface enhanced the interface stability and conductivity. - Abstract: Using first-principles plane-wave calculations within density functional theory, we theoretically studied the perfect WZ-CIS (1 0 0)/MoS{sub 2} (−1 0 0) interface, including the atomic structure, bonding energy and electronic properties. After relaxation the atomic positions and the bond lengths change slightly on the interface. The WZ-CIS/MoS{sub 2} interface can exist stably with the interface bonding energy about −0.65 J/m{sup 2}. Via analysis density of states, difference charge density and Bader charges we find that the electrons are largely redistributed on the interface, and there are some interface states near the Fermi level, which are mainly caused by In-5s orbital in the WZ-CIS region and S-3p orbital in the MoS{sub 2} region. On the interface the orbital hybridizations of different interfacial atoms highly enhance the bonding ability of the atoms. Electron transformation and orbital hybridization together promote the bonding between atoms and increase the adhesion energy of the interface.
Haussener, S.
2007-03-15
A solar reactor for the first step of the zinc/zinc-oxide thermochemical redox cycle is analysed and dimensioned in terms of maximization of efficiency and reaction conversion. Zinc-oxide particles carried in an inert carrier gas, in our case argon, enter the reactor in absorber tubes and are heated by concentrated solar radiation mainly due to radiative heat transfer. The particles dissociate and, in case of complete conversion, a gas mixture of argon, zinc and oxygen leaves the reactor. The aim of this study is to find an optimal design of the reactor regarding efficiency, materials and economics. The number of absorber tubes and their dimensions, the cavity dimension and its material as well as the operating conditions should be determined. Therefore 2D and 3D simulations of an 8 kW reactor are implemented. The gases are modeled as ideal gases with temperature-dependent properties. Absorption and scattering of the particle gas mixture are calculated by Mie-theory. Radiative heat transfer is included in the simulation and implemented with the aid of the discrete ordinates (DO) method. The mixture is modeled as ideal mixture and the reaction with an Arrhenius-type ansatz. Temperature distribution, reaction efficiency (heat used for zinc-oxide reaction divided by input) and tube efficiency (heat going into absorber tubes divided by input) as well as reaction conversion are analyzed to find the most promising reactor design. The results show that the most significant factors for efficiencies, conversion and absorber fluid temperature are concentration of the solar incoming radiation, zinc-oxide mass flow, the number of tubes and their dimension. Higher concentration leads to solely positive effects. Zinc-oxide mass flow variations indicate the existence of an optimal flow rate for each reactor design which maximizes efficiencies and conversion. Higher zinc-oxide mass flow leads, on one hand, to higher tube efficiency but on the other hand to lower temperatures in
Marsodi
2006-01-01
Full Text Available Calculation and analysis of B/T (Burning and/or Transmutation rate of MA (minor actinides and Pu (Plutonium has been performed in fast B/T reactor. The study was based on the assumption that the spectrum shift of neutron flux to higher side of neutron energy had a potential significance for designing the fast B/T reactor and a remarkable effect for increasing the B/T rate of MA and/or Pu. The spectrum shifts of neutron have been performed by change MOX to metallic fuel. Blending fraction of MA and or Pu in B/T fuel and the volume ratio of fuel to coolant in the reactor core were also considered. Here, the performance of fast B/T reactor was evaluated theoretically based on the calculation results of the neutronics and burn-up analysis. In this study, the B/T rate of MA and/or Pu increased by increasing the blending fraction of MA and or Pu and by changing the F/C ratio. According to the results, the total B/T rate, i.e. [B/T rate]MA + [B/T rate]Pu, could be kept nearly constant under the critical condition, if the sum of the MA and Pu inventory in the core is nearly constant. The effect of loading structure was examined for inner or outer loading of concentric geometry and for homogeneous loading. Homogeneous loading of B/T fuel was the good structure for obtaining the higher B/T rate, rather than inner or outer loading
Alvarez Cardona, Caridad M.; Guerra Valdes, Ramiro; Lopez Aldama, Daniel [Centro de Tecnologia Nuclear, La Habana (Cuba)
1996-07-01
Burnable absorbers are not used in current operating WWERs, but in order to optimize the fuel cycle and enhance operational safety, one should also introduce gadolinium or a similar burnable absorber in these reactors. For this purpose adequate tools for properly calculating local effects in hexagonal geometries should be developed and validated. The present gives main results in validating the WIMS-D/4 lattice code for Gd burnable absorber bearing WWER lattices. To validate the code experimental and calculational benchmarks proposed in a IAEA Coordinated Research Program were solved. A code system for the optimization of the Gd axial distribution in a WWER reactor was developed and it also presented here. (author)
Hernandez Valle, S.; Lopez Aldama, D. [Centro de Investigaciones Nucleares, Tecnologicas y Ambientales, La Habana (Cuba). E-mail: svalle@ctn.isctn.edu.cu
2000-07-01
The following paper contributes to validate the Nuclear Engineering Department methods to carry out calculations in TRIGA reactors solving a Benchmark. The benchmark is analyzed with the WIMS-D/4-SNAP/3D code system and using the cross section library WIMS-TRIGA. A brief description of the DSN method is presented used in WIMS/d{sup 4} code and also the SNAP-3d code is shortly explained. The results are presented and compared with the experimental values. In other hand the possible error sources are analyzed. (author)
Preparation macroconstants to simulate the core of VVER-1000 reactor
Seleznev, V. Y.
2017-01-01
Dynamic model is used in simulators of VVER-1000 reactor for training of operating staff and students. As a code for the simulation of neutron-physical characteristics is used DYNCO code that allows you to perform calculations of stationary, transient and emergency processes in real time to a different geometry of the reactor lattices [1]. To perform calculations using this code, you need to prepare macroconstants for each FA. One way of getting macroconstants is to use the WIMS code, which is based on the use of its own 69-group macroconstants library. This paper presents the results of calculations of FA obtained by the WIMS code for VVER-1000 reactor with different parameters of fuel and coolant, as well as the method of selection of energy groups for further calculation macroconstants.
Investigation of the Performance of D2O-Cooled High-Conversion Reactors for Fuel Cycle Calculations
Hikaru Hiruta; Gilles Youinou
2013-09-01
This report presents FY13 activities for the analysis of D2O cooled tight-pitch High-Conversion PWRs (HCPWRs) with U-Pu and Th-U fueled cores aiming at break-even or near breeder conditions while retaining the negative void reactivity. The analyses are carried out from several aspects which could not be covered in FY12 activities. SCALE 6.1 code system is utilized, and a series of simple 3D fuel pin-cell models are developed in order to perform Monte Carlo based criticality and burnup calculations. The performance of U-Pu fueled cores with axial and internal blankets is analyzed in terms of their impact on the relative fissile Pu mass balance, initial Pu enrichment, and void coefficient. In FY12, Pu conversion performances of D2O-cooled HCPWRs fueled with MOX were evaluated with small sized axial/internal DU blankets (approximately 4cm of axial length) in order to ensure the negative void reactivity, which evidently limits the conversion performance of HCPWRs. In this fiscal year report, the axial sizes of DU blankets are extended up to 30 cm in order to evaluate the amount of DU necessary to reach break-even and/or breeding conditions. Several attempts are made in order to attain the milestone of the HCPWR designs (i.e., break-even condition and negative void reactivity) by modeling of HCPWRs under different conditions such as boiling of D2O coolant, MOX with different 235U enrichment, and different target burnups. A similar set of analyses are performed for Th-U fueled cores. Several promising characteristics of 233U over other fissile like 239Pu and 235U, most notably its higher fission neutrons per absorption in thermal and epithermal ranges combined with lower ___ in the fast range than 239Pu allows Th-U cores to be taller than MOX ones. Such an advantage results in 4% higher relative fissile mass balance than that of U-Pu fueled cores while retaining the negative void reactivity until the target burnup of 51 GWd/t. Several other distinctions between U-Pu and
Murshed, M. Mangir, E-mail: murshed@uni-bremen.de [Chemische Kristallographie fester Stoffe, Institut für Anorganische Chemie, Universität Bremen, Leobener Straße, D-28359 Bremen (Germany); Mendive, Cecilia B.; Curti, Mariano [Departamento de Química, Facultad de Ciencias Exactas y Naturales, Universidad Nacional de Mar del Plata, Dean Funes 3350, B7600AYL, Mar del Plata (Argentina); Nénert, Gwilherm [Institut Laue-Langevin, 6 rue Jules Horowitz, 38042 Grenoble (France); Kalita, Patricia E. [Department of Physics and Astronomy and High-Pressure Science and Engineering Center, University of Nevada Las Vegas, Box 4002, Las Vegas, NV 89154-4002 (United States); Lipinska, Kris [Department of Mechanical Engineering, University of Nevada Las Vegas, 4505 Maryland Parkway, Box 454009, Las Vegas, NV 89154-4009 (United States); Cornelius, Andrew L. [Department of Physics and Astronomy and High-Pressure Science and Engineering Center, University of Nevada Las Vegas, Box 4002, Las Vegas, NV 89154-4002 (United States); Huq, Ashfia [Chemical and Engineering Materials Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6475 (United States); Gesing, Thorsten M. [Chemische Kristallographie fester Stoffe, Institut für Anorganische Chemie, Universität Bremen, Leobener Straße, D-28359 Bremen (Germany)
2014-11-15
Highlights: • Mullite-type PbFeBO{sub 4} shows uni-axial negative coefficient of thermal expansion. • Anisotropic thermal expansion of the metric parameters was modeled using modified Grüneisen approximation. • The model includes harmonic, quasi-harmonic and intrinsic anharmonic contributions to the internal energy. • DFT calculation, temperature- and pressure-dependent Raman spectra help understand the phonon decay and associated anharmonicity. - Abstract: The lattice thermal expansion of mullite-type PbFeBO{sub 4} is presented in this study. The thermal expansion coefficients of the metric parameters were obtained from composite data collected from temperature-dependent neutron and X-ray powder diffraction between 10 K and 700 K. The volume thermal expansion was modeled using extended Grüneisen first-order approximation to the zero-pressure equation of state. The additive frame of the model includes harmonic, quasi-harmonic and intrinsic anharmonic potentials to describe the change of the internal energy as a function of temperature. The unit-cell volume at zero-pressure and 0 K was optimized during the DFT simulations. Harmonic frequencies of the optical Raman modes at the Γ-point of the Brillouin zone at 0 K were also calculated by DFT, which help to assign and crosscheck the experimental frequencies. The low-temperature Raman spectra showed significant anomaly in the antiferromagnetic regions, leading to softening or hardening of some phonons. Selected modes were analyzed using a modified Klemens model. The shift of the frequencies and the broadening of the line-widths helped to understand the anharmonic vibrational behaviors of the PbO{sub 4}, FeO{sub 6} and BO{sub 3} polyhedra as a function of temperature.
Blink, J.A.
1985-03-01
In this manual we describe the use of the FORIG computer code to solve isotope-generation and depletion problems in fusion and fission reactors. FORIG runs on a Cray-1 computer and accepts more extensive activation cross sections than ORIGEN2 from which it was adapted. This report is an updated and a combined version of the previous ORIGEN2 and FORIG manuals. 7 refs., 15 figs., 13 tabs.
Kurth, S.
2002-09-04
The renormalised quark mass in the Schroedinger functional is studied perturbatively with a non-vanishing background field. The framework in which the calculations are done is the Schroedinger functional. Its definition and basic properties are reviewed and it is shown how to make the theory converge faster towards its continuum limit by O(a) improvement. It is explained how the Schroedinger functional scheme avoids the implications of treating a large energy range on a single lattice in order to determine the scale dependence of renormalised quantities. The description of the scale dependence by the step scaling function is introduced both for the renormalised coupling and the renormalised quark masses. The definition of the renormalised coupling in the Schroedinger functional is reviewed, and the concept of the renormalised mass being defined by the axial current and density via the PCAC-relation is explained. The running of the renormalised mass described by its step scaling function is presented as a consequence of the fact that the renormalisation constant of the axial density is scale dependent. The central part of the thesis is the expansion of several correlation functions up to 1-loop order. The expansion coefficients are used to compute the critical quark mass at which the renormalised mass vanishes, as well as the 1-loop coefficient of the renormalisation constant of the axial density. Using the result for this renormalisation constant, the 2-loop anomalous dimension is obtained by conversion from the MS-scheme. Another important application of perturbation theory carried out in this thesis is the determination of discretisation errors. The critical quark mass at 1-loop order is used to compute the deviation of the coupling's step scaling function from its continuum limit at 2-loop order. Several lattice artefacts of the current quark mass, defined by the PCAC relation with the unrenormalised axial current and density, are computed at 1-loop order
Demaziere, C. [CEA Centre d' Etudes de Cadarache, Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires
1999-07-01
This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). The tools that are available to perform a modeling in the Department of Reactor Physics in Chalmers are CASMO-4/TABLES-3/SIMULATE-3 from Studsvik of America. These CMS (Core Management System) programs have been extensively compared with both measurements and reference codes. Nevertheless some data are proprietary in particular the comparison of the calculated nuclide concentrations versus experiments (because of the cost of this kind of experimental study). This is why this report describes such a comparative investigation carried out with a General Electric 7x7 BWR bundle. Unfortunately, since some core history parameters were unknown, a lot of hypotheses have been adopted. This invokes sometimes a significant discrepancy in the results without being able to determine the origin of the differences between calculations and experiments. Yet one can assess that, except for four nuclides - Plutonium-238, Curium-243, Curium-244 and Cesium-135 - for which the approximate power history (history effect) can be invoked, the accuracy of the calculated nuclide concentrations is rather good if one takes the numerous approximations into account.
Short, S.M.; Luksic, A.T.; Schutz, M.E.
1989-06-01
Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly that is also radioactive and required disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report presents a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volume 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1.
Alipchenkov, V. M.; Belikov, V. V.; Davydov, A. V.; Emel'yanov, D. A.; Mosunova, N. A.
2013-05-01
Closing relations describing friction pressure drop during the motion of two-phase flows that are widely applied in thermal-hydraulic codes and in calculations of the parameters characterizing the flow of water coolant in the loops of reactor installations used at nuclear power stations and in other thermal power systems are reviewed. A new formula developed by the authors of this paper is proposed. The above-mentioned relations are implemented in the HYDRA-IBRAE thermal-hydraulic computation code developed at the Nuclear Safety Institute of the Russian Academy of Sciences. A series of verification calculations is carried out for a wide range of pressures, flowrates, and heat fluxes typical for transient and emergency operating conditions of nuclear power stations equipped with VVER reactors. Advantages and shortcomings of different closing relations are revealed, and recommendations for using them in carrying out thermal-hydraulic calculations of coolant flow in the loops of VVER-based nuclear power stations are given.
Short, S.M.; Luksic, A.T.; Lotz, T.L.; Schutz, M.E.
1989-06-01
Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly remains that is also radioactive and requires disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report present a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from Laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volumes 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1.
Berman, M.; Byers, R.K.; Steck, G.P.
1979-01-01
A statistical study is presented of the blowdown phase of a design basis accident (double-ended cold leg guillotine break) in the Zion pressurized water reactor. The response surface method was employed to generate a polynomial approximation of the peak clad temperatures calculated by RELAP4/MOD6. The nodalization was a modification of the RELAP model of Zion developed in the BE/EM study. Twenty one variables were initially selected for the study. These variables, their ranges and distributions resulted from the best engineering judgement of NRC, Sandia, INFL, and other interested and knowledgeable investigators.
Martinez C, E.
2011-07-01
One of the main objectives related to the safety of any nuclear power plant, including the nuclear power plant of Laguna Verde is to ensure the structural integrity of reactor pressure vessel. To identify and quantify the damage caused by neutron irradiation in the vessel of any nuclear reactor, it is necessary to know both the neutron flux and the neutron fluence that the vessel has been receiving during its operation lifetime, and that the damage observed by mechanical testing are products of microstructural effects induced by neutron irradiation; therefore, it is important the study and prediction of the neutron flux in order to have a better understanding of the damage that these materials are receiving. The calculation here described uses the DORT code, which solves the neutron transport equation in discrete ordinates in two dimensions (x-y, r-{theta} and r-z), according to a regulatory guide, it should make an approximation of the neutron flux in three dimensions by the so called synthesis method. It is called in that way because it achieves a representation of 3 Dimensional neutron flux combining or summarizing the fluxes calculated by DORT r-{theta}, r-z and r. This work presents the application of synthesis method, according to Regulatory Guide 1190, to determine the 3 Dimensional fluxes in internal BWR reactor using three different spatial meshes. The results of the neutron flux and fluence, using three different meshes in the directions r, {theta} and z were compared with results reported in the literature obtaining a difference not larger than 9.61%, neutron flux reached its maximum, 1.58 E + 12 n/cm{sup 2}s, at a height H 4 (239.07 cm) and angle 32.236{sup o} in the core shroud and 4.00 E + 09 n/cm{sup 2}s at a height H 4 and angle 35.27{sup o} in the inner wall of the reactor vessel, positions that are consistent to within {+-}10% over the ones reported in the literature. (Author)
Chakrabarti, J; Bagchi, B; Chakrabarti, Jayprokas; Basu, Asis; Bagchi, Bijon
2000-01-01
Fermions on the lattice have bosonic excitations generated from the underlying periodic background. These, the lattice bosons, arise near the empty band or when the bands are nearly full. They do not depend on the nature of the interactions and exist for any fermion-fermion coupling. We discuss these lattice boson solutions for the Dirac Hamiltonian.
Gholamzadeh Zohreh
2014-12-01
Full Text Available Decrease of the economically accessible uranium resources and the inherent proliferation resistance of thorium fuel motivate its application in nuclear power systems. Estimation of the nuclear reactor’s neutronic parameters during different operational situations is of key importance for the safe operation of nuclear reactors. In the present research, thorium oxide fuel burn-up calculations for a demonstrative model of a heavy water- -cooled reactor have been performed using MCNPX 2.6 code. Neutronic parameters for three different thorium fuel matrices loaded separately in the modelled thermal core have been investigated. 233U, 235U and 239Pu isotopes have been used as fissile element in the thorium oxide fuel, separately. Burn-up of three different fuels has been calculated at 1 MW constant power. 135X and 149Sm concentration variations have been studied in the modelled core during 165 days burn-up. Burn-up of thorium oxide enriched with 233U resulted in the least 149Sm and 135Xe productions and net fissile production of 233U after 165 days. The negative fuel, coolant and void reactivity of the used fuel assures safe operation of the modelled thermal core containing (233U-Th O2 matrix. Furthermore, utilisation of thorium breeder fuel demonstrates several advantages, such as good neutronic economy, 233U production and less production of long-lived α emitter high radiotoxic wastes in biological internal exposure point of view
Khattab, K; Sulieman, I
2009-04-01
The MCNP-4C code, based on the probabilistic approach, was used to model the 3D configuration of the core of the Syrian miniature neutron source reactor (MNSR). The continuous energy neutron cross sections from the ENDF/B-VI library were used to calculate the thermal and fast neutron fluxes in the inner and outer irradiation sites of MNSR. The thermal fluxes in the MNSR inner irradiation sites were also measured experimentally by the multiple foil activation method ((197)Au (n, gamma) (198)Au and (59)Co (n, gamma) (60)Co). The foils were irradiated simultaneously in each of the five MNSR inner irradiation sites to measure the thermal neutron flux and the epithermal index in each site. The calculated and measured results agree well.
Farkas, Istvan; Hutli, Ezddin; Faekas, Tatiana; Takacs, Antal; Guba, Attila; Toth, Ivan [Dept. of Thermohydraulics, Centre for Energy Research, Hungarian Academy of Sciences, Budapest (Hungary)
2016-08-15
The aim of this work is to simulate the thermohydraulic consequences of a main steam line break and to compare the obtained results with Rossendorf Coolant Mixing Model (ROCOM) 1.1 experimental results. The objective is to utilize data from steady-state mixing experiments and computational fluid dynamics (CFD) calculations to determine the flow distribution and the effect of thermal mixing phenomena in the primary loops for the improvement of normal operation conditions and structural integrity assessment of pressurized water reactors. The numerical model of ROCOM was developed using the FLUENT code. The positions of the inlet and outlet boundary conditions and the distribution of detailed velocity/turbulence parameters were determined by preliminary calculations. The temperature fields of transient calculation were averaged in time and compared with time-averaged experimental data. The perforated barrel under the core inlet homogenizes the flow, and therefore, a uniform temperature distribution is formed in the pressure vessel bottom. The calculated and measured values of lowest temperature were equal. The inlet temperature is an essential parameter for safety assessment. The calculation predicts precisely the experimental results at the core inlet central region. CFD results showed a good agreement (both qualitatively and quantitatively) with experimental results.
Oigawa, Hiroyuki; Iijima, Susumu; Sakurai, Takeshi; Okajima, Shigeaki; Andoh, Masaki; Nemoto, Tatsuo; Kato, Yuichi; Osugi, Toshitaka [Dept. of Nuclear Energy System, Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)
2000-02-01
In order to assess the validity of the cross section library for fast reactor physics, a set of benchmark calculation is proposed. The benchmark calculation is based upon mock-up experiments at three FCA cores with various compositions of central test regions, two of which were mock-ups of metallic fueled LMFBR's, and the other was a mock-up of a mixed oxide fueled LMFBR. One of the metallic cores included enriched uranium in the test region, while the others did not. Physics parameters to be calculated are criticality, reaction rate ratios, plutonium and B{sub 4}C sample worth, sodium void reactivity worth, and Doppler reactivity worth of {sup 238}U. Homogenized atomic number densities and various correction factors are given so that anyone can easily perform diffusion calculation in two-dimensional RZ-model and compare the results with the experiments. The validity of the correction factors are proved by changing the calculation method and used nuclear data file. (author)
Lewis, Randy
2014-01-01
Several collaborations have recently performed lattice calculations aimed specifically at dark matter, including work with SU(2), SU(3), SU(4) and SO(4) gauge theories to represent the dark sector. Highlights of these studies are presented here, after a reminder of how lattice calculations in QCD itself are helping with the hunt for dark matter.
Hogerton, John
1964-01-01
This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.
Lattice Boltzmann Stokesian dynamics.
Ding, E J
2015-11-01
Lattice Boltzmann Stokesian dynamics (LBSD) is presented for simulation of particle suspension in Stokes flows. This method is developed from Stokesian dynamics (SD) with resistance and mobility matrices calculated using the time-independent lattice Boltzmann algorithm (TILBA). TILBA is distinguished from the traditional lattice Boltzmann method (LBM) in that a background matrix is generated prior to the calculation. The background matrix, once generated, can be reused for calculations for different scenarios, thus the computational cost for each such subsequent calculation is significantly reduced. The LBSD inherits the merits of the SD where both near- and far-field interactions are considered. It also inherits the merits of the LBM that the computational cost is almost independent of the particle shape.
Sandrin, Ch.
2010-04-15
This PhD Thesis aims to achieve a method for the modelling of the reflector surrounding the core for neutronics core calculations. This method should consider the EPR reactor specificities (steel reflector) and the increased demand in precision. In neutronics core calculations, the reflector can be represented either by albedos boundary conditions (current ratios) or by one or several media, surrounding the core, characterised by homogenized parameters. Those parameters (cross sections and diffusion coefficients) should be obtained using equivalence so that they allow a good reproduction of the reference albedos in a representative situation. During this PhD, such an equivalence method has been developed in the APOLLO-2 code with the minimization of a functional of the differences between the reference albedos and those computed with the equivalent parameters. Because of the positiveness constraints, a local minimization, such as Newton-like methods, is not always possible and we have therefore also implemented a Particle Swarm Optimization Algorithm for more than two energy groups' problems. The parameters obtained have been used in two dimensions EPR core calculations with the CRONOS-2 code for various fuel loadings in two to eight groups diffusion. Those core calculation have been validated against reference Monte-Carlo calculations and against core calculations with albedos boundary conditions. In addition to the increased easiness of utilization, the implemented equivalence method has yielded an improvement of the results for the two groups calculation. With a higher energy groups number, the use of a unique equivalent reflector does not account correctly for the two dimensions effects; a modelling with different reflector meshes has improved the results. The modelling of the reflector by two dimensions albedos boundary conditions is the more suited for the representation of the boundary conditions and, therefore, should the two dimensions albedos
Nuclear reactions from lattice QCD
Briceño, Raúl A.; Davoudi, Zohreh; Luu, Thomas C.
2015-02-01
One of the overarching goals of nuclear physics is to rigorously compute properties of hadronic systems directly from the fundamental theory of strong interactions, quantum chromodynamics (QCD). In particular, the hope is to perform reliable calculations of nuclear reactions which will impact our understanding of environments that occur during big bang nucleosynthesis, the evolution of stars and supernovae, and within nuclear reactors and high energy/density facilities. Such calculations, being truly ab initio, would include all two-nucleon and three-nucleon (and higher) interactions in a consistent manner. Currently, lattice quantum chromodynamics (LQCD) provides the only reliable option for performing calculations of some of the low-energy hadronic observables. With the aim of bridging the gap between LQCD and nuclear many-body physics, the Institute for Nuclear Theory held a workshop on Nuclear Reactions from LQCD on March 2013. In this review article, we report on the topics discussed in this workshop and the path planned to move forward in the upcoming years.
Effect of DUPIC Cycle on CANDU Reactor Safety Parameters
Nader M.A. Mohamed
2016-10-01
Full Text Available Although, the direct use of spent pressurized water reactor (PWR fuel in CANda Deuterium Uranium (CANDU reactors (DUPIC cycle is still under investigation, DUPIC cycle is a promising method for uranium utilization improvement, for reduction of high level nuclear waste, and for high degree of proliferation resistance. This paper focuses on the effect of DUPIC cycle on CANDU reactor safety parameters. MCNP6 was used for lattice cell simulation of a typical 3,411 MWth PWR fueled by UO2 enriched to 4.5w/o U-235 to calculate the spent fuel inventories after a burnup of 51.7 MWd/kgU. The code was also used to simulate the lattice cell of CANDU-6 reactor fueled with spent fuel after its fabrication into the standard 37-element fuel bundle. It is assumed a 5-year cooling time between the spent fuel discharges from the PWR to the loading into the CANDU-6. The simulation was carried out to calculate the burnup and the effect of DUPIC fuel on: (1 the power distribution amongst the fuel elements of the bundle; (2 the coolant void reactivity; and (3 the reactor point-kinetics parameters.
Effect of DUPIC cycle on CANDU reactor safety parameters
Mohamed, Nader M. A. [Atomic Energy Authority, ETRR-2, Cairo (Egypt); Badawi, Alya [Dept. of Nuclear and Radiation Engineering, Alexandria University, Alexandria (Egypt)
2016-10-15
Although, the direct use of spent pressurized water reactor (PWR) fuel in CANda Deuterium Uranium (CANDU) reactors (DUPIC) cycle is still under investigation, DUPIC cycle is a promising method for uranium utilization improvement, for reduction of high level nuclear waste, and for high degree of proliferation resistance. This paper focuses on the effect of DUPIC cycle on CANDU reactor safety parameters. MCNP6 was used for lattice cell simulation of a typical 3,411 MWth PWR fueled by UO{sub 2} enriched to 4.5w/o U-235 to calculate the spent fuel inventories after a burnup of 51.7 MWd/kgU. The code was also used to simulate the lattice cell of CANDU-6 reactor fueled with spent fuel after its fabrication into the standard 37-element fuel bundle. It is assumed a 5-year cooling time between the spent fuel discharges from the PWR to the loading into the CANDU-6. The simulation was carried out to calculate the burnup and the effect of DUPIC fuel on: (1) the power distribution amongst the fuel elements of the bundle; (2) the coolant void reactivity; and (3) the reactor point-kinetics parameters.
Verification of spectral burn-up codes on 2D fuel assemblies of the GFR demonstrator ALLEGRO reactor
Čerba, Štefan, E-mail: stefan.cerba@stuba.sk [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia); Vrban, Branislav; Lüley, Jakub [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia); Dařílek, Petr [VUJE a.s., Okružná 5, 918 64 Trnava (Slovakia); Zajac, Radoslav, E-mail: radoslav.zajac@vuje.sk [VUJE a.s., Okružná 5, 918 64 Trnava (Slovakia); Nečas, Vladimír; Haščik, Ján [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia)
2014-02-15
Highlights: • Verification of the MCNPX, HELIOS and SCALE codes. • MOX and ceramic fuel assembly. • Gas-cooled fast reactor. • Burnup calculation. - Abstract: The gas-cooled fast reactor, which is one of the six GEN IV reactor concepts, is characterized by high operational temperatures and a hard neutron spectrum. The utilization of commonly used spectral codes, developed mainly for LWR reactors operated in the thermal/epithermal neutron spectrum, may be connected with systematic deviations since the main development effort of these codes has been focused on the thermal part of the neutron spectrum. To be able to carry out proper calculations for fast systems the used codes have to account for neutron resonances including the self-shielding effect. The presented study aims at verifying the spectral HELIOS, MCNPX and SCALE codes on the basis of depletion calculations of 2D MOX and ceramic fuel assemblies of the ALLEGRO gas-cooled fast reactor demonstrator in infinite lattice.
Berkhan, Ana; Starflinger, Joerg [Stuttgart Univ. (Germany). Inst. fuer Kernenergetik und Energiesysteme (IKE)
2013-07-01
The computer code MEWA is used for the description of severe accident sequences in light-water reactors. During the reactor accident with core disruption the solidified core fragments are displaced into the lower plenum of the reactor pressure vessel (RPV) or in case of RPV failure into the water filled reactor sump. For the progress or cessation of the severe accident the cooling of the packed bed is of main importance. With the 3D version of the code it is possible to study spatially complex packed beds with respect to their coolability. Further extension of the MEWA code will include the optimization for the improvement of the calculation efficiency and reduction of computation time. The validation will be performed by re-calculation of experiments (for instance DEBRIS experiments at the IKE) and the comparison with results of the 2D version.
Dzhalandinov, A.; Tsofin, V.; Kochkin, V.; Panferov, P.; Timofeev, A.; Reshetnikov, A.; Makhotin, D.; Erak, D.; Voloschenko, A.
2016-02-01
Usually the synthesis of two-dimensional and one-dimensional discrete ordinate calculations is used to evaluate neutron fluence on VVER-1000 reactor pressure vessel (RPV) for prognosis of radiation embrittlement. But there are some cases when this approach is not applicable. For example the latest projects of VVER-1000 have upgraded surveillance program. Containers with surveillance specimens are located on the inner surface of RPV with fast neutron flux maximum. Therefore, the synthesis approach is not suitable enough for calculation of local disturbance of neutron field in RPV inner surface behind the surveillance specimens because of their complicated and heterogeneous structure. In some cases the VVER-1000 core loading consists of fuel assemblies with different fuel height and the applicability of synthesis approach is also ambiguous for these fuel cycles. Also, the synthesis approach is not enough correct for the neutron fluence estimation at the RPV area above core top. Because of these reasons only the 3D neutron transport codes seem to be satisfactory for calculation of neutron fluence on the VVER-1000 RPV. The direct 3D calculations are also recommended by modern regulations.
Dzhalandinov A.
2016-01-01
Full Text Available Usually the synthesis of two-dimensional and one-dimensional discrete ordinate calculations is used to evaluate neutron fluence on VVER-1000 reactor pressure vessel (RPV for prognosis of radiation embrittlement. But there are some cases when this approach is not applicable. For example the latest projects of VVER-1000 have upgraded surveillance program. Containers with surveillance specimens are located on the inner surface of RPV with fast neutron flux maximum. Therefore, the synthesis approach is not suitable enough for calculation of local disturbance of neutron field in RPV inner surface behind the surveillance specimens because of their complicated and heterogeneous structure. In some cases the VVER-1000 core loading consists of fuel assemblies with different fuel height and the applicability of synthesis approach is also ambiguous for these fuel cycles. Also, the synthesis approach is not enough correct for the neutron fluence estimation at the RPV area above core top. Because of these reasons only the 3D neutron transport codes seem to be satisfactory for calculation of neutron fluence on the VVER-1000 RPV. The direct 3D calculations are also recommended by modern regulations.
Akie, H.; Sugo, Y.; Okawa, R.
2003-06-01
A rock-like oxide (ROX) fuel - light water reactor (LWR) burning system has been studied for efficient plutonium transmutation. For the improvement of small negative reactivity coefficients and severe transient behaviors of ROX fueled LWRs, a partial loading core of ROX fuel assemblies with conventional UO 2 assemblies was considered. As a result, although the reactivity coefficients could be improved, the power peaking tends to be large in this heterogeneous core configuration. The reactivity initiated accident (RIA) and loss of coolant accident (LOCA) behaviors were not sufficiently improved. In order to reduce the power peaking, the fuel composition and the assembly design of the ROX fuel were modified. Firstly, erbium burnable poison was added as Er 2O 3 in the ROX fuel to reduce the burnup reactivity swing. Then pin-by-pin Pu enrichment and Er content distributions within the ROX fuel assembly were considered. In addition, the Er content distribution was also considered in the axial direction of the ROX fuel pin. With these modifications, a power peaking factor even lower than the one in a conventional UO 2 fueled core can be obtained. The RIA and LOCA analyses of the modified core have also shown the comparable transient behaviors of ROX partial loading core to those of the UO 2 core.
Brisbois, J.; Vergnaud, T.; Oceraies, Y
1967-12-01
In a graphite pile, EDF or Inca type reactor, it is necessary to know the value of the intermediate neutron flux at the output of the lateral reflector in order to determine more precisely the neutron flux at the level of ionisation chambers. A sub critical pile of graphite and natural uranium was built, allowing to reconstitute the geometry of the radiation sources and the disposition of inferior and lateral protections of these piles. This pile is supplied with thermal neutrons coming from the Nereide light water type reactor. Some measurements of intermediate neutron flux have been made in this pile in order to establish a formalism for neutron flux calculation in slowing down in a whole core-lateral reflector, from the distribution of the thermal neutrons flux in the core. The flux calculation is done by age theory in three dimensions, in two homogenous media, separated by an axially semi infinite and laterally finite plane. One of these media includes a distribution of source. The constants are modified in order to take into account the presence of empty channels in the stacking. These calculations are done by the Malaga code. The checking of the formalism has been made in a greater complex geometry of these reactors that introduces an uncertainty factor in the comparison of results. We can however tell that we estimate correctly the variation of the intermediate neutrons flux in the core as well as its descending in a holed lateral reflector. The ratio between the calculation and the experiment is inferior to 2 or 3. Most of the time to a factor 2. [French] Dans une pile a graphite, du type EdF ou Inca, il est necessaire de connaitre la valeur du flux de neutrons intermediaires a la sortie du reflecteur lateral, afin de determiner avec plus de precision le flux de neutrons au niveau des chambres d'ionisation. Il a ete construit un empilement sous-critique, graphite uranium naturel, qui permet de reconstituer la geometrie des sources de rayonnement et la
Lee, C.E.; Apperson, C.E. Jr.; Foley, J.E.
1976-10-01
The report describes an analytic containment building model that is used for calculating the leakage into the environment of each isotope of an arbitrary radioactive decay chain. The model accounts for the source, the buildup, the decay, the cleanup, and the leakage of isotopes that are gas-borne inside the containment building.
Perez, Daniel Milian; Lorenzo, Daniel E. Milian; Garcia, Lorena P. Rodriguez; Llanes, Jesus Salomon; Hernandez, Carlos R. Garcia, E-mail: dperez@instec.cu, E-mail: dmilian@instec.cu, E-mail: lorenapilar@instec.cu, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba); Lira, Carlos A. Brayner de Oliveira, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife (Brazil); Rodriguez, Manuel Cadavid, E-mail: mcadavid2001@yahoo.com [Tecnologia Nuclear Medica Spa, TNM (Chile)
2015-07-01
{sup 99m}Tc is the most common radioisotope used in nuclear medicine. It is a very useful radioisotope, which is used in about 30-40 million procedures worldwide every year. Medical diagnostic imaging techniques using {sup 99m}Tc represent approximately 80% of all nuclear medicine procedures. Although {sup 99m}Tc can be produced directly on a cyclotron or other type of particle accelerator, currently is almost exclusively produced from the beta-decay of its 66-h parent {sup 99}Mo. {sup 99}Mo production system in an Aqueous Homogeneous Reactor (AHR) is potentially advantageous because of its low cost, small critical mass, inherent passive safety, and simplified fuel handling, processing and purification characteristics. In this paper, an AHR conceptual design using Low Enriched Uranium (LEU) is studied and optimized for the production of {sup 99}Mo. Aspects related with the neutronic behavior such as optimal reflector thickness, critical height, medical isotopes production and the reactivity feedback introduced in the solution by the volumetric expansion of the fuel solution due to thermal expansion of the fuel solution and the void volume generated by radiolytic gas bubbles were evaluated. Thermal-hydraulics studies were carried out in order to show that sufficient cooling capacity exists to prevent fuel overheating. The neutronic and thermal-hydraulics calculations have been performed with the MCNPX computational code and the version 14 of ANSYS CFX respectively. The neutronic calculations demonstrated that the reactor is able to produce 370 six-day curies of {sup 99}Mo in 5 days operation cycles and the CFD simulation demonstrated that the heat removal systems provide sufficient cooling capacity to prevent fuel overheating, the maximum temperature reached by the fuel (89.29 deg C) was smaller to the allowable temperature limit (90 deg C). (author)
Jingang Liang
2016-06-01
Full Text Available Because of prohibitive data storage requirements in large-scale simulations, the memory problem is an obstacle for Monte Carlo (MC codes in accomplishing pin-wise three-dimensional (3D full-core calculations, particularly for whole-core depletion analyses. Various kinds of data are evaluated and quantificational total memory requirements are analyzed based on the Reactor Monte Carlo (RMC code, showing that tally data, material data, and isotope densities in depletion are three major parts of memory storage. The domain decomposition method is investigated as a means of saving memory, by dividing spatial geometry into domains that are simulated separately by parallel processors. For the validity of particle tracking during transport simulations, particles need to be communicated between domains. In consideration of efficiency, an asynchronous particle communication algorithm is designed and implemented. Furthermore, we couple the domain decomposition method with MC burnup process, under a strategy of utilizing consistent domain partition in both transport and depletion modules. A numerical test of 3D full-core burnup calculations is carried out, indicating that the RMC code, with the domain decomposition method, is capable of pin-wise full-core burnup calculations with millions of depletion regions.
Kenneth Wilson and lattice QCD
Ukawa, Akira
2015-01-01
We discuss the physics and computation of lattice QCD, a space-time lattice formulation of quantum chromodynamics, and Kenneth Wilson's seminal role in its development. We start with the fundamental issue of confinement of quarks in the theory of the strong interactions, and discuss how lattice QCD provides a framework for understanding this phenomenon. A conceptual issue with lattice QCD is a conflict of space-time lattice with chiral symmetry of quarks. We discuss how this problem is resolved. Since lattice QCD is a non-linear quantum dynamical system with infinite degrees of freedom, quantities which are analytically calculable are limited. On the other hand, it provides an ideal case of massively parallel numerical computations. We review the long and distinguished history of parallel-architecture supercomputers designed and built for lattice QCD. We discuss algorithmic developments, in particular the difficulties posed by the fermionic nature of quarks, and their resolution. The triad of efforts toward b...
Ku, L.P.; Hendel, H.W.; Liew, S.L.
1989-02-01
Neutron transport simulations have been carried out to calculate the absolute detection efficiency of a moderated /sup 235/U neutron detector which is used on the TFTR as a part of the primary fission detector diagnostic system for measuring fusion power yields. Transport simulations provide a means by which the effects of variations in various shielding and geometrical parameters can be explored. These effects are difficult to study in calibration experiments. The calculational model, benchmarked against measurements, can be used to complement future detector calibrations, when the high level of radioactivity resulting from machine operation may severely restrict access to the tokamak. We present a coupled forward-adjoint algorithm, employing both the deterministic and Monte Carlo sampling methods, to model the neutron transport in the complex tokamak and detector geometries. Sensitivities of the detector response to the major and minor radii, and angular anisotropy of the neutron emission are discussed. A semi-empirical model based on matching the calculational results with a small set of experiments produces good agreement (+-15%) for a wide range of source energies and geometries. 20 refs., 6 figs., 4 tabs.
De Raedt, C
2000-07-01
The Reactor Physics and 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 on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research 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. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.
Hadron Structure and Spectrum from the Lattice
Lang, C B
2015-01-01
Lattice calculations for hadrons are now entering the domain of resonances and scattering, necessitating a better understanding of the observed discrete energy spectrum. This is a reviewing survey about recent lattice QCD results, with some emphasis on spectrum and scattering.
Schmitz, Tobias; Blaickner, Matthias; Schütz, Christian; Wiehl, Norbert; Kratz, Jens V; Bassler, Niels; Holzscheiter, Michael H; Palmans, Hugo; Sharpe, Peter; Otto, Gerd; Hampel, Gabriele
2010-10-01
To establish Boron Neutron Capture Therapy (BNCT) for non-resectable liver metastases and for in vitro experiments at the TRIGA Mark II reactor at the University of Mainz, Germany, it is necessary to have a reliable dose monitoring system. The in vitro experiments are used to determine the relative biological effectiveness (RBE) of liver and cancer cells in our mixed neutron and gamma field. We work with alanine detectors in combination with Monte Carlo simulations, where we can measure and characterize the dose. To verify our calculations we perform neutron flux measurements using gold foil activation and pin-diodes. Material and methods. When L-α-alanine is irradiated with ionizing radiation, it forms a stable radical which can be detected by electron spin resonance (ESR) spectroscopy. The value of the ESR signal correlates to the amount of absorbed dose. The dose for each pellet is calculated using FLUKA, a multipurpose Monte Carlo transport code. The pin-diode is augmented by a lithium fluoride foil. This foil converts the neutrons into alpha and tritium particles which are products of the (7)Li(n,α)(3)H-reaction. These particles are detected by the diode and their amount correlates to the neutron fluence directly. Results and discussion. Gold foil activation and the pin-diode are reliable fluence measurement systems for the TRIGA reactor, Mainz. Alanine dosimetry of the photon field and charged particle field from secondary reactions can in principle be carried out in combination with MC-calculations for mixed radiation fields and the Hansen & Olsen alanine detector response model. With the acquired data about the background dose and charged particle spectrum, and with the acquired information of the neutron flux, we are capable of calculating the dose to the tissue. Conclusion. Monte Carlo simulation of the mixed neutron and gamma field of the TRIGA Mainz is possible in order to characterize the neutron behavior in the thermal column. Currently we also
Wang, Da-Wei; Zhu, Shi-Yao; Scully, Marlan O
2014-01-01
We show that the timed Dicke states of a collection of three-level atoms can form a tight-binding lattice in the momentum space. This lattice, coined the superradiance lattice (SL), can be constructed based on an electromagnetically induced transparency (EIT) system. For a one-dimensional SL, we need the coupling field of the EIT system to be a standing wave. The detuning between the two components of the standing wave introduces an effective electric field. The quantum behaviours of electrons in lattices, such as Bloch oscillations, Wannier-Stark ladders, Bloch band collapsing and dynamic localization can be observed in the SL. The SL can be extended to two, three and even higher dimensions where no analogous real space lattices exist and new physics are waiting to be explored.
Heat transfer calculation of esterification reactor jacket%酯化反应釜夹套的传热计算
马欣
2001-01-01
By calculating the heat transfer coefficents of liquid heating medium jacket and gas heating medium chilling jacket,quantitative comparison was done.The advantages of gas heating medium chilling jacket were pointed out.The design of the jacket has been discussed.%通过计算液相热媒夹套和气相热媒冷凝夹套的给热系数,进行定量的比较.指出应用气相热媒冷凝夹套的优点,并对它的设计进行了讨论.
Fort, E. E-mail: eric.fort@cea.fr; Rimpault, G.; Bosq, J-C.; Camous, B.; Zammit, V.; Dupont, E.; Jacqmin, R.; Smith, P.; Biron, D. E-mail: didier.biron@edf.fr; Verrier, D. E-mail: dverrier@framatome.fr
2003-12-01
A single consistent scheme of calculational methods and nuclear data called ERANOS-ERALIB1 was produced in 1996 to calculate fast reactor neutronic parameters. It represents a significant improvement on previous schemes such as CARNAVAL-IV, PROPANE and VASCO, each of which were required in order to calculate one specific application. The nuclear data library ERALIB1 has been obtained by a consistent statistical adjustment based on 355 integral data from 71 different systems. The performance of ERALIB1 is excellent, as demonstrated during its validation for which all the k{sub eff} SUPER-PHENIX data were reproduced to within 70 pcm. The only restriction on this satisfactory performance is related to the rather poor prediction of the sodium void reactivity effect. This was due to very bad nuclear data for {sup 23}Na, and the unsatisfactory methods used to calculate the sensitivity coefficients for the sodium void reactivity variation {delta}{rho}{sub Na}. To improve the performance relative to this point and to enlarge the domain of validation several actions have been undertaken: - a revision of the formalism and algorithms used to calculate the derivatives of {delta}{rho}{sub Na} to the sodium cross section data,; - a significant enlargement of the integral data base related to this aspect of the sodium void effect. Compared to the initial data base established in support of ERALIB1, several additional (18) sodium void configurations corresponding to voids of different volumes at different core locations have been studied. In order to broaden the range of application of the improved library, which will be called ERALIB1.A, significant effort has been devoted to additional configurations which have firstly been evaluated, and then if judged suitable, included in the adjustment process. They are related to two specifically targeted experimental programmes: - a study of neutron deep penetration. Several configurations of the JANUS experimental programme (shielding
Yamamoto, Arata
2016-01-01
We propose the lattice QCD calculation of the Berry phase which is defined by the ground state of a single fermion. We perform the ground-state projection of a single-fermion propagator, construct the Berry link variable on a momentum-space lattice, and calculate the Berry phase. As the first application, the first Chern number of the (2+1)-dimensional Wilson fermion is calculated by the Monte Carlo simulation.
Multifractal behaviour of -simplex lattic
Sanjay Kumar; Debaprasad Giri; Sujata Krishna
2000-06-01
We study the asymptotic behaviour of resistance scaling and ﬂuctuation of resistance that give rise to ﬂicker noise in an -simplex lattice. We propose a simple method to calculate the resistance scaling and give a closed-form formula to calculate the exponent, , associated with resistance scaling, for any . Using current cumulant method we calculate the exact noise exponent for -simplex lattices.
Chernyshev, V. A.; Petrov, V. P.; Nikiforov, A. E.
2015-05-01
The ab initio calculation has been performed for the crystal structure and the phonon spectrum of titanates with the structure of pyrochlore R 2Ti2O7 ( R = Gd-Lu). The frequencies and types of fundamental vibrations have been found. For R = Tb, Tm, and Yb, this calculation has been carried out for the first time; furthermore, there is no available information on experimental studies of the phonon spectrum for Tm and Yb. The influence of hydrostatic pressure to 35 GPa on the structure, dynamics, and elastic properties of the Gd2Ti2O7 lattice has been investigated. The dependence of the phonon frequencies on the pressure has been obtained. The calculations have predicted that the relative change in the pyrochlore structure volume during compression at pressures to 35 GPa is well described by the third-order Birch-Murnaghan equation of states. The results of the calculations agree with the available experimental data. It has been shown that the structural, dynamic, and elastic properties of the R 2Ti2O7 crystal lattice can be adequately described in the case where the inner shells of the RE ion up to 4 f are replaced by the pseudopotential.
Veli ÇAPALI
2016-05-01
Full Text Available BeO is one of the most common moderator material for neutron moderation; due to its high density, neutron capture cross section and physical-chemical properties that provides usage at elevated temperatures. As it’s known, for various applications in the field of reactor design and neutron capture, reaction cross–section data are required. The cross–sections of (n,α, (n,2n, (n,t, (n,EL and (n,TOT reactions for 9Be and 16O nuclei have been calculated by using TALYS 1.6 Two Component Exciton model and EMPIRE 3.2 Exciton model in this study. Hadronic interactions of low energetic neutrons and generated isotopes–particles have been investigated for a situation in which BeO was used as a neutron moderator by using GEANT4, which is a powerful simulation software. In addition, energy deposition along BeO material has been obtained. Results from performed calculations were compared with the experimental nuclear reaction data exist in EXFOR.
Domingos, Douglas Borges; Silva, Antonio Teixeira e; Umbehaun, Pedro Ernesto; Silva, Jose Eduardo Rosa da; Conti, Thadeu das Neves; Yamaguchi, Mitsuo [Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), Sao Paulo, SP (Brazil)], e-mail: douglasborgesdomingos@yahoo.com.br
2009-07-01
Neutronic, thermal-hydraulics and accident analysis calculations were developed to estimate the safety of an irradiation device placed in the IEA-R1 reactor core. The irradiation device will be used to receive miniplates of U{sub 3}O{sub 8}-Al e U{sub 3}Si{sub 2}-Al dispersion fuels, LEU type (19.9% of {sup 235}U), with uranium densities of, respectively, 3.0 gU/cm{sup 3} and 4.8gU/cm{sup 3}. The fuel miniplates will be irradiated to nominal {sup 235}U burnup levels of 50% and 80%, in order to qualify the above high-density dispersion fuels to be used in the Brazilian Multipurpose Reactor, now in the conception phase. For the neutronic calculation, the computer code CITATION was utilized. The computer code FLOW was used to calculate the coolant flow rate in the irradiation device, allowing the determination of the fuel miniplate temperatures with the computer model MTRCR-IEA-R1. A postulated Loss of Coolant Accident (LOCA) was analyzed with the computer codes LOSS and TEMPLOCA, allowing the calculation of the fuel miniplate temperatures after the reactor pool draining. The calculations showed that the irradiation of the fuel miniplates will happen without any adverse consequence in the IEA-R1 reactor. (author)
Charmed baryons on the lattice
Padmanath, M
2015-01-01
We discuss the significance of charm baryon spectroscopy in hadron physics and review the recent developments of the spectra of charmed baryons in lattice calculations. Special emphasis is given on the recent studies of highly excited charm baryon states. Recent precision lattice measurements of the low lying charm and bottom baryons are also reviewed.
Hwang, R N; Toppel, B J; Henryson, H II
1980-10-01
Motivated by a need for an economical yet rigorous tool which can address the computation of the structural material Doppler effect, an extremely efficient improved RABANL capability has been developed utilizing the fact that the Doppler broadened line shape functions become essentially identical to the natural line shape functions or Lorentzian limits beyond about 100 Doppler widths from the resonance energy, or when the natural width exceeds about 200 Doppler widths. The computational efficiency has been further enhanced by preprocessing or screening a significant number of selected resonances during library preparation into composition and temperature independent smooth background cross sections. The resonances which are suitable for such pre-processing are those which are either very broad or those which are very weak. The former contribute very little to the Doppler effect and their self-shielding effect can readily be averaged into slowly varying background cross section data, while the latter contribute very little to either the Doppler or to self-shielding effects. To illustrate the accuracy and efficiency of the improved RABANL algorithms and resonance screening techniques, calculations have been performed for two systems, the first with a composition typical of the STF converter region and the second typical of an LMFBR core composition. Excellent agreement has been found for RABANL compared to the reference Monte Carlo solution obtained using the code VIM, and improved results have also been obtained for the narrow resonance approximation in the ultra-fine-group option of MC/sup 2/-2.
Verification of HELIOS/MASTER Nuclear Analysis System for SMART Research Reactor, Rev. 1.0
Lee, Kyung Hoon; Kim, Kang Seog; Cho, Jin Young; Lee, Chung Chan; Zee, Sung Quun
2005-12-15
Nuclear design for the SMART reactor is performed by using the transport lattice code HELIOS and the core analysis code MASTER. HELIOS code developed by Studsvik Scandpower in Norway is a transport lattice code for the neutron and gamma behavior, and is used to generate few group constants. MASTER code is a nodal diffusion code developed by KAERI, and is used to analyze reactor physics. This nuclear design code package requires verification. Since the SMART reactor is unique, it is impossible to verify this code system through the comparison of the calculation results with the measured ones. Therefore, the uncertainties for the nuclear physics parameters calculated by HELIOS/MASTER have been evaluated indirectly. Since Monte Carlo calculation includes least approximations an assumptions to simulate a neutron behavior, HELIOS/MASTER has been verified by this one. Monte Carlo code has been verified by the Kurchatov critical experiments similar to SMART reactor, and HELIOS/MASTER code package has been verified by Monte Carlo calculations for the SMART research reactor.
Verification of HELIOS/MASTER Nuclear Analysis System for SMART Research Reactor
Kim, Kang Seog; Cho, Jin Young; Lee, Chung Chan; Zee, Sung Quun
2005-07-15
Nuclear design for the SMART reactor is performed by using the transport lattice code HELIOS and the core analysis code MASTER. HELIOS code developed by Studsvik Scandpower in Norway is a transport lattice code for the neutron and gamma behavior, and is used to generate few group constants. MASTER code is a nodal diffusion code developed by KAERI, and is used to analyze reactor physics. This nuclear design code package requires verification. Since the SMART reactor is unique, it is impossible to verify this code system through the comparison of the calculation results with the measured ones. Therefore, the uncertainties for the nuclear physics parameters calculated by HELIOS/MASTER have been evaluated indirectly. Since Monte Carlo calculation includes least approximations an assumptions to simulate a neutron behavior, HELIOS/MASTER has been verified by this one. Monte Carlo code has been verified by the Kurchatov critical experiments similar to SMART reactor, and HELIOS/MASTER code package has been verified by Monte Carlo calculations for the SMART research reactor.
de Raedt, Hans; von der Linden, W.; Binder, K
1995-01-01
In this chapter we review methods currently used to perform Monte Carlo calculations for quantum lattice models. A detailed exposition is given of the formalism underlying the construction of the simulation algorithms. We discuss the fundamental and technical difficulties that are encountered and gi
Chang, G. S.; Lillo, M. A.
2009-08-01
The National Nuclear Security Administrations (NNSA) Reduced Enrichment for Research and Test Reactors (RERTR) program assigned to the Idaho National Laboratory (INL) the responsibility of developing and demonstrating high uranium density research reactor fuel forms to enable the use of low enriched uranium (LEU) in research and test reactors around the world. A series of full-size fuel plate experiments have been proposed for irradiation testing in the center flux trap (CFT) position of the Advanced Test Reactor (ATR). These full-size fuel plate tests are designated as the AFIP tests. The AFIP nominal fuel zone is rectangular in shape having a designed length of 21.5-in (54.61-cm), width of 1.6-in (4.064-cm), and uniform thickness of 0.014-in (0.03556-cm). This gives a nominal fuel zone volume of 0.482 in3 (7.89 cm3) per fuel plate. The AFIP test assembly has two test positions. Each test position is designed to hold 2 full-size plates, for a total of 4 full-size plates per test assembly. The AFIP test plates will be irradiated at a peak surface heat flux of about 350 W/cm2 and discharged at a peak U-235 burn-up of about 70 at.%. Based on limited irradiation testing of the monolithic (U-10Mo) fuel form, it is desirable to keep the peak fuel temperature below 250°C to achieve this, it will be necessary to keep plate heat fluxes below 500 W/cm2. Due to the heavy U-235 loading and a plate width of 1.6-in (4.064-cm), the neutron self-shielding will increase the local-to-average-ratio (L2AR) fission power near the sides of the fuel plates. To demonstrate that the AFIP experiment will meet the ATR safety requirements, a very detailed 2-dimensional (2D) Y-Z fission power profile was evaluated in order to best predict the fuel plate temperature distribution. The ability to accurately predict fuel plate power and burnup are essential to both the design of the AFIP tests as well as evaluation of the irradiated fuel performance. To support this need, a detailed MCNP Y
Cross section generation for LWR pin lattices simulations
Velasquez, Carlos E.; Macedo, Anderson A.P.; Cardoso, Fabiano S.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Costa, Antonella L. [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil); Instituto Nacional de Ciencia e Tecnologia de Reatores Nucleares Inovadores/CNPq, Brasilia, DF (Brazil); Barros, Graiciany de P. [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)
2015-07-01
The majority of the neutron data library provided with the MCNP code is set at room temperature. Therefore, it is important to generate continuous energy cross section library for MCNP code for different temperatures. To evaluate the methodology used, the criticality calculations obtained using MCNP with the cross section generated at DEN/UFMG, are compared with the criticality data from the International Handbook of Evaluated Reactor Physics Benchmarks Experiments about the PIN lattices for light water reactors. It was used nuclear data from the ENDF-VII.1, JEFF-3.1 and TENDL-2014, which were processed using the NJOY99 code for different energies and temperatures. This code provides the nuclear data in ACE libraries, which then are added to MCNP libraries to perform the simulations. The results indicate the methodology efficiency developed by DEN/UFMG. (author)
Application of fully ceramic microencapsulated fuels in light water reactors
Gentry, C.; George, N.; Maldonado, I. [Dept. of Nuclear Engineering, Univ. of Tennessee-Knoxville, Knoxville, TN 37996-2300 (United States); Godfrey, A.; Terrani, K.; Gehin, J. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)
2012-07-01
This study performs a preliminary evaluation of the feasibility of incorporation of Fully Ceramic Microencapsulated (FCM) fuels in light water reactors (LWRs). In particular, pin cell, lattice, and full core analyses are carried out on FCM fuel in a pressurized water reactor (PWR). Using uranium-based fuel and Pu/Np-based fuel in TRistructural isotropic (TRISO) particle form, each fuel design was examined using the SCALE 6.1 analytical suite. In regards to the uranium-based fuel, pin cell calculations were used to determine which fuel material performed best when implemented in the fuel kernel as well as the size of the kernel and surrounding particle layers. The higher fissile material density of uranium mononitride (UN) proved to be favorable, while the parametric studies showed that the FCM particle fuel design with 19.75% enrichment would need roughly 12% additional fissile material in comparison to that of a standard UO{sub 2} rod in order to match the lifetime of an 18-month PWR cycle. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a 'color-set' array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In regards to the Pu/Np-based fuel, lattice calculations were performed to determine an optimal lattice design based on reactivity behavior, pin power peaking, and isotopic content. After obtaining a satisfactory lattice design, the feasibility of core designs fully loaded with Pu/Np FCM lattices was demonstrated using the NESTLE three-dimensional core simulator. (authors)
Application of Fully Ceramic Microencapsulated Fuels in Light Water Reactors
Gentry, Cole A [ORNL; George, Nathan M [ORNL; Maldonado, G Ivan [ORNL; Godfrey, Andrew T [ORNL; Terrani, Kurt A [ORNL; Gehin, Jess C [ORNL
2012-01-01
This study aims to perform a preliminary evaluation of the feasibility of incorporation of Fully Ceramic Microencapsulated (FCM) fuels in Light Water Reactors (LWRs). In particular pin cell, lattice, and full core analyses are carried out on FCM fuel in a pressurized water reactor. Using uranium-based fuel and transuranic (TRU) based fuel in TRistructural ISOtropic (TRISO) particle form, each fuel design was examined using the SCALE 6.1 analytical suite. In regards to the uranium-based fuel, pin cell calculations were used to determine which fuel material performed best when implemented in the fuel kernel as well as the size of the kernel and surrounding particle layers. The higher physical density of uranium mononitride (UN) proved to be favorable, while the parametric studies showed that the FCM particle fuel design would need roughly 12% additional fissile material in comparison to that of a standard UO2 rod in order to match the lifetime of an 18-month PWR cycle. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a color-set array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In regards to the TRU based fuel, lattice calculations were performed to determine an optimal lattice design based on reactivity behavior, pin power peaking, and isotopic content. After obtaining a satisfactory lattice design, feasibility of core designs fully loaded with TRU FCM lattices was demonstrated using the NESTLE three-dimensional core simulator.
Lattice Studies of Hyperon Spectroscopy
Richards, David G. [Thomas Jefferson National Accelerator Facility (TJNAF), Newport News, VA (United States)
2016-04-01
I describe recent progress at studying the spectrum of hadrons containing the strange quark through lattice QCD calculations. I emphasise in particular the richness of the spectrum revealed by lattice studies, with a spectrum of states at least as rich as that of the quark model. I conclude by prospects for future calculations, including in particular the determination of the decay amplitudes for the excited states.
Beland, Laurent Karim; El-Mellouhi, Fedwa; Mousseau, Normand
2010-03-01
Using a topological classification of eventsfootnotetextB. D. McKay, Congressus Numerantium 30, 45 (1981). combined with the Activation-Relaxation Technique (ART nouveau) for the generation of diffusion pathways, the kinetic ART (k-ART)footnotetextF. El-Mellouhi, N. Mousseau and L. J. Lewis, Phys Rev B, 78,15 (2008). lifts many restrictions generally associated with standard kinetic Monte Carlo algorithms. In particular, it can treat on and off-lattice atomic positions and handles exactly long-range elastic deformation. Here we introduce a set of modifications to k-ART that reduce the computational cost of the algorithm to near order 1 and show applications of the algorithm to the diffusion of vacancy and interstitial complexes in large models of crystalline Si (100 000 atoms).
Delayed Fission Product Gamma-Ray Transmission Through Low Enriched UO_{2} Fuel Pin Lattices in Air
Trumbull, TH [Rensselaer Polytechnic Inst., Troy, NY (United States)
2004-10-18
The transmission of delayed fission-product gamma rays through various arrangements of low-enriched UO2 fuel pin lattices in an air medium was studied. Experimental measurements, point-kernel and Monte Carlo photon transport calculations were performed to demonstrate the shielding effect of ordered lattices of fuel pins on the resulting gamma-ray dose to a detector outside the lattice. The variation of the gamma-ray dose on the outside of the lattice as a function of radial position, the so-called “channeling” effect, was analyzed. Techniques for performing experimental measurements and data reduction at Rensselaer Polytechnic Institute’s Reactor Critical Facility (RCF) were derived. An experimental apparatus was constructed to hold the arrangements of fuel pins for the measurements. A gamma-ray spectroscopy system consisting of a sodium-iodide scintillation detector was used to collect data. Measurements were made with and without a collimator installed. A point-kernel transport code was developed to map the radial dependence of the gamma-ray flux. Input files for the Monte Carlo code, MCNP, were also developed to accurately model the experimental measurements. The results of the calculations were compared to the experimental measurements. In order to determine the delayed fission-product gamma-ray source for the calculations, a technique was developed using a previously written code, DELBG and the reactor state-point data obtained during the experimental measurements. Calculations were performed demonstrating the effects of material homogenization on the gamma-ray transmission through the fuel pin lattice.Homogeneous and heterogeneous calculations were performed for all RCF fuel pin lattices as well as for a typical commercial pressurized water reactor fuel bundle. The results of the study demonstrated the effectiveness of the experimental measurements to isolate the channeling effect of delayed fission-product gamma-rays through lattices of RCF fuel pins
Flavor Physics and Lattice QCD
Bouchard, C M
2013-01-01
Our ability to resolve new physics effects is, largely, limited by the precision with which we calculate. The calculation of observables in the Standard (or a new physics) Model requires knowledge of associated hadronic contributions. The precision of such calculations, and therefore our ability to leverage experiment, is typically limited by hadronic uncertainties. The only first-principles method for calculating the nonperturbative, hadronic contributions is lattice QCD. Modern lattice calculations have controlled errors, are systematically improvable, and in some cases, are pushing the sub-percent level of precision. I outline the role played by, highlight state of the art efforts in, and discuss possible future directions of lattice calculations in flavor physics.
Cadenas Mendicoa, A. M.; Benito Hernandez, M.; Barreira Pereira, P.
2012-07-01
This study involves the development of the methodology and three-dimensional models to estimate the damage to the vessel internals of a commercial PWR reactor from irradiation history of operating cycles.
Donnellan, Thomas; Maxwell, E A; Plumpton, C
1968-01-01
Lattice Theory presents an elementary account of a significant branch of contemporary mathematics concerning lattice theory. This book discusses the unusual features, which include the presentation and exploitation of partitions of a finite set. Organized into six chapters, this book begins with an overview of the concept of several topics, including sets in general, the relations and operations, the relation of equivalence, and the relation of congruence. This text then defines the relation of partial order and then partially ordered sets, including chains. Other chapters examine the properti
Hanford reactor and separations facility advantages
1963-06-27
This document describes the advantages and limitations of Hanford production facilities. In addition to summarizing the technical parameters of the reactors and separations plants and their mechanical features, the unique aspects of these facilities to the production of special materials in which the Commission may be interested have been discussed. As the primary difference between the B-C-D-DR-F-H reactors and the K reactors and the K reactors is in the number and length of process channels. This report is addressed primarily to the 2000-tube reactors. K reactor characteristics are within the range of lattice and flexibility parameters described.
Lattice Boltzmann Model for Compressible Fluid on a Square Lattice
SUN Cheng-Hai
2000-01-01
A two-level four-direction lattice Boltzmann model is formulated on a square lattice to simulate compressible flows with a high Mach number. The particle velocities are adaptive to the mean velocity and internal energy. Therefore, the mean flow can have a high Mach number. Due to the simple form of the equilibrium distribution, the 4th order velocity tensors are not involved in the calculations. Unlike the standard lattice Boltzmann model, o special treatment is need for the homogeneity of 4th order velocity tensors on square lattices. The Navier-Stokes equations were derived by the Chapman-Enskog method from the BGK Boltzmann equation. The model can be easily extended to three-dimensional cubic lattices. Two-dimensional shock-wave propagation was simulated
Uhl, Elmar [Instituto de Quimica, Departamento de Fisico-Quimica, Universidade Federal do Rio de Janeiro, Cidade Universitaria, CT Bloco A. Rio de Janeiro, 21941-909 Rio de Janeiro (Brazil); Leitao, Alexandre A. [Departamento de Quimica, Universidade Federal de Juiz de Fora, Campus Universitario, Juiz de Fora, MG 36036-900 (Brazil); Rocha, Alexandre B., E-mail: rocha@iq.ufrj.br [Instituto de Quimica, Departamento de Fisico-Quimica, Universidade Federal do Rio de Janeiro, Cidade Universitaria, CT Bloco A. Rio de Janeiro, 21941-909 Rio de Janeiro (Brazil)
2011-11-07
Graphical abstract: Temperature dependence of oscillator strengths calculated through vibronic coupling for electronic transitions of Cu{sup +} impurity in NaF host, described by embedded cluster model. Highlights: Black-Right-Pointing-Pointer Embedded cluster model for impurity levels in the NaF:Cu{sup +} system. Black-Right-Pointing-Pointer Oscillator strengths (OSs) calculated by direct vibronic coupling method. Black-Right-Pointing-Pointer The dependence of the OS on temperature is reported. Black-Right-Pointing-Pointer OS and transition energies calculated at CASSCF and CASSCF/SOCI level. - Abstract: An embedded cluster model is used to describe electronic structure of Cu{sup +} ion in NaF host. Transition energies and oscillator strengths are calculated for the 3d{sup 10} {yields} 3d{sup 9}4s{sup 1} Cu{sup +} ligand field transitions. These are forbidden by dipole selection rules, which can, though, be broken by vibronic coupling. The basic model consists of a [CuF{sub 6}]{sup 5-} cluster surrounded by total ion potentials representing second, third and fourth neighbors to the central Cu{sup +}. The resulting structure is placed inside a cube of point charges to take long distance Coulomb interactions into account. Variations of this basic model needed especially to the calculation of transition energy. The oscillator strengths are calculated by the direct vibronic coupling method we have previously proposed. The effect of temperature on the value of the oscillator strength is calculated for the first time as well as their absolute value. Results are in good agreement with available experiment.
Webber, J. Beau W., E-mail: J.B.W.Webber@kent.ac.uk
2013-05-15
Neutron scattering offers a length-scale-independent method of probing structured matter on an atomic scale through nano-scale to meso-scale. A protocol is presented that provides a versatile method of determining structure, by comparison of measured and calculated neutron scattering, for any structural distribution that can be described algebraically or numerically, requiring no particular model other than the model of the structure, and needing no adjustable parameters other than the scale and other parameters describing the physical model. The method enables the direct comparison of measured and calculated scattering from structured matter: from simple finite and infinite bodies, from extended regular array of pores, or from extended arrays of pores with a partially randomised character. Examples are given for the radial distributions of a range of regular bodies, of large arrays of highly ordered porous materials such as templated SBA-15 and MCM-41 silicas, as well as for more disordered materials such as sol–gel silicas. Monte Carlo integration of the calculated scattering for ensembles of up to about 100,000 pores has been studied using these techniques. The method enables the calculation of the solid–solid density correlation function G(r) for model systems, and hence, by Fourier transformation, the expected scattering. Example measured scattering is compared with the calculated scattering, with further data presented in a related paper. The technique allows the direct calculation and comparison with measurement of all three of the main pore structural parameters: lattice spacing, pore diameter, and pore-wall thickness. Example SBA-15 wide and small angle neutron scattering (SANS) data, measured on NIMROD (the Near and InterMediate Range Order Diffractometer at ISIS), is used as an initial evaluation of the applicability of the techniques. The method is also applicable to determining structure by comparing calculating with measured diffraction broadening
Lattice models of ionic systems
Kobelev, Vladimir; Kolomeisky, Anatoly B.; Fisher, Michael E.
2002-05-01
A theoretical analysis of Coulomb systems on lattices in general dimensions is presented. The thermodynamics is developed using Debye-Hückel theory with ion-pairing and dipole-ion solvation, specific calculations being performed for three-dimensional lattices. As for continuum electrolytes, low-density results for simple cubic (sc), body-centered cubic (bcc), and face-centered cubic (fcc) lattices indicate the existence of gas-liquid phase separation. The predicted critical densities have values comparable to those of continuum ionic systems, while the critical temperatures are 60%-70% higher. However, when the possibility of sublattice ordering as well as Debye screening is taken into account systematically, order-disorder transitions and a tricritical point are found on sc and bcc lattices, and gas-liquid coexistence is suppressed. Our results agree with recent Monte Carlo simulations of lattice electrolytes.
Wang, Xianlong, E-mail: WangXianlong@uestc.edu.cn, E-mail: pbeckman@brynmawr.edu [Key Laboratory for NeuroInformation of Ministry of Education, School of Life Science and Technology, University of Electronic Science and Technology of China, 4 North Jianshe Rd., 2nd Section, Chengdu 610054 (China); Mallory, Frank B. [Department of Chemistry, Bryn Mawr College, 101 North Merion Ave., Bryn Mawr, Pennsylvania 19010-2899 (United States); Mallory, Clelia W. [Department of Chemistry, Bryn Mawr College, 101 North Merion Ave., Bryn Mawr, Pennsylvania 19010-2899 (United States); Department of Chemistry, University of Pennsylvania, Philadelphia, Pennsylvania 19104-6323 (United States); Odhner, Hosanna R.; Beckmann, Peter A., E-mail: WangXianlong@uestc.edu.cn, E-mail: pbeckman@brynmawr.edu [Department of Physics, Bryn Mawr College, 101 North Merion Ave., Bryn Mawr, Pennsylvania 19010-2899 (United States)
2014-05-21
We report ab initio density functional theory electronic structure calculations of rotational barriers for t-butyl groups and their constituent methyl groups both in the isolated molecules and in central molecules in clusters built from the X-ray structure in four t-butyl aromatic compounds. The X-ray structures have been reported previously. We also report and interpret the temperature dependence of the solid state {sup 1}H nuclear magnetic resonance spin-lattice relaxation rate at 8.50, 22.5, and 53.0 MHz in one of the four compounds. Such experiments for the other three have been reported previously. We compare the computed barriers for methyl group and t-butyl group rotation in a central target molecule in the cluster with the activation energies determined from fitting the {sup 1}H NMR spin-lattice relaxation data. We formulate a dynamical model for the superposition of t-butyl group rotation and the rotation of the t-butyl group's constituent methyl groups. The four compounds are 2,7-di-t-butylpyrene, 1,4-di-t-butylbenzene, 2,6-di-t-butylnaphthalene, and 3-t-butylchrysene. We comment on the unusual ground state orientation of the t-butyl groups in the crystal of the pyrene and we comment on the unusually high rotational barrier of these t-butyl groups.
汤晓斌; 谢芹; 耿长冉; 陈达
2012-01-01
超临界水堆是国际第Ⅳ代核能系统论坛推荐的六种第Ⅳ代核电反应堆堆型之一,与现有的轻水堆相比,具有热效率高、系统结构简单、造价低等优点.建立了MCNP程序下的超临界水堆堆芯物理计算模型,解决了燃料组件几何结构过于复杂精细难以建模的技术难题；考虑了堆芯轴向冷却剂密度的不均匀分布,计算并分析各区域的中子能谱分布；对失水事故下的超临界水冷堆安全性进行了分析,研究了不同区域冷却剂丢失程度对反应性及有效增殖系数的影响,表明所设计堆型具有较高的安全性；分析处理失水事故的应对措施,验证了使用注入硼水措施处理超临界水冷堆失水事故的可行性.%The supercritical water reactor is one of the six reactors recommended by Generation IV International Forum, Compared with existing light water reactors, the supercritical water reactor has advantages of high thermal efficiency, simplified system structure and low cost. The physical model of the supercritical water reactor is established with MCNP program in this paper, which solves the problem of intricate geometry of fuel assembly. The change of coolant density along the axis is considered and the neutron spectrum distribution of different regions of the core is calculated. The safety in loss of coolant accident for the supercritical water reactor and the effect of missing coolant in different regions on the reactivity and effective multiplication factor analyzed. The results show the supercritical water reactor core has high security. The countermeasures of loss of coolant accident is studied and the effectiveness of boron water cooling is validated. The research not only provide important reference for the construction and security analysis of the supercritical water reactor, but also has great significance for the application and development of the supercritical water reactor.
Reactivity feedback coefficients Pakistan research reactor-1 using PRIDE code
Mansoor, Ali; Ahmed, Siraj-ul-Islam; Khan, Rustam [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Inam-ul-Haq [Comsats Institute of Information Technology, Islamabad (Pakistan). Dept. of Physics
2017-05-15
Results of the analyses performed for fuel, moderator and void's temperature feedback reactivity coefficients for the first high power core configuration of Pakistan Research Reactor - 1 (PARR-1) are summarized. For this purpose, a validated three dimensional model of PARR-1 core was developed and confirmed against the reference results for reactivity calculations. The ''Program for Reactor In-Core Analysis using Diffusion Equation'' (PRIDE) code was used for development of global (3-dimensional) model in conjunction with WIMSD4 for lattice cell modeling. Values for isothermal fuel, moderator and void's temperature feedback reactivity coefficients have been calculated. Additionally, flux profiles for the five energy groups were also generated.
Chiral Four-Dimensional Heterotic Covariant Lattices
Beye, Florian
2014-01-01
In the covariant lattice formalism, chiral four-dimensional heterotic string vacua are obtained from certain even self-dual lattices which completely decompose into a left-mover and a right-mover lattice. The main purpose of this work is to classify all right-mover lattices that can appear in such a chiral model, and to study the corresponding left-mover lattices using the theory of lattice genera. In particular, the Smith-Minkowski-Siegel mass formula is employed to calculate a lower bound on the number of left-mover lattices. Also, the known relationship between asymmetric orbifolds and covariant lattices is considered in the context of our classification.
Lattice dynamics of Cs2NaYbF6 and Cs2NaYF6 elpasolites: Ab initio calculation
Chernyshev, V. A.; Petrov, V. P.; Nikiforov, A. E.; Zakir'yanov, D. O.
2015-06-01
The ab initio calculations of the crystal structure and the phonon spectrum of Cs2NaYbF6 and Cs2NaYF6 crystals with the elpasolite structure have been performed. The frequencies and types of fundamental vibrations have been determined. The calculations have been performed in the framework of the density functional theory using the molecular orbital method with hybrid functionals in the CRYSTAL09 program developed for the simulation of periodic structures. The outer 5 s and 5 p shells of the rare-earth ion have been described in Gaussian-type basis sets. The influence of inner shells, including 4 f electron shells, on the outer shells has been described using the pseudopotential. It has been shown that this approach allows the description of the phonon spectrum with the inclusion of the splitting of the longitudinal and transverse optical modes.
Introduction to lattice gauge theory
Gupta, R.
The lattice formulation of Quantum Field Theory (QFT) can be exploited in many ways. We can derive the lattice Feynman rules and carry out weak coupling perturbation expansions. The lattice then serves as a manifestly gauge invariant regularization scheme, albeit one that is more complicated than standard continuum schemes. Strong coupling expansions: these give us useful qualitative information, but unfortunately no hard numbers. The lattice theory is amenable to numerical simulations by which one calculates the long distance properties of a strongly interacting theory from first principles. The observables are measured as a function of the bare coupling g and a gauge invariant cut-off approx. = 1/alpha, where alpha is the lattice spacing. The continuum (physical) behavior is recovered in the limit alpha yields 0, at which point the lattice artifacts go to zero. This is the more powerful use of lattice formulation, so in these lectures the author focuses on setting up the theory for the purpose of numerical simulations to get hard numbers. The numerical techniques used in Lattice Gauge Theories have their roots in statistical mechanics, so it is important to develop an intuition for the interconnection between quantum mechanics and statistical mechanics.
Theory of vortex-lattice melting in a one-dimensional optical lattice
Snoek, M.; Stoof, H.T.C.
2006-01-01
We investigate quantum and temperature fluctuations of a vortex lattice in a one-dimensional optical lattice. We discuss in particular the Bloch bands of the Tkachenko modes and calculate the correlation function of the vortex positions along the direction of the optical lattice. Because of the
Lattice Boltzmann model for nanofluids
Xuan Yimin; Yao Zhengping [Nanjing University of Science and Technology, School of Power Engineering, Nanjing (China)
2005-01-01
A nanofluid is a particle suspension that consists of base liquids and nanoparticles and has great potential for heat transfer enhancement. By accounting for the external and internal forces acting on the suspended nanoparticles and interactions among the nanoparticles and fluid particles, a lattice Boltzmann model is proposed for simulating flow and energy transport processes inside the nanofluids. First, we briefly introduce the conventional lattice Boltzmann model for multicomponent systems. Then, we discuss the irregular motion of the nanoparticles and inherent dynamic behavior of nanofluids and describe a lattice Boltzmann model for simulating nanofluids. Finally, we conduct some calculations for the distribution of the suspended nanoparticles. (orig.)
Lattice Platonic Solids and their Ehrhart polynomial
Ionascu, Eugen J
2011-01-01
First, we calculate the Ehrhart polynomial associated to an arbitrary cube with integer coordinates for its vertices. Then, we use this result to derive relationships between the Ehrhart polynomials for regular lattice tetrahedrons and those for regular lattice octahedrons. These relations allow one to reduce the calculation of these polynomials to only one coefficient.
Preliminary design concept of a subcritical reactor using available resources
Churnetski, E.L. [Oak Ridge Y-12 Plant, TN (United States); Hoyny, V.; Chaudhuri, B.R.; Taprantzis, A.; Yavas, A. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering
1993-12-31
During the Fall 1993 semester, a project was initiated within the Nuclear Engineering Department of the University of Tennessee with the objective of developing a design for a subcritical reactor with maximized multiplication factor using materials currently available. Such a device, if constructed, would serve as a teaching tool for the Department of Nuclear Engineering. Design work was conducted as a large number of computer calculations, with trial pile configurations based on fundamental nuclear engineering principles, in an effort to maximize multiplication factor through fuel element geometry, moderator type, fissile/moderator ratio, and reflector character. The principal objective of the design group for the early phase of this project was to present several possible ``baseline`` reactor designs and identify directions for improvements. For the sake of calculational ease, the cores analyzes to date have been of nearly cubic shape. The SCALE CSAS25 software which runs KENO.Va, a Monte Carlo code, was used for all neutronics calculations. The baseline reactors resulting from work to date are cuboidal in shape and graphite reflected. Two types of fuel element geometries are proposed, a typical triangular pitch rod lattice and an arrangement of discrete fuel slugs placed in a lattice corresponding to body centered cubic packing. The latter arrangement provides slightly higher multiplication factors than the former. Calculations were performed for both graphite and heavy water moderation with heavy water moderation producing considerably higher multiplication factors, as expected. In general, the maximum k{sub eff} for the reactors are in the range of 0.5 to 0.9, well subcritical, except in the cases of the extreme possible values of fuel assay where critical configurations are possible. In these cases, designs with reduced fuel loading are recommended to assure a subcritical multiplication factor.
赵文清
2014-01-01
Reactor vessel internal upper support assembly adopt shell element combined with beam ele-ment to create model,shell element to create model and solid element to create model,which carry through respective finite element analysis calculation and stress evaluation.The stress calculation effect of reactor vessel internal upper support assembly adopting solid element model created is conservative and high precision by means of modification and simplification about preexistent stress analysis method related to reactor vessel internal,which can satisfy the demand of RCC-M criterion.The calculation method of reactor vessel internal upper support assembly adopting solid element to create model is simplification and practicality by comparing with different creating model mode about stress calculation mentioned above, which may apply to the stress analysis and evaluation of reactor vessel internal assembly related to differ-ent reactor core.%堆内构件上支承组件采用不同的建模方法，分别采用壳单元和梁单元相组合的建模模式、壳单元和壳单元相组合的建模模式、实体单元建模的模式，对堆内构件上支承组件进行了有限元应力计算，比较了不同建模模式下应力计算的各自特点，堆内构件上支承组件实体单元建模模式应力计算结果精确并能满足RC C-M规范应力评定要求，壳单元和梁单元相组合的建模模式、壳单元和壳单元相组合的建模模式应力计算结果保守且应力评定需等效处理其计算结果。堆内构件上支承组件采用整体实体单元全模型建模的计算方法，计算精确且应力评定简单直接，它可应用于其他工况和不同堆芯堆内构件应力计算及其应力评定。
网壳结构寿命周期总费用的计算方法研究%Study on the calculation of life-cycle cost of latticed shell structures
杜文风; 高博青; 董石麟
2011-01-01
The huge losses of modern structural systems caused by disasters have confirmed the necessity of evaluating the life-cycle cost in structural design and guiding the construction. The calculation method of life-cycle cost of latticed shell structures is studied. The formula for initial building cost, maintenance cost and loss cost is proposed, and the formula for life-cycle cost is formulated considering the occurrence probability of load in the life-cycle of latticed shell structures. The validity of the calculation method and procedure is tested, and the results show that the added invest in the initial building cost of latticed shell structures mil be compensated by decrease of loss cost in structure life, and that the minimum of life-cycle loss is the optimum state of structure design.%现代结构系统在灾害中产生巨大经济损失的特点,使人们意识到在结构设计之初就对结构的寿命周期总费用进行有效评估、并反过来指导工程设计具有重要意义.对网壳结构寿命周期总费用的计算方法进行研究,分别提出初始造价、维护费用和失效损失的计算公式,并考虑寿命周期内网壳结构可能受到的荷载发生的概率,建立寿命周期总费用的计算方法,编制网壳结构寿命周期总费用的计算程序.通过一具体的网壳结构工程算例验证上述方法和程序的有效性.研究结果表明,网壳结构初始造价的追加投资会由于结构使用过程中失效损失费用的减小而得到补偿,而寿命周期总费用最小正是结构设计的最优状态,按照现行规范设计方法得到的结构配置方案不一定是寿命周期总费用最小的方案.
Baryon spectroscopy in lattice QCD
Derek B. Leinweber; Wolodymyr Melnitchouk; David Richards; Anthony G. Williams; James Zanotti
2004-04-01
We review recent developments in the study of excited baryon spectroscopy in lattice QCD. After introducing the basic methods used to extract masses from correlation functions, we discuss various interpolating fields and lattice actions commonly used in the literature. We present a survey of results of recent calculations of excited baryons in quenched QCD, and outline possible future directions in the study of baryon spectra.
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...
Beane, Silas
2009-01-01
Recent studies by the NPLQCD collaboration of hadronic interactions using lattice QCD are reviewed, with an emphasis on a recent calculation of meson-baryon scattering lengths. Ongoing high-statistics calculations of baryon interactions are also reviewed. In particular, new insights into the signal/noise problems that plague correlation functions involving baryons are discussed.
Stable kagome lattices from group IV elements
Leenaerts, O.; Schoeters, B.; Partoens, B.
2015-03-01
A thorough investigation of three-dimensional kagome lattices of group IV elements is performed with first-principles calculations. The investigated kagome lattices of silicon and germanium are found to be of similar stability as the recently proposed carbon kagome lattice. Carbon and silicon kagome lattices are both direct-gap semiconductors but they have qualitatively different electronic band structures. While direct optical transitions between the valence and conduction bands are allowed in the carbon case, no such transitions can be observed for silicon. The kagome lattice of germanium exhibits semimetallic behavior but can be transformed into a semiconductor after compression.
Development of Burnup Calculation Function in Reactor Monte Carlo Code RMC%堆用蒙卡程序燃耗计算功能开发
佘顶; 王侃; 余纲林
2012-01-01
This paper presents the burnup calculation capability of RMC, which is a new Monte Carlo (MC) neutron transport code developed by Reactor Engineering Analysis Laboratory (REAL) in Tsinghua university of China. Unlike most of existing MC depletion codes which explicitly couple the depletion module, RMC incorporates ORIGEN 2.1 in an implicit way. Different burn step strategies, including the middle-of-step approximation and the predictor-corrector method, are adopted by RMC to assure the accuracy under large burnup step size. RMC employs a spectrum-based method of tallying one-group cross section, which can considerably saves computational time with negligible accuracy loss. According to the validation results of benchmarks and examples, it is proved that the burnup function of RMC performs quite well in accuracy and efficiency.%堆用蒙卡程序(RMC)是由清华大学工程物理系REAL实验室自主开发的用于反应堆物理分析的中子输运蒙卡程序,本文主要介绍其燃耗计算功能的开发与验证.RMC的燃耗计算功能具有的特点:内部耦合ORIGEN,相比于外耦合方式,更加灵活和高效；使用基于能谱的单群截面统计方法,可在保证精度的前提下,显著提高计算效率；采取预估修正和中点近似等多种燃耗步策略,减小大燃耗步长时的计算误差.通过计算压水堆栅元、沸水堆组件、快堆等一系列基准题和算例,验证了RMC燃耗计算的正确性和速度优势.
Working Group Report: Lattice Field Theory
Blum, T.; et al.,
2013-10-22
This is the report of the Computing Frontier working group on Lattice Field Theory prepared for the proceedings of the 2013 Community Summer Study ("Snowmass"). We present the future computing needs and plans of the U.S. lattice gauge theory community and argue that continued support of the U.S. (and worldwide) lattice-QCD effort is essential to fully capitalize on the enormous investment in the high-energy physics experimental program. We first summarize the dramatic progress of numerical lattice-QCD simulations in the past decade, with some emphasis on calculations carried out under the auspices of the U.S. Lattice-QCD Collaboration, and describe a broad program of lattice-QCD calculations that will be relevant for future experiments at the intensity and energy frontiers. We then present details of the computational hardware and software resources needed to undertake these calculations.
Advances in Lattice Quantum Chromodynamics
McGlynn, Greg
In this thesis we make four contributions to the state of the art in numerical lattice simulations of quantum chromodynamics (QCD). First, we present the most detailed investigation yet of the autocorrelations of topological observations in hybrid Monte Carlo simulations of QCD and of the effects of the boundary conditions on these autocorrelations. This results in a numerical criterion for deciding when open boundary conditions are useful for reducing these autocorrelations, which are a major barrier to reliable calculations at fine lattice spacings. Second, we develop a dislocation-enhancing determinant, and demonstrate that it reduces the autocorrelation time of the topological charge. This alleviates problems with slow topological tunneling at fine lattice spacings, enabling simulations on fine lattices to be completed with much less computational effort. Third, we show how to apply the recently developed zMobius technique to hybrid Monte Carlo evolutions with domain wall fermions, achieving nearly a factor of two speedup in the light quark determinant, the single most expensive part of the calculation. The dislocation-enhancing determinant and the zMobius technique have enabled us to begin simulations of fine ensembles with four flavors of dynamical domain wall quarks. Finally, we show how to include the previously-neglected G1 operator in nonperturbative renormalization of the DeltaS = 1 effective weak Hamiltonian on the lattice. This removes an important systematic error in lattice calculations of weak matrix elements, in particular the important K → pipi decay.
Kazantsev, A. A., E-mail: kazantsevanatoly@gmail.com [Experimental Scientific Research and Methodology Center Simulation Systems (Russian Federation); Sergeev, V. V. [Leipunsky Institute of Physics and Power Engineering (Russian Federation); Kochnov, O. Yu. [Karpov Institute of Physical Chemistry (Obninsk Branch) (Russian Federation)
2015-12-15
The temperature regime is calculated for two different designs of containers with uranium-bearing material for the upgraded VVR-Ts research reactor facility (IVV.10M). The containers are to be used in the production of {sup 99}Mo. It is demonstrated that the modification of the container design leads to a considerable temperature reduction and an increase in the near-wall boiling margin and allows one to raise the amount of material loaded into the container. The calculations were conducted using the international thermohydraulic contour code TRAC intended to analyze the technical safety of water-cooled nuclear power units.
Schaefer, Stefan [DESY (Germany). Neumann Inst. for Computing
2016-11-01
These configurations are currently in use in many on-going projects carried out by researchers throughout Europe. In particular this data will serve as an essential input into the computation of the coupling constant of QCD, where some of the simulations are still on-going. But also projects computing the masses of hadrons and investigating their structure are underway as well as activities in the physics of heavy quarks. As this initial project of gauge field generation has been successful, it is worthwhile to extend the currently available ensembles with further points in parameter space. These will allow to further study and control systematic effects like the ones introduced by the finite volume, the non-physical quark masses and the finite lattice spacing. In particular certain compromises have still been made in the region where pion masses and lattice spacing are both small. This is because physical pion masses require larger lattices to keep the effects of the finite volume under control. At light pion masses, a precise control of the continuum extrapolation is therefore difficult, but certainly a main goal of future simulations. To reach this goal, algorithmic developments as well as faster hardware will be needed.
Schmitz, T.; Blaickner, M.; Schütz, C.
2010-01-01
To establish Boron Neutron Capture Therapy (BNCT) for non-resectable liver metastases and for in vitro experiments at the TRIGA Mark II reactor at the University of Mainz, Germany, it is necessary to have a reliable dose monitoring system. The in vitro experiments are used to determine the relative...... to the neutron fluence directly. Results and discussion. Gold foil activation and the pin-diode are reliable fluence measurement systems for the TRIGA reactor, Mainz. Alanine dosimetry of the photon field and charged particle field from secondary reactions can in principle be carried out in combination with MC...
Palmrose, D.E. (EG and G Idaho, Inc., Idaho Falls, ID (United States)); Mandl, R.M. (Siemens AG, Berlin (Germany))
1991-01-01
There has been recent interest in the United States concerning the loss of residual heat removal system (RHRS) under reduced coolant inventory conditions for pressurized water reactors. This issue is also of interest in the Federal Republic of Germany and an experiment was performed in the integral PKL-HI experimental facility at Siemens-KWU to supply applicable data. Recently, an NRC-sponsored effort has been undertaken at the Idaho-National Engineering Laboratory to identify and analyze the important thermal-hydraulic phenomena in pressurized water reactors following the long term loss-of-RHRS during reduced inventory operation. The thermal-hydraulic response of a closed reactor coolant system during such a transient is investigated in this report. Some of the specific processes investigated include: reflux condensation in the steam generators, the corresponding pressure increase in the reactor coolant system, and void fraction distributions on the primary side of the system. Mathematical models of these and other physical processes Experiment B4.5.
Design and calculation for the main shielding layer of researching reactor%反应堆主屏蔽的设计与计算
钟文发; 胡永明; 钟兆鹏
2001-01-01
为使反应堆处于运行状态时，对辐射源的屏蔽满足辐射安全的要求，以及对堆的各部件和材料满足辐射限制的要求，必须设计堆的主屏蔽层。介绍了主屏蔽的设计与计算方法，以研究堆为设计实例，给出了主屏蔽的主要计算结果表明，以池水和重混凝土作生物屏蔽能满足辐射安全限值的要求，设计的主屏蔽层是适宜的。%To satisfy the national safety criteria of the radiationshielding when the reactor is in operation, We must complete the design of shielding for the main body of reactor. This paper introduces the design and computing methods of the shielding for the main body of reactor, using the researching reactor as an example, and gives the results of the computing for the radiation shielding. The results show that it can satisfy the safety criteria of radiation to use water and heavy concrete as biological layer for radiation, so the design of the main layer for radiation shielding is reliable.
Liu, Hong-Xia; Tang, Fu-Ling; Xue, Hong-Tao; Zhang, Yu; Cheng, Yu-Wen; Feng, Yu-Dong
2016-12-01
Using the first-principles plane-wave calculations within density functional theory, the perfect bi-layer and monolayer terminated WZ-CIS (100)/WZ-CdS (100) interfaces are investigated. After relaxation the atomic positions and the bond lengths change slightly on the two interfaces. The WZ-CIS/WZ-CdS interfaces can exist stably, when the interface bonding energies are -0.481 J/m2 (bi-layer terminated interface) and -0.677 J/m2 (monolayer terminated interface). Via analysis of the density of states, difference charge density and Bader charges, no interface state is found near the Fermi level. The stronger adhesion of the monolayer terminated interface is attributed to more electron transformations and orbital hybridizations, promoting stable interfacial bonds between atoms than those on a bi-layer terminated interface. Project supported by the National Natural Science Foundation of China (Grant Nos. 11164014 and 11364025) and the Gansu Science and Technology Pillar Program, China (Grant No. 1204GKCA057).
Gupta, R.
1998-12-31
The goal of the lectures on lattice QCD (LQCD) is to provide an overview of both the technical issues and the progress made so far in obtaining phenomenologically useful numbers. The lectures consist of three parts. The author`s charter is to provide an introduction to LQCD and outline the scope of LQCD calculations. In the second set of lectures, Guido Martinelli will discuss the progress they have made so far in obtaining results, and their impact on Standard Model phenomenology. Finally, Martin Luescher will discuss the topical subjects of chiral symmetry, improved formulation of lattice QCD, and the impact these improvements will have on the quality of results expected from the next generation of simulations.
Lattice Quantum Chromodynamics
Sachrajda, C. T.
2016-10-01
I review the the application of the lattice formulation of QCD and large-scale numerical simulations to the evaluation of non-perturbative hadronic effects in Standard Model Phenomenology. I present an introduction to the elements of the calculations and discuss the limitations both in the range of quantities which can be studied and in the precision of the results. I focus particularly on the extraction of the QCD parameters, i.e. the quark masses and the strong coupling constant, and on important quantities in flavour physics. Lattice QCD is playing a central role in quantifying the hadronic effects necessary for the development of precision flavour physics and its use in exploring the limits of the Standard Model and in searches for inconsistencies which would signal the presence of new physics.
Lattices of dielectric resonators
Trubin, Alexander
2016-01-01
This book provides the analytical theory of complex systems composed of a large number of high-Q dielectric resonators. Spherical and cylindrical dielectric resonators with inferior and also whispering gallery oscillations allocated in various lattices are considered. A new approach to S-matrix parameter calculations based on perturbation theory of Maxwell equations, developed for a number of high-Q dielectric bodies, is introduced. All physical relationships are obtained in analytical form and are suitable for further computations. Essential attention is given to a new unified formalism of the description of scattering processes. The general scattering task for coupled eigen oscillations of the whole system of dielectric resonators is described. The equations for the expansion coefficients are explained in an applicable way. The temporal Green functions for the dielectric resonator are presented. The scattering process of short pulses in dielectric filter structures, dielectric antennas and lattices of d...
Borsanyi, Sz; Kampert, K H; Katz, S D; Kawanai, T; Kovacs, T G; Mages, S W; Pasztor, A; Pittler, F; Redondo, J; Ringwald, A; Szabo, K K
2016-01-01
We present a full result for the equation of state (EoS) in 2+1+1 (up/down, strange and charm quarks are present) flavour lattice QCD. We extend this analysis and give the equation of state in 2+1+1+1 flavour QCD. In order to describe the evolution of the universe from temperatures several hundreds of GeV to several tens of MeV we also include the known effects of the electroweak theory and give the effective degree of freedoms. As another application of lattice QCD we calculate the topological susceptibility (chi) up to the few GeV temperature region. These two results, EoS and chi, can be used to predict the dark matter axion's mass in the post-inflation scenario and/or give the relationship between the axion's mass and the universal axionic angle, which acts as a initial condition of our universe.
Lattice Quantum Chromodynamics
Sachrajda, C T
2016-01-01
I review the the application of the lattice formulation of QCD and large-scale numerical simulations to the evaluation of non-perturbative hadronic effects in Standard Model Phenomenology. I present an introduction to the elements of the calculations and discuss the limitations both in the range of quantities which can be studied and in the precision of the results. I focus particularly on the extraction of the QCD parameters, i.e. the quark masses and the strong coupling constant, and on important quantities in flavour physics. Lattice QCD is playing a central role in quantifying the hadronic effects necessary for the development of precision flavour physics and its use in exploring the limits of the Standard Model and in searches for inconsistencies which would signal the presence of new physics.
Development of improved methods for the LWR lattice physics code EPRI-CELL
Williams, M.L.; Wright, R.Q.; Barhen, J.
1982-07-01
A number of improvements have been made by ORNL to the lattice physics code EPRI-CELL (E-C) which is widely used by utilities for analysis of power reactors. The code modifications were made mainly in the thermal and epithermal routines and resulted in improved reactor physics approximations and more efficient running times. The improvements in the thermal flux calculation included implementation of a group-dependent rebalance procedure to accelerate the iterative process and a more rigorous calculation of interval-to-interval collision probabilities. The epithermal resonance shielding methods used in the code have been extensively studied to determine its major approximations and to examine the sensitivity of computed results to these approximations. The study has resulted in several improvements in the original methodology.
Lattice QCD for nuclear physics
Meyer, Harvey
2015-01-01
With ever increasing computational resources and improvements in algorithms, new opportunities are emerging for lattice gauge theory to address key questions in strongly interacting systems, such as nuclear matter. Calculations today use dynamical gauge-field ensembles with degenerate light up/down quarks and the strange quark and it is possible now to consider including charm-quark degrees of freedom in the QCD vacuum. Pion masses and other sources of systematic error, such as finite-volume and discretization effects, are beginning to be quantified systematically. Altogether, an era of precision calculation has begun, and many new observables will be calculated at the new computational facilities. The aim of this set of lectures is to provide graduate students with a grounding in the application of lattice gauge theory methods to strongly interacting systems, and in particular to nuclear physics. A wide variety of topics are covered, including continuum field theory, lattice discretizations, hadron spect...
Irreversible stochastic processes on lattices
Nord, R.S.
1986-01-01
Models for irreversible random or cooperative filling of lattices are required to describe many processes in chemistry and physics. Since the filling is assumed to be irreversible, even the stationary, saturation state is not in equilibrium. The kinetics and statistics of these processes are described by recasting the master equations in infinite hierarchical form. Solutions can be obtained by implementing various techniques: refinements in these solution techniques are presented. Programs considered include random dimer, trimer, and tetramer filling of 2D lattices, random dimer filling of a cubic lattice, competitive filling of two or more species, and the effect of a random distribution of inactive sites on the filling. Also considered is monomer filling of a linear lattice with nearest neighbor cooperative effects and solve for the exact cluster-size distribution for cluster sizes up to the asymptotic regime. Additionally, a technique is developed to directly determine the asymptotic properties of the cluster size distribution. Finally cluster growth is considered via irreversible aggregation involving random walkers. In particular, explicit results are provided for the large-lattice-size asymptotic behavior of trapping probabilities and average walk lengths for a single walker on a lattice with multiple traps. Procedures for exact calculation of these quantities on finite lattices are also developed.
R-matrix parameters in reactor applications
Hwang, R.N.
1992-01-01
The key role of the resonance phenomena in reactor applications manifests through the self-shielding effect. The basic issue involves the application of the microscopic cross sections in the macroscopic reactor lattices consisting of many nuclides that exhibit resonance behavior. To preserve the fidelity of such a effect requires the accurate calculations of the cross sections and the neutron flux in great detail. This clearly not possible without viable resonance data. Recently released ENDF/B VI resonance data in the resolved range especially reflect the dramatic improvement in two important areas; namely, the significant extension of the resolved resonance ranges accompanied by the availability of the R-matrix parameters of the Reich-Moore type. Aside from the obvious increase in computing time required for the significantly greater number of resonances, the main concern is the compatibility of the Riech-Moore representation to the existing reactor processing codes which, until now, are based on the traditional cross section formalisms. This purpose of this paper is to summarize our recent efforts to facilitate implementation of the proposed methods into the production codes at ANL.
Nuclear Physics from Lattice QCD
William Detmold, Silas Beane, Konstantinos Orginos, Martin Savage
2011-01-01
We review recent progress toward establishing lattice Quantum Chromodynamics as a predictive calculational framework for nuclear physics. A survey of the current techniques that are used to extract low-energy hadronic scattering amplitudes and interactions is followed by a review of recent two-body and few-body calculations by the NPLQCD collaboration and others. An outline of the nuclear physics that is expected to be accomplished with Lattice QCD in the next decade, along with estimates of the required computational resources, is presented.
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.
Jo, YuGwon; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)
2015-05-15
In this paper, the FSS iteration method is applied to the fast reactor where the neutron mean-free-path is around 10 times longer than that in the thermal reactor. The FSS iteration method with domain-based parallelism is tested on a two-dimensional continuous-energy fast reactor test problem. The multiplication factor and the pinwise fission-rate distributions of the FSS iteration method show good agreements with those of the conventional power method. A local domain is chosen as a cluster of 19 assemblies, taking into account the longer neutron mean-free-path. In the future, another type of local domain can be defined to take into account reflector assemblies and shield assemblies with appropriate boundary conditions. In the test problem, the multiplication factor and the pinwise fission-rate distributions of the FSS iteration method show good agreements with those of the conventional power method. Although the domain decomposition is easily achieved by the FSS iteration method, load-imbalance of local problems causes idle times in the processors. Applying the source splitting scheme and assigning different numbers of processors to local problems will reduce this problem.
Nucleon structure from lattice QCD
Dinter, Simon
2012-11-13
In this thesis we compute within lattice QCD observables related to the structure of the nucleon. One part of this thesis is concerned with moments of parton distribution functions (PDFs). Those moments are essential elements for the understanding of nucleon structure and can be extracted from a global analysis of deep inelastic scattering experiments. On the theoretical side they can be computed non-perturbatively by means of lattice QCD. However, since the time lattice calculations of moments of PDFs are available, there is a tension between these lattice calculations and the results from a global analysis of experimental data. We examine whether systematic effects are responsible for this tension, and study particularly intensively the effects of excited states by a dedicated high precision computation. Moreover, we carry out a first computation with four dynamical flavors. Another aspect of this thesis is a feasibility study of a lattice QCD computation of the scalar quark content of the nucleon, which is an important element in the cross-section of a heavy particle with the nucleon mediated by a scalar particle (e.g. Higgs particle) and can therefore have an impact on Dark Matter searches. Existing lattice QCD calculations of this quantity usually have a large error and thus a low significance for phenomenological applications. We use a variance-reduction technique for quark-disconnected diagrams to obtain a precise result. Furthermore, we introduce a new stochastic method for the calculation of connected 3-point correlation functions, which are needed to compute nucleon structure observables, as an alternative to the usual sequential propagator method. In an explorative study we check whether this new method is competitive to the standard one. We use Wilson twisted mass fermions at maximal twist in all our calculations, such that all observables considered here have only O(a{sup 2}) discretization effects.
Kaon fluctuations from lattice QCD
Noronha-Hostler, Jacquelyn; Gunther, Jana; Parotto, Paolo; Pasztor, Attila; Vazquez, Israel Portillo; Ratti, Claudia
2016-01-01
We show that it is possible to isolate a set of kaon fluctuations in lattice QCD. By means of the Hadron Resonance Gas (HRG) model, we calculate the actual kaon second-to-first fluctuation ratio, which receives contribution from primordial kaons and resonance decays, and show that it is very close to the one obtained for primordial kaons in the Boltzmann approximation. The latter only involves the strangeness and electric charge chemical potentials, which are functions of $T$ and $\\mu_B$ due to the experimental constraint on strangeness and electric charge, and can therefore be calculated on the lattice. This provides an unambiguous method to extract the kaon freeze-out temperature, by comparing the lattice results to the experimental values for the corresponding fluctuations.
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
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
姜志鹏; 文习山; 王羽; 陈瑞珍; 曹继丰; 陈图腾
2015-01-01
为了研究特高压干式空心平波电抗器的温升分布特性,该文基于计算流体力学和传热学理论,建立了电抗器稳态流体与固体耦合温度场的数学计算模型.采用有限容积法对三维模型进行稳态流体场与温度场直接求解,获得其温度场分布特性,研究了包封轴向及径向温度分布规律.最后采用光纤测温法对自然对流下的电抗器进行温升测量.对比分析表明,计算与试验结果吻合较好,验证温度场数值计算的合理性和准确性,为特高压干式空心平波电抗器温升监测提供参考.%To research the distribution characteristics of temperature rise for UHV dry-type air-core smoothing reactor, according to computational fluid dynamics and heat transfer theory, this paper presented the mathematical model of temperature field coupling steady fluid and solid for the reactor. The finite volume method was employed to solve the steady flow and temperature fields of 3D model directly, and the temperature distribution characteristics of the reactor were obtained. Then the axial and radial temperature distributions of encapsulations were studied separately. Finally, optical fiber temperature measurement method was used to test temperature rise for the reactor under natural convection condition. Comparative analysis shows that the calculated results are in good agreement with the experiment, which verifies the rationality and accuracy of the temperature field numerical calculation. And it can provide references for the temperature rise monitoring of UHV dry-type air-core smoothing reactor.
1D Burnup Calculation of Fusion-Fission Hybrid Energy Reactor%聚变-裂变混合能源堆一维计算模型燃耗分析
李茂生; 师学明; 伊炜伟
2012-01-01
Fusion-fission hybrid energy reactor is driven by Tokamak fusion source for energy production. Its subcritical zone uses the natural uranium as fuel and water as coolant. The neutron multiplication constant keff, energy multiplication factor M and tritium breeding ratio TBR of the ID hybrid energy reactor model were calculated by transport burnup code MCORGS. The neutron spectrum and nuclear density changing as a function of time show the characteristics of the hybrid energy reactors, which differs from the hybrid reactor for breed nuclear fuel and for spent fuel transmutation. The definition and results may be a reference to the other conceptual analysis.%聚变-裂变混合能源堆包括聚变中子源和以天然铀为燃料、水为冷却剂的次临界包层,主要目标是生产电力.利用输运燃耗耦合程序系统MCORGS计算了混合能源堆一维模型的燃耗,给出了中子有效增殖因数keff、能量放大倍数M、氚增殖比TBR等物理量随时间的变化.通过分析能谱和重要核素随燃耗时间的变化,说明混合能源堆与核燃料增殖、核废料嬗变混合堆的不同特点.本文给出的结果可作为混合堆中子输运、燃耗分析程序校验的参考数据,为混合堆概念研究提供了基础数据.
Lattice dynamics of ferromagnetic superconductor UGe2
Satyam Shinde; Prafulla K Jha
2008-11-01
This paper reports the lattice dynamical study of the UGe2 using a lattice dynamical model theory based on pairwise interactions under the framework of the shell model. The calculated phonon dispersion curves and phonon density of states are in good agreement with the measured data.
Lattice QCD simulations beyond the quenched approximation
Ukawa, A. (European Organization for Nuclear Research, Geneva (Switzerland). Theory Div.)
1989-07-01
Present status of lattice QCD simulations incorporating the effects of dynamical quarks is presented. After a brief review of the formalism of lattice QCD, the dynamical fermion algorithms in use today are described. Recent attempts at the hadron mass calculation are discussed in relation to the quenched results, and current understanding on the finite temperature behavior of QCD is summarized. (orig.).
Lattice Studies for hadron spectroscopy and interactions
Aoki, Sinya
2014-01-01
Recent progresses of lattice QCD studies for hadron spectroscopy and interactions are briefly reviewed. Some emphasis are given on a new proposal for a method, which enable us to calculate potentials between hadrons. As an example of the method, the extraction of nuclear potential in lattice QCD is discussed in detail.
刘大庆; 吴济民; 陈莹
2002-01-01
We develop a new approach to constructing the lattice operators for the calculation of the glueball mass, which is based on the connection between the continuum limit of the chosen operator and the quantum number JPc of the state. The spin of the state is then determined uniquely and directly in numerical simulation. Furthermore, the approach can be applied to the calculation of the mass of glueball states with any spin. J. Under the quenched approximation, we present our preliminary results in SU(3) pure gauge theory for the mass of 0++ state and 2++ state, which are 1754(85)(86) MeV and 2417(56)(117) MeV, respectively.%发展了一种为了计算胶球质量而构造格点算符的新途径.基于所选用算符的连续极限与状态量子数JPC两者之间的联系,状态的自旋就可以在数值模拟中唯一和直接地被确定下来.进而,这一途径可以被应用于计算任意自旋J的胶球质量.在淬火近似下,给出在SU(3)纯规范场中0++态和2++态胶球质量的初步结果,它们分别是1754(85)(86)MeV和2417(56)(17)MeV.
Rocklin, Gabriel J. [Department of Pharmaceutical Chemistry, University of California San Francisco, 1700 4th St., San Francisco, California 94143-2550, USA and Biophysics Graduate Program, University of California San Francisco, 1700 4th St., San Francisco, California 94143-2550 (United States); Mobley, David L. [Departments of Pharmaceutical Sciences and Chemistry, University of California Irvine, 147 Bison Modular, Building 515, Irvine, California 92697-0001, USA and Department of Chemistry, University of New Orleans, 2000 Lakeshore Drive, New Orleans, Louisiana 70148 (United States); Dill, Ken A. [Laufer Center for Physical and Quantitative Biology, 5252 Stony Brook University, Stony Brook, New York 11794-0001 (United States); Hünenberger, Philippe H., E-mail: phil@igc.phys.chem.ethz.ch [Laboratory of Physical Chemistry, Swiss Federal Institute of Technology, ETH, 8093 Zürich (Switzerland)
2013-11-14
The calculation of a protein-ligand binding free energy based on molecular dynamics (MD) simulations generally relies on a thermodynamic cycle in which the ligand is alchemically inserted into the system, both in the solvated protein and free in solution. The corresponding ligand-insertion free energies are typically calculated in nanoscale computational boxes simulated under periodic boundary conditions and considering electrostatic interactions defined by a periodic lattice-sum. This is distinct from the ideal bulk situation of a system of macroscopic size simulated under non-periodic boundary conditions with Coulombic electrostatic interactions. This discrepancy results in finite-size effects, which affect primarily the charging component of the insertion free energy, are dependent on the box size, and can be large when the ligand bears a net charge, especially if the protein is charged as well. This article investigates finite-size effects on calculated charging free energies using as a test case the binding of the ligand 2-amino-5-methylthiazole (net charge +1 e) to a mutant form of yeast cytochrome c peroxidase in water. Considering different charge isoforms of the protein (net charges −5, 0, +3, or +9 e), either in the absence or the presence of neutralizing counter-ions, and sizes of the cubic computational box (edges ranging from 7.42 to 11.02 nm), the potentially large magnitude of finite-size effects on the raw charging free energies (up to 17.1 kJ mol{sup −1}) is demonstrated. Two correction schemes are then proposed to eliminate these effects, a numerical and an analytical one. Both schemes are based on a continuum-electrostatics analysis and require performing Poisson-Boltzmann (PB) calculations on the protein-ligand system. While the numerical scheme requires PB calculations under both non-periodic and periodic boundary conditions, the latter at the box size considered in the MD simulations, the analytical scheme only requires three non
Nucleon structure using lattice QCD
Alexandrou, C.; Kallidonis, C. [Cyprus Univ., Nicosia (Cyprus). Dept. of Physics; The Cyprus Institute, Nicosia (Cyprus). Computational-Based Science and technology Research Center; Constantinou, M.; Hatziyiannakou, K. [Cyprus Univ., Nicosia (Cyprus). Dept. of Physics; Drach, V. [DESY Zeuthen (Germany). John von Neumann-Institut fuer Computing NIC; Jansen, K. [DESY Zeuthen (Germany). John von Neumann-Institut fuer Computing NIC; Cyprus Univ., Nicosia (Cyprus). Dept. of Physics; Koutsou, G.; Vaquero, A. [The Cyprus Institute, Nicosia (Cyprus). Computational-Based Science and technology Research Center; Leontiou, T. [Frederick Univ, Nicosia (Cyprus). General Dept.
2013-03-15
A review of recent nucleon structure calculations within lattice QCD is presented. The nucleon excited states, the axial charge, the isovector momentum fraction and helicity distribution are discussed, assessing the methods applied for their study, including approaches to evaluate the disconnected contributions. Results on the spin carried by the quarks in the nucleon are also presented.
From lattice gases to polymers
Frenkel, D.
1990-01-01
The modification of a technique that was developed to study time correlations in lattice-gas cellular automata to facilitate the numerical simulation of chain molecules is described. As an example, the calculation of the excess chemical potential of an ideal polymer in a dense colloidal
Gray S. Chang
2005-11-01
The currently being developed advanced High Temperature gas-cooled Reactors (HTR) is able to achieve a simplification of safety through reliance on innovative features and passive systems. One of the innovative features in these HTRs is reliance on ceramic-coated fuel particles to retain the fission products even under extreme accident conditions. Traditionally, the effect of the random fuel kernel distribution in the fuel pebble / block is addressed through the use of the Dancoff correction factor in the resonance treatment. However, the Dancoff correction factor is a function of burnup and fuel kernel packing factor, which requires that the Dancoff correction factor be updated during Equilibrium Fuel Cycle (EqFC) analysis. An advanced KbK-sph model and whole pebble super lattice model (PSLM), which can address and update the burnup dependent Dancoff effect during the EqFC analysis. The pebble homogeneous lattice model (HLM) is verified by the burnup characteristics with the double-heterogeneous KbK-sph lattice model results. This study summarizes and compares the KbK-sph lattice model and HLM burnup analyzed results. Finally, we discuss the Monte-Carlo coupling with a fuel depletion and buildup code - ORIGEN-2 as a fuel burnup analysis tool and its PSLM calculated results for the HTR EqFC burnup analysis.
On the lattice rotations accompanying slip
Wronski, M.; Wierzbanowski, K.; Leffers, Torben
2013-01-01
of the crystal lattices, and this texture may have a strong effect on the properties of the materials. The texture is introduced by lattice rotations in the individual grains during processing. The present critical assessment deals with the lattice rotations during rolling of face centred cubic (fcc) metals...... and alloys. Sixteen years ago, a modification of the traditional procedure for the calculation of these lattice rotations was suggested, a modification that would permit a realistic modelling of the development of the brass type texture, one of the two types of texture developed during rolling of fcc...
SuPer-Homogenization (SPH) Corrected Cross Section Generation for High Temperature Reactor
Sen, Ramazan Sonat [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hummel, Andrew John [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hiruta, Hikaru [Idaho National Lab. (INL), Idaho Falls, ID (United States)
2017-03-01
The deterministic full core simulators require homogenized group constants covering the operating and transient conditions over the entire lifetime. Traditionally, the homogenized group constants are generated using lattice physics code over an assembly or block in the case of prismatic high temperature reactors (HTR). For the case of strong absorbers that causes strong local depressions on the flux profile require special techniques during homogenization over a large volume. Fuel blocks with burnable poisons or control rod blocks are example of such cases. Over past several decades, there have been a tremendous number of studies performed for improving the accuracy of full-core calculations through the homogenization procedure. However, those studies were mostly performed for light water reactor (LWR) analyses, thus, may not be directly applicable to advanced thermal reactors such as HTRs. This report presents the application of SuPer-Homogenization correction method to a hypothetical HTR core.
Lattice dislocation in Si nanowires
Omar, M.S., E-mail: dr_m_s_omar@yahoo.co [Department of Physics, College of Science, University of Salahaddin, Arbil, Iraqi Kurdistan (Iraq); Taha, H.T. [Department of Physics, College of Science, University of Salahaddin, Arbil, Iraqi Kurdistan (Iraq)
2009-12-15
Modified formulas were used to calculate lattice thermal expansion, specific heat and Bulk modulus for Si nanowires with diameters of 115, 56, 37 and 22 nm. From these values and Gruneisen parameter taken from reference, mean lattice volumes were found to be as 20.03 A{sup 3} for the bulk and 23.63, 29.91, 34.69 and 40.46 A{sup 3} for Si nanowire diameters mentioned above, respectively. Their mean bonding length was calculated to be as 0.235 nm for the bulk and 0.248, 0.269, 0.282 and 0.297 nm for the nanowires diameter mentioned above, respectively. By dividing the nanowires diameter on the mean bonding length, number of layers per each nanowire size was found to be as 230, 104, 65 and 37 for the diameters mentioned above, respectively. Lattice dislocations in 22 nm diameter wire were found to be from 0.00324 nm for the 1st central lattice to 0.2579 nm for the last surface lattice. Such dislocation was smaller for larger wire diameters. Dislocation concentration found to change in Si nanowires according to the proportionalities of surface thickness to nanowire radius ratios.
潘昕怿; 兰兵; 张春明; 靖剑平; 攸国顺
2016-01-01
Background: The uncertainty of nuclear data is one of the key factors resulting in the uncertainty of reactor physics calculation. Purpose: The influence of multigroup nuclear data uncertainties on the reactor core physics calculation was studied in this paper. Methods:The stochastic sampling modular SAMP based on covariance matrix of nuclear data was developed, and the hybrid method and stochastic sampling method were realized using SCALE (Standardized Computer Analyses for Licensing Evaluation) software package. The two methods were validated using 3×3 hypothetical core and then applied to the first cycle of Almaraz pressurized-water reactor (PWR) in the IAEA (International Atomic Energy Agency) fuel management benchmark. Results: Results of the two methods are in good agreement. The uncertainty of core effective multiplication factor is about 0.5%, and the maximum uncertainties of the radial and axial power are about 1.9% and 0.45% respectively in Almaraz PWR. Conclusion:The two-step method and stochastic sampling method can both be used for the uncertainty analysis of reactor core calculation.%核数据不确定性是造成反应堆物理计算结果不确定性的重要因素之一。基于所需抽样核数据的协方差矩阵开发了随机抽样模块(Stochastic Sampling, SAMP)，在此基础上利用SCALE (Standardized Computer Analyses for Licensing Evaluation)软件包实现了混合法和随机抽样法两种不确定性分析方法，以研究多群核数据不确定性对堆芯物理计算的影响。以3×3假想堆芯为对象，对两种方法进行了验证，然后应用于国际原子能机构(International Atomic Energy Agency, IAEA)燃料管理基准题中的Almaraz核电厂首循环堆芯。分析结果表明，两种方法结果符合良好，Almaraz核电厂堆芯kef 不确定性约为0.5%，堆芯径向和轴向功率的最大不确定性分别为1.9%和0.45%。
Innovations in Lattice QCD Algorithms
Konstantinos Orginos
2006-06-25
Lattice QCD calculations demand a substantial amount of computing power in order to achieve the high precision results needed to better understand the nature of strong interactions, assist experiment to discover new physics, and predict the behavior of a diverse set of physical systems ranging from the proton itself to astrophysical objects such as neutron stars. However, computer power alone is clearly not enough to tackle the calculations we need to be doing today. A steady stream of recent algorithmic developments has made an important impact on the kinds of calculations we can currently perform. In this talk I am reviewing these algorithms and their impact on the nature of lattice QCD calculations performed today.
Soo-Bong Kim; Thierry Lasserre; Yifang Wang
2013-01-01
We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very ...
Untermyer, S.
1962-04-10
A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)
朱贵凤; 邹杨; 李明海; 严睿; 彭红花; 徐洪杰
2015-01-01
The burnup calculation code PBRE coupling MCNP5 and ORIGEN2 was developed for pebble‐bed high temperature reactor at equilibrium state ,and it can be used to analyze the neutronic performance of equilibrium core .The iteration method was optimized in order to save Monte Carlo calculation time ,and the convergence can be reached in 10 iterative steps .The average discharged burnup for HTR‐10 is consistent with literature ,and it indicates that the PBRE is suitable to analyze the burnup for pebble‐bed reactor at equilibrium state .%基于MCNP5和ORIGEN2耦合方法，开发了平衡态下球床高温堆的燃耗计算程序PBRE ，用于堆的性能价值分析。为节省蒙特卡罗计算时间，对迭代收敛的方法进行优化，使之可在10个迭代步内收敛。使用PBRE对清华大学H T R‐10进行建模计算，得到的平均卸料燃耗深度与文献报道值一致，表明PBRE程序适用于球床堆平衡态的燃耗分析。
Gauntt, Randall O.; Ross, Kyle W. (Los Alamos National Laboratory, Los Alamos, NM); Smith, James Dean; Longmire, Pamela
2010-04-01
The Oak Ridge National Laboratory computer code, ORIGEN2.2 (CCC-371, 2002), was used to obtain the elemental composition of irradiated low-enriched uranium (LEU)/mixed-oxide (MOX) pressurized-water reactor fuel assemblies. Described in this report are the input parameters for the ORIGEN2.2 calculations. The rationale for performing the ORIGEN2.2 calculation was to generate inventories to be used to populate MELCOR radionuclide classes. Therefore the ORIGEN2.2 output was subsequently manipulated. The procedures performed in this data reduction process are also described herein. A listing of the ORIGEN2.2 input deck for two-cycle MOX is provided in the appendix. The final output from this data reduction process was three tables containing the radionuclide inventories for LEU/MOX in elemental form. Masses, thermal powers, and activities were reported for each category.
LeSage, L.G.; McKnight, R.D.; Wade, D.C.; Freese, K.E.; Collins, P.J.
1980-08-01
The results of an international comparison calculation of a large (1250 MWe) LMFBR benchmark model are presented and discussed. Eight reactor configurations were calculated. Parameters included with the comparison were: eigenvalue, k/sub infinity/, neutron balance data, breeding reaction rate ratios, reactivity worths, central control rod worth, regional sodium void reactivity, core Doppler and effective delayed neutron fraction. Ten countries participated in the comparison, and sixteen solutions were contributed. The discussion focuses on the variation in parameter values, the degree of consistency among the various parameters and solutions, and the identification of unexpected results. The results are displayed and discussed both by individual participants and by groupings of participants (e.g., results from adjusted data sets versus non-adjusted data sets).
Santocanale, Luigi
2002-01-01
A μ-lattice is a lattice with the property that every unary polynomial has both a least and a greatest fix-point. In this paper we define the quasivariety of μ-lattices and, for a given partially ordered set P, we construct a μ-lattice JP whose elements are equivalence classes of games in a preor...
Hadron structure from lattice QCD
Green, Jeremy [Institut für Kernphysik, Johannes Gutenberg-Universität Mainz, D-55099 Mainz (Germany)
2016-01-22
Recent progress in lattice QCD calculations of nucleon structure will be presented. Calculations of nucleon matrix elements and form factors have long been difficult to reconcile with experiment, but with advances in both methodology and computing resources, this situation is improving. Some calculations have produced agreement with experiment for key observables such as the axial charge and electromagnetic form factors, and the improved understanding of systematic errors will help to increase confidence in predictions of unmeasured quantities. The long-omitted disconnected contributions are now seeing considerable attention and some recent calculations of them will be discussed.
Hadron Structure from Lattice QCD
Green, Jeremy
2014-01-01
Recent progress in lattice QCD calculations of nucleon structure will be presented. Calculations of nucleon matrix elements and form factors have long been difficult to reconcile with experiment, but with advances in both methodology and computing resources, this situation is improving. Some calculations have produced agreement with experiment for key observables such as the axial charge and electromagnetic form factors, and the improved understanding of systematic errors will help to increase confidence in predictions of unmeasured quantities. The long-omitted disconnected contributions are now seeing considerable attention and some recent calculations of them will be discussed.
Understanding Parton Distributions from Lattice QCD
Renner, Dru B.
2005-01-01
I examine the past lattice QCD calculations of three representative observables, the transverse quark distribution, momentum fraction, and axial charge, and emphasize the prospects for not only quantitative comparison with experiment but also qualitative understanding of QCD.
Unconventional superconductivity in honeycomb lattice
P Sahebsara
2013-03-01
Full Text Available The possibility of symmetrical s-wave superconductivity in the honeycomb lattice is studied within a strongly correlated regime, using the Hubbard model. The superconducting order parameter is defined by introducing the Green function, which is obtained by calculating the density of the electrons . In this study showed that the superconducting order parameter appears in doping interval between 0 and 0.5, and x=0.25 is the optimum doping for the s-wave superconductivity in honeycomb lattice.
Effective Field Theories and Lattice QCD
Bernard, C
2015-01-01
I describe some of the many connections between lattice QCD and effective field theories, focusing in particular on chiral effective theory, and, to a lesser extent, Symanzik effective theory. I first discuss the ways in which effective theories have enabled and supported lattice QCD calculations. Particular attention is paid to the inclusion of discretization errors, for a variety of lattice QCD actions, into chiral effective theory. Several other examples of the usefulness of chiral perturbation theory, including the encoding of partial quenching and of twisted boundary conditions, are also described. In the second part of the talk, I turn to results from lattice QCD for the low energy constants of the two- and three-flavor chiral theories. I concentrate here on mesonic quantities, but the dependence of the nucleon mass on the pion mass is also discussed. Finally I describe some recent preliminary lattice QCD calculations by the MILC Collaboration relating to the three-flavor chiral limit.
Uncertainties in the Anti-neutrino Production at Nuclear Reactors
Djurcic, Z.(Argonne National Laboratory, Argonne, Illinois, 60439, U.S.A.); Detwiler, J. A.; Piepke, A.; Foster Jr., V. R.; Miller, L.; Gratta, G.
2008-01-01
Anti-neutrino emission rates from nuclear reactors are determined from thermal power measurements and fission rate calculations. The uncertainties in these quantities for commercial power plants and their impact on the calculated interaction rates in electron anti-neutrino detectors is examined. We discuss reactor-to-reactor correlations between the leading uncertainties and their relevance to reactor anti-neutrino experiments.
Membrane reactor. Membrane reactor
Shindo, Y.; Wakabayashi, K. (National Chemical Laboratory for Industry, Tsukuba (Japan))
1990-08-05
Many reaction examples were introduced of membrane reactor, to be on the point of forming a new region in the field of chemical technology. It is a reactor to exhibit excellent function, by its being installed with membrane therein, and is generally classified into catalyst function type and reaction promotion type. What firstly belongs to the former is stabilized zirconia, where oxygen, supplied to the cathodic side of membrane with voltage, impressed thereon, becomes O {sup 2 {minus}} to be diffused through the membrane and supplied, as variously activated oxygenous species, on the anodic side. Examples with many advantages can be given such as methane coupling, propylene oxidation, methanating reaction of carbon dioxide, etc. Apart, palladium film and naphion film also belong to the former. While examples of the latter comprise, among others, decomposition of hydrogen sulfide by porous glass film and dehydrogenation of cyclohexane or palladium alloy film, which are expected to be developed and materialized in the industry. 33 refs., 8 figs.
Lattice QCD and the Jefferson Laboratory Program
Jozef Dudek, Robert Edwards, David Richards, Konstantinos Orginos
2011-06-01
Lattice gauge theory provides our only means of performing \\textit{ab initio} calculations in the non-perturbative regime. It has thus become an increasing important component of the Jefferson Laboratory physics program. In this paper, we describe the contributions of lattice QCD to our understanding of hadronic and nuclear physics, focusing on the structure of hadrons, the calculation of the spectrum and properties of resonances, and finally on deriving an understanding of the QCD origin of nuclear forces.
Equation of state and more from lattice regularized QCD
Karsch, Frithjof
2008-01-01
We present results from a calculation of the QCD equation of state with two light (up, down) and one heavier (strange) quark mass performed on lattices with three different values of the lattice cut-off. We show that also on the finest lattice analyzed by us observables sensitive to deconfinement and chiral symmetry restoration, respectively, vary most rapidly in the same temperature regime.
Reactor Antineutrino Signals at Morton and Boulby
Dye, Steve
2016-01-01
Increasing the distance from which an antineutrino detector is capable of monitoring the operation of a registered reactor, or discovering a clandestine reactor, strengthens the Non-Proliferation of Nuclear Weapons Treaty. This report presents calculations of reactor antineutrino interactions, from quasi-elastic neutrino-proton scattering and elastic neutrino-electron scattering, in a water-based detector operated >10 km from a commercial power reactor. It separately calculates signal from the proximal reactor and background from all other registered reactors. The main results are interaction rates and kinetic energy distributions of charged leptons scattered from quasi-elastic and elastic processes. Comparing signal and background distributions evaluates reactor monitoring capability. Scaling the results to detectors of different sizes, target media, and standoff distances is straightforward. Calculations are for two examples of a commercial reactor (P_th~3 GW) operating nearby (L~20 km) an underground facil...
Ground-state phase diagram of the Kondo lattice model on triangular-to-kagome lattices
Akagi, Yutaka; Motome, Yukitoshi
2012-01-01
We investigate the ground-state phase diagram of the Kondo lattice model with classical localized spins on triangular-to-kagome lattices by using a variational calculation. We identify the parameter regions where a four-sublattice noncoplanar order is stable with a finite spin scalar chirality while changing the lattice structure from triangular to kagome continuously. Although the noncoplanar spin states appear in a wide range of parameters, the spin configurations on the kagome network beco...
New Lattice Results for Parton Distributions
Alexandrou, Constantia; Constantinou, Martha; Hadjiyiannakou, Kyriakos; Jansen, Karl; Steffens, Fernanda; Wiese, Christian
2016-01-01
We provide a high statistics analysis of the $x$-dependence of the bare unpolarized, helicity and transversity iso-vector parton distribution functions (PDFs) from lattice calculations employing (maximally) twisted mass fermions. The $x$-dependence of the calculated PDFs resembles those of the phenomenological parameterizations, a feature that makes this approach promising despite the lack of a full renormalization program for them. Furthermore, we apply momentum smearing for the relevant matrix elements to compute the lattice PDFs and find a large improvement factor when compared to conventional Gaussian smearing. This allows us to extend the lattice computation of the distributions to higher values of the nucleon momentum.
Shin, Chang Hwan; Seo, Kyong Won; Chun, Tae Hyun; Kim, Kang Seog
2005-03-15
Code coupling activities have so far focused on coupling the neutronics modules with the CFD module. An interface module for the CFD-ACE/DeCART coupling was established as an alternative to the original STAR-CD/DeCART interface. The interface module for DeCART/CFD-ACE was validated by single-pin model. The optimized CFD mesh was decided through the calculation of multi-pin model. It was important to consider turbulent mixing of subchannels for calculation of fuel temperature. For the parallel calculation, the optimized decompose process was necessary to reduce the calculation costs and setting of the iteration and convergence criterion for each code was important, too.
Analysis of High Temperature Reactor Control Rod Worth for the Initial and Full Core
Oktajianto, Hammam; Setiawati, Evi; Anam, Khoirul; Sugito, Heri
2017-01-01
Control rod is one important component in a nuclear reactor. In nuclear reactor operations the control rod functions to shut down the reactor. This research analyses ten control rods worth of HTR (High Temperature Reactor) at initial and full core. The HTR in this research adopts HTR-10 China and HTR- of pebble bed. Core calculations are performed by using MCNPX code after modelling the entire parts of core in condition of ten control rods fully withdrawn, all control rods in with 20 cm ranges of depth and the use of one control rod. Pebble bed and moderator balls are distributed in the core zone using a Body Centred Cubic (BCC) lattice by ratio of 57:43. The research results are obtained that the use of one control rod will decrease the reactor criticality of 2.04±0.12 %Δk/k at initial core and 1.57±0.10 %Δk/k at full core. The deeper control rods are in, the lesser criticality of reactor is with reactivity of ten control rods of 16.41±0.11 %Δk/k at initial core and 15.43±0.11 %Δk/k at full core. The results show that the use of ten control rods at full core will keep achieving subcritical condition even though the reactivity is smaller than reactivity at initial core.
Logarithmic divergent thermal conductivity in two-dimensional nonlinear lattices.
Wang, Lei; Hu, Bambi; Li, Baowen
2012-10-01
Heat conduction in three two-dimensional (2D) momentum-conserving nonlinear lattices are numerically calculated via both nonequilibrium heat-bath and equilibrium Green-Kubo algorithms. It is expected by mainstream theories that heat conduction in such 2D lattices is divergent and the thermal conductivity κ increases with lattice length N logarithmically. Our simulations for the purely quartic lattice firmly confirm it. However, very robust finite-size effects are observed in the calculations for the other two lattices, which well explain some existing studies and imply the extreme difficulties in observing their true asymptotic behaviors with affordable computation resources.
Frolov, A. A.; Sedov, A. A.
2016-08-01
A method for combined 3D/1D-modeling of thermohydraulics of a once-through steam generator (SG) based on the joint analysis of three-dimensional thermo- and hydrodynamics of a single-phase heating coolant in the intertube space and one-dimensional thermohydraulics of steam-generating channels (tubes) with the use of well-known friction and heat-transfer correlations under various boiling conditions is discussed. This method allows one to determine the spatial distribution of temperatures and heat fluxes of heat-exchange surfaces of SGs with a single-phase heating coolant in the intertube space and with steam generation within tubes. The method was applied in the analytical investigation of typical operation of a once-through SG of a nuclear power installation with an RBEC fast-neutron heavy-metal reactor that is being designed by Kurchatov Institute in collaboration with OKB GIDROPRESS and Leipunsky Institute of Physics and Power Engineering. Flow pattern and temperature fields were obtained for the heavy-metal heating coolant in the intertube space. Nonuniformities of heating of the steam-water coolant in different heat-exchange tubes and nonuniformities in the distribution of heat fluxes at SG heat-exchange surfaces were revealed.
刘一哲; 薛秀丽; 许义军; 冯预恒; 侯志峰
2012-01-01
Based on the core and primary circuit design of China Experimental Fast Reactor(CEFR), a multiple-channel thermal-hydraulic analysis code DAEMON was developed to calculate the core flow rate distribution and unsymmetric coefficient in different conditions. In the commissioning stage, a series of full-scale tests for reactor core were performed in CEFR with a permanent-magnet sodium flow meter. The numerical results of code DAEMON showed a good agreement with test data. The core hydraulic design was also validated with a view to the requirements of design criteria, commissioning and operation specifications.%针对中国实验快堆(CEFR)堆芯和一回路的设计特点,开发水力特性计算程序DAEMON,完成不同工况下的全堆芯流量分配计算,给出流量分配不均匀性等参数.在反应堆调试阶段,进行全堆芯流量分配试验.结果表明,程序计算值与试验值符合较好.在此基础上,验证了CEFR堆芯的流体力学设计,并为反应堆调试和运行提供了基础数据.
Lattice thermal conductivity evaluated using elastic properties
Jia, Tiantian; Chen, Gang; Zhang, Yongsheng
2017-04-01
Lattice thermal conductivity is one of the most important thermoelectric parameters in determining the energy conversion efficiency of thermoelectric materials. However, the lattice thermal conductivity evaluation requires time-consuming first-principles (quasi)phonon calculations, which limits seeking high-performance thermoelectric materials through high-throughput computations. Here, we establish a methodology to determine the Debye temperature Θ , Grüneisen parameter γ , and lattice thermal conductivity κ using computationally feasible elastic properties (the bulk and shear moduli). For 39 compounds with three different prototypes (the cubic isotropic rocksalt and zinc blende, and the noncubic anisotropic wurtzite), the theoretically calculated Θ ,γ , and κ are in reasonable agreement with those determined using (quasi)harmonic phonon calculations or experimental measurements. Our results show that the methodology is an efficient tool to predict the anharmonicity and the lattice thermal conductivity.
Keney, G.S.
1981-08-01
A computer code has been written to calculate neutron induced activation of neutral-beam injector components and the corresponding dose rates as a function of geometry, component composition, and time after shutdown. The code, ACDOS1, was written in FORTRAN IV to calculate both activity and dose rates for up to 30 target nuclides and 50 neutron groups. Sufficient versatility has also been incorporated into the code to make it applicable to a variety of general activation problems due to neutrons of energy less than 20 MeV.
Micro reactor physics of MOX fueled LWR
Takeda, Toshikazu [Osaka Univ. (Japan)
2001-09-01
Upon the background that the LWR fuels become complicated in recent years because of the introduction of high burnup fuels, high density Gd fuels, MOX fuels, the author proposes the Micro Reactor Physics. He intends to investigate the behaviors of neutrons and reactions in a pin rod that have not yet been paid attention. Conventionally the resonance absorption has been evaluated by assuming the uniform effective cross sections in a pin rod. However, due to the self-shielding, the neutron spectrum near the surface of the rod is quite different with that of the center of rod. This fact affects the spatial distributions of Pu isotopes produced during burnup. The spatial distribution of temperature in a rod affects the Doppler coefficient. He solved this problem by the multi-band method. In the case where MOX rods are adjacent with U rods, the spectrum of the current from MOX rods to U rods is different with that of U to MOX. That makes the spatial distribution of azimuthal direction together with that of the infinite lattice. He solved this problem by a cell calculation based on the characteristic method. This report introduces several numerical results of his Micro Reactor Physics. One of the important results is the indication that the conventional Doppler coefficient gives 20% higher (not conservative) value. (K. Tsuchihashi)
Micro reactor physics of MOX fueled LWR
Takeda, Toshikazu [Osaka Univ. (Japan)
2001-09-01
Upon the background that the LWR fuels become complicated in recent years because of the introduction of high burnup fuels, high density Gd fuels, MOX fuels, the author proposes the Micro Reactor Physics. He intends to investigate the behaviors of neutrons and reactions in a pin rod that have not yet been paid attention. Conventionally the resonance absorption has been evaluated by assuming the uniform effective cross sections in a pin rod. However, due to the self-shielding, the neutron spectrum near the surface of the rod is quite different with that of the center of rod. This fact affects the spatial distributions of Pu isotopes produced during burnup. The spatial distribution of temperature in a rod affects the Doppler coefficient. He solved this problem by the multi-band method. In the case where MOX rods are adjacent with U rods, the spectrum of the current from MOX rods to U rods is different with that of U to MOX. That makes the spatial distribution of azimuthal direction together with that of the infinite lattice. He solved this problem by a cell calculation based on the characteristic method. This report introduces several numerical results of his Micro Reactor Physics. One of the important results is the indication that the conventional Doppler coefficient gives 20% higher (not conservative) value. (K. Tsuchihashi)
SUBGR: A Program to Generate Subgroup Data for the Subgroup Resonance Self-Shielding Calculation
Kim, Kang Seog [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
2016-06-06
The Subgroup Data Generation (SUBGR) program generates subgroup data, including levels and weights from the resonance self-shielded cross section table as a function of background cross section. Depending on the nuclide and the energy range, these subgroup data can be generated by (a) narrow resonance approximation, (b) pointwise flux calculations for homogeneous media; and (c) pointwise flux calculations for heterogeneous lattice cells. The latter two options are performed by the AMPX module IRFFACTOR. These subgroup data are to be used in the Consortium for Advanced Simulation of Light Water Reactors (CASL) neutronic simulator MPACT, for which the primary resonance self-shielding method is the subgroup method.
Campos, R G; Campos, Rafael G.; Tututi, Eduardo S.
2002-01-01
It is shown that the nonlocal Dirac operator yielded by a lattice model that preserves chiral symmetry and uniqueness of fields, approaches to an ultralocal and invariant under translations operator when the size of the lattice tends to zero.
New integrable lattice hierarchies
Pickering, Andrew [Area de Matematica Aplicada, ESCET, Universidad Rey Juan Carlos, c/ Tulipan s/n, 28933 Mostoles, Madrid (Spain); Zhu Zuonong [Departamento de Matematicas, Universidad de Salamanca, Plaza de la Merced 1, 37008 Salamanca (Spain) and Department of Mathematics, Shanghai Jiao Tong University, Shanghai 200030 (China)]. E-mail: znzhu2@yahoo.com.cn
2006-01-23
In this Letter we give a new integrable four-field lattice hierarchy, associated to a new discrete spectral problem. We obtain our hierarchy as the compatibility condition of this spectral problem and an associated equation, constructed herein, for the time-evolution of eigenfunctions. We consider reductions of our hierarchy, which also of course admit discrete zero curvature representations, in detail. We find that our hierarchy includes many well-known integrable hierarchies as special cases, including the Toda lattice hierarchy, the modified Toda lattice hierarchy, the relativistic Toda lattice hierarchy, and the Volterra lattice hierarchy. We also obtain here a new integrable two-field lattice hierarchy, to which we give the name of Suris lattice hierarchy, since the first equation of this hierarchy has previously been given by Suris. The Hamiltonian structure of the Suris lattice hierarchy is obtained by means of a trace identity formula.
none,
1989-01-01
This paper discuses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics.
Fast breeder reactors an engineering introduction
Judd, A M
1981-01-01
Fast Breeder Reactors: An Engineering Introduction is an introductory text to fast breeder reactors and covers topics ranging from reactor physics and design to engineering and safety considerations. Reactor fuels, coolant circuits, steam plants, and control systems are also discussed. This book is comprised of five chapters and opens with a brief summary of the history of fast reactors, with emphasis on international and the prospect of making accessible enormous reserves of energy. The next chapter deals with the physics of fast reactors and considers calculation methods, flux distribution,
Sober Topological Molecular Lattices
张德学; 李永明
2003-01-01
A topological molecular lattice (TML) is a pair (L, T), where L is a completely distributive lattice and r is a subframe of L. There is an obvious forgetful functor from the category TML of TML's to the category Loc of locales. In this note,it is showed that this forgetful functor has a right adjoint. Then, by this adjunction,a special kind of topological molecular lattices called sober topological molecular lattices is introduced and investigated.
Kelmers, A.D.; Hightower, J.R.
1987-05-01
Emplacement of contaminated reactor components involves disposal in lined and unlined auger holes in soil above the water table. The radionuclide inventory of disposed components was calculated. Information on the composition and weight of the components, as well as reasonable assumptions for the neutron flux fueling use, the time of neutron exposure, and radioactive decay after discharge, were employed in the inventory calculation. Near-field release rates of /sup 152/Eu, /sup 154/Eu, and /sup 155/Eu from control plates and cylinders were calculated for 50 years after emplacement. Release rates of the europium isotopes were uncertain. Two release-rate-limiting models were considered and a range of reasonable values were assumed for the time-to-failure of the auger-hole linear and aluminum cladding and europium solubility in SWSA-6 groundwater. The bounding europium radionuclide near-field release rates peaked at about 1.3 Ci/year total for /sup 152,154,155/Eu in 1987 for the lower bound, and at about 420 Ci/year in 1992 for the upper bound. The near-field release rates of /sup 55/Fe, /sup 59/Ni, /sup 60/Co, and /sup 63/Ni from stainless steel and cobalt alloy components, as well as of /sup 10/Be, /sup 41/Ca, and /sup 55/Fe from beryllium reflectors, were calculated for the next 100 years, assuming bulk waste corrosion was the release-rate-limiting step. Under the most conservative assumptions for the reflectors, the current (1986) total radionuclide release rate was calculated to be about 1.2 x 10/sup -4/ Ci/year, decreasing by 1992 to a steady release of about 1.5 x 10/sup -5/ Ci/year due primarily to /sup 41/Ca. 50 refs., 13 figs., 8 tabs.
Lattice Regularization and Symmetries
Hasenfratz, Peter; Von Allmen, R; Allmen, Reto von; Hasenfratz, Peter; Niedermayer, Ferenc
2006-01-01
Finding the relation between the symmetry transformations in the continuum and on the lattice might be a nontrivial task as illustrated by the history of chiral symmetry. Lattice actions induced by a renormalization group procedure inherit all symmetries of the continuum theory. We give a general procedure which gives the corresponding symmetry transformations on the lattice.
Analysis of quantum spin models on hyperbolic lattices and Bethe lattice
Daniška, Michal; Gendiar, Andrej
2016-04-01
The quantum XY, Heisenberg, and transverse field Ising models on hyperbolic lattices are studied by means of the tensor product variational formulation algorithm. The lattices are constructed by tessellation of congruent polygons with coordination number equal to four. The calculated ground-state energies of the XY and Heisenberg models and the phase transition magnetic field of the Ising model on the series of lattices are used to estimate the corresponding quantities of the respective models on the Bethe lattice. The hyperbolic lattice geometry induces mean-field-like behavior of the models. The ambition to obtain results on the non-Euclidean lattice geometries has been motivated by theoretical studies of the anti-de Sitter/conformal field theory correspondence.
Qcd Thermodynamics On A Lattice
Levkova, L A
2004-01-01
Numerical simulations of full QCD on anisotropic lattices provide a convenient way to study QCD thermodynamics with fixed physics scales and reduced lattice spacing errors. We report results from calculations with two flavors of dynamical staggered fermions, where all bare parameters and the renormalized anisotropy are kept constant and the temperature is changed in small steps by varying only the number of time slices. Including results from zero- temperature scale setting simulations, which determine the Karsch coefficients, allows for the calculation of the Equation of State at finite temperatures. We also report on studies of the chiral properties of dynamical domain-wall fermions combined with the DBW2 gauge action for different gauge couplings and fermion masses. For quenched theories, the DBW2 action gives a residual chiral symmetry breaking much smaller than what was found with more traditional choices for the gauge action. Our goal is to investigate the possibilities which this and further improvemen...
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
Screening in graphene antidot lattices
Schultz, Marco Haller; Jauho, A. P.; Pedersen, T. G.
2011-01-01
We compute the dynamical polarization function for a graphene antidot lattice in the random-phase approximation. The computed polarization functions display a much more complicated structure than what is found for pristine graphene (even when evaluated beyond the Dirac-cone approximation...... the plasmon dispersion law and find an approximate square-root dependence with a suppressed plasmon frequency as compared to doped graphene. The plasmon dispersion is nearly isotropic and the developed approximation schemes agree well with the full calculation....
Contributions to the neutronic analysis of a gas-cooled fast reactor
Martin-del-Campo, Cecilia, E-mail: cecilia.martin.del.campo@gmail.com [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico); Reyes-Ramirez, Ricardo, E-mail: ricarera@yahoo.com.mx [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico); Francois, Juan-Luis, E-mail: juan.luis.francois@gmail.com [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico); Reinking-Cejudo, Arturo G., E-mail: reinking@servidor.unam.mx [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico)
2011-06-15
Highlights: > Differences on reactivity with MCNPX and TRIPOLI-4 are negligible. > Fuel lattice and core criticality calculations were done. > A higher Doppler coefficient than coolant density coefficient. > Zirconium carbide is a better reflector than silicon carbide. > Adequate active height, radial size and reflector thickness were obtained. - Abstract: In this work the Monte Carlo codes MCNPX and TRIPOLI-4 were used to perform the criticality calculations of the fuel assembly and the core configuration of a gas-cooled fast reactor (GFR) concept, currently in development. The objective is to make contributions to the neutronic analysis of a gas-cooled fast reactor. In this study the fuel assembly is based on a hexagonal lattice of fuel-pins. The materials used are uranium and plutonium carbide as fuel, silicon carbide as cladding, and helium gas as coolant. Criticality calculations were done for a fuel assembly where the axial reflector thickness was varied in order to find the optimal thickness. In order to determine the best material to be used as a reflector, in the reactor core with neutrons of high energy spectrum, criticality calculations were done for three reflector materials: zirconium carbide, silicon carbide and natural uranium. It was found that the zirconium carbide provides the best neutron reflection. Criticality calculations using different active heights were done to determine the optimal height, and the reflector thickness was adjusted. Core criticality calculations were performed with different radius sizes to determine the active radial dimension of the core. A negative temperature coefficient of reactivity was verified for the fuel. The effect on reactivity produced by changes in the coolant density was also evaluated. We present the main neutronic characteristics of a preliminary fuel and core designs for the GFR concept. ENDF-VI cross-sections libraries were used in both the MCNPX and TRIPOLI-4 codes, and we verified that the obtained
Isospin Breaking Effects on the Lattice
Tantalo, Nazario
2013-01-01
Isospin symmetry is not exact and the corrections to the isosymmetric limit are, in general, at the percent level. For gold plated quantities, such as pseudoscalar meson masses or the kaon leptonic and semileptonic decay rates, these effects are of the same order of magnitude of the errors quoted in nowadays lattice calculations and cannot be neglected any longer. In this talk I discuss the methods that have been developed in the last few years to calculate isospin breaking corrections by starting from first principles lattice simulations. In particular, I discuss how to perform a combined QCD+QED lattice simulation and a renormalization prescription to be used in order to separate QCD from QED isospin breaking effects. A brief review of recent lattice results of isospin breaking effects on the hadron spectrum is also included.
DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-TEMPERATURE GAS-COOLED TEST REACTOR
Sterbentz, James; Bayless, Paul; Strydom, Gerhard; Kumar, Akansha; Gougar, Hans
2016-11-01
Uncertainty and sensitivity analysis is an indispensable element of any substantial attempt in reactor simulation validation. The quantification of uncertainties in nuclear engineering has grown more important and the IAEA Coordinated Research Program (CRP) on High-Temperature Gas Cooled Reactor (HTGR) initiated in 2012 aims to investigate the various uncertainty quantification methodologies for this type of reactors. The first phase of the CRP is dedicated to the estimation of cell and lattice model uncertainties due to the neutron cross sections co-variances. Phase II is oriented towards the investigation of propagated uncertainties from the lattice to the coupled neutronics/thermal hydraulics core calculations. Nominal results for the prismatic single block (Ex.I-2a) and super cell models (Ex.I-2c) have been obtained using the SCALE 6.1.3 two-dimensional lattice code NEWT coupled to the TRITON sequence for cross section generation. In this work, the TRITON/NEWT-flux-weighted cross sections obtained for Ex.I-2a and various models of Ex.I-2c is utilized to perform a sensitivity analysis of the MHTGR-350 core power densities and eigenvalues. The core solutions are obtained with the INL coupled code PHISICS/RELAP5-3D, utilizing a fixed-temperature feedback for Ex. II-1a.. It is observed that the core power density does not vary significantly in shape, but the magnitude of these variations increases as the moderator-to-fuel ratio increases in the super cell lattice models.
Fuel lattice design using heuristics and new strategies
Ortiz S, J. J.; Castillo M, J. A.; Torres V, M.; Perusquia del Cueto, R. [ININ, Carretera Mexico-Toluca s/n, Ocoyoacac 52750, Estado de Mexico (Mexico); Pelta, D. A. [ETS Ingenieria Informatica y Telecomunicaciones, Universidad de Granada, Daniel Saucedo Aranda s/n, 18071 Granada (Spain); Campos S, Y., E-mail: juanjose.ortiz@inin.gob.m [IPN, Escuela Superior de Fisica y Matematicas, Unidad Profesional Adolfo Lopez Mateos, Edif. 9, 07738 Mexico D. F. (Mexico)
2010-10-15
This work show some results of the fuel lattice design in BWRs when some allocation pin rod rules are not taking into account. Heuristics techniques like Path Re linking and Greedy to design fuel lattices were used. The scope of this work is to search about how do classical rules in design fuel lattices affect the heuristics techniques results and the fuel lattice quality. The fuel lattices quality is measured by Power Peaking Factor and Infinite Multiplication Factor at the beginning of the fuel lattice life. CASMO-4 code to calculate these parameters was used. The analyzed rules are the following: pin rods with lowest uranium enrichment are only allocated in the fuel lattice corner, and pin rods with gadolinium cannot allocated in the fuel lattice edge. Fuel lattices with and without gadolinium in the main diagonal were studied. Some fuel lattices were simulated in an equilibrium cycle fuel reload, using Simulate-3 to verify their performance. So, the effective multiplication factor and thermal limits can be verified. The obtained results show a good performance in some fuel lattices designed, even thought, the knowing rules were not implemented. A fuel lattice performance and fuel lattice design characteristics analysis was made. To the realized tests, a dell workstation was used, under Li nux platform. (Author)
Rare Kaon Decays on the Lattice
Isidori, Gino; Turchetti, P; Isidori, Gino; Martinelli, Guido; Turchetti, Paolo
2006-01-01
We show that long distance contributions to the rare decays K -> pi nu nu-bar and K -> pi l+ l- can be computed using lattice QCD. The proposed approach requires well established methods, successfully applied in the calculations of electromagnetic and semileptonic form factors. The extra power divergences, related to the use of weak four-fermion operators, can be eliminated using only the symmetries of the lattice action without ambiguities or complicated non-perturbative subtractions. We demonstrate that this is true even when a lattice action with explicit chiral symmetry breaking is employed. Our study opens the possibility of reducing the present uncertainty in the theoretical predictions for these decays.
AN EQUIVALENT CONTINUUM METHOD OF LATTICE STRUCTURES
Fan Hualin; Yang Wei
2006-01-01
An equivalent continuum method is developed to analyze the effective stiffness of three-dimensional stretching dominated lattice materials. The strength and three-dimensional plastic yield surfaces are calculated for the equivalent continuum. A yielding model is formulated and compared with the results of other models. The bedding-in effect is considered to include the compliance of the lattice joints. The predicted stiffness and strength are in good agreement with the experimental data, validating the present model in the prediction of the mechanical properties of stretching dominated lattice structures.
Investigating jet quenching on the lattice
Panero, Marco; Schäfer, Andreas
2014-01-01
Due to the dynamical, real-time, nature of the phenomenon, the study of jet quenching via lattice QCD simulations is not straightforward. In this contribution, however, we show how one can extract information about the momentum broadening of a hard parton moving in the quark-gluon plasma, from lattice calculations. After discussing the basic idea (originally proposed by Caron-Huot), we present a recent study, in which we estimated the jet quenching parameter non-perturbatively, from the lattice evaluation of a particular set of gauge-invariant operators.
Electronic properties of graphene antidot lattices
Fürst, Joachim Alexander; Pedersen, Jesper Goor; Flindt, C.
2009-01-01
Graphene antidot lattices constitute a novel class of nano-engineered graphene devices with controllable electronic and optical properties. An antidot lattice consists of a periodic array of holes that causes a band gap to open up around the Fermi level, turning graphene from a semimetal...... into a semiconductor. We calculate the electronic band structure of graphene antidot lattices using three numerical approaches with different levels of computational complexity, efficiency and accuracy. Fast finite-element solutions of the Dirac equation capture qualitative features of the band structure, while full...
Properties of the Quark Gluon Plasma: A lattice perspective
Karsch, Frithjof
2007-01-01
We discuss results from lattice calculations for a few observables that are sensitive to different length scales in the high temperature phase of QCD and can give insight into its non-perturbative structure. We compare lattice results with perturbative calculations at high temperature obtained for vanishing and non-vanishing quark chemical potential.
Determinant of twisted chiral Dirac Operator on the Lattice
Fosco, C. D.; Randjbar-Daemi, S.
1995-01-01
Using the overlap formulation, we calculate the fermionic determinant on the lattice for chiral fermions with twisted boundary conditions in two dimensions. When the lattice spacing tends to zero we recover the results of the usual string-theory continuum calculations.