Differential and complex geometry origins, abstractions and embeddings

CERN Document Server

Wells, Jr , Raymond O

2017-01-01

Differential and complex geometry are two central areas of mathematics with a long and intertwined history. This book, the first to provide a unified historical perspective of both subjects, explores their origins and developments from the sixteenth to the twentieth century. Providing a detailed examination of the seminal contributions to differential and complex geometry up to the twentieth century embedding theorems, this monograph includes valuable excerpts from the original documents, including works of Descartes, Fermat, Newton, Euler, Huygens, Gauss, Riemann, Abel, and Nash. Suitable for beginning graduate students interested in differential, algebraic or complex geometry, this book will also appeal to more experienced readers.

The Kerr geometry, complex world lines and hyperbolic strings

International Nuclear Information System (INIS)

Burinskii, A.Ya.

1994-01-01

In the Lind-Newman representation the Kerr geometry is created by a source moving along an analytical complex world line. An equivalence of the complex world line and complex (hyperbolic) string is considered. Therefore the hyperbolic string may play the role of the complex source of the Kerr geometry. The Kerr solution with the complex string source acquires Regge behavior of the angular momentum. (orig.)

International conference on Algebraic and Complex Geometry

CERN Document Server

Kloosterman, Remke; Schütt, Matthias

2014-01-01

Several important aspects of moduli spaces and irreducible holomorphic symplectic manifolds were highlighted at the conference “Algebraic and Complex Geometry” held September 2012 in Hannover, Germany. These two subjects of recent ongoing progress belong to the most spectacular developments in Algebraic and Complex Geometry. Irreducible symplectic manifolds are of interest to algebraic and differential geometers alike, behaving similar to K3 surfaces and abelian varieties in certain ways, but being by far less well-understood. Moduli spaces, on the other hand, have been a rich source of open questions and discoveries for decades and still continue to be a hot topic in itself as well as with its interplay with neighbouring fields such as arithmetic geometry and string theory. Beyond the above focal topics this volume reflects the broad diversity of lectures at the conference and comprises 11 papers on current research from different areas of algebraic and complex geometry sorted in alphabetic order by the ...

MOCUM: A two-dimensional method of characteristics code based on constructive solid geometry and unstructured meshing for general geometries

International Nuclear Information System (INIS)

Yang Xue; Satvat, Nader

2012-01-01

Highlight: ► A two-dimensional numerical code based on the method of characteristics is developed. ► The complex arbitrary geometries are represented by constructive solid geometry and decomposed by unstructured meshing. ► Excellent agreement between Monte Carlo and the developed code is observed. ► High efficiency is achieved by parallel computing. - Abstract: A transport theory code MOCUM based on the method of characteristics as the flux solver with an advanced general geometry processor has been developed for two-dimensional rectangular and hexagonal lattice and full core neutronics modeling. In the code, the core structure is represented by the constructive solid geometry that uses regularized Boolean operations to build complex geometries from simple polygons. Arbitrary-precision arithmetic is also used in the process of building geometry objects to eliminate the round-off error from the commonly used double precision numbers. Then, the constructed core frame will be decomposed and refined into a Conforming Delaunay Triangulation to ensure the quality of the meshes. The code is fully parallelized using OpenMP and is verified and validated by various benchmarks representing rectangular, hexagonal, plate type and CANDU reactor geometries. Compared with Monte Carlo and deterministic reference solution, MOCUM results are highly accurate. The mentioned characteristics of the MOCUM make it a perfect tool for high fidelity full core calculation for current and GenIV reactor core designs. The detailed representation of reactor physics parameters can enhance the safety margins with acceptable confidence levels, which lead to more economically optimized designs.

Geometric Transitions, Topological Strings, and Generalized Complex Geometry

International Nuclear Information System (INIS)

Chuang, Wu-yen

2007-01-01

Mirror symmetry is one of the most beautiful symmetries in string theory. It helps us very effectively gain insights into non-perturbative worldsheet instanton effects. It was also shown that the study of mirror symmetry for Calabi-Yau flux compactification leads us to the territory of ''Non-Kaehlerity''. In this thesis we demonstrate how to construct a new class of symplectic non-Kaehler and complex non-Kaehler string theory vacua via generalized geometric transitions. The class admits a mirror pairing by construction. From a variety of sources, including super-gravity analysis and KK reduction on SU(3) structure manifolds, we conclude that string theory connects Calabi-Yau spaces to both complex non-Kaehler and symplectic non-Kaehler manifolds and the resulting manifolds lie in generalized complex geometry. We go on to study the topological twisted models on a class of generalized complex geometry, bi-Hermitian geometry, which is the most general target space for (2, 2) world-sheet theory with non-trivial H flux turned on. We show that the usual Kaehler A and B models are generalized in a natural way. Since the gauged supergravity is the low energy effective theory for the compactifications on generalized geometries, we study the fate of flux-induced isometry gauging in N = 2 IIA and heterotic strings under non-perturbative instanton effects. Interestingly, we find we have protection mechanisms preventing the corrections to the hyper moduli spaces. Besides generalized geometries, we also discuss the possibility of new NS-NS fluxes in a new doubled formalism

Geometric Transitions, Topological Strings, and Generalized Complex Geometry

Energy Technology Data Exchange (ETDEWEB)

Chuang, Wu-yen; /SLAC /Stanford U., Phys. Dept.

2007-06-29

Mirror symmetry is one of the most beautiful symmetries in string theory. It helps us very effectively gain insights into non-perturbative worldsheet instanton effects. It was also shown that the study of mirror symmetry for Calabi-Yau flux compactification leads us to the territory of ''Non-Kaehlerity''. In this thesis we demonstrate how to construct a new class of symplectic non-Kaehler and complex non-Kaehler string theory vacua via generalized geometric transitions. The class admits a mirror pairing by construction. From a variety of sources, including super-gravity analysis and KK reduction on SU(3) structure manifolds, we conclude that string theory connects Calabi-Yau spaces to both complex non-Kaehler and symplectic non-Kaehler manifolds and the resulting manifolds lie in generalized complex geometry. We go on to study the topological twisted models on a class of generalized complex geometry, bi-Hermitian geometry, which is the most general target space for (2, 2) world-sheet theory with non-trivial H flux turned on. We show that the usual Kaehler A and B models are generalized in a natural way. Since the gauged supergravity is the low energy effective theory for the compactifications on generalized geometries, we study the fate of flux-induced isometry gauging in N = 2 IIA and heterotic strings under non-perturbative instanton effects. Interestingly, we find we have protection mechanisms preventing the corrections to the hyper moduli spaces. Besides generalized geometries, we also discuss the possibility of new NS-NS fluxes in a new doubled formalism.

Numerical simulation and geometry optimization of hot-gas mixing in lower plenum of high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Wang Hang; Wang Jie; Laurien, E.

2010-01-01

The lower plenum in high temperature gas-cooled reactor was designed to mix the gas of different temperatures from the reactor core. Previous researches suggest the current geometry of the lower plenum to be improved for better mixing capability and lower pressure drop. In the presented work, a series of varied geometries were investigated with numerical simulation way. The choice of appropriate mesh type and size used in the geometry variation was discussed with the reference of experimental data. The original thin ribs in the current design were merged into thicker ones, and a junction located at the starting end of the outlet pipe was introduced. After comparing several potential optimization methods, an improved geometry was selected with the merged ribs increasing the pre-defined mixing coefficient and the junction reducing the pressure drop. Future work was discussed based on the simulation of real reactor case. The work shows a direction for design improvements of the lower plenum geometry. (authors)

Complex geometry and quantum string theory

International Nuclear Information System (INIS)

Belavin, A.A.; Knizhnik, V.G.

1986-01-01

Summation over closed oriented surfaces of genus p ≥ 2 (p - loop vacuum amplitudes in boson string theory) in a critical dimensions D=26 is reduced to integration over M p space of complex structures of Riemann surfaces of genus p. The analytic properties of the integration measure as a function of the complex coordinates on M p are studied. It is shown that the measure multiplied by (det Im τ-circumflex) 13 (τ-circumflex is the surface period matrix) is the square of the modulus of a function which is holomorphic on M p and does not vanish anywhere. The function has a second order pole at infinity of compactified space of moduli M p . These properties define the measure uniquely up to a constant multiple and this permits one to set up explicitformulae for p=2,3 in terms of the theta-constants. Power and logarithmic divergences connected with renormalization of the tachyon wave function and of the slope respectively are involved in the theory. Quantum geometry of critical strings turns out to be a complex geometry

Complex geometries in wood

DEFF Research Database (Denmark)

Tamke, Martin; Ramsgaard Thomsen, Mette; Riiber Nielsen, Jacob

2009-01-01

The versatility of wood constructions and traditional wood joints for the production of non standard elements was in focus of a design based research. Herein we established a seamless process from digital design to fabrication. A first research phase centered on the development of a robust...... parametric model and a generic design language a later explored the possibilities to construct complex shaped geometries with self registering joints on modern wood crafting machines. The research was carried out as collaboration with industrial partners....

Application of finite element numerical technique to nuclear reactor geometries

Energy Technology Data Exchange (ETDEWEB)

Rouai, N M [Nuclear engineering department faculty of engineering Al-fateh universty, Tripoli (Libyan Arab Jamahiriya)

1995-10-01

Determination of the temperature distribution in nuclear elements is of utmost importance to ensure that the temperature stays within safe limits during reactor operation. This paper discusses the use of Finite element numerical technique (FE) for the solution of the two dimensional heat conduction equation in geometries related to nuclear reactor cores. The FE solution stats with variational calculus which considers transforming the heat conduction equation into an integral equation I(O) and seeks a function that minimizes this integral and hence gives the solution to the heat conduction equation. In this paper FE theory as applied to heat conduction is briefly outlined and a 2-D program is used to apply the theory to simple shapes and to two gas cooled reactor fuel elements. Good results are obtained for both cases with reasonable number of elements. 7 figs.

Application of finite element numerical technique to nuclear reactor geometries

International Nuclear Information System (INIS)

Rouai, N. M.

1995-01-01

Determination of the temperature distribution in nuclear elements is of utmost importance to ensure that the temperature stays within safe limits during reactor operation. This paper discusses the use of Finite element numerical technique (FE) for the solution of the two dimensional heat conduction equation in geometries related to nuclear reactor cores. The FE solution stats with variational calculus which considers transforming the heat conduction equation into an integral equation I(O) and seeks a function that minimizes this integral and hence gives the solution to the heat conduction equation. In this paper FE theory as applied to heat conduction is briefly outlined and a 2-D program is used to apply the theory to simple shapes and to two gas cooled reactor fuel elements. Good results are obtained for both cases with reasonable number of elements. 7 figs

Pressure loss coefficient evaluation based on CFD analysis for simple geometries and PWR reactor vessel without geometry simplification

International Nuclear Information System (INIS)

Ko II, B.; Park, J. P.; Jeong, J. H.

2008-01-01

Nuclear vendors and utilities perform lots of simulations and analyses in order to ensure the safe operation of nuclear power plants (NPPs). In general, the simulations are carried out using vendor-specific design codes and best-estimate system analysis codes and most of them were developed based on 1-dimensional lumped parameter models. These thermal-hydraulic system analysis codes require user input for pressure loss coefficient, k-factor; since they numerically solve Euler-equation. In spite of its high impact on the safety analysis results, there has not been good validation method for the selection of loss coefficient. During the past decade, however; computers, parallel computation methods, and 3-dimensional computational fluid dynamics (CFD) codes have been dramatically enhanced. It is believed to be beneficial to take advantage of advanced commercial CFD codes in safety analysis and design of NPP5. The present work aims to validate pressure loss coefficient evaluation for simple geometries and k-factor calculation for PWR based on CFD. The performances of standard k-ε model, RNG k-ε model, Reynolds stress model (RSM) on the simulation of pressure drop for simple geometry such as, or sudden-expansion, and sudden-contraction are evaluated. The calculated value was compared with pressure loss coefficient in handbook of hydraulic resistance. Then the present work carried out analysis for flow distribution in downcomer and lower plenum of Korean standard nuclear power plants (KSNPs) using STAR-CD. The lower plenum geometry of a PWR is very complicated since there are so many reactor internals, which hinders in CFD analysis for real reactor geometry up to now. The present work takes advantage of 3D CAD model so that real geometry of lower plenum is used. The results give a clear figure about flow fields in the reactor vessel, which is one of major safety concerns. The calculated pressure drop across downcomer and lower plenum appears to be in good agreement

Practice of calculation of neutron-physical characteristics of reactors and radiating shielding in structure SNPS with program complex MCNP

International Nuclear Information System (INIS)

Krotov, A.D.; Son'ko, A.V.

2009-01-01

Calculation of neutron-physical properties and radiation protection of space power reactor was made by means of the MCNP code allowing simulation of neutron, γ- and electron transport by the Monte Carlo method in the systems with combined geometry. Universality of the MCNP code has been demonstrated both for the calculation of reactor-converter so for the optimization of radiation protection that allows to reserve a new level of complex simulation of SNPS [ru

Advancing the manufacture of complex geometry GFRC for today's building envelopes

Directory of Open Access Journals (Sweden)

Thomas Henriksen

2017-06-01

With this research the current architectural knowledge base has been advanced in terms of complex geometry thin-walled GFRC for building envelopes. The identified solutions should allow building with complex geometries to be realised using thin-walled GFRC as the envelope cladding.

An introduction to complex analysis and geometry

CERN Document Server

D'Angelo, John P

2010-01-01

An Introduction to Complex Analysis and Geometry provides the reader with a deep appreciation of complex analysis and how this subject fits into mathematics. The book developed from courses given in the Campus Honors Program at the University of Illinois Urbana-Champaign. These courses aimed to share with students the way many mathematics and physics problems magically simplify when viewed from the perspective of complex analysis. The book begins at an elementary level but also contains advanced material. The first four chapters provide an introduction to complex analysis with many elementary

Experimental and numerical investigation of shock wave propagation through complex geometry, gas continuous, two-phase media

Energy Technology Data Exchange (ETDEWEB)

Chien-Chih Liu, James [Univ. of California, Berkeley, CA (United States)

1993-01-01

The work presented here investigates the phenomenon of shock wave propagation in gas continuous, two-phase media. The motivation for this work stems from the need to understand blast venting consequences in the HYLIFE inertial confinement fusion (ICF) reactor. The HYLIFE concept utilizes lasers or heavy ion beams to rapidly heat and compress D-T targets injected into the center of a reactor chamber. A segmented blanket of falling molten lithium or Li₂BeF₄ (Flibe) jets encircles the reactor`s central cavity, shielding the reactor structure from radiation damage, absorbing the fusion energy, and breeding more tritium fuel. X-rays from the fusion microexplosion will ablate a thin layer of blanket material from the surfaces which face toward the fusion site. This generates a highly energetic vapor, which mostly coalesces in the central cavity. The blast expansion from the central cavity generates a shock which propagates through the segmented blanket - a complex geometry, gas-continuous two-phase medium. The impulse that the blast gives to the liquid as it vents past, the gas shock on the chamber wall, and ultimately the liquid impact on the wall are all important quantities to the HYLIFE structural designers.

3D Printer-Manufacturing of Complex Geometry Elements

Science.gov (United States)

Ciubară, A.; Burlea, Ș L.; Axinte, M.; Cimpoeșu, R.; Chicet, D. L.; Manole, V.; Burlea, G.; Cimpoeșu, N.

2018-06-01

In the last 5-10 years the process of 3D printing has an incredible advanced in all the fields with a tremendous number of applications. Plastic materials exhibit highly beneficial mechanical properties while delivering complex designs impossible to achieve using conventional manufacturing. In this article the printing process (filling degree, time, complications and details finesse) of few plastic elements with complicated geometry and fine details was analyzed and comment. 3D printing offers many of the thermoplastics and industrial materials found in conventional manufacturing. The advantages and disadvantages of 3D printing for plastic parts are discussed. Time of production for an element with complex geometry, from the design to final cut, was evaluated.

Modular fabrication and characterization of complex silicon carbide composite structures Advanced Reactor Technologies (ART) Research Final Report (Feb 2015 – May 2017)

Energy Technology Data Exchange (ETDEWEB)

Khalifa, Hesham [General Atomics, San Diego, CA (United States)

2017-08-03

Advanced ceramic materials exhibit properties that enable safety and fuel cycle efficiency improvements in advanced nuclear reactors. In order to fully exploit these desirable properties, new processing techniques are required to produce the complex geometries inherent to nuclear fuel assemblies and support structures. Through this project, the state of complex SiC-SiC composite fabrication for nuclear components has advanced significantly. New methods to produce complex SiC-SiC composite structures have been demonstrated in the form factors needed for in-core structural components in advanced high temperature nuclear reactors. Advanced characterization techniques have been employed to demonstrate that these complex SiC-SiC composite structures provide the strength, toughness and hermeticity required for service in harsh reactor conditions. The complex structures produced in this project represent a significant step forward in leveraging the excellent high temperature strength, resistance to neutron induced damage, and low neutron cross section of silicon carbide in nuclear applications.

The simplified P3 approach on a trigonal geometry in the nodal reactor code DYN3D

International Nuclear Information System (INIS)

Duerigen, S.; Fridman, E.

2011-01-01

DYN3D is a three-dimensional nodal diffusion code for steady-state and transient analyses of Light-Water Reactors with square and hexagonal fuel assembly geometries. Currently, several versions of the DYN3D code are available including a multi-group diffusion and a simplified P 3 (SP 3 ) neutron transport option. In this work, the multi-group SP 3 method based on trigonal-z geometry was developed. The method is applicable to the analysis of reactor cores with hexagonal fuel assemblies and allows flexible mesh refinement, which is of particular importance for WWER-type Pressurized Water Reactors as well as for innovative reactor concepts including block type High-Temperature Reactors and Sodium Fast Reactors. In this paper, the theoretical background for the trigonal SP 3 methodology is outlined and the results of a preliminary verification analysis are presented by means of a simplified WWER-440 core test example. The accordant cross sections and reference solutions were produced by the Monte Carlo code SERPENT. The DYN3D results are in good agreement with the reference solutions. The average deviation in the nodal power distribution is about 1%. (Authors)

Control rod interaction models for use in 2D and 3D reactor geometries

International Nuclear Information System (INIS)

Bannerman, R.C.

1985-11-01

Control rod interaction models are developed for use in two-dimensional and three-dimensional reactor geometries. These models allow the total worth of any combination of control rods inserted to be determined from the individual worths in conjunction with an algorithm representing interaction effects between them. The validity of the assumptions is demonstrated for fast and thermal systems showing modelling errors of 1#percent# or less in inserted control rod worths. The models are ideally suited for most reactor systems. (UK)

Geometry and quadratic nonlinearity of charge transfer complexes in solution: A theoretical study

International Nuclear Information System (INIS)

Mukhopadhyay, S.; Ramasesha, S.; Pandey, Ravindra; Das, Puspendu K.

2011-01-01

In this paper, we have computed the quadratic nonlinear optical (NLO) properties of a class of weak charge transfer (CT) complexes. These weak complexes are formed when the methyl substituted benzenes (donors) are added to strong acceptors like chloranil (CHL) or di-chloro-di-cyano benzoquinone (DDQ) in chloroform or in dichloromethane. The formation of such complexes is manifested by the presence of a broad absorption maximum in the visible range of the spectrum where neither the donor nor the acceptor absorbs. The appearance of this visible band is due to CT interactions, which result in strong NLO responses. We have employed the semiempirical intermediate neglect of differential overlap (INDO/S) Hamiltonian to calculate the energy levels of these CT complexes using single and double configuration interaction (SDCI). The solvent effects are taken into account by using the self-consistent reaction field (SCRF) scheme. The geometry of the complex is obtained by exploring different relative molecular geometries by rotating the acceptor with respect to the fixed donor about three different axes. The theoretical geometry that best fits the experimental energy gaps, β HRS and macroscopic depolarization ratios is taken to be the most probable geometry of the complex. Our studies show that the most probable geometry of these complexes in solution is the parallel displaced structure with a significant twist in some cases.

Complex quantum network geometries: Evolution and phase transitions

Science.gov (United States)

Bianconi, Ginestra; Rahmede, Christoph; Wu, Zhihao

2015-08-01

Networks are topological and geometric structures used to describe systems as different as the Internet, the brain, or the quantum structure of space-time. Here we define complex quantum network geometries, describing the underlying structure of growing simplicial 2-complexes, i.e., simplicial complexes formed by triangles. These networks are geometric networks with energies of the links that grow according to a nonequilibrium dynamics. The evolution in time of the geometric networks is a classical evolution describing a given path of a path integral defining the evolution of quantum network states. The quantum network states are characterized by quantum occupation numbers that can be mapped, respectively, to the nodes, links, and triangles incident to each link of the network. We call the geometric networks describing the evolution of quantum network states the quantum geometric networks. The quantum geometric networks have many properties common to complex networks, including small-world property, high clustering coefficient, high modularity, and scale-free degree distribution. Moreover, they can be distinguished between the Fermi-Dirac network and the Bose-Einstein network obeying, respectively, the Fermi-Dirac and Bose-Einstein statistics. We show that these networks can undergo structural phase transitions where the geometrical properties of the networks change drastically. Finally, we comment on the relation between quantum complex network geometries, spin networks, and triangulations.

Generalized complex geometry, generalized branes and the Hitchin sigma model

International Nuclear Information System (INIS)

Zucchini, Roberto

2005-01-01

Hitchin's generalized complex geometry has been shown to be relevant in compactifications of superstring theory with fluxes and is expected to lead to a deeper understanding of mirror symmetry. Gualtieri's notion of generalized complex submanifold seems to be a natural candidate for the description of branes in this context. Recently, we introduced a Batalin-Vilkovisky field theoretic realization of generalized complex geometry, the Hitchin sigma model, extending the well known Poisson sigma model. In this paper, exploiting Gualtieri's formalism, we incorporate branes into the model. A detailed study of the boundary conditions obeyed by the world sheet fields is provided. Finally, it is found that, when branes are present, the classical Batalin-Vilkovisky cohomology contains an extra sector that is related non trivially to a novel cohomology associated with the branes as generalized complex submanifolds. (author)

A dissipative particle dynamics method for arbitrarily complex geometries

Science.gov (United States)

Li, Zhen; Bian, Xin; Tang, Yu-Hang; Karniadakis, George Em

2018-02-01

Dissipative particle dynamics (DPD) is an effective Lagrangian method for modeling complex fluids in the mesoscale regime but so far it has been limited to relatively simple geometries. Here, we formulate a local detection method for DPD involving arbitrarily shaped geometric three-dimensional domains. By introducing an indicator variable of boundary volume fraction (BVF) for each fluid particle, the boundary of arbitrary-shape objects is detected on-the-fly for the moving fluid particles using only the local particle configuration. Therefore, this approach eliminates the need of an analytical description of the boundary and geometry of objects in DPD simulations and makes it possible to load the geometry of a system directly from experimental images or computer-aided designs/drawings. More specifically, the BVF of a fluid particle is defined by the weighted summation over its neighboring particles within a cutoff distance. Wall penetration is inferred from the value of the BVF and prevented by a predictor-corrector algorithm. The no-slip boundary condition is achieved by employing effective dissipative coefficients for liquid-solid interactions. Quantitative evaluations of the new method are performed for the plane Poiseuille flow, the plane Couette flow and the Wannier flow in a cylindrical domain and compared with their corresponding analytical solutions and (high-order) spectral element solution of the Navier-Stokes equations. We verify that the proposed method yields correct no-slip boundary conditions for velocity and generates negligible fluctuations of density and temperature in the vicinity of the wall surface. Moreover, we construct a very complex 3D geometry - the "Brown Pacman" microfluidic device - to explicitly demonstrate how to construct a DPD system with complex geometry directly from loading a graphical image. Subsequently, we simulate the flow of a surfactant solution through this complex microfluidic device using the new method. Its

The AFEN Method in Cylindrical (r,θ ,z) Geometry for Pebble Bed Reactors -Incorporation of Acceleration and Discontinuity Factor

International Nuclear Information System (INIS)

Lee, Jaejun; Cho, Namzin

2007-01-01

Most existing methods of nuclear design analysis for pebble bed reactors (PBRs) are based on old finite difference solvers or on statistical methods. These methods require very long computer times. Therefore, there is strong desire of making available high fidelity coarse-mesh nodal computer codes. Recently, we extended the analytic function expansion nodal (AFEN) method developed quite extensively in Cartesian (x,y,z) geometry and in hexagonal-z geometry to the treatment of the full three dimensional cylindrical (r,θ,z) geometry for pebble bed reactors(PBRs). The AFEN methodology in this geometry as in hexagonal geometry is 'robust', due to the unique feature of the AFEN method that it does not use the transverse integration. This paper presents an acceleration scheme based on the coarse-group rebalance (CGR) concept and provides test results verifying the method and its implementation in the TOPS code. Also, we implemented discontinuity factors in the TOPS code and tested on benchmark problems. The TOPS results are in excellent agreement with those of the VENTURE code, using significantly less computer time

2-DB, 2-D Multigroup Diffusion, X-Y, R-Theta, Hexagonal Geometry Fast Reactor, Criticality Search

International Nuclear Information System (INIS)

Little, W.W. Jr.; Hardie, R.W.; Hirons, T.J.; O'Dell, R.D.

1969-01-01

1 - Description of problem or function: 2DB is a flexible, two- dimensional (x-y, r-z, r-theta, hex geometry) diffusion code for use in fast reactor analyses. The code can be used to: (a) Compute fuel burnup using a flexible material shuffling scheme. (b) Perform criticality searches on time absorption (alpha), material concentrations, and region dimensions using a regular or adjoint model. Criticality searches can be performed during burnup to compensate for fuel depletion. (c) Compute flux distributions for an arbitrary extraneous source. 2 - Method of solution: Standard source-iteration techniques are used. Group re-balancing and successive over-relaxation with line inversion are used to accelerate convergence. Material burnup is by reactor zone. The burnup rate is determined by the zone and energy (group) averaged cross sections which are recomputed after each time-step. The isotopic chains, which can contain any number of isotopes, are formed by the user. The code does not contain built-in or internal chains. 3 - Restrictions on the complexity of the problem: Since variable dimensioning is employed, no simple bounds can be stated. The current 1108 version, however, is nominally restricted to 50 energy groups in a 65 K memory. In the 6600 version the power fraction, average burnup rate, and breeding ratio calculations are limited to reactors with a maximum of 50 zones

Optimization geometries of a vortex gliding-arc reactor for partial oxidation of methane

International Nuclear Information System (INIS)

Guofeng, Xu; Xinwei, Ding

2012-01-01

The effects of the geometry of gliding-arc reactor – such as distance between the electrodes, outlet diameter, and inlet position – on the reactor characteristics (methane conversion, hydrogen yield, and energy efficiency) have not been fully investigated. In this paper, AC gliding-arc reactors including the vortex flow configuration are designed to produce hydrogen from the methane by partial oxidation. The influence of vortex flow configuration on the reactor characteristics is also studied by varying the inlet position. When the inlet of the gliding-arc reactor is positioned close to the outlet, reverse vortex flow reactor (RVFR), the maximum energy efficiency reaches 50% and the yields of hydrogen and carbon monoxide are 40% and 65%, respectively. As the distance between electrodes increases from 5 mm to 15 mm, both hydrogen yield and energy efficiency increase approximately 10% for the RVFR. The energy efficiency and hydrogen yield are highest when the ratio of the outlet diameter to the inner diameter is 0.5 for the RVFR. Experimental results indicate that the flow field in the plasma reactor has an important influence on the reactor performance. Furthermore, hydrogen production increases as the number of feed gas flows in contact with the plasma zone increases. -- Highlights: ► Gliding-arc reactors were designed to produce hydrogen for studying the characteristics of the vortex flow reactor. ► Hydrogen yield of reverse vortex flow reactor was 10% higher than that of forward vortex flow reactor. ► Maximum energy efficiency was 50% for reverse vortex flow reactor. ► If discharge power was supplied to the reactors, the reactor performance increased with increasing distance between electrodes. ► Optimum ratio of the outlet and inner diameter was 1/2.

Seismic analysis of fast breeder reactor block

International Nuclear Information System (INIS)

Gantenbein, F.

1990-01-01

Seismic analysis of LMFBR reactor block is complex due mainly to the fluid structure interaction and the 3D geometry of the structure. Analytical methods which have been developed for this analysis will be briefly described in the paper and applications to a geometry similar to SPX1 will be shown

Numerical simulations and mathematical models of flows in complex geometries

DEFF Research Database (Denmark)

Hernandez Garcia, Anier

The research work of the present thesis was mainly aimed at exploiting one of the strengths of the Lattice Boltzmann methods, namely, the ability to handle complicated geometries to accurately simulate flows in complex geometries. In this thesis, we perform a very detailed theoretical analysis...... and through the Chapman-Enskog multi-scale expansion technique the dependence of the kinetic viscosity on each scheme is investigated. Seeking for optimal numerical schemes to eciently simulate a wide range of complex flows a variant of the finite element, off-lattice Boltzmann method [5], which uses...... the characteristic based integration is also implemented. Using the latter scheme, numerical simulations are conducted in flows of different complexities: flow in a (real) porous network and turbulent flows in ducts with wall irregularities. From the simulations of flows in porous media driven by pressure gradients...

A co-ordinate system for reactor physics calculations in hexagonal geometry

International Nuclear Information System (INIS)

Burte, D.P.

1990-01-01

A method for generating all the geometric information concerning typical reactor physics calculations for a basically hexagonal reactor core or its sector involving any of the possible symmetries is presented. The geometrically allowed symmetries for regular hexagons are discussed. The approach is based on the choice of a suitable co-ordinate system, viz. one using three coplanar (including one redundant) axes, each at 120 0 with its cyclically preceding one. A code named KEKULE' is developed for a 2-D, finite difference, one-group diffusion analysis of a hexagonal core using the approach. It can cater to a full hexagonal core as well as to any symmetric sectorial part of it. The main feature of the code is that the input concerning geometry is a bare minimum. It is hoped that the approach presented will be useful even for the calculations for hexagonal fuel assemblies. (author)

Culham conceptual Tokamak reactor MkII. Conceptual layout of buildings for a twin reactor power station

International Nuclear Information System (INIS)

Guthrie, J.A.S.; Harding, N.H.

1981-01-01

This paper discusses the conceptual design of the nuclear complex of a 2400 MWe twin fusion reactor power station utilising common services and a single containment building. The design is based upon environmental and maintenance logistical requirements, the provision of adequate storage, workshop and construction facilities and the constraints imposed by the geometry of the main and auxiliary reactor coolant systems. (author)

Calculation of neutron spectra in the reactor cell of the RA experimental reactor in Vinca

International Nuclear Information System (INIS)

Bosevski, T.; Altiparmakov, D.; Marinkovic, N.

1974-01-01

In the frame of neutron properties of RA experimental reactor the study of energy neutron spectra in the reactor cell are planned. Complex reactor cell geometry, nine cylindrical regions causes high space-energy variations of neutron flux with a significant gradient both in energy and space variables. Treatment of such a complex problem needs adequate methodology which ensures reliable results and control of accuracy. This paper describes in detail the method for calculating group constants based on lattice cell calculation for the need of calculation of reactor core parameters. In 26 group approximation for the energy region from 0 - 10.5 MeV, values of neutron spectra are obtained in 18 space points chosen to describe, with high accuracy, integral reactor cell parameters of primary importance for the reactor core calculation. Obtained space-energy distribution of neutron flux in the reactor cell is up to now unique in the study of neutron properties of Ra reactor [sr

Conference on Complex Geometry and Mirror Symmetry

CERN Document Server

Vinet, Luc; Yau, Shing-Tung; Mirror Symmetry III

1999-01-01

This book presents surveys from a workshop held during the theme year in geometry and topology at the Centre de recherches mathématiques (CRM, University of Montréal). The volume is in some sense a sequel to Mirror Symmetry I (1998) and Mirror Symmetry II (1996), copublished by the AMS and International Press. Included are recent developments in the theory of mirror manifolds and the related areas of complex and symplectic geometry. The long introductory articles explain the key physical ideas and motivation, namely conformal field theory, supersymmetry, and string theory. Open problems are emphasized. Thus the book provides an efficient way for a very broad audience of mathematicians and physicists to reach the frontier of research in this fast expanding area. - See more at: http://bookstore.ams.org/amsip-10#sthash.DbxEFJDx.dpuf

Phased array ultrasound testing on complex geometry

International Nuclear Information System (INIS)

Tuan Arif Tuan Mat; Khazali Mohd Zin

2009-01-01

Phase array ultrasonic inspection is used to investigate its response to complex welded joints geometry. A 5 MHz probe with 64 linear array elements was employed to scan mild steel T-joint, nozzle and node samples. These samples contain many defects such as cracks, lack of penetration and lack of fusion. Ultrasonic respond is analysed and viewed using the Tomoview software. The results show the actual phase array images on respective types of defect. (author)

Verification of a neutronic code for transient analysis in reactors with Hex-z geometry

Energy Technology Data Exchange (ETDEWEB)

Gonzalez-Pintor, S.; Verdu, G. [Departamento de Ingenieria Quimica Y Nuclear, Universitat Politecnica de Valencia, Cami de Vera, 14, 46022. Valencia (Spain); Ginestar, D. [Departamento de Matematica Aplicada, Universitat Politecnica de Valencia, Cami de Vera, 14, 46022. Valencia (Spain)

2012-07-01

Due to the geometry of the fuel bundles, to simulate reactors such as VVER reactors it is necessary to develop methods that can deal with hexagonal prisms as basic elements of the spatial discretization. The main features of a code based on a high order finite element method for the spatial discretization of the neutron diffusion equation and an implicit difference method for the time discretization of this equation are presented and the performance of the code is tested solving the first exercise of the AER transient benchmark. The obtained results are compared with the reference results of the benchmark and with the results provided by PARCS code. (authors)

Full Core modeling techniques for research reactors with irregular geometries using Serpent and PARCS applied to the CROCUS reactor

International Nuclear Information System (INIS)

Siefman, Daniel J.; Girardin, Gaëtan; Rais, Adolfo; Pautz, Andreas; Hursin, Mathieu

2015-01-01

Highlights: • Modeling of research reactors. • Serpent and PARCS coupling. • Lattice physics codes modeling techniques. - Abstract: This paper summarizes the results of modeling methodologies developed for the zero-power (100 W) teaching and research reactor CROCUS located in the Laboratory for Reactor Physics and Systems Behavior (LRS) at the Swiss Federal Institute of Technology in Lausanne (EPFL). The study gives evidence that the Monte Carlo code Serpent can be used effectively as a lattice physics tool for small reactors. CROCUS’ core has an irregular geometry with two fuel zones of different lattice pitches. This and the reactor’s small size necessitate the use of nonstandard cross-section homogenization techniques when modeling the full core with a 3D nodal diffusion code (e.g. PARCS). The primary goal of this work is the development of these techniques for steady-state neutronics and future transient neutronics analyses of not only CROCUS, but research reactors in general. In addition, the modeling methods can provide useful insight for analyzing small modular reactor concepts based on light water technology. Static computational models of CROCUS with the codes Serpent and MCNP5 are presented and methodologies are analyzed for using Serpent and SerpentXS to prepare macroscopic homogenized group cross-sections for a pin-by-pin model of CROCUS with PARCS. The most accurate homogenization scheme lead to a difference in terms of k eff of 385 pcm between the Serpent and PARCS model, while the MCNP5 and Serpent models differed in terms of k eff by 13 pcm (within the statistical error of each simulation). Comparisons of the axial power profiles between the Serpent model as a reference and a set of PARCS models using different homogenization techniques showed a consistent root-mean-square deviation of ∼8%, indicating that the differences are not due to the homogenization technique but rather arise from the definition of the diffusion coefficients

GPU-accelerated depth map generation for X-ray simulations of complex CAD geometries

Science.gov (United States)

Grandin, Robert J.; Young, Gavin; Holland, Stephen D.; Krishnamurthy, Adarsh

2018-04-01

Interactive x-ray simulations of complex computer-aided design (CAD) models can provide valuable insights for better interpretation of the defect signatures such as porosity from x-ray CT images. Generating the depth map along a particular direction for the given CAD geometry is the most compute-intensive step in x-ray simulations. We have developed a GPU-accelerated method for real-time generation of depth maps of complex CAD geometries. We preprocess complex components designed using commercial CAD systems using a custom CAD module and convert them into a fine user-defined surface tessellation. Our CAD module can be used by different simulators as well as handle complex geometries, including those that arise from complex castings and composite structures. We then make use of a parallel algorithm that runs on a graphics processing unit (GPU) to convert the finely-tessellated CAD model to a voxelized representation. The voxelized representation can enable heterogeneous modeling of the volume enclosed by the CAD model by assigning heterogeneous material properties in specific regions. The depth maps are generated from this voxelized representation with the help of a GPU-accelerated ray-casting algorithm. The GPU-accelerated ray-casting method enables interactive (> 60 frames-per-second) generation of the depth maps of complex CAD geometries. This enables arbitrarily rotation and slicing of the CAD model, leading to better interpretation of the x-ray images by the user. In addition, the depth maps can be used to aid directly in CT reconstruction algorithms.

Geometry and Framework Interactions of Zeolite-Encapsulated Copper(II)-Histidine Complexes

NARCIS (Netherlands)

Weckhuysen, B.M.; Grommen, R.; Manikandan, P.; Gao, Y.; Shane, T.; Shane, J.J.; Schoonheydt, R.A.; Goldfarb, D.

2000-01-01

The coordination geometry of zeolite-encapsulated copper(II)-histidine (CuHis) complexes, prepared by ion exchange of the complexes from aqueous solutions into zeolite NaY, was determined by a combination of UV-vis-NIR diffuse reflectance spectroscopy (DRS), X-band EPR, electron-spin-echo envelope

Natural circulation in a VVER reactor geometry: Experiments with the PACTEL facility and Cathare simulations

Energy Technology Data Exchange (ETDEWEB)

Raussi, P.; Kainulainen, S. [Lappeenranta Univ. of Technology, Lappeenranta (Finland); Kouhia, J. [VTT Energy, Lappeenranta (Finland)

1995-09-01

There are some 40 reactors based on the VVER design in use. Database available for computer code assessment for VVER reactors is rather limited. Experiments were conducted to study natural circulation behaviour in the PACTEL facility, a medium-scale integral test loop patterned after VVER pressurized water reactors. Flow behaviour over a range of coolant inventories was studied with a small-break experiment. In the small-break experiments, flow stagnation and system repressurization were observed when the water level in the upper plenum fell below the entrances to the hot legs. The cause was attributed to the hot leg loop seals, which are a unique feature of the VVER geometry. At low primary inventories, core cooling was achieved through the boiler-condenser mode. The experiment was simulated using French thermalhydraulic system code CATHARE.

Presentation of geometries and transient results of TRAC-calculations

International Nuclear Information System (INIS)

Lutz, A.; Lang, U.; Ruehle, R.

1985-02-01

The computer code TRAC is used to analyze the transient behaviour of nuclear reactors. The input of a TRAC-Calculation, as well as the produced result files serve for the graphical presentation of the geometries and transient results. This supports the search for errors during input generation and the understanding of complex processes by dynamic presentation of calculational result in colour. (orig.) [de

Chemotactic droplet swimmers in complex geometries

Science.gov (United States)

Jin, Chenyu; Hokmabad, Babak V.; Baldwin, Kyle A.; Maass, Corinna C.

2018-02-01

Chemotaxis1 and auto-chemotaxis are key mechanisms in the dynamics of micro-organisms, e.g. in the acquisition of nutrients and in the communication between individuals, influencing the collective behaviour. However, chemical signalling and the natural environment of biological swimmers are generally complex, making them hard to access analytically. We present a well-controlled, tunable artificial model to study chemotaxis and autochemotaxis in complex geometries, using microfluidic assays of self-propelling oil droplets in an aqueous surfactant solution (Herminghaus et al 2014 Soft Matter 10 7008-22 Krüger et al 2016 Phys. Rev. Lett. 117). Droplets propel via interfacial Marangoni stresses powered by micellar solubilisation. Moreover, filled micelles act as a chemical repellent by diffusive phoretic gradient forces. We have studied these chemotactic effects in a series of microfluidic geometries, as published in Jin et al (2017 Proc. Natl Acad. Sci. 114 5089-94): first, droplets are guided along the shortest path through a maze by surfactant diffusing into the maze from the exit. Second, we let auto-chemotactic droplet swimmers pass through bifurcating microfluidic channels and record anticorrelations between the branch choices of consecutive droplets. We present an analytical Langevin model matching the experimental data. In a previously unpublished experiment, pillar arrays of variable sizes and shapes provide a convex wall interacting with the swimmer and, in the case of attachment, bending its trajectory and forcing it to revert to its own trail. We observe different behaviours based on the interplay of wall curvature and negative autochemotaxis, i.e. no attachment for highly curved interfaces, stable trapping at large pillars, and a narrow transition region where negative autochemotaxis makes the swimmers detach after a single orbit.

A computational approach to modeling cellular-scale blood flow in complex geometry

Science.gov (United States)

Balogh, Peter; Bagchi, Prosenjit

2017-04-01

We present a computational methodology for modeling cellular-scale blood flow in arbitrary and highly complex geometry. Our approach is based on immersed-boundary methods, which allow modeling flows in arbitrary geometry while resolving the large deformation and dynamics of every blood cell with high fidelity. The present methodology seamlessly integrates different modeling components dealing with stationary rigid boundaries of complex shape, moving rigid bodies, and highly deformable interfaces governed by nonlinear elasticity. Thus it enables us to simulate 'whole' blood suspensions flowing through physiologically realistic microvascular networks that are characterized by multiple bifurcating and merging vessels, as well as geometrically complex lab-on-chip devices. The focus of the present work is on the development of a versatile numerical technique that is able to consider deformable cells and rigid bodies flowing in three-dimensional arbitrarily complex geometries over a diverse range of scenarios. After describing the methodology, a series of validation studies are presented against analytical theory, experimental data, and previous numerical results. Then, the capability of the methodology is demonstrated by simulating flows of deformable blood cells and heterogeneous cell suspensions in both physiologically realistic microvascular networks and geometrically intricate microfluidic devices. It is shown that the methodology can predict several complex microhemodynamic phenomena observed in vascular networks and microfluidic devices. The present methodology is robust and versatile, and has the potential to scale up to very large microvascular networks at organ levels.

Measurements of plume geometry and argon-41 radiation field at the BR1 reactor in Mol, Belgium

International Nuclear Information System (INIS)

Drews, M.; Joergensen, H.; Lauritzen, Bent; Mikkelsen, T.; Aage, H.K.; Korsbech, U.; Bargholz, K.; Rojas-Palma, C.; Ammel, R. van

2002-02-01

An atmospheric dispersion experiment was conducted using a visible tracer along with the routine releases of 41 Ar from the BR1 air-cooled research reactor in Mol. In the experiment, simultaneous measurements of the radiation field from the 41 Ar decay, the meteorology, the 41 Ar source term and plume geometry were performed. The visible tracer was injected into the reactor emission stack, and the plume cross section determined by Lidar scanning of the released aerosols. The data collected in the exercise provide a valuable resource for atmospheric dispersion and dose rate modeling. (au)

Probability-neighbor method of accelerating geometry treatment in reactor Monte Carlo code RMC

International Nuclear Information System (INIS)

She, Ding; Li, Zeguang; Xu, Qi; Wang, Kan; Yu, Ganglin

2011-01-01

Probability neighbor method (PNM) is proposed in this paper to accelerate geometry treatment of Monte Carlo (MC) simulation and validated in self-developed reactor Monte Carlo code RMC. During MC simulation by either ray-tracking or delta-tracking method, large amounts of time are spent in finding out which cell one particle is located in. The traditional way is to search cells one by one with certain sequence defined previously. However, this procedure becomes very time-consuming when the system contains a large number of cells. Considering that particles have different probability to enter different cells, PNM method optimizes the searching sequence, i.e., the cells with larger probability are searched preferentially. The PNM method is implemented in RMC code and the numerical results show that the considerable time of geometry treatment in MC calculation for complicated systems is saved, especially effective in delta-tracking simulation. (author)

Phased Array Imaging of Complex-Geometry Composite Components.

Science.gov (United States)

Brath, Alex J; Simonetti, Francesco

2017-10-01

Progress in computational fluid dynamics and the availability of new composite materials are driving major advances in the design of aerospace engine components which now have highly complex geometries optimized to maximize system performance. However, shape complexity poses significant challenges to traditional nondestructive evaluation methods whose sensitivity and selectivity rapidly decrease as surface curvature increases. In addition, new aerospace materials typically exhibit an intricate microstructure that further complicates the inspection. In this context, an attractive solution is offered by combining ultrasonic phased array (PA) technology with immersion testing. Here, the water column formed between the complex surface of the component and the flat face of a linear or matrix array probe ensures ideal acoustic coupling between the array and the component as the probe is continuously scanned to form a volumetric rendering of the part. While the immersion configuration is desirable for practical testing, the interpretation of the measured ultrasonic signals for image formation is complicated by reflection and refraction effects that occur at the water-component interface. To account for refraction, the geometry of the interface must first be reconstructed from the reflected signals and subsequently used to compute suitable delay laws to focus inside the component. These calculations are based on ray theory and can be computationally intensive. Moreover, strong reflections from the interface can lead to a thick dead zone beneath the surface of the component which limits sensitivity to shallow subsurface defects. This paper presents a general approach that combines advanced computing for rapid ray tracing in anisotropic media with a 256-channel parallel array architecture. The full-volume inspection of complex-shape components is enabled through the combination of both reflected and transmitted signals through the part using a pair of arrays held in a yoke

Six-Coordinate Ln(III Complexes with Various Coordination Geometries Showing Distinct Magnetic Properties

Directory of Open Access Journals (Sweden)

Mei Guo

2018-01-01

Full Text Available The syntheses, structural characterization, and magnetic properties of three lanthanide complexes with formulas [Ln(L13] (Ln = Dy (1Dy; Er (1Er; and [Dy(L22] (2Dy were reported. Complexes 1Dy and 1Er are isostructural with the metal ion in distorted trigonal-prismatic coordination geometry, but exhibit distinct magnetic properties due to the different shapes of electron density for DyIII (oblate and ErIII (prolate ions. Complex 1Dy shows obvious SMM behavior under a zero direct current (dc field with an effective energy barrier of 31.4 K, while complex 1Er only features SMM behavior under a 400 Oe external field with an effective energy barrier of 23.96 K. In stark contrast, complex 2Dy with the octahedral geometry only exhibits the frequency dependence of alternating current (ac susceptibility signals without χ″ peaks under a zero dc field.

Neutronic study of a nuclear reactor of fused salts; Estudio neutronico de un reactor nuclear de sales fundidas

Energy Technology Data Exchange (ETDEWEB)

Garcia B, F. B.; Francois L, J. L., E-mail: faviolabelen@gmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

2012-10-15

The reactors of fused salts called Molten Salt Reactor have presented a resurgence of interest in the last decade, due to they have a versatility in particular to operate, either with a thermal or fast neutrons spectrum. The most active development was by the middle of 1950 and principles of 1970 in the Oak Ridge National Laboratory. In this work some developed models are presented particularly and studied with the help of the MCNPX code, for the development of the neutronic study of this reactor, starting of proposed models and from a simple and homogeneous geometry until other more complex models and approximate to more real cases. In particular the geometry conditions and criticality of each model were analyzed, the isotopic balance, as well as the concentrations of the salts and different assigned fuel types. (Author)

Solar Proton Transport within an ICRU Sphere Surrounded by a Complex Shield: Combinatorial Geometry

Science.gov (United States)

Wilson, John W.; Slaba, Tony C.; Badavi, Francis F.; Reddell, Brandon D.; Bahadori, Amir A.

2015-01-01

The 3DHZETRN code, with improved neutron and light ion (Z (is) less than 2) transport procedures, was recently developed and compared to Monte Carlo (MC) simulations using simplified spherical geometries. It was shown that 3DHZETRN agrees with the MC codes to the extent they agree with each other. In the present report, the 3DHZETRN code is extended to enable analysis in general combinatorial geometry. A more complex shielding structure with internal parts surrounding a tissue sphere is considered and compared against MC simulations. It is shown that even in the more complex geometry, 3DHZETRN agrees well with the MC codes and maintains a high degree of computational efficiency.

Surface complexation of selenite on goethite: MO/DFT geometry and charge distribution

NARCIS (Netherlands)

Hiemstra, T.; Rietra, R.P.J.J.; Riemsdijk, van W.H.

2007-01-01

The adsorption of selenite on goethite (alpha-FeOOH) has been analyzed with the charge distribution (CD) and the multi-site surface complexation (MUSIC) model being combined with an extended Stem (ES) layer model option. The geometry of a set of different types of hydrated iron-selenite complexes

Complex Structure of the Four-Dimensional Kerr Geometry: Stringy System, Kerr Theorem, and Calabi-Yau Twofold

Directory of Open Access Journals (Sweden)

Alexander Burinskii

2013-01-01

Full Text Available The 4D Kerr geometry displays many wonderful relations with quantum world and, in particular, with superstring theory. The lightlike structure of fields near the Kerr singular ring is similar to the structure of Sen solution for a closed heterotic string. Another string, open and complex, appears in the complex representation of the Kerr geometry initiated by Newman. Combination of these strings forms a membrane source of the Kerr geometry which is parallel to the structure of M-theory. In this paper we give one more evidence of this relationship, emergence of the Calabi-Yau twofold (K3 surface in twistorial structure of the Kerr geometry as a consequence of the Kerr theorem. Finally, we indicate that the Kerr stringy system may correspond to a complex embedding of the critical N = 2 superstring.

Algebrodynamics over complex space and phase extension of the Minkowski geometry

International Nuclear Information System (INIS)

Kassandrov, V. V.

2009-01-01

First principles should predetermine physical geometry and dynamics both together. In the 'algebrodynamics' they follow solely from the properties of biquaternion algebra B and the analysis over B. We briefly present the algebrodynamics over Minkowski background based on a nonlinear generalization to B of the Cauchi-Riemann analyticity conditions. Further, we consider the effective real geometry uniquely resulting from the structure of B multiplication and found it to be of the Minkowski type, with an additional phase invariant. Then we pass to study the primordial dynamics that takes place in the complex B space and brings into consideration a number of remarkable structures: an ensemble of identical correlated matter pre-elements ('duplicons'), caustic-like signals (interaction carriers), a concept of random complex time resulting in irreversibility of physical time at macrolevel, etc. In partucular, the concept of 'dimerous electron' naturally arises in the framework of complex algebrodynamics and, together with the above-mentioned phase invariant, allows for a novel approach to explanation of quantum interference phenomena alternative to recently accepted wave-particle dualism paradigm.

Identification of different coordination geometries by XAFS in copper(II) complexes with trimesic acid

Science.gov (United States)

Gaur, A.; Klysubun, W.; Soni, Balram; Shrivastava, B. D.; Prasad, J.; Srivastava, K.

2016-10-01

X-ray absorption spectroscopy (XAS) is very useful in revealing the information about geometric and electronic structure of a transition-metal absorber and thus commonly used for determination of metal-ligand coordination. But XAFS analysis becomes difficult if differently coordinated metal centers are present in a system. In the present investigation, existence of distinct coordination geometries around metal centres have been studied by XAFS in a series of trimesic acid Cu(II) complexes. The complexes studied are: Cu3(tma)2(im)6 8H2O (1), Cu3(tma)2(mim)6 17H2O (2), Cu3(tma)2(tmen)3 8.5H2O (3), Cu3(tma) (pmd)3 6H2O (ClO4)3 (4) and Cu3(tma)2 3H2O (5). These complexes have not only Cu metal centres with different coordination but in complexes 1-3, there are multiple coordination geometries present around Cu centres. Using XANES spectra, different coordination geometries present in these complexes have been identified. The variation observed in the pre-edge features and edge features have been correlated with the distortion of the specific coordination environment around Cu centres in the complexes. XANES spectra have been calculated for the distinct metal centres present in the complexes by employing ab-initio calculations. These individual spectra have been used to resolve the spectral contribution of the Cu centres to the particular XANES features exhibited by the experimental spectra of the multinuclear complexes. Also, the variation in the 4p density of states have been calculated for the different Cu centres and then correlated with the features originated from corresponding coordination of Cu. Thus, these spectral features have been successfully utilized to detect the presence of the discrete metal centres in a system. The inferences about the coordination geometry have been supported by EXAFS analysis which has been used to determine the structural parameters for these complexes.

Two-dimensional DORT discrete ordinates X-Y geometry neutron flux calculations for the Halden Heavy Boiling Water Reactor core configurations

Energy Technology Data Exchange (ETDEWEB)

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.

Method and program for complex calculation of heterogeneous reactor

International Nuclear Information System (INIS)

Kalashnikov, A.G.; Glebov, A.P.; Elovskaya, L.F.; Kuznetsova, L.I.

1988-01-01

An algorithm and the GITA program for complex one-dimensional calculation of a heterogeneous reactor which permits to conduct calculations for the reactor and its cell simultaneously using the same algorithm are described. Multigroup macrocross sections for reactor zones in the thermal energy range are determined according to the technique for calculating a cell with complicate structure and then the continuous multi group calculation of the reactor in the thermal energy range and in the range of neutron thermalization is made. The kinetic equation is solved using the Pi- and DSn- approximations [fr

Steady-state and transient heat transfer through fins of complex geometry

Directory of Open Access Journals (Sweden)

Taler Dawid

2014-06-01

Full Text Available Various methods for steady-state and transient analysis of temperature distribution and efficiency of continuous-plate fins are presented. For a constant heat transfer coefficient over the fin surface, the plate fin can be divided into imaginary rectangular or hexangular fins. At first approximate methods for determining the steady-state fin efficiency like the method of equivalent circular fin and the sector method are discussed. When the fin geometry is complex, thus transient temperature distribution and fin efficiency can be determined using numerical methods. A numerical method for transient analysis of fins with complex geometry is developed. Transient temperature distributions in continuous fins attached to oval tubes is computed using the finite volume - finite element methods. The developed method can be used in the transient analysis of compact heat exchangers to calculate correctly the heat flow rate transferred from the finned tubes to the fluid.

Solar proton exposure of an ICRU sphere within a complex structure Part I: Combinatorial geometry.

Science.gov (United States)

Wilson, John W; Slaba, Tony C; Badavi, Francis F; Reddell, Brandon D; Bahadori, Amir A

2016-06-01

The 3DHZETRN code, with improved neutron and light ion (Z≤2) transport procedures, was recently developed and compared to Monte Carlo (MC) simulations using simplified spherical geometries. It was shown that 3DHZETRN agrees with the MC codes to the extent they agree with each other. In the present report, the 3DHZETRN code is extended to enable analysis in general combinatorial geometry. A more complex shielding structure with internal parts surrounding a tissue sphere is considered and compared against MC simulations. It is shown that even in the more complex geometry, 3DHZETRN agrees well with the MC codes and maintains a high degree of computational efficiency. Published by Elsevier Ltd.

A numerical calculation method for flow discretisation in complex geometry with body-fitted grids

International Nuclear Information System (INIS)

Jin, X.

2001-04-01

A numerical calculation method basing on body fitted grids is developed in this work for computational fluid dynamics in complex geometry. The method solves the conservation equations in a general nonorthogonal coordinate system which matches the curvilinear boundary. The nonorthogonal, patched grid is generated by a grid generator which solves algebraic equations. By means of an interface its geometrical data can be used by this method. The conservation equations are transformed from the Cartesian system to a general curvilinear system keeping the physical Cartesian velocity components as dependent variables. Using a staggered arrangement of variables, the three Cartesian velocity components are defined on every cell surface. Thus the coupling between pressure and velocity is ensured, and numerical oscillations are avoided. The contravariant velocity for calculating mass flux on one cell surface is resulting from dependent Cartesian velocity components. After the discretisation and linear interpolation, a three dimensional 19-point pressure equation is found. Using the explicit treatment for cross-derivative terms, it reduces to the usual 7-point equation. Under the same data and process structure, this method is compatible with the code FLUTAN using Cartesian coordinates. In order to verify this method, several laminar flows are simulated in orthogonal grids at tilted space directions and in nonorthogonal grids with variations of cell angles. The simulated flow types are considered like various duct flows, transient heat conduction, natural convection in a chimney and natural convection in cavities. Their results achieve very good agreement with analytical solutions or empirical data. Convergence for highly nonorthogonal grids is obtained. After the successful validation of this method, it is applied for a reactor safety case. A transient natural convection flow for an optional sump cooling concept SUCO is simulated. The numerical result is comparable with the

Advances in Spectral Nodal Methods applied to SN Nuclear Reactor Global calculations in Cartesian Geometry

International Nuclear Information System (INIS)

Barros, R.C.; Filho, H.A.; Oliveira, F.B.S.; Silva, F.C. da

2004-01-01

Presented here are the advances in spectral nodal methods for discrete ordinates (SN) eigenvalue problems in Cartesian geometry. These coarse-mesh methods are based on three ingredients: (i) the use of the standard discretized spatial balance SN equations; (ii) the use of the non-standard spectral diamond (SD) auxiliary equations in the multiplying regions of the domain, e.g. fuel assemblies; and (iii) the use of the non-standard spectral Green's function (SGF) auxiliary equations in the non-multiplying regions of the domain, e.g., the reflector. In slab-geometry the hybrid SD-SGF method generates numerical results that are completely free of spatial truncation errors. In X,Y-geometry, we obtain a system of two 'slab-geometry' SN equations for the node-edge average angular fluxes by transverse-integrating the X,Y-geometry SN equations separately in the y- and then in the x-directions within an arbitrary node of the spatial grid set up on the domain. In this paper, we approximate the transverse leakage terms by constants. These are the only approximations considered in the SD-SGF-constant nodal method, as the source terms, that include scattering and eventually fission events, are treated exactly. Moreover, we describe in this paper the progress of the approximate SN albedo boundary conditions for substituting the non-multiplying regions around the nuclear reactor core. We show numerical results to typical model problems to illustrate the accuracy of spectral nodal methods for coarse-mesh SN criticality calculations. (Author)

Simulation of biological flow and transport in complex geometries using embedded boundary/volume-of-fluid methods

International Nuclear Information System (INIS)

Trebotich, David

2007-01-01

We have developed a simulation capability to model multiscale flow and transport in complex biological systems based on algorithms and software infrastructure developed under the SciDAC APDEC CET. The foundation of this work is a new hybrid fluid-particle method for modeling polymer fluids in irregular microscale geometries that enables long-time simulation of validation experiments. Both continuum viscoelastic and discrete particle representations have been used to model the constitutive behavior of polymer fluids. Complex flow environment geometries are represented on Cartesian grids using an implicit function. Direct simulation of flow in the irregular geometry is then possible using embedded boundary/volume-of-fluid methods without loss of geometric detail. This capability has been used to simulate biological flows in a variety of application geometries including biomedical microdevices, anatomical structures and porous media

Innovation Study for Laser Cutting of Complex Geometries with Paper Materials

Science.gov (United States)

Happonen, A.; Stepanov, A.; Piili, H.; Salminen, A.

Even though technology for laser cutting of paper materials has existed for over 30 years, it seems that results of applications of this technology and possibilities of laser cutting systems are not easily available. The aim of this study was to analyze the feasibility of the complex geometry laser cutting of paper materials and to analyze the innovation challenges and potential of current laser cutting technologies offer. This research studied the potential and possible challenges in applying CO2 laser cutting technology for cutting of paper materials in current supply chains trying to fulfil the changing needs of customer in respect of shape, fast response during rapid delivery cycle. The study is focused on examining and analyzing the different possibilities of laser cutting of paper material in application area of complex low volume geometry cutting. The goal of this case was to analyze the feasibility of the laser cutting from technical, quality and implementation points of view and to discuss availability of new business opportunities. It was noticed that there are new business models still available within laser technology applications in complex geometry cutting. Application of laser technology, in business-to-consume markets, in synergy with Internet service platforms can widen the customer base and offer new value streams for technology and service companies. Because of this, existing markets and competition has to be identified, and appropriate new and innovative business model needs to be developed. And to be competitive in the markets, models like these need to include the earning logic and the stages from production to delivery as discussed in the paper.

The water desalination complex based on ABV-type reactor plant

International Nuclear Information System (INIS)

Panov, Yu.K.; Fadeev, Yu.P.; Vorobiev, V.M.; Baranaev, Yu.D.

1997-01-01

A floating nuclear desalination complex with two barges, one for ABV type reactor plant, with twin reactor 2 x 6 MW(e), and one for reverse osmosis desalination plant, was described. The principal specifications of the ABV type reactor plant and desalination barge were given. The ABV type reactor has a traditional two-circuit layout using an integral type reactor vessel with all mode natural convection of primary coolant. The desalted water cost was estimated to be around US $0.86 per cubic meter. R and D work has been performed and preparations for commercial production are under way. (author)

Action Memorandum for Decommissioning the Engineering Test Reactor Complex under the Idaho Cleanup Project

International Nuclear Information System (INIS)

A. B. Culp

2007-01-01

This Action Memorandum documents the selected alternative for decommissioning of the Engineering Test Reactor at the Idaho National Laboratory under the Idaho Cleanup Project. Since the missions of the Engineering Test Reactor Complex have been completed, an engineering evaluation/cost analysis that evaluated alternatives to accomplish the decommissioning of the Engineering Test Reactor Complex was prepared and released for public comment. The scope of this Action Memorandum is to encompass the final end state of the Complex and disposal of the Engineering Test Reactor vessel. The selected removal action includes removing and disposing of the vessel at the Idaho CERCLA Disposal Facility and demolishing the reactor building to ground surface

Neutronic study of a nuclear reactor of fused salts

International Nuclear Information System (INIS)

Garcia B, F. B.; Francois L, J. L.

2012-10-01

The reactors of fused salts called Molten Salt Reactor have presented a resurgence of interest in the last decade, due to they have a versatility in particular to operate, either with a thermal or fast neutrons spectrum. The most active development was by the middle of 1950 and principles of 1970 in the Oak Ridge National Laboratory. In this work some developed models are presented particularly and studied with the help of the MCNPX code, for the development of the neutronic study of this reactor, starting of proposed models and from a simple and homogeneous geometry until other more complex models and approximate to more real cases. In particular the geometry conditions and criticality of each model were analyzed, the isotopic balance, as well as the concentrations of the salts and different assigned fuel types. (Author)

Inspection of complex geometry pieces with an intelligent contact transducer

International Nuclear Information System (INIS)

Chatillon, S.; Roy, O.; Mahaut, St.

2000-01-01

A new multi-element contact transducer has been developed to improve the inspection of components with complex geometry. The emitting surface is flexible in order to optimize the contact with pieces. An algorithm, based on a simplified geometric model, has been used to determine the delays law which allows to control the focal characteristics of the transmitted field. Acquisition data lead in transmission with an articulated transducer validate the behavior provided by simulation. Thus the optimization of the delays law ensures the transmission of a beam which is homogeneous and controlled during the moving of the transducer. Inspections in echo-pulse mode are implemented on a sample simulating a component controlled on site. Results show that the dynamical adaptation of the delays law to the geometry of the piece leads to very good performances

The effect of reactor geometry on the synthesis of graphene materials in plasma jets

Science.gov (United States)

Shavelkina, M. B.; Amirov, R. H.; Shatalova, T. B.

2017-05-01

The possibility of synthesis of graphene and graphane (hydrogenated graphene) using the decomposition of hydrocarbons by thermal plasma has been investigated. Investigations of the influence of the plasma-forming gas on the efficiency of synthesis and the morphology of graphene materials were carried out. The synthesis products have been characterized by the methods of scanning microscopy, Raman spectroscopy and thermal analysis. It is found that the morphology of graphene materials is affected by the geometry of the reactor. It was demonstrated that the obtained graphene materials are uniformly distributed in the volume of plastic based on cyanate ester resins under mixing.

Comparison of low enriched uranium (UAlx-Al and U-Ni) targets with different geometries for the production of molybdenum-99 in the RMB (Brazilian multipurpose reactor)

International Nuclear Information System (INIS)

Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Silva, Jose Eduardo R. da; Angelo, Gabriel; Fedorenko, Giuliana G.; Nishiyama, Pedro J.B. de O.

2011-01-01

The Brazilian Multipurpose Reactor (RMB), now in the conception design phase, is being designed in Brazil to attend the demand of radiopharmaceuticals in the country and conduct researches in various areas. The new reactor, planned for 30 MW, will replace the IEA-R1 reactor of IPEN-CNEN/SP. Low enriched uranium ( 235 U) UAl x dispersed in Al (plate geometry) and metallic uranium foil targets (plate and cylinder geometries) are being considered for production of Molybdenum-99 ( 99 Mo) by fission. Neutronic and thermal-hydraulics calculations were performed to compare the production of 99 Mo for these targets in the RMB. For the neutronic calculations were utilized the computer codes Hammer-Technion, Citation and Scale and for the thermal-hydraulics calculations were utilized the computer code MTRCR-IEAR1 and ANSYS CFX. (author)

Structural design of SBWR reactor building complex using microcomputers

International Nuclear Information System (INIS)

Mandagi, K.; Rajagopal, R.S.; Sawhney, P.S.; Gou, P.F.

1993-01-01

The design concept of Simplified Boiling Water Reactor (SBWR) plant is based on simplicity and passive features to enhance safety and reliability, improve performance, and increase economic viability. The SBWR utilizes passive systems such as Gravity Driven Core-Cooling System (GDCS) and Passive Containment Cooling System (PCCS). To suit these design features the Reactor Building (RB) complex of the SBWR is configured as an integrated structure consisting of a cylindrical Reinforced Concrete Containment Vessel (RCCV) surrounded by square reinforced concrete safety envelope and outer box structures, all sharing a common reinforced concrete basemat. This paper describes the structural analysis and design aspects of the RB complex. A 3D STARDYNE finite element model has been developed for the structural analysis of the complex using a PC Compaq 486/33L microcomputer. The structural analysis is performed for service and factored load conditions for the applicable loading combinations. The dynamic responses of containment structures due to pool hydrodynamic loads have been calculated by an axisymmetric shell model using COSMOS/M program. The RCCV is designed in accordance with ASME Section 3, Division 2 Code. The rest of the RB which is classified as Seismic Category 1 structure is designed in accordance with the ACI 349 Code. This paper shows that microcomputers can be efficiently used for the analysis and design of large and complex structures such as RCCV and Reactor Building complex. The use of microcomputers can result in significant savings in the computational cost compared with that of mainframe computers

The new electricity of France PWR: calculation scheme of neutron leakages from the reactor cavity

International Nuclear Information System (INIS)

Vergnaud, T.; Bourdet, L.; Nimal, J.C.; Brandicourt, G.; Champion, G.

1987-04-01

A new calculation scheme is adapted to evaluate neutron fluxes in the reactor cavity and the containment of next french PWR. In this scheme a large part is given to Monte Carlo method, coupled with SN-method, in order to take into account multiple neutron diffusions and the complexity of the reactor geometry

MURLI, 1-D Flux, Reaction Rate in Cylindrical Geometry Thermal Reactor Lattice by Transport

International Nuclear Information System (INIS)

Huria, H.C.

1985-01-01

1 - Description of problem or function: MURLI is an integral transport theory code to calculate fluxes and reaction rates in one- dimensional cylindrical geometry lattice cells of a thermal reactor. For a specified buckling, it computes k-effective using few-group diffusion theory and a few-group collapsed set of Cross sections. The code can optionally be used to solve a first order differential equation for the number density of fissile, fertile and fission product nuclei as a function of time, and to recalculate fluxes, reaction rates and k-effective at different stages of burnup. A 27-group cross section data library is included. There are four pseudo-fission products each associated with the decay chains of plutonium and uranium isotopes in addition to Rh-105, Xe-135, Np-239, U-236, Am-241, Am-242 and Am-243. There is also data for one lumped pseudo-fission product. 2 - Method of solution: Multiple collision probabilities and escape probabilities are calculated for each cylindrical shell region assuming protons are born uniformly and isotropically over the entire region volume. The equations of integral transport theory can then be solved for neutron flux. The first order differential burnup equation is solved by a fourth order Runge-Kutta method. 3 - Restrictions on the complexity of the problem: There are maxima of 8 fissionable elements, 8 resonant elements, and 20 spatial regions

Fluid flow and heat transfer investigation of pebble bed reactors using mesh-adaptive LES

International Nuclear Information System (INIS)

Pavlidis, Dimitrios; Lathouwers, Danny

2013-01-01

The very high temperature reactor is one of the designs currently being considered for nuclear power generation. One its variants is the pebble bed reactor in which the coolant passes through complex geometries (pores) at high Reynolds numbers. A computational fluid dynamics model with anisotropic mesh adaptivity is used to investigate coolant flow and heat transfer in such reactors. A novel method for implicitly incorporating solid boundaries based on multi-fluid flow modelling is adopted. The resulting model is able to resolve and simulate flow and heat transfer in randomly packed beds, regardless of the actual geometry, starting off with arbitrarily coarse meshes. The model is initially evaluated using an orderly stacked square channel of channel-height-to-particle diameter ratio of unity for a range of Reynolds numbers. The model is then applied to the face-centred cubical geometry. coolant flow and heat transfer patterns are investigated

Calibration of Ge gamma-ray spectrometers for complex sample geometries and matrices

Energy Technology Data Exchange (ETDEWEB)

Semkow, T.M., E-mail: thomas.semkow@health.ny.gov [Wadsworth Center, New York State Department of Health, Empire State Plaza, Albany, NY 12201 (United States); Department of Environmental Health Sciences, School of Public Health, University at Albany, State University of New York, Rensselaer, NY 12144 (United States); Bradt, C.J.; Beach, S.E.; Haines, D.K.; Khan, A.J.; Bari, A.; Torres, M.A.; Marrantino, J.C.; Syed, U.-F. [Wadsworth Center, New York State Department of Health, Empire State Plaza, Albany, NY 12201 (United States); Kitto, M.E. [Wadsworth Center, New York State Department of Health, Empire State Plaza, Albany, NY 12201 (United States); Department of Environmental Health Sciences, School of Public Health, University at Albany, State University of New York, Rensselaer, NY 12144 (United States); Hoffman, T.J. [Wadsworth Center, New York State Department of Health, Empire State Plaza, Albany, NY 12201 (United States); Curtis, P. [Kiltel Systems, Inc., Clyde Hill, WA 98004 (United States)

2015-11-01

A comprehensive study of the efficiency calibration and calibration verification of Ge gamma-ray spectrometers was performed using semi-empirical, computational Monte-Carlo (MC), and transfer methods. The aim of this study was to evaluate the accuracy of the quantification of gamma-emitting radionuclides in complex matrices normally encountered in environmental and food samples. A wide range of gamma energies from 59.5 to 1836.0 keV and geometries from a 10-mL jar to 1.4-L Marinelli beaker were studied on four Ge spectrometers with the relative efficiencies between 102% and 140%. Density and coincidence summing corrections were applied. Innovative techniques were developed for the preparation of artificial complex matrices from materials such as acidified water, polystyrene, ethanol, sugar, and sand, resulting in the densities ranging from 0.3655 to 2.164 g cm{sup −3}. They were spiked with gamma activity traceable to international standards and used for calibration verifications. A quantitative method of tuning MC calculations to experiment was developed based on a multidimensional chi-square paraboloid. - Highlights: • Preparation and spiking of traceable complex matrices in extended geometries. • Calibration of Ge gamma spectrometers for complex matrices. • Verification of gamma calibrations. • Comparison of semi-empirical, computational Monte Carlo, and transfer methods of Ge calibration. • Tuning of Monte Carlo calculations using a multidimensional paraboloid.

Calibration of Ge gamma-ray spectrometers for complex sample geometries and matrices

International Nuclear Information System (INIS)

Semkow, T.M.; Bradt, C.J.; Beach, S.E.; Haines, D.K.; Khan, A.J.; Bari, A.; Torres, M.A.; Marrantino, J.C.; Syed, U.-F.; Kitto, M.E.; Hoffman, T.J.; Curtis, P.

2015-01-01

A comprehensive study of the efficiency calibration and calibration verification of Ge gamma-ray spectrometers was performed using semi-empirical, computational Monte-Carlo (MC), and transfer methods. The aim of this study was to evaluate the accuracy of the quantification of gamma-emitting radionuclides in complex matrices normally encountered in environmental and food samples. A wide range of gamma energies from 59.5 to 1836.0 keV and geometries from a 10-mL jar to 1.4-L Marinelli beaker were studied on four Ge spectrometers with the relative efficiencies between 102% and 140%. Density and coincidence summing corrections were applied. Innovative techniques were developed for the preparation of artificial complex matrices from materials such as acidified water, polystyrene, ethanol, sugar, and sand, resulting in the densities ranging from 0.3655 to 2.164 g cm −3 . They were spiked with gamma activity traceable to international standards and used for calibration verifications. A quantitative method of tuning MC calculations to experiment was developed based on a multidimensional chi-square paraboloid. - Highlights: • Preparation and spiking of traceable complex matrices in extended geometries. • Calibration of Ge gamma spectrometers for complex matrices. • Verification of gamma calibrations. • Comparison of semi-empirical, computational Monte Carlo, and transfer methods of Ge calibration. • Tuning of Monte Carlo calculations using a multidimensional paraboloid

URDME: a modular framework for stochastic simulation of reaction-transport processes in complex geometries.

Science.gov (United States)

Drawert, Brian; Engblom, Stefan; Hellander, Andreas

2012-06-22

Experiments in silico using stochastic reaction-diffusion models have emerged as an important tool in molecular systems biology. Designing computational software for such applications poses several challenges. Firstly, realistic lattice-based modeling for biological applications requires a consistent way of handling complex geometries, including curved inner- and outer boundaries. Secondly, spatiotemporal stochastic simulations are computationally expensive due to the fast time scales of individual reaction- and diffusion events when compared to the biological phenomena of actual interest. We therefore argue that simulation software needs to be both computationally efficient, employing sophisticated algorithms, yet in the same time flexible in order to meet present and future needs of increasingly complex biological modeling. We have developed URDME, a flexible software framework for general stochastic reaction-transport modeling and simulation. URDME uses Unstructured triangular and tetrahedral meshes to resolve general geometries, and relies on the Reaction-Diffusion Master Equation formalism to model the processes under study. An interface to a mature geometry and mesh handling external software (Comsol Multiphysics) provides for a stable and interactive environment for model construction. The core simulation routines are logically separated from the model building interface and written in a low-level language for computational efficiency. The connection to the geometry handling software is realized via a Matlab interface which facilitates script computing, data management, and post-processing. For practitioners, the software therefore behaves much as an interactive Matlab toolbox. At the same time, it is possible to modify and extend URDME with newly developed simulation routines. Since the overall design effectively hides the complexity of managing the geometry and meshes, this means that newly developed methods may be tested in a realistic setting already at

Modelling and simulation of gas explosions in complex geometries

Energy Technology Data Exchange (ETDEWEB)

Saeter, Olav

1998-12-31

This thesis presents a three-dimensional Computational Fluid Dynamics (CFD) code (EXSIM94) for modelling and simulation of gas explosions in complex geometries. It gives the theory and validates the following sub-models : (1) the flow resistance and turbulence generation model for densely packed regions, (2) the flow resistance and turbulence generation model for single objects, and (3) the quasi-laminar combustion model. It is found that a simple model for flow resistance and turbulence generation in densely packed beds is able to reproduce the medium and large scale MERGE explosion experiments of the Commission of European Communities (CEC) within a band of factor 2. The model for a single representation is found to predict explosion pressure in better agreement with the experiments with a modified k-{epsilon} model. This modification also gives a slightly improved grid independence for realistic gas explosion approaches. One laminar model is found unsuitable for gas explosion modelling because of strong grid dependence. Another laminar model is found to be relatively grid independent and to work well in harmony with the turbulent combustion model. The code is validated against 40 realistic gas explosion experiments. It is relatively grid independent in predicting explosion pressure in different offshore geometries. It can predict the influence of ignition point location, vent arrangements, different geometries, scaling effects and gas reactivity. The validation study concludes with statistical and uncertainty analyses of the code performance. 98 refs., 96 figs, 12 tabs.

XAFS study of copper(II) complexes with square planar and square pyramidal coordination geometries

Science.gov (United States)

Gaur, A.; Klysubun, W.; Nitin Nair, N.; Shrivastava, B. D.; Prasad, J.; Srivastava, K.

2016-08-01

X-ray absorption fine structure of six Cu(II) complexes, Cu2(Clna)4 2H2O (1), Cu2(ac)4 2H2O (2), Cu2(phac)4 (pyz) (3), Cu2(bpy)2(na)2 H2O (ClO4) (4), Cu2(teen)4(OH)2(ClO4)2 (5) and Cu2(tmen)4(OH)2(ClO4)2 (6) (where ac, phac, pyz, bpy, na, teen, tmen = acetate, phenyl acetate, pyrazole, bipyridine, nicotinic acid, tetraethyethylenediamine, tetramethylethylenediamine, respectively), which were supposed to have square pyramidal and square planar coordination geometries have been investigated. The differences observed in the X-ray absorption near edge structure (XANES) features of the standard compounds having four, five and six coordination geometry points towards presence of square planar and square pyramidal geometry around Cu centre in the studied complexes. The presence of intense pre-edge feature in the spectra of four complexes, 1-4, indicates square pyramidal coordination. Another important XANES feature, present in complexes 5 and 6, is prominent shoulder in the rising part of edge whose intensity decreases in the presence of axial ligands and thus indicates four coordination in these complexes. Ab initio calculations were carried out for square planar and square pyramidal Cu centres to observe the variation of 4p density of states in the presence and absence of axial ligands. To determine the number and distance of scattering atoms around Cu centre in the complexes, EXAFS analysis has been done using the paths obtained from Cu(II) oxide model and an axial Cu-O path from model of a square pyramidal complex. The results obtained from EXAFS analysis have been reported which confirmed the inference drawn from XANES features. Thus, it has been shown that these paths from model of a standard compound can be used to determine the structural parameters for complexes having unknown structure.

Blast venting through blanket material in the HYLIFE ICF reactor

International Nuclear Information System (INIS)

Liu, J.C.; Peterson, P.F.; Schrock, V.E.

1992-01-01

This work presents a numerical study of blast venting through various blanket configurations in the HYLIFE ICF reactor design. The study uses TSUNAMI -- a multi-dimensional, high-resolution, shock capturing code -- to predict the momentum exchange and gas dynamics for blast venting in complex geometries. In addition, the study presents conservative predictions of wall loading by gas shock and impulse delivered to the protective liquid blanket. Configurations used in the study include both 2700 MJ and 350 MJ fusion yields per pulse for 5 meter and 3 meter radius reactor chambers. For the former, an annular jet array is used for the blanket geometry, while in the latter, both annular jet array as well as slab geometries are used. Results of the study indicate that blast venting and wall loading may be manageable in the HYLIFE-II design by a judicious choice of blanket configuration

Fuel management optimization in pressure water reactors with hexagonal geometry using hill climbing method

International Nuclear Information System (INIS)

Andres Diaz, J.; Quintero, Ruben; Melian, Manuel; Rosete, Alejandro

2000-01-01

In this work the general-purpose optimization method, Hill Climbing, was applied to the Fuel Management Optimization problem in PWR reactors, WWER type. They were carried out a series of experiments in order to study the performance of Hill Climbing. It was proven two starting point for initialize the search: a reload configuration by project and a reload configuration generated with the application of a minimal knowledge of the problem. It was also studied the effect of imposing constraints based on the physics of the reactor in order to reduce the number of possible solutions to be generated. The operator used in Hill Climbing was defined as a binary exchange of fuel assemblies. For the simulation of each generated configuration, the tridimensional simulator program SPPS-1 was used. It was formulated an objective function with power peaking constraint to guide the search. As results, a methodology ws proposed for the In-core Fuel Management Optimization in hexagonal geometry, and the feasibility of the application of the Hill Climbing to this type of problem was demonstrated. (author)

MODEL SIMULATION OF GEOMETRY AND STRESS-STRAIN VARIATION OF BATAN FUEL PIN PROTOTYPE DURING IRRADIATION TEST IN RSG-GAS REACTOR

Directory of Open Access Journals (Sweden)

Suwardi Suwardi

2015-03-01

Full Text Available MODEL SIMULATION OF GEOMETRY AND STRESS-STRAIN VARIATION OF BATAN FUEL PIN PROTOTYPE DURING IRRADIATION TEST IN RSG-GAS REACTOR*. The first short fuel pin containing natural UO2 pellet in Zry4 cladding has been prepared at the CNFT (Center for Nuclear Fuel Technology then a ramp test will be performed. The present work is part of designing first irradiation experiments in the PRTF (Power Ramp Test Facility of RSG-GAS 30 MW reactor. The thermal mechanic of the pin during irradiation has simulated. The geometry variation of pellet and cladding is modeled by taking into account different phenomena such as thermal expansion, densification, swelling by fission product, thermal creep and radiation growth. The cladding variation is modeled by thermal expansion, thermal and irradiation creeps. The material properties are modeled by MATPRO and standard numerical parameter of TRANSURANUS code. Results of irradiation simulation with 9 kW/m LHR indicates that pellet-clad contacts onset from 0.090 mm initial gaps after 806 d, when pellet radius expansion attain 0.015 mm while inner cladding creep-down 0.075 mm. A newer computation data show that the maximum measured LHR of n-UO2 pin in the PRTF 12.4 kW/m. The next simulation will be done with a higher LHR, up to ~ 25 kW/m. MODEL SIMULASI VARIASI GEOMETRI DAN STRESS-STRAIN DARI PROTOTIP BAHAN BAKAR PIN BATAN SELAMA UJI IRADIASI DI REAKTOR RSG-GAS. Pusat Teknologi Bahan Bakar Nuklir (PTBBN telah menyiapkan tangkai (pin bahan bakar pendek perdana yang berisi pelet UO2 alam dalam kelongsong paduan zircaloy untuk dilakukan uji iradiasi daya naik. Penelitian ini merupakan bagian dari perancangan percobaan iradiasi pertama di PRTF (Power Ramp Test Fasility yang terpasang di reaktor serbaguna RSG-GAS berdaya 30 MW. Telah dilakukan pemodelan dan simulasi kinerja termal mekanikal pin selama iradiasi. Variasi geometri pelet dan kelongsong selama pengujian dimodelkan dengan memperhatikan fenomena ekspansi termal

Neutronics analysis of Dalat Nuclear Research Reactor by MVP/GMVP code

International Nuclear Information System (INIS)

Nguyen Kien Cuong; Toru Obara

2008-01-01

The paper presents neutronics calculation for Dalat Nuclear Research Reactor (DNRR) to validate MVP/GMVP Code. Beside fresh core calculation, burnt core and burn up distribution were also carried out and compared with experimental data or result obtained from other codes. With complex geometry and operating history like DNRR, burn up calculation by Monte Carlo Method is the better choice owing to the use of exact geometry description and continuous neutron energy in calculation. The discrepancy between calculated data and experimental data is good to compare. By using Monte Carlo method, continuous neutron energy from JENDL3.3 library and combined with burn up calculation, MVP/GMVP Code is a very useful tool for reactor calculation. (author)

Analysis of possibility to apply new mathematical methods (R-function theory) in Monte Carlo simulation of complex geometry

International Nuclear Information System (INIS)

Altiparmakov, D.

1988-12-01

This analysis is part of the report on ' Implementation of geometry module of 05R code in another Monte Carlo code', chapter 6.0: establishment of future activity related to geometry in Monte Carlo method. The introduction points out some problems in solving complex three-dimensional models which induce the need for developing more efficient geometry modules in Monte Carlo calculations. Second part include formulation of the problem and geometry module. Two fundamental questions to be solved are defined: (1) for a given point, it is necessary to determine material region or boundary where it belongs, and (2) for a given direction, all cross section points with material regions should be determined. Third part deals with possible connection with Monte Carlo calculations for computer simulation of geometry objects. R-function theory enables creation of geometry module base on the same logic (complex regions are constructed by elementary regions sets operations) as well as construction geometry codes. R-functions can efficiently replace functions of three-value logic in all significant models. They are even more appropriate for application since three-value logic is not typical for digital computers which operate in two-value logic. This shows that there is a need for work in this field. It is shown that there is a possibility to develop interactive code for computer modeling of geometry objects in parallel with development of geometry module [sr

Relaxed geometries and dipole moments of positron complexes with diatomic molecules

Energy Technology Data Exchange (ETDEWEB)

Assafrao, Denise; Mohallem, Jose R, E-mail: rachid@fisica.ufmg.b [Laboratorio de Atomos e Moleculas Especiais, Departamento de Fisica, ICEx, Universidade Federal de Minas Gerais, CP 702, 30123-970, Belo Horizonte, MG (Brazil)

2010-01-01

Relaxed geometries and dipole moments of diatomic molecules interacting with a slow positron are reported as functions of the positron distance to the more electronegative atom. A molecular model for the complex that allows applications to large systems is used. The electron population on the positron is proposed as a weighting function to calculate the average quantities. Results show Self-Consistent-Field quality or better.

Observations on the CANDLE burn-up in various geometries

International Nuclear Information System (INIS)

Seifritz, W.

2007-01-01

We have looked at all geometrical conditions under which an auto catalytically propagating burnup wave (CANDLE burn-up) is possible. Thereby, the Sine Gordon equation finds a new place in the burn-up theory of nuclear fission reactors. For a practical reactor design the axially burning 'spaghetti' reactor and the azimuthally burning 'pancake' reactor, respectively, seem to be the most promising geometries for a practical reactor design. Radial and spherical burn-waves in cylindrical and spherical geometry, respectively, are principally impossible. Also, the possible applicability of such fission burn-waves on the OKLO-phenomenon and the GEOREACTOR in the center of Earth, postulated by Herndon, is discussed. A fast CANDLE-reactor can work with only depleted uranium. Therefore, uranium mining and uranium-enrichment are not necessary anymore. Furthermore, it is also possible to dispense with reprocessing because the uranium utilization factor is as high as about 40%. Thus, this completely new reactor type can open a new era of reactor technology

Low enriched uranium foil targets with different geometries for the production of Molybdenum-99 in the BMR (Brazilian Multipurpose Reactor)

International Nuclear Information System (INIS)

Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Muniz, Rafael O.R.; Coelho, Talita S.

2011-01-01

A new research reactor is being planned in Brazil to take care of the demand of radiopharmaceuticals in the country and conduct research in various areas. This new reactor, the Brazilian Multipurpose Reactor (RMB), planned for 30 MW, is now in the conception design phase. Two low enriched ( 235 U) metallic uranium foil targets (cylinder and plate geometries) are being considered for production of Molybdenum-99 ( 99 Mo) by fission. Neutronic and thermal-hydraulics calculations were performed to compare the production of 99 Mo for these targets in the RMB and to determine the temperatures achieved in the targets. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION, CITATION and SCALE and for the thermal-hydraulics calculations were utilized the computer codes MTRCR-IEA-R1 and ANSYS CFX. (author)

Thermal plume above a simulated sitting person with different complexity of body geometry

DEFF Research Database (Denmark)

Zukowska, Daria; Melikov, Arsen Krikor; Popiolek, Zbigniew J.

2007-01-01

Occupants are one of the main heat sources in rooms. They generate thermal plumes with characteristics, which depend on geometry, surface temperature and area of the human body in contact with the surrounding air as well as temperature, velocity and turbulence intensity distribution in the room....... The characteristics of the thermal plume generated by a sitting person were studied using four human body simulators with different complexity of geometry but equal surface area: a vertical cylinder, a rectangular box, a dummy, and a thermal manikin. The results show that the dummy and the thermal manikin generate...

CIME school “Fully Nonlinear PDEs in Real and Complex Geometry and Optics”

CERN Document Server

Capogna, Luca; Gutiérrez, Cristian E; Montanari, Annamaria

2014-01-01

The purpose of this CIME summer school was to present current areas of research arising both in the theoretical and applied setting that involve fully nonlinear partial different equations. The equations presented in the school stem from the fields of Conformal Mapping Theory, Differential Geometry, Optics, and Geometric Theory of Several Complex Variables. The school consisted of four courses: Extremal problems for quasiconformal mappings in space by Luca Capogna, Fully nonlinear equations in geometry by Pengfei Guan, Monge-Ampere type equations and geometric optics by Cristian E. Gutiérrez, and On the Levi Monge Ampere equation by Annamaria Montanari.

Calculation of Added Mass for Submerged Reactor with Complex Shape

Energy Technology Data Exchange (ETDEWEB)

Sun, Jong-Oh; Kim, Gyeongho; Choo, Yeon-Seok; Yoo, Yeon-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Kijang Research Reactor (KJRR) is currently under construction. Its reactor is located on the bottom of a reactor pool which is filled with water to a depth of 12m. Some components are installed on or inside the reactor and their structural integrity and safety performance need to be verified under seismic situations. For the verification, time history data or Floor Response Spectrum (FRS) on their support location, which is the reactor, should be obtained. A Finite Element (FE) model with fluid elements can give very accurate results for the matter; however, it costs too many resources and takes too much time for the transient analyses. In order to make the model more efficient and simple, added masses are often used to simulate the effect of water instead of the fluid elements. Many literatures introduce methods to calculate the added mass according to the exterior shape of structures. In this paper, how to calculate added masses for complex shaped structure was suggested. The proposed method was applied to RSA for KJRR and its accuracy was verified through comparison of the natural frequencies of RSA with fluid elements and the added masses. They showed the differences less than 1.5% between two models. Finally, it is concluded that the proposed method is quite useful to obtain added masses for complex shaped structure.

The relation between geometry, hydrology and stability of complex hillslopes examined using low-dimensional hydrological models

NARCIS (Netherlands)

Talebi, A.

2008-01-01

Key words: Hillslope geometry, Hillslope hydrology, Hillslope stability, Complex hillslopes, Modeling shallow landslides, HSB model, HSB-SM model.

The hydrologic response of a hillslope to rainfall involves a complex, transient saturated-unsaturated interaction that usually leads to a

From Stein to Weinstein and back symplectic geometry of affine complex manifolds

CERN Document Server

Cieliebak, Kai

2013-01-01

A beautiful and comprehensive introduction to this important field. -Dusa McDuff, Barnard College, Columbia University This excellent book gives a detailed, clear, and wonderfully written treatment of the interplay between the world of Stein manifolds and the more topological and flexible world of Weinstein manifolds. Devoted to this subject with a long history, the book serves as a superb introduction to this area and also contains the authors' new results. -Tomasz Mrowka, MIT This book is devoted to the interplay between complex and symplectic geometry in affine complex manifolds. Affine co

Hazards of nuclear reactors and other major industrial complexes

International Nuclear Information System (INIS)

Farmer, F.R.

1982-01-01

Some of the problems of quantified risk analysis of the hazards of nuclear reactors and other major industrial complexes are raised particularly as seen by the proponents and opponents of atomic energy. These are exemplified by discussing the chemical accidents at Flixborough and Canvey Island and the Light Water Reactor Studies. The role of risk analysis in improving knowledge of the systems studies, improving methods of analysis, identifying weaknesses in systems and in improving engineering/maintenance/operation is also stressed. (U.K.)

Modelling of turbulence and combustion for simulation of gas explosions in complex geometries

Energy Technology Data Exchange (ETDEWEB)

Arntzen, Bjoern Johan

1998-12-31

This thesis analyses and presents new models for turbulent reactive flows for CFD (Computational Fluid Dynamics) simulation of gas explosions in complex geometries like offshore modules. The course of a gas explosion in a complex geometry is largely determined by the development of turbulence and the accompanying increased combustion rate. To be able to model the process it is necessary to use a CFD code as a starting point, provided with a suitable turbulence and combustion model. The modelling and calculations are done in a three-dimensional finite volume CFD code, where complex geometries are represented by a porosity concept, which gives porosity on the grid cell faces, depending on what is inside the cell. The turbulent flow field is modelled with a k-{epsilon} turbulence model. Subgrid models are used for production of turbulence from geometry not fully resolved on the grid. Results from laser doppler anemometry measurements around obstructions in steady and transient flows have been analysed and the turbulence models have been improved to handle transient, subgrid and reactive flows. The combustion is modelled with a burning velocity model and a flame model which incorporates the burning velocity into the code. Two different flame models have been developed: SIF (Simple Interface Flame model), which treats the flame as an interface between reactants and products, and the {beta}-model where the reaction zone is resolved with about three grid cells. The flame normally starts with a quasi laminar burning velocity, due to flame instabilities, modelled as a function of flame radius and laminar burning velocity. As the flow field becomes turbulent, the flame uses a turbulent burning velocity model based on experimental data and dependent on turbulence parameters and laminar burning velocity. The laminar burning velocity is modelled as a function of gas mixture, equivalence ratio, pressure and temperature in reactant. Simulations agree well with experiments. 139

On geometric simulating in nuclear reactor calculations by the Monte-Carlo method

International Nuclear Information System (INIS)

Ostashenko, S.V.

1988-01-01

Analysis of existing geometric modules makes it possible to reveal their disadvantages and to formulate requirements list, which should be satisfied by any usefull geometry system. Short description of GDL language used for complex reactor systems simulating is given. GDL language applies hierarchical representation scheme to assemblies, which aids to reduce significantly amount of input data. The language is part of GDL geometry system designed for MCU package and implemented on ES computers

An algorithm for solving thermalhydraulic equations in complex geometries: the Astec code

International Nuclear Information System (INIS)

Lonsdale, R.D.

1987-01-01

By applying a finite volume approach to a finite element mesh, the ASTEC computer code allows three-dimensional incompressible fluid flow and heat transfer in complex geometries to be simulated realistically, without making excessive demands on computing resources. The methods used in the code are described, and examples of the application of the code are presented

Hyperbolic Metamaterials with Complex Geometry

DEFF Research Database (Denmark)

Lavrinenko, Andrei; Andryieuski, Andrei; Zhukovsky, Sergei

2016-01-01

We investigate new geometries of hyperbolic metamaterialssuch as highly corrugated structures, nanoparticle monolayer assemblies, super-structured or vertically arranged multilayersand nanopillars. All structures retain basic propertiesof hyperbolic metamaterials, but have functionality improved...

Solar Proton Transport Within an ICRU Sphere Surrounded by a Complex Shield: Ray-trace Geometry

Science.gov (United States)

Slaba, Tony C.; Wilson, John W.; Badavi, Francis F.; Reddell, Brandon D.; Bahadori, Amir A.

2015-01-01

A computationally efficient 3DHZETRN code with enhanced neutron and light ion (Z is less than or equal to 2) propagation was recently developed for complex, inhomogeneous shield geometry described by combinatorial objects. Comparisons were made between 3DHZETRN results and Monte Carlo (MC) simulations at locations within the combinatorial geometry, and it was shown that 3DHZETRN agrees with the MC codes to the extent they agree with each other. In the present report, the 3DHZETRN code is extended to enable analysis in ray-trace geometry. This latest extension enables the code to be used within current engineering design practices utilizing fully detailed vehicle and habitat geometries. Through convergence testing, it is shown that fidelity in an actual shield geometry can be maintained in the discrete ray-trace description by systematically increasing the number of discrete rays used. It is also shown that this fidelity is carried into transport procedures and resulting exposure quantities without sacrificing computational efficiency.

Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1

International Nuclear Information System (INIS)

Reyes F, M. C.; Del Valle G, E.; Filio L, C.

2013-10-01

The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S 2 and P 1 . Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

Low enriched uranium foil targets with different geometries for the production of Molybdenum-99 in the BMR (Brazilian Multipurpose Reactor)

Energy Technology Data Exchange (ETDEWEB)

Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Muniz, Rafael O.R.; Coelho, Talita S., E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

A new research reactor is being planned in Brazil to take care of the demand of radiopharmaceuticals in the country and conduct research in various areas. This new reactor, the Brazilian Multipurpose Reactor (RMB), planned for 30 MW, is now in the conception design phase. Two low enriched (<20% {sup 235}U) metallic uranium foil targets (cylinder and plate geometries) are being considered for production of Molybdenum-99 ({sup 99}Mo) by fission. Neutronic and thermal-hydraulics calculations were performed to compare the production of {sup 99}Mo for these targets in the RMB and to determine the temperatures achieved in the targets. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION, CITATION and SCALE and for the thermal-hydraulics calculations were utilized the computer codes MTRCR-IEA-R1 and ANSYS CFX. (author)

Analysis of short-term reactor cavity transient

International Nuclear Information System (INIS)

Cheng, T.C.; Fischer, S.R.

1981-01-01

Following the transient of a hypothetical loss-of-coolant accident (LOCA) in a nuclear reactor, peak pressures are reached within the first 0.03 s at different locations inside the reactor cavity. Due to the complicated multidimensional nature of the reactor cavity, the short-term analysis of the LOCA transient cannot be performed by using traditional containment codes, such as CONTEMPT. The advanced containment code, BEACON/MOD3, developed at the Idaho National Engineering Laboratory (INEL), can be adapted for such analysis. This code provides Eulerian, one and two-dimensional, nonhomogeneous, nonequilibrium flow modeling as well as lumped parameter, homogeneous, equilibrium flow modeling for the solution of two-component, two-phase flow problems. The purpose of this paper is to demonstrate the capability of the BEACON code to analyze complex containment geometry such as a reactor cavity

Multiregion reactors

International Nuclear Information System (INIS)

Moura Neto, C. de; Nair, R.P.K.

1979-08-01

The study of reflected reactors can be done employing the multigroup diffusion method. The neutron conservation equations, inside the intervals, can be written by fluxes and group constants. A reflected reactor (one and two groups) for a slab geometry is studied, aplying the continuity of flux and current in the interface. At the end, the appropriated solutions for a infinite cylindrical reactor and for a spherical reactor are presented. (Author) [pt

Comparison of low enriched uranium (UAl{sub x}-Al and U-Ni) targets with different geometries for the production of molybdenum-99 in the RMB (Brazilian multipurpose reactor)

Energy Technology Data Exchange (ETDEWEB)

Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Silva, Jose Eduardo R. da; Angelo, Gabriel; Fedorenko, Giuliana G., E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Nishiyama, Pedro J.B. de O., E-mail: pedro.julio@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil)

2011-07-01

The Brazilian Multipurpose Reactor (RMB), now in the conception design phase, is being designed in Brazil to attend the demand of radiopharmaceuticals in the country and conduct researches in various areas. The new reactor, planned for 30 MW, will replace the IEA-R1 reactor of IPEN-CNEN/SP. Low enriched uranium (<20% {sup 235}U) UAl{sub x} dispersed in Al (plate geometry) and metallic uranium foil targets (plate and cylinder geometries) are being considered for production of Molybdenum-99 ({sup 99}Mo) by fission. Neutronic and thermal-hydraulics calculations were performed to compare the production of {sup 99}Mo for these targets in the RMB. For the neutronic calculations were utilized the computer codes Hammer-Technion, Citation and Scale and for the thermal-hydraulics calculations were utilized the computer code MTRCR-IEAR1 and ANSYS CFX. (author)

A new concept of smart flexible phased array transducer to inspect component of complex geometry

International Nuclear Information System (INIS)

Roy, O.; Mauhaut, S.; Casula, O.; Cattiaux, G.

2001-01-01

In most of industries as aeronautics, aerospace and nuclear, the main part of the non destructive testing is carried out directly in touch with the inspected component. Among others, the cooling piping of French pressurized water reactor comprises many welding components with complex geometry: elbows, butt welds, nozzles. In service inspections of such components performed with conventional ultrasonic contact transducers present limited performances. First, variations in sensitivity, due to unmatched contact on depressions or irregular surface are observed, resulting in poor detection performances. In addition, the beam orientation transmitted through complex interfaces cannot be totally controlled, because of the disorientations suffered by the transducer during its displacement. As a result, the possible defect cannot be correctly detected, positioned and characterized. To overcome these difficulties and to improve the performances of such inspections, the French Atomic Energy Commission has developed a new concept of transducer, allowing both to take into account the varying profile of the tested component and to efficiently compensate these effects. This transducer is a flexible phased array able to match the surface of the inspected specimen and to efficiently compensate the deformation of its own surface, in order to preserve the ultrasonic beam characteristics in spite of the profile variations encountered during the scanning. This ability is achieved thanks to a specific instrumentation, which measures the deformation of the transducer radiating surface, made of individual ultrasonic elements mechanically jointed to fit the actual surface of the component being inspected. Inspections in pulse-echo mode have been performed on a specimen with an irregular profile containing artificial embedded reflectors. The comparison with inspection carried out using conventional transducer shows the efficiency of the system to characterize defects under such complex

Molecular geometry

CERN Document Server

Rodger, Alison

1995-01-01

Molecular Geometry discusses topics relevant to the arrangement of atoms. The book is comprised of seven chapters that tackle several areas of molecular geometry. Chapter 1 reviews the definition and determination of molecular geometry, while Chapter 2 discusses the unified view of stereochemistry and stereochemical changes. Chapter 3 covers the geometry of molecules of second row atoms, and Chapter 4 deals with the main group elements beyond the second row. The book also talks about the complexes of transition metals and f-block elements, and then covers the organometallic compounds and trans

Coherent neutrino scattering with low temperature bolometers at Chooz reactor complex

International Nuclear Information System (INIS)

Billard, J; Gascon, J; Jesus, M De; Carr, R; Formaggio, J A; Heine, S T; Johnston, J; Leder, A; Sibille, V; Winslow, L; Dawson, J; Lasserre, T; Figueroa-Feliciano, E; Palladino, K J; Vivier, M

2017-01-01

We present the potential sensitivity of a future recoil detector for a first detection of the process of coherent elastic neutrino nucleus scattering (CE ν NS). We use the Chooz reactor complex in France as our luminous source of reactor neutrinos. Leveraging the ability to cleanly separate the rate correlated with the reactor thermal power against (uncorrelated) backgrounds, we show that a 10 kg cryogenic bolometric array with 100 eV threshold should be able to extract a CE ν NS signal within one year of running. (paper)

SCAP-82, Single Scattering, Albedo Scattering, Point-Kernel Analysis in Complex Geometry

International Nuclear Information System (INIS)

Disney, R.K.; Vogtman, S.E.

1987-01-01

1 - Description of problem or function: SCAP solves for radiation transport in complex geometries using the single or albedo scatter point kernel method. The program is designed to calculate the neutron or gamma ray radiation level at detector points located within or outside a complex radiation scatter source geometry or a user specified discrete scattering volume. Geometry is describable by zones bounded by intersecting quadratic surfaces within an arbitrary maximum number of boundary surfaces per zone. Anisotropic point sources are describable as pointwise energy dependent distributions of polar angles on a meridian; isotropic point sources may also be specified. The attenuation function for gamma rays is an exponential function on the primary source leg and the scatter leg with a build- up factor approximation to account for multiple scatter on the scat- ter leg. The neutron attenuation function is an exponential function using neutron removal cross sections on the primary source leg and scatter leg. Line or volumetric sources can be represented as a distribution of isotropic point sources, with un-collided line-of-sight attenuation and buildup calculated between each source point and the detector point. 2 - Method of solution: A point kernel method using an anisotropic or isotropic point source representation is used, line-of-sight material attenuation and inverse square spatial attenuation between the source point and scatter points and the scatter points and detector point is employed. A direct summation of individual point source results is obtained. 3 - Restrictions on the complexity of the problem: - The SCAP program is written in complete flexible dimensioning so that no restrictions are imposed on the number of energy groups or geometric zones. The geometric zone description is restricted to zones defined by boundary surfaces defined by the general quadratic equation or one of its degenerate forms. The only restriction in the program is that the total

A review on research activities using the SANS spectrometer in transmission geometry at ET-RR-1 reactor

International Nuclear Information System (INIS)

Adib, M.

1999-01-01

The phased double rotor facility operating at ET-RR-1 reactor (2MW) was rearranged to operate as SANS spectrometer in transmission geometry. The rotors are suspended in magnetic fields and are spinning up to 16,000 rpm producing bursts of polyenergetic neutrons with wavelengths from 0.2 nm to 6.5 nm and beam divergence of 17' on the sample. The review on research activities using the SANS spectrometer and its applications for powder particle size determination and the long wavelength fluctuation of magnetization of the Fe-Ni alloys are discussed. (author)

FORE-2, Thermohydraulics and Space-Independent Reactor Kinetics for Transients

International Nuclear Information System (INIS)

Fox, J.N.; Lawler, B.E.; Butz, H.R.; Heames, T.J.

1984-01-01

1 - Description of problem or function: FORE2 is a coupled thermal hydraulics-point kinetics digital computer code designed to calculate significant reactor parameters under steady-state conditions, or as functions of time during transients. The transients may result from a programmed reactivity insertion or a power change. Variable inlet coolant flow rate and temperature are considered. The code calculates the reactor power, the individual reactivity feedbacks, and the temperature of coolant, cladding, fuel, structure, and additional material for up to seven axial positions in three channel types which represent radial zones of the reactor. The heat of fusion, accompanying fuel melting, the liquid metal voiding reactivity, and the spatial and the time variation of the fuel cladding gap coefficient due to changes in gap size are considered. 2 - Method of solution: FORE2 input consists of property data, geometry, power and flow distribution factors, external time varying functions, experimental coefficients, and termination data. The differential equations for fluid flow, heat transfer, and point neutronics are solved by explicit finite-difference procedures. 3 - Restrictions on the complexity of the problem: Reactor excursions which can be calculated are restricted to those transients in which the reactor is not substantially destroyed. As a general rule, changes in reactor geometry and composition during an excursion are limited to those cases in which the reactivity effects of the changes may be considered as small perturbations of the initial system. Thus, accidents involving large-scale disassembly and bulk meltdown of a core are not covered by FORE2. FORE2 is valid only while the core retains its initial geometry

A study of complexity of oral mucosa using fractal geometry

Directory of Open Access Journals (Sweden)

S R Shenoi

2017-01-01

Full Text Available Background: The oral mucosa lining the oral cavity is composed of epithelium supported by connective tissue. The shape of the epithelial-connective tissue interface has traditionally been used to describe physiological and pathological changes in the oral mucosa. Aim: The aim is to evaluate the morphometric complexity in normal, dysplastic, well-differentiated, and moderately differentiated squamous cell carcinoma (SCC of the oral mucosa using fractal geometry. Materials and Methods: A total of 80 periodic acid–Schiff stained histological images of four groups: normal mucosa, dysplasia, well-differentiated SCC, and moderately differentiated SCC were verified by the gold standard. These images were then subjected to fractal analysis. Statistical Analysis: ANOVA and post hoc test: Bonferroni was applied. Results: Fractal dimension (FD increases as the complexity increases from normal to dysplasia and then to SCC. Normal buccal mucosa was found to be significantly different from dysplasia and the two grades of SCC (P < 0.05. ANOVA of fractal scores of four morphometrically different groups of buccal mucosa was significantly different with F (3,76 = 23.720 and P< 0.01. However, FD of dysplasia was not significantly different from well-differentiated and moderately differentiated SCC (P = 1.000 and P = 0.382, respectively. Conclusion: This study establishes FD as a newer tool in differentiating normal tissue from dysplastic and neoplastic tissue. Fractal geometry is useful in the study of both physiological and pathological changes in the oral mucosa. A new grading system based on FD may emerge as an adjuvant aid in cancer diagnosis.

Critical Parameters of Complex Geometry Intersecting Cylinders Containing Uranyl Nitrate Solution

Energy Technology Data Exchange (ETDEWEB)

Rothe, Robert Emil; Briggs, Joseph Blair

1999-06-01

About three dozen previously unreported critical configurations are presented for very complex geometries filled with high concentration enriched uranyl nitrate solution. These geometries resemble a tall, thin Central Column (or trunk of a "tree") having long, thin arms (or "branches") extending up to four directions off the column. Arms are equally spaced from one another in vertical planes; and that spacing ranges from arms in contact to quite wide spacings. Both the Central Column and the many different arms are critically safe by themselves when each, alone, is filled with fissile solution; but, in combination, criticality occurs due to the interactions between arms and the column. Such neutronic interactions formed the principal focus of this study. While these results are fresh to the nuclear criticality safety industry and to those seeking novel experiments against which to validate computer codes, the experiments, themselves, are not recent. Over 100 experiments were performed at the Rocky Flats Critical Mass Laboratory between September, 1967, and February of the following year.

Complex nonlinear behaviour of a fixed bed reactor with reactant recycle

DEFF Research Database (Denmark)

Recke, Bodil; Jørgensen, Sten Bay

1999-01-01

The fixed bed reactor with reactant recycle investigated in this paper can exhibit periodic solutions. These solutions bifurcate from the steady state in a Hopf bifurcation. The Hopf bifurcation encountered at the lowest value of the inlet concentration turns the steady state unstable and marks......,that the dynamic behaviour of a fixed bed reactor with reactant recycle is much more complex than previously reported....

Solar proton exposure of an ICRU sphere within a complex structure part II: Ray-trace geometry.

Science.gov (United States)

Slaba, Tony C; Wilson, John W; Badavi, Francis F; Reddell, Brandon D; Bahadori, Amir A

2016-06-01

A computationally efficient 3DHZETRN code with enhanced neutron and light ion (Z ≤ 2) propagation was recently developed for complex, inhomogeneous shield geometry described by combinatorial objects. Comparisons were made between 3DHZETRN results and Monte Carlo (MC) simulations at locations within the combinatorial geometry, and it was shown that 3DHZETRN agrees with the MC codes to the extent they agree with each other. In the present report, the 3DHZETRN code is extended to enable analysis in ray-trace geometry. This latest extension enables the code to be used within current engineering design practices utilizing fully detailed vehicle and habitat geometries. Through convergence testing, it is shown that fidelity in an actual shield geometry can be maintained in the discrete ray-trace description by systematically increasing the number of discrete rays used. It is also shown that this fidelity is carried into transport procedures and resulting exposure quantities without sacrificing computational efficiency. Published by Elsevier Ltd.

A tokamak reactor with servicing capability

International Nuclear Information System (INIS)

Mitchell, J.T.D.; Hollis, A.

1976-01-01

A conceptual design for a Tokamak reactor with practical facilities for the regular replacement of blanket components after the inevitable damage from neutron irradiation, and fatigue is described. This essential facility has been largely ignored in published fusion reactor designs. One exception is the inertially-confined Saturn proposal. Tokamak and other toroidal closed-line systems have very complex geometries and sub-system requirements, which result in blanket servicing being a very difficult problem. In the concept described the magnet shield is divided into two structures - an outer permanent one with access doors and an inner shield, part of and supporting the blanket inside. Servicing access is horizontally between the toroidal magnet coils, after moving some outer poloidal magnet coils. The reactor, reactor hall, workshops and remote-handling facilities are described, and the servicing requirements discussed. The important servicing operation is the remote replacement of radiation damaged blanket and shield - divided in this design into 20 sectors, each weighing 75-100 tons and 11-12 metres high. Analysis of the operation indicates that if one sector can be replaced during a single weekend - i.e. a period of low power demand - then the annual reactor-generator availability allowing as well for the general plant servicing should be >0.9. This level of availability should meet the requirements of generating authorities but the facilities, equipment and workshops necessary may be complex and expensive

MM99.81 Projection welding of complex geometries

DEFF Research Database (Denmark)

Kristensen, Lars

The objective of this work has been to establish a profound knowledge about design rules for projection welding geometries dependent of the actual material combination.Design rules and recommendations for geometries and projections in projection welding given in literature is summarised...... and these are catalogued into geometry-classes. A simulation software, SORPAS, based on the finite element method (FEM) is chosen as tool to investigate projection weld quality. SORPAS needs input of the material flow stress as function of strain, strain rate and temperature. Flow stress experiments are performed using...... been investigated.Two different welding geometries, disc with triangular ring projection welded to ring and hat welded to inside hole in ring, are both experimentally and numerically used to investigate the influence of different geometric parameters (thicknesses and angles) on weldability and weld...

Rotating reactors : a review

NARCIS (Netherlands)

Visscher, F.; Schaaf, van der J.; Nijhuis, T.A.; Schouten, J.C.

2013-01-01

This review-perspective paper describes the current state-of-the-art in the field of rotating reactors. The paper has a focus on rotating reactor technology with applications at lab scale, pilot scale and industrial scale. Rotating reactors are classified and discussed according to their geometry:

Two Step Procedure Using a 1-D Slab Spectral Geometry in a Pebble Bed Reactor Core Analysis

International Nuclear Information System (INIS)

Lee, Hyun Chul; Kim, Kang Seog; Noh, Jae Man; Joo, Hyung Kook

2005-01-01

A strong spectral interaction between the core and the reflector has been one of the main concerns in the analysis of pebble bed reactor cores. To resolve this problem, VSOP adopted iteration between the spectrum calculation in a spectral zone and the global core calculation. In VSOP, the whole problem domain is divided into many spectral zones in which the fine group spectrum is calculated using bucklings for fast groups and albedos for thermal groups from the global core calculation. The resulting spectrum in each spectral zone is used to generate broad group cross sections of the spectral zone for the global core calculation. In this paper, we demonstrate a two step procedure in a pebble bed reactor core analysis. In the first step, we generate equivalent cross sections from a 1-D slab spectral geometry model with the help of the equivalence theory. The equivalent cross sections generated in this way include the effect of the spectral interaction between the core and the reflector. In the second step, we perform a diffusion calculation using the equivalent cross sections generated in the first step. A simple benchmark problem derived from the PMBR-400 Reactor was introduced to verify this approach. We compared the two step solutions with the Monte Carlo (MC) solutions for the problem

Highly Manufacturable Deep (Sub-Millimeter) Etching Enabled High Aspect Ratio Complex Geometry Lego-Like Silicon Electronics

KAUST Repository

Ghoneim, Mohamed T.; Hussain, Muhammad Mustafa

2017-01-01

A highly manufacturable deep reactive ion etching based process involving a hybrid soft/hard mask process technology shows high aspect ratio complex geometry Lego-like silicon electronics formation enabling free-form (physically flexible

Accurate simulation of geometry, singlet-singlet and triplet-singlet excitation of cyclometalated iridium(III) complex.

Science.gov (United States)

Wang, Jian; Bai, Fu-Quan; Xia, Bao-Hui; Zhang, Hong-Xing; Cui, Tian

2014-03-01

In the current contribution, we present a critical study of the theoretical protocol used for the determination of the electronic spectra properties of luminescent cyclometalated iridium(III) complex, [Ir(III)(ppy)₂H₂dcbpy]⁺ (where, ppy = 2-phenylpyridine, H₂dcbpy = 2,2'-bipyridine-4,4'-dicarboxylic acid), considered as a representative example of the various problems related to the prediction of electronic spectra of transition metal complex. The choice of the exchange-correlation functional is crucial for the validity of the conclusions that would be drawn from the numerical results. The influence of the exchange-correlation on geometry parameter and absorption/emission band, the role of solvent effects on time-dependent density function theory (TD-DFT) calculations, as well as the importance of the chosen proper procedure to optimize triplet excited geometry, have been thus examined in detail. From the obtained results, some general conclusions and guidelines are presented: i) PBE0 functional is the most accurate in prediction of ground state geometry; ii) the well-established B3LYP, B3P86, PBE0, and X3LYP have similar accuracy in calculation of absorption spectrum; and iii) the hybrid approach TD-DFT//CIS gives out excellent agreement in the evaluation of triplet excitation energy.

High performance ultrasonic field simulation on complex geometries

Science.gov (United States)

Chouh, H.; Rougeron, G.; Chatillon, S.; Iehl, J. C.; Farrugia, J. P.; Ostromoukhov, V.

2016-02-01

Ultrasonic field simulation is a key ingredient for the design of new testing methods as well as a crucial step for NDT inspection simulation. As presented in a previous paper [1], CEA-LIST has worked on the acceleration of these simulations focusing on simple geometries (planar interfaces, isotropic materials). In this context, significant accelerations were achieved on multicore processors and GPUs (Graphics Processing Units), bringing the execution time of realistic computations in the 0.1 s range. In this paper, we present recent works that aim at similar performances on a wider range of configurations. We adapted the physical model used by the CIVA platform to design and implement a new algorithm providing a fast ultrasonic field simulation that yields nearly interactive results for complex cases. The improvements over the CIVA pencil-tracing method include adaptive strategies for pencil subdivisions to achieve a good refinement of the sensor geometry while keeping a reasonable number of ray-tracing operations. Also, interpolation of the times of flight was used to avoid time consuming computations in the impulse response reconstruction stage. To achieve the best performance, our algorithm runs on multi-core superscalar CPUs and uses high performance specialized libraries such as Intel Embree for ray-tracing, Intel MKL for signal processing and Intel TBB for parallelization. We validated the simulation results by comparing them to the ones produced by CIVA on identical test configurations including mono-element and multiple-element transducers, homogeneous, meshed 3D CAD specimens, isotropic and anisotropic materials and wave paths that can involve several interactions with interfaces. We show performance results on complete simulations that achieve computation times in the 1s range.

Cleaning of small components of complex geometry by means of the sodium-alcohol reaction

International Nuclear Information System (INIS)

De Luca, B.; Grasso, C.; Spadoni, M.

1978-01-01

The results of some experiments on the vacuum reaction between butylcellosolve and sodium, contained in small diameter capillaries, are reported. The effects on the cleaning rate of the temperature, amount of solvent, diameter and position of the capillaries are analyzed. The facility, used for the cleaning of small components of complex geometry, is described. (author)

Comparison of surface extraction techniques performance in computed tomography for 3D complex micro-geometry dimensional measurements

DEFF Research Database (Denmark)

Torralba, Marta; Jiménez, Roberto; Yagüe-Fabra, José A.

2018-01-01

micro-geometries as well (i.e., in the sub-mm dimensional range). However, there are different factors that may influence the CT process performance, being one of them the surface extraction technique used. In this paper, two different extraction techniques are applied to measure a complex miniaturized......The number of industrial applications of computed tomography (CT) for dimensional metrology in 100–103 mm range has been continuously increasing, especially in the last years. Due to its specific characteristics, CT has the potential to be employed as a viable solution for measuring 3D complex...... dental file by CT in order to analyze its contribution to the final measurement uncertainty in complex geometries at the mm to sub-mm scales. The first method is based on a similarity analysis: the threshold determination; while the second one is based on a gradient or discontinuity analysis: the 3D...

A study on axial and torsional resonant mode matching for a mechanical system with complex nonlinear geometries

Science.gov (United States)

Watson, Brett; Yeo, Leslie; Friend, James

2010-06-01

Making use of mechanical resonance has many benefits for the design of microscale devices. A key to successfully incorporating this phenomenon in the design of a device is to understand how the resonant frequencies of interest are affected by changes to the geometric parameters of the design. For simple geometric shapes, this is quite easy, but for complex nonlinear designs, it becomes significantly more complex. In this paper, two novel modeling techniques are demonstrated to extract the axial and torsional resonant frequencies of a complex nonlinear geometry. The first decomposes the complex geometry into easy to model components, while the second uses scaling techniques combined with the finite element method. Both models overcome problems associated with using current analytical methods as design tools, and enable a full investigation of how changes in the geometric parameters affect the resonant frequencies of interest. The benefit of such models is then demonstrated through their use in the design of a prototype piezoelectric ultrasonic resonant micromotor which has improved performance characteristics over previous prototypes.

The estimation of collision probabilities in complicated geometries

International Nuclear Information System (INIS)

Roth, M.J.

1969-04-01

This paper demonstrates how collision probabilities in complicated geometries may be estimated. It is assumed that the reactor core may be divided into a number of cells each with simple geometry so that a collision probability matrix can be calculated for each cell by standard methods. It is then shown how these may be joined together. (author)

Critical Parameters of Complex Geometries of Intersecting Cylinders Containing Uranyl Nitrate Solution

International Nuclear Information System (INIS)

Rothe, R. E.

1999-01-01

About three dozen previously unreported critical configurations are presented for very complex geometries filled with high concentration enriched uranyl nitrate solution. These geometries resemble a tall, thin Central Column (or trunk of a ''tree'') having long, thin arms (or ''branches'') extending up to four directions off the column. Arms are equally spaced from one another in vertical planes, and that spacing ranges from arms in contact to quite wide spacings. Both the Central Column and the many different arms are critically safe by themselves with each, alone, is filled with fissile solution; but, in combination, criticality occurs due to the interactions between arms and the column. Such neutronic interactions formed the principal focus of this study. While these results are fresh to the nuclear criticality safety industry and to those seeking novel experiments against which to validate computer codes, the experiments, themselves, are not recent. Over 100 experiments were performed at the Rocky Flats Critical Mass Laboratory between September, 1967, and February of the following year

Highly Manufacturable Deep (Sub-Millimeter) Etching Enabled High Aspect Ratio Complex Geometry Lego-Like Silicon Electronics

KAUST Repository

Ghoneim, Mohamed T.

2017-02-01

A highly manufacturable deep reactive ion etching based process involving a hybrid soft/hard mask process technology shows high aspect ratio complex geometry Lego-like silicon electronics formation enabling free-form (physically flexible, stretchable, and reconfigurable) electronic systems.

A numerical calculation method for flow discretisation in complex geometry with body-fitted grids; Rechenverfahren zur Diskretisierung von Stroemungen in komplexer Geometrie mittels koerperangepasster Gitter

Energy Technology Data Exchange (ETDEWEB)

Jin, X.

2001-04-01

A numerical calculation method basing on body fitted grids is developed in this work for computational fluid dynamics in complex geometry. The method solves the conservation equations in a general nonorthogonal coordinate system which matches the curvilinear boundary. The nonorthogonal, patched grid is generated by a grid generator which solves algebraic equations. By means of an interface its geometrical data can be used by this method. The conservation equations are transformed from the Cartesian system to a general curvilinear system keeping the physical Cartesian velocity components as dependent variables. Using a staggered arrangement of variables, the three Cartesian velocity components are defined on every cell surface. Thus the coupling between pressure and velocity is ensured, and numerical oscillations are avoided. The contravariant velocity for calculating mass flux on one cell surface is resulting from dependent Cartesian velocity components. After the discretisation and linear interpolation, a three dimensional 19-point pressure equation is found. Using the explicit treatment for cross-derivative terms, it reduces to the usual 7-point equation. Under the same data and process structure, this method is compatible with the code FLUTAN using Cartesian coordinates. In order to verify this method, several laminar flows are simulated in orthogonal grids at tilted space directions and in nonorthogonal grids with variations of cell angles. The simulated flow types are considered like various duct flows, transient heat conduction, natural convection in a chimney and natural convection in cavities. Their results achieve very good agreement with analytical solutions or empirical data. Convergence for highly nonorthogonal grids is obtained. After the successful validation of this method, it is applied for a reactor safety case. A transient natural convection flow for an optional sump cooling concept SUCO is simulated. The numerical result is comparable with the

A model of gas cavity breakup behind a blockage in fast breeder reactor subassembly geometry

International Nuclear Information System (INIS)

Fukuzawa, Y.

1980-05-01

A semi-empirical model has been developed to describe the transient behaviour of a gas cavity due to breakup behind a blockage in Liquid Metal Fast Breeder Reactor subassembly geometry. The main mechanisms assumed for gas cavity breakup in the present model are as follows: The gas cavity is broken up by the pressure fluctuation at the interface due to turbulence in the liquid. The centrifugal force on the liquid opposes breakup. The model is able to describe experimental results on the transient behaviour of a gas cavity due to breakup after the termination of gas injection. On the basis of the present model the residence time of a gas cavity behind a blockage in sodium is predicted and the dependence of the residence time on blockage size is discussed. (orig.) [de

An Automated Approach to Very High Order Aeroacoustic Computations in Complex Geometries

Science.gov (United States)

Dyson, Rodger W.; Goodrich, John W.

2000-01-01

Computational aeroacoustics requires efficient, high-resolution simulation tools. And for smooth problems, this is best accomplished with very high order in space and time methods on small stencils. But the complexity of highly accurate numerical methods can inhibit their practical application, especially in irregular geometries. This complexity is reduced by using a special form of Hermite divided-difference spatial interpolation on Cartesian grids, and a Cauchy-Kowalewslci recursion procedure for time advancement. In addition, a stencil constraint tree reduces the complexity of interpolating grid points that are located near wall boundaries. These procedures are used to automatically develop and implement very high order methods (>15) for solving the linearized Euler equations that can achieve less than one grid point per wavelength resolution away from boundaries by including spatial derivatives of the primitive variables at each grid point. The accuracy of stable surface treatments is currently limited to 11th order for grid aligned boundaries and to 2nd order for irregular boundaries.

Multi-Physics Demonstration Problem with the SHARP Reactor Simulation Toolkit

Energy Technology Data Exchange (ETDEWEB)

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

2015-12-21

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

Loading pattern optimization in hexagonal geometry using PANTHER

International Nuclear Information System (INIS)

Parks, G.T.; Knight, M.P.

1996-01-01

The extension of the loading pattern optimization capability of Nuclear Electric's reactor physics code PANTHER to hexagonal geometry cores is described. The variety of search methods available and the code's performance are illustrated by an example in which three search different methods are used in turn in order to find an optimal reload design for a sample hexagonal geometry problem. (author)

Groundwater Monitoring Plan for the Reactor Technology Complex Operable Unit 2-13

International Nuclear Information System (INIS)

Richard P. Wells

2007-01-01

This Groundwater Monitoring Plan describes the objectives, activities, and assessments that will be performed to support the on-going groundwater monitoring requirements at the Reactor Technology Complex, formerly the Test Reactor Area (TRA). The requirements for groundwater monitoring were stipulated in the Final Record of Decision for Test Reactor Area, Operable Unit 2-13, signed in December 1997. The monitoring requirements were modified by the First Five-Year Review Report for the Test Reactor Area, Operable Unit 2-13, at the Idaho National Engineering and Environmental Laboratory to focus on those contaminants of concern that warrant continued surveillance, including chromium, tritium, strontium-90, and cobalt-60. Based upon recommendations provided in the Annual Groundwater Monitoring Status Report for 2006, the groundwater monitoring frequency was reduced to annually from twice a year

Check and visualization of input geometry data using the geometrical module of the Monte Carlo code MCU: WWER-440 pressure vessel dosimetry benchmarks

International Nuclear Information System (INIS)

Gurevich, M.; Zaritsky, S.; Osmera, B.; Mikus, J.

1997-01-01

The Monte Carlo method gives the opportunity to conduct the calculations of neutron and photon flux without any simplifications of the 3-D geometry of the nuclear power and experimental devices. So, each graduated Monte Carlo code includes the combinatorial geometry module and tools for the geometry description giving a possibility to describe very complex systems with a number of hierarchy levels of the geometrical objects. Such codes as usual have special modules for the visual checking of geometry input information. These geometry opportunities could be used for all cases when the accurate 3-D description of the complex geometry becomes a necessity. The description (specification) of benchmark experiments is one of the such cases. Such accurate and uniform description detects all mistakes and ambiguities in the starting information of various kinds (drawings, reports etc.). Usually the quality of different parts of the starting information (generally produced by different persons during the different stages of the device elaboration and operation) is different. After using the above mentioned modules and tools, the resultant geometry description can be used as a standard for this device. One can automatically produce any type of the device figure. The detail geometry description can be used as input for different calculation models carrying out (not only for Monte Carlo). The application of that method to the description of the WWER-440 mock-ups is represented in the report. The mock-ups were created on the reactor LR-O (NRI) and the reactor vessel dosimetry benchmarks were developed on the basis of these mock-up experiments. The NCG-8 module of the Russian Monte Carlo code MCU was used. It is the combinatorial multilingual universal geometrical module. The MCU code was certified by Russian Nuclear Regulatory Body. Almost all figures for mentioned benchmarks specifications were made by the MCU visualization code. The problem of the automatic generation of the

Integrative shell of the program complex MARS (Version 1.0) radiation transfer in three-dimensional geometries

International Nuclear Information System (INIS)

Degtyarev, I.I.; Lokhovitskij, A.E.; Maslov, M.A.; Yazynin, I.A.

1994-01-01

The first version of integrative shell of the program complex MARS is written for calculating radiation transfer in the three-dimensional geometries. The integrative shell allows the user to work in convenient form with complex MARS, creat input files data and get graphic visualization of calculated functions. Version 1.0 is adapted for personal computers of types IBM-286,386,486 with operative size memory not smaller than 500K. 5 refs

Computational synthetic geometry

CERN Document Server

Bokowski, Jürgen

1989-01-01

Computational synthetic geometry deals with methods for realizing abstract geometric objects in concrete vector spaces. This research monograph considers a large class of problems from convexity and discrete geometry including constructing convex polytopes from simplicial complexes, vector geometries from incidence structures and hyperplane arrangements from oriented matroids. It turns out that algorithms for these constructions exist if and only if arbitrary polynomial equations are decidable with respect to the underlying field. Besides such complexity theorems a variety of symbolic algorithms are discussed, and the methods are applied to obtain new mathematical results on convex polytopes, projective configurations and the combinatorics of Grassmann varieties. Finally algebraic varieties characterizing matroids and oriented matroids are introduced providing a new basis for applying computer algebra methods in this field. The necessary background knowledge is reviewed briefly. The text is accessible to stud...

Extension of the comet method to 2-D hexagonal geometry

International Nuclear Information System (INIS)

Connolly, Kevin John; Rahnema, Farzad; Zhang, Dingkang

2011-01-01

The capability of the heterogeneous coarse mesh radiation transport (COMET) method developed at Georgia Tech has been expanded. COMET is now able to treat hexagonal geometry in two dimensions, allowing reactor problems to be solved for those next-generation reactors which utilize prismatic block structure and hexagonal lattice geometry in their designs. The COMET method is used to solve whole core reactor analysis problems without resorting to homogenization or low-order transport approximations. The eigenvalue and fission density distribution of the reactor are determined iteratively using response functions. The method has previously proven accurate in solving PWR, BWR, and CANDU eigenvalue problems. In this paper, three simple test cases inspired by high temperature test reactor material cross sections and fuel block geometry are presented. These cases are given not in an attempt to model realistic nuclear power systems, but in order to test the ability of the improved method. Solutions determined by the new hexagonal version of COMET, COMET-Hex, are compared with solutions determined by MCNP5, and the results show the accuracy and efficiency of the improved COMET-Hex method in calculating the eigenvalue and fuel pin fission density in sample full-core problems. COMETHex determines the eigenvalues of these simple problems to an order of within 50 pcm of the reference solutions and all pin fission densities to an average error of 0.2%, and it requires fewer than three minutes to produce these results. (author)

Dynamic Complexity Study of Nuclear Reactor and Process Heat Application Integration

International Nuclear Information System (INIS)

Taylor, J'Tia Patrice; Shropshire, David E.

2009-01-01

This paper describes the key obstacles and challenges facing the integration of nuclear reactors with process heat applications as they relate to dynamic issues. The paper also presents capabilities of current modeling and analysis tools available to investigate these issues. A pragmatic approach to an analysis is developed with the ultimate objective of improving the viability of nuclear energy as a heat source for process industries. The extension of nuclear energy to process heat industries would improve energy security and aid in reduction of carbon emissions by reducing demands for foreign derived fossil fuels. The paper begins with an overview of nuclear reactors and process application for potential use in an integrated system. Reactors are evaluated against specific characteristics that determine their compatibility with process applications such as heat outlet temperature. The reactor system categories include light water, heavy water, small to medium, near term high-temperature, and far term high temperature reactors. Low temperature process systems include desalination, district heating, and tar sands and shale oil recovery. High temperature processes that support hydrogen production include steam reforming, steam cracking, hydrogen production by electrolysis, and far-term applications such as the sulfur iodine chemical process and high-temperature electrolysis. A simple static matching between complementary systems is performed; however, to gain a true appreciation for system integration complexity, time dependent dynamic analysis is required. The paper identifies critical issues arising from dynamic complexity associated with integration of systems. Operational issues include scheduling conflicts and resource allocation for heat and electricity. Additionally, economic and safety considerations that could impact the successful integration of these systems are considered. Economic issues include the cost differential arising due to an integrated system

Ultrasonic non-destructive testing of pieces of complex geometry with a flexible phased array transducer

Science.gov (United States)

Chatillon; Cattiaux; Serre; Roy

2000-03-01

Ultrasonic non-destructive testing of components of complex geometry in the nuclear industry faces several difficulties: sensitivity variations due to unmatched contact, inaccurate localization of defects due to variations of transducer orientation, and uncovered area of the component. To improve the performances of such testing and defect characterization, we propose a new concept of ultrasonic contact phased array transducer. The phased array transducer has a flexible radiating surface able to fit the actual surface of the piece to optimize the contact and thus the sensitivity of the test. To control the transmitted field, and therefore to improve the defect characterization, a delay law optimizing algorithm is developed. To assess the capability of such a transducer, the Champ-Sons model, developed at the French Atomic Energy Commission for predicting field radiated by arbitrary transducers into pieces, has to be extended to sources directly in contact with pieces of complex geometry. The good behavior of this new type of probe predicted by computations is experimentally validated with a jointed transducer positioned on pieces of various profiles.

Assessment of capability for modeling the core degradation in 2D geometry with ASTEC V2 integral code for VVER type of reactor

International Nuclear Information System (INIS)

Dimov, D.

2011-01-01

The ASTEC code is progressively becoming the reference European severe accident integral code through in particular the intensification of research activities carried out since 2004. The purpose of this analysis is to assess ASTEC code modelling of main phenomena arising during hypothetical severe accidents and particularly in-vessel degradation in 2D geometry. The investigation covers both early and late phase of degradation of reactor core as well as determination of corium which will enter the reactor cavity. The initial event is station back-out. In order to receive severe accident condition, failure of all active component of emergency core cooling system is apply. The analysis is focus on ICARE module of ASTEC code and particularly on so call MAGMA model. The aim of study is to determine the capability of the integral code to simulate core degradation and to determine the corium composition entering the reactor cavity. (author)

Critical Parameters of Complex Geometries of Intersecting Cylinders Containing Uranyl Nitrate Solution

Energy Technology Data Exchange (ETDEWEB)

J. B. Briggs (INEEL POC); R. E. Rothe

1999-06-14

About three dozen previously unreported critical configurations are presented for very complex geometries filled with high concentration enriched uranyl nitrate solution. These geometries resemble a tall, thin Central Column (or trunk of a ''tree'') having long, thin arms (or ''branches'') extending up to four directions off the column. Arms are equally spaced from one another in vertical planes, and that spacing ranges from arms in contact to quite wide spacings. Both the Central Column and the many different arms are critically safe by themselves with each, alone, is filled with fissile solution; but, in combination, criticality occurs due to the interactions between arms and the column. Such neutronic interactions formed the principal focus of this study. While these results are fresh to the nuclear criticality safety industry and to those seeking novel experiments against which to validate computer codes, the experiments, themselves, are not recent. Over 100 experiments were performed at the Rocky Flats Critical Mass Laboratory between September, 1967, and February of the following year.

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

Energy Technology Data Exchange (ETDEWEB)

Wilson, Paul; Evans, Thomas; Tautges, Tim

2012-12-24

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

CALIPSO - a computer code for the calculation of fluiddynamics, thermohydraulics and changes of geometry in failing fuel elements of a fast breeder reactor

International Nuclear Information System (INIS)

Kedziur, F.

1982-07-01

The computer code CALIPSO was developed for the calculation of a hypothetical accident in an LMFBR (Liquid Metal Fast Breeder Reactor), where the failure of fuel pins is assumed. It calculates two-dimensionally the thermodynamics, fluiddynamics and changes in geometry of a single fuel pin and its coolant channel in a time period between failure of the pin and a state, at which the geometry is nearly destroyed. The determination of temperature profiles in the fuel pin cladding and the channel wall make it possible to take melting and freezing processes into account. Further features of CALIPSO are the variable channel cross section in order to model disturbances of the channel geometry as well as the calculation of two velocity fields including the consideration of virtual mass effects. The documented version of CALIPSO is especially suited for the calculation of the SIMBATH experiments carried out at the Kernforschungszentrum Karlsruhe, which simulate the above-mentioned accident. The report contains the complete documentation of the CALIPSO code: the modeling of the geometry, the equations used, the structure of the code and the solution procedure as well as the instructions for use with an application example. (orig.) [de

Highly Manufacturable Deep (Sub-Millimeter) Etching Enabled High Aspect Ratio Complex Geometry Lego-Like Silicon Electronics.

Science.gov (United States)

Ghoneim, Mohamed Tarek; Hussain, Muhammad Mustafa

2017-04-01

A highly manufacturable deep reactive ion etching based process involving a hybrid soft/hard mask process technology shows high aspect ratio complex geometry Lego-like silicon electronics formation enabling free-form (physically flexible, stretchable, and reconfigurable) electronic systems. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

CFD Modeling of Flow and Ion Exchange Kinetics in a Rotating Bed Reactor System

DEFF Research Database (Denmark)

Larsson, Hilde Kristina; Schjøtt Andersen, Patrick Alexander; Byström, Emil

2017-01-01

A rotating bed reactor (RBR) has been modeled using computational fluid dynamics (CFD). The flow pattern in the RBR was investigated and the flow through the porous material in it was quantified. A simplified geometry representing the more complex RBR geometry was introduced and the simplified...... model was able to reproduce the main characteristics of the flow. Alternating reactor shapes were investigated, and it was concluded that the use of baffles has a very large impact on the flows through the porous material. The simulations suggested, therefore, that even faster reaction rates could...... be achieved by making the baffles deeper. Two-phase simulations were performed, which managed to reproduce the deflection of the gas–liquid interface in an unbaffled system. A chemical reaction was implemented in the model, describing the ion-exchange phenomena in the porous material using four different...

Theory for stationary nonlinear wave propagation in complex magnetic geometry

International Nuclear Information System (INIS)

Watanabe, T.; Hojo, H.; Nishikawa, Kyoji.

1977-08-01

We present our recent efforts to derive a systematic calculation scheme for nonlinear wave propagation in the self-consistent plasma profile in complex magnetic-field geometry. Basic assumptions and/or approximations are i) use of the collisionless two-fluid model with an equation of state; ii) restriction to a steady state propagation and iii) existence of modified magnetic surface, modification due to Coriolis' force. We discuss four situations: i) weak-field propagation without static flow, ii) arbitrary field strength with flow in axisymmetric system, iii) weak field limit of case ii) and iv) arbitrary field strength in nonaxisymmetric torus. Except for case iii), we derive a simple variation principle, similar to that of Seligar and Whitham, by introducing appropriate coordinates. In cases i) and iii), we derive explicit results for quasilinear profile modification. (auth.)

Tetrahydropentalenyl-phosphazene constrained geometry complexes of rare-earth metal alkyls.

Science.gov (United States)

Hangaly, Noa K; Petrov, Alexander R; Elfferding, Michael; Harms, Klaus; Sundermeyer, Jörg

2014-05-21

Reactions of Cp™HPPh2 (1, diphenyl(4,4,6,6-tetramethyl-1,4,5,6-tetrahydropentalen-2-yl)phosphane) with the organic azides AdN3 and DipN3 (Ad = 1-adamantyl; Dip = 2,6-di-iso-propylphenyl) led to the formation of two novel CpPN ligands: P-amino-cyclopentadienylidene-phosphorane (Cp™PPh2NHAd; L(Ad)H) and P-cyclopentadienyl-iminophosphorane (Cp™HPPh2NDip; L(Dip)H). Both were characterized by NMR spectroscopy and X-ray structure analysis. For both compounds only one isomer was observed. Neither possesses any detectable prototropic or elementotropic isomers. Reactions of these ligands with [Lu(CH2SiMe3)3(thf)2] or with rare-earth metal halides and three equivalents of LiCH2SiMe3 produced the desired bis(alkyl) Cp™PN complexes: [{Cp™PN}M(CH2SiMe3)2] (M = Sc (1(Ad), 1(Dip)), Lu (2(Ad), 2(Dip)), Y (3(Ad), 3(Dip)), Sm (4(Ad)), Nd (5(Ad)), Pr (6(Ad)), Yb (7(Ad))). These complexes were characterized by extensive NMR studies for the diamagnetic and the paramagnetic complexes with full signal assignment. An almost mirror inverted order of the paramagnetic shifts has been observed for ytterbium complex 7(Ad) compared to 4(Ad), 5(Ad) and 6(Ad). For the assignment of the NMR signals [{η(1) : η(5)-C5Me4PMe2NAd}Yb(CH2SiMe3)2] 7 was synthesized, characterized and the (1)H NMR signals were compared to 7(Ad) and to other paramagnetic lanthanide complexes with the same ligand. 1(Ad), 2(Ad), 2(Dip), 3(Ad) and 3(Dip) were characterized by X-ray structure analysis revealing a sterically congested constrained geometry structure.

Convex-based void filling method for CAD-based Monte Carlo geometry modeling

International Nuclear Information System (INIS)

Yu, Shengpeng; Cheng, Mengyun; Song, Jing; Long, Pengcheng; Hu, Liqin

2015-01-01

Highlights: • We present a new void filling method named CVF for CAD based MC geometry modeling. • We describe convex based void description based and quality-based space subdivision. • The results showed improvements provided by CVF for both modeling and MC calculation efficiency. - Abstract: CAD based automatic geometry modeling tools have been widely applied to generate Monte Carlo (MC) calculation geometry for complex systems according to CAD models. Automatic void filling is one of the main functions in the CAD based MC geometry modeling tools, because the void space between parts in CAD models is traditionally not modeled while MC codes such as MCNP need all the problem space to be described. A dedicated void filling method, named Convex-based Void Filling (CVF), is proposed in this study for efficient void filling and concise void descriptions. The method subdivides all the problem space into disjointed regions using Quality based Subdivision (QS) and describes the void space in each region with complementary descriptions of the convex volumes intersecting with that region. It has been implemented in SuperMC/MCAM, the Multiple-Physics Coupling Analysis Modeling Program, and tested on International Thermonuclear Experimental Reactor (ITER) Alite model. The results showed that the new method reduced both automatic modeling time and MC calculation time

Water-chemical regime of a fast reactor ower complex

International Nuclear Information System (INIS)

Musikhin, R.N.; Piskunov, E.M.; Samarkin, A.A.; Yurchenko, D.S.

1983-01-01

Some peculiarities of water-chemical regime of a power compleX in Shevchenko are considered. The complex comprises a desalination unit, a gas-masout heating-and-power plant and the BN-350 reactor. The compleX is used for the production of electric and thermal energy and fresh water. The power complex peculiarity is the utilization of disalinated seawater in a technological cycle along with highly mineralized seawater with a total salt content of 13.5 g/l (for cooling) in heat exchanges. A regime of ammoniacal correction of feed water was used as a basic water-chemical regime in the initial period of the BN-350 steam generator operation. Deposits composed mainly of iron oxide slime were observed on steam generator surfaces during the operation under these conditions. A conclusion is made that the regime with chelating agent providing steam generator safe operation without chemical cleaning is the most expedient one

ARGO, 1-D Neutron Diffusion in Slab, Cylindrical, Spherical Geometry from JAERI Fast-Set, ABBN, RCBN

International Nuclear Information System (INIS)

Ikawa, Koji

1971-01-01

1 - Nature of physical problem solved: ARGO is a one-dimensional (slab, cylinder or sphere), multigroup diffusion code for use in fast reactor criticality and kinetic parameter analysis. Three cross section sets, i.e., JAERI-Fast-Set, ABBN-Set and RCBN-Set, of 25 groups are prepared for the code as its library tapes. 2 - Method of solution: Eigenvalues are computed by ordinary source-iteration techniques with ordinary acceleration methods for convergence. 3 - Restrictions on the complexity of the problem: Sphere geometry

Experimental approach for the uncertainty assessment of 3D complex geometry dimensional measurements using computed tomography at the mm and sub-mm scales

DEFF Research Database (Denmark)

Jiménez, Roberto; Torralba, Marta; Yagüe-Fabra, José A.

2017-01-01

The dimensional verification of miniaturized components with 3D complex geometries is particularly challenging. Computed Tomography (CT) can represent a suitable alternative solution to micro metrology tools based on optical and tactile techniques. However, the establishment of CT systems......’ traceability when measuring 3D complex geometries is still an open issue. In this work, an alternative method for the measurement uncertainty assessment of 3D complex geometries by using CT is presented. The method is based on the micro-CT system Maximum Permissible Error (MPE) estimation, determined...... experimentally by using several calibrated reference artefacts. The main advantage of the presented method is that a previous calibration of the component by a more accurate Coordinate Measuring System (CMS) is not needed. In fact, such CMS would still hold all the typical limitations of optical and tactile...

Information geometry

CERN Document Server

Ay, Nihat; Lê, Hông Vân; Schwachhöfer, Lorenz

2017-01-01

The book provides a comprehensive introduction and a novel mathematical foundation of the field of information geometry with complete proofs and detailed background material on measure theory, Riemannian geometry and Banach space theory. Parametrised measure models are defined as fundamental geometric objects, which can be both finite or infinite dimensional. Based on these models, canonical tensor fields are introduced and further studied, including the Fisher metric and the Amari-Chentsov tensor, and embeddings of statistical manifolds are investigated. This novel foundation then leads to application highlights, such as generalizations and extensions of the classical uniqueness result of Chentsov or the Cramér-Rao inequality. Additionally, several new application fields of information geometry are highlighted, for instance hierarchical and graphical models, complexity theory, population genetics, or Markov Chain Monte Carlo. The book will be of interest to mathematicians who are interested in geometry, inf...

The role of metal complexes in nuclear reactor decontamination

Energy Technology Data Exchange (ETDEWEB)

Prince, A.A.M.; Raghavan, P.S.; Gopalan, R. [Madras Christian College, Tambaram, Chennai (India); Velmurugan, S.; Narasimhan, S.V. [Bhabha Atomic Research Center (BARC) (IN). Water and Steam Chemistry Lab. (WSCL)

2006-07-15

Chemical decontamination is the process of removal of radioactivity from corrosion products formed on structural materials in the nuclear reactors. These corrosion products cause problems for the operation and maintenance of the plants. Removal of the radioactive contaminants can be achieved by dissolving the oxide from the system surface using organic complexing agents in low concentrations known as dilute chemical decontamination (DCD) formulations. These organic complexing agents attack the oxide surface and form metal complexes, which further accelerate the dissolution process. The stability of the complexes plays an important role in dissolving the radioactive contaminated oxides. In addition, the DCD process is operated through ion exchange resins for the removal of the dissolved metal ions and radioactive nuclides. In the present study, the kinetics of dissolution of various model corrosion products such as magnetite (Fe{sub 3}O{sub 4}), hematite ({alpha}-Fe{sub 2}O{sub 3}) and maghemite ({gamma}-Fe{sub 2}O{sub 3}) have been studied in the presence of complexing agents such as ethylenediaminetetraacetic acid (EDTA), nitrilotriacetic acid (NTA), hydroxyethylethylenediaminepentaacetic acid (HEEDTA), and 2,6 pyridinedicarboxylic acid (PDCA). The reductive roles of metal complexes and organic reducing agents are discussed. (orig.)

Development of M3C code for Monte Carlo reactor physics criticality calculations

International Nuclear Information System (INIS)

Kumar, Anek; Kannan, Umasankari; Krishanani, P.D.

2015-06-01

The development of Monte Carlo code (M3C) for reactor design entails use of continuous energy nuclear data and Monte Carlo simulations for each of the neutron interaction processes. BARC has started a concentrated effort for developing a new general geometry continuous energy Monte Carlo code for reactor physics calculation indigenously. The code development required a comprehensive understanding of the basic continuous energy cross section sets. The important features of this code are treatment of heterogeneous lattices by general geometry, use of point cross sections along with unionized energy grid approach, thermal scattering model for low energy treatment, capability of handling the microscopic fuel particles dispersed randomly. The capability of handling the randomly dispersed microscopic fuel particles which is very useful for the modeling of High-Temperature Gas-Cooled reactor fuels which are composed of thousands of microscopic fuel particle (TRISO fuel particle), randomly dispersed in a graphite matrix. The Monte Carlo code for criticality calculation is a pioneering effort and has been used to study several types of lattices including cluster geometries. The code has been verified for its accuracy against more than 60 sample problems covering a wide range from simple (like spherical) to complex geometry (like PHWR lattice). Benchmark results show that the code performs quite well for the criticality calculation of the system. In this report, the current status of the code, features of the code, some of the benchmark results for the testing of the code and input preparation etc. are discussed. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Duerigen, Susan

2013-05-15

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

Dynamic Complexity Study of Nuclear Reactor and Process Heat Application Integration

Energy Technology Data Exchange (ETDEWEB)

J' Tia Patrice Taylor; David E. Shropshire

2009-09-01

Abstract This paper describes the key obstacles and challenges facing the integration of nuclear reactors with process heat applications as they relate to dynamic issues. The paper also presents capabilities of current modeling and analysis tools available to investigate these issues. A pragmatic approach to an analysis is developed with the ultimate objective of improving the viability of nuclear energy as a heat source for process industries. The extension of nuclear energy to process heat industries would improve energy security and aid in reduction of carbon emissions by reducing demands for foreign derived fossil fuels. The paper begins with an overview of nuclear reactors and process application for potential use in an integrated system. Reactors are evaluated against specific characteristics that determine their compatibility with process applications such as heat outlet temperature. The reactor system categories include light water, heavy water, small to medium, near term high-temperature, and far term high temperature reactors. Low temperature process systems include desalination, district heating, and tar sands and shale oil recovery. High temperature processes that support hydrogen production include steam reforming, steam cracking, hydrogen production by electrolysis, and far-term applications such as the sulfur iodine chemical process and high-temperature electrolysis. A simple static matching between complementary systems is performed; however, to gain a true appreciation for system integration complexity, time dependent dynamic analysis is required. The paper identifies critical issues arising from dynamic complexity associated with integration of systems. Operational issues include scheduling conflicts and resource allocation for heat and electricity. Additionally, economic and safety considerations that could impact the successful integration of these systems are considered. Economic issues include the cost differential arising due to an integrated

Geometry of surfaces a practical guide for mechanical engineers

CERN Document Server

Radzevich, Stephen P

2012-01-01

Presents an in-depth analysis of geometry of part surfaces and provides the tools for solving complex engineering problems Geometry of Surfaces: A Practical Guide for Mechanical Engineers is a comprehensive guide to applied geometry of surfaces with focus on practical applications in various areas of mechanical engineering. The book is divided into three parts on Part Surfaces, Geometry of Contact of Part Surfaces and Mapping of the Contacting Part Surfaces. Geometry of Surfaces: A Practical Guide for Mechanical Engineers combines differential geometry and gearing theory and presents new developments in the elementary theory of enveloping surfaces. Written by a leading expert of the field, this book also provides the reader with the tools for solving complex engineering problems in the field of mechanical engineering. Presents an in-depth analysis of geometry of part surfaces Provides tools for solving complex engineering problems in the field of mechanical engineering Combines differential geometry an...

Material and geometry options and performance characteristics for a test reactor

International Nuclear Information System (INIS)

Jahshan, S.N.; Fletcher, C.D.; Terry, W.K.

1993-01-01

For the past 3 yr, an Idaho National Engineering Laboratory (INEL) design team has studied design options for a new test reactor to provide continued testing services after several aging test reactors in the United States are decommissioned. This new reactor, the Broad Application Test Reactor (BATR), would also fill other currently unmet needs, such as medical isotope production and space reactor component testing. Consideration of user needs, safety requirements, developmental uncertainties, and other factors led to the selection of an evolutionary design with plate fuel and several independently cooled test loops. The fuel would be cooled by light water, but most neutron moderation would come from heavy water or beryllium. The BATR design was tentatively scaled to the Advanced Test Reactor (ATR), an existing reactor at INEL: The power output of BATR is 250 MW(thermal), and the active core heights is 1 m. For safety in loss-of-flow events, the coolant flows upward through the core. The BATR design has one large test loop (with a test space diameter of 15.0 cm) along the central axis of the core and six smaller test loops (with test space diameters of 8.0 cm) centered at 6-deg azimuthal intervals on a 24.71-cm-diam circle around the central core axis

SERPENT Monte Carlo reactor physics code

International Nuclear Information System (INIS)

Leppaenen, J.

2010-01-01

SERPENT is a three-dimensional continuous-energy Monte Carlo reactor physics burnup calculation code, developed at VTT Technical Research Centre of Finland since 2004. The code is specialized in lattice physics applications, but the universe-based geometry description allows transport simulation to be carried out in complicated three-dimensional geometries as well. The suggested applications of SERPENT include generation of homogenized multi-group constants for deterministic reactor simulator calculations, fuel cycle studies involving detailed assembly-level burnup calculations, validation of deterministic lattice transport codes, research reactor applications, educational purposes and demonstration of reactor physics phenomena. The Serpent code has been publicly distributed by the OECD/NEA Data Bank since May 2009 and RSICC in the U. S. since March 2010. The code is being used in some 35 organizations in 20 countries around the world. This paper presents an overview of the methods and capabilities of the Serpent code, with examples in the modelling of WWER-440 reactor physics. (Author)

Electrochemical machining - manufacturing of turbine and reactor components

International Nuclear Information System (INIS)

Otto, K.

1987-01-01

Electrochemical machining is a shaping process for metallic workpieces with complex geometries. Using an electrode corresponding to the negative of the desired shape, the material to be removed is dissolved anodically at erosion rates of up to 10 mm/min. The reproducible shape accuracy lies between 0,02 and 0,08 mm, depending on the machining problem. Surface finishes of less than 18 μm are attained. The hardness of the material has no influence on the metal removal process. The workpiece is not subjected to any thermal stressing during machining. The process is well suited for quantity production of complex parts and is used inter alia for turbine blades and components for nuclear reactors. (orig.) [de

Developments in special geometry

International Nuclear Information System (INIS)

Mohaupt, Thomas; Vaughan, Owen

2012-01-01

We review the special geometry of N = 2 supersymmetric vector and hypermultiplets with emphasis on recent developments and applications. A new formulation of the local c-map based on the Hesse potential and special real coordinates is presented. Other recent developments include the Euclidean version of special geometry, and generalizations of special geometry to non-supersymmetric theories. As applications we discuss the proof that the local r-map and c-map preserve geodesic completeness, and the construction of four- and five-dimensional static solutions through dimensional reduction over time. The shared features of the real, complex and quaternionic version of special geometry are stressed throughout.

After Action Report: Advanced Test Reactor Complex 2015 Evaluated Drill October 6, 2015

Energy Technology Data Exchange (ETDEWEB)

Holmes, Forest Howard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-11-01

The Advanced Test Reactor (ATR) Complex, operated by Battelle Energy Alliance, LLC, at the Idaho National Laboratory (INL) conducted an evaluated drill on October 6, 2015, to allow the ATR Complex emergency response organization (ERO) to demonstrate the ability to respond to and mitigate an emergency by implementing the requirements of DOE O 151.1C, “Comprehensive Emergency Management System.”

Reactor prospects and present status of field-reversed configurations

International Nuclear Information System (INIS)

Hoffman, A.L.

1995-01-01

Field-Reversed Configurations (FRC) have an ideal geometry for a reactor, combining high beta toroidal confinement, with a linear external geometry. Present small diameter FRCs are thought to be stabilized by kinetic effects, but recent experiments in the Large s Experiment (LSX) have demonstrated stability as well into the MHD regime. Present empirical transport coefficients are already sufficient for a small pulsed reactor, but small steady state reactors will require about an order of magnitude reduction in plasma diffusivity. 13 refs., 4 figs., 1 tab

Numerical optimization of laboratory combustor geometry for NO suppression

International Nuclear Information System (INIS)

Mazaheri, Karim; Shakeri, Alireza

2016-01-01

Highlights: • A five-step kinetics for NO and CO prediction is extracted from GRI-3.0 mechanism. • Accuracy and applicability of this kinetics for numerical optimization were shown. • Optimized geometry for a combustor was determined using the combined process. • NO emission from optimized geometry is found 10.3% lower than the basis geometry. - Abstract: In this article, geometry optimization of a jet stirred reactor (JSR) combustor has been carried out for minimum NO emissions in methane oxidation using a combined numerical algorithm based on computational fluid dynamics (CFD) and differential evolution (DE) optimization. The optimization algorithm is also used to find a fairly accurate reduced mechanism. The combustion kinetics is based on a five-step mechanism with 17 unknowns which is obtained using an optimization DE algorithm for a PSR–PFR reactor based on GRI-3.0 full mechanism. The optimization design variables are the unknowns of the five-step mechanism and the cost function is the concentration difference of pollutants obtained from the 5-step mechanism and the full mechanism. To validate the flow solver and the chemical kinetics, the computed NO at the outlet of the JSR is compared with experiments. To optimize the geometry of a combustor, the JSR combustor geometry is modeled using three parameters (i.e., design variables). An integrated approach using a flow solver and the DE optimization algorithm produces the lowest NO concentrations. Results show that the exhaust NO emission for the optimized geometry is 10.3% lower than the original geometry, while the inlet temperature of the working fluid and the concentration of O_2 are operating constraints. In addition, the concentration of CO pollutant is also much less than the original chamber.

Integral Design Methodology of Photocatalytic Reactors for Air Pollution Remediation

Directory of Open Access Journals (Sweden)

Claudio Passalía

2017-06-01

Full Text Available An integral reactor design methodology was developed to address the optimal design of photocatalytic wall reactors to be used in air pollution control. For a target pollutant to be eliminated from an air stream, the proposed methodology is initiated with a mechanistic derived reaction rate. The determination of intrinsic kinetic parameters is associated with the use of a simple geometry laboratory scale reactor, operation under kinetic control and a uniform incident radiation flux, which allows computing the local superficial rate of photon absorption. Thus, a simple model can describe the mass balance and a solution may be obtained. The kinetic parameters may be estimated by the combination of the mathematical model and the experimental results. The validated intrinsic kinetics obtained may be directly used in the scaling-up of any reactor configuration and size. The bench scale reactor may require the use of complex computational software to obtain the fields of velocity, radiation absorption and species concentration. The complete methodology was successfully applied to the elimination of airborne formaldehyde. The kinetic parameters were determined in a flat plate reactor, whilst a bench scale corrugated wall reactor was used to illustrate the scaling-up methodology. In addition, an optimal folding angle of the corrugated reactor was found using computational fluid dynamics tools.

Direct numerical simulation of reactor two-phase flows enabled by high-performance computing

Energy Technology Data Exchange (ETDEWEB)

Fang, Jun; Cambareri, Joseph J.; Brown, Cameron S.; Feng, Jinyong; Gouws, Andre; Li, Mengnan; Bolotnov, Igor A.

2018-04-01

Nuclear reactor two-phase flows remain a great engineering challenge, where the high-resolution two-phase flow database which can inform practical model development is still sparse due to the extreme reactor operation conditions and measurement difficulties. Owing to the rapid growth of computing power, the direct numerical simulation (DNS) is enjoying a renewed interest in investigating the related flow problems. A combination between DNS and an interface tracking method can provide a unique opportunity to study two-phase flows based on first principles calculations. More importantly, state-of-the-art high-performance computing (HPC) facilities are helping unlock this great potential. This paper reviews the recent research progress of two-phase flow DNS related to reactor applications. The progress in large-scale bubbly flow DNS has been focused not only on the sheer size of those simulations in terms of resolved Reynolds number, but also on the associated advanced modeling and analysis techniques. Specifically, the current areas of active research include modeling of sub-cooled boiling, bubble coalescence, as well as the advanced post-processing toolkit for bubbly flow simulations in reactor geometries. A novel bubble tracking method has been developed to track the evolution of bubbles in two-phase bubbly flow. Also, spectral analysis of DNS database in different geometries has been performed to investigate the modulation of the energy spectrum slope due to bubble-induced turbulence. In addition, the single-and two-phase analysis results are presented for turbulent flows within the pressurized water reactor (PWR) core geometries. The related simulations are possible to carry out only with the world leading HPC platforms. These simulations are allowing more complex turbulence model development and validation for use in 3D multiphase computational fluid dynamics (M-CFD) codes.

Low Complexity Connectivity Driven Dynamic Geometry Compression for 3D Tele-Immersion

NARCIS (Netherlands)

R.N. Mekuria (Rufael); D.C.A. Bulterman (Dick); P.S. Cesar Garcia (Pablo Santiago)

2014-01-01

htmlabstractGeometry based 3D Tele-Immersion is a novel emerging media application that involves on the fly reconstructed 3D mesh geometry. To enable real-time communication of such live reconstructed mesh geometry over a bandwidth limited link, fast dynamic geometry compression is needed. However,

Nuclear power desalinating complex with IRIS reactor plant and Russian distillation desalinating unit

International Nuclear Information System (INIS)

Kostin, V. I.; Panov, Yu.K.; Polunichev, V. I.; Fateev, S. A.; Gureeva, L. V.

2004-01-01

This paper has been prepared as a result of Russian activities on the development of nuclear power desalinating complex (NPDC) with the IRIS reactor plant (RP). The purpose of the activities was to develop the conceptual design of power desalinating complex (PDC) and to evaluate technical and economical indices, commercial attractiveness and economical efficiency of PDC based on an IRIS RP with distillation desalinating plants. The paper presents the main results of studies as applied to dual-purpose PDC based on IRIS RP with different types of desalinating plants, namely: characteristics of nuclear power desalinating complex based on IRIS reactor plant using Russian distillation desalinating technologies; prospective options of interface circuits of the IRIS RP with desalinating plants; evaluations of NPDC with IRIS RP output based on selected desalinating technologies for water and electric power supplied to the grid; cost of water generated by NPDC for selected interface circuits made by the IAEA DEEP code as well as by the Russian TEO-INVEST code; cost evaluation results for desalinated water of PDC operating on fossil fuel and conditions for competitiveness of the nuclear PDC based on IRIS RP compared with analog desalinating complexes operating on fossil fuel.(author)

Modelization of physical phenomena in research reactors with the help of new developments in transport methods, and methodology validation with experimental data

International Nuclear Information System (INIS)

Rauck, St.

2000-10-01

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

Challenges associated with the current processes for ultrasonic inspection of CANDU reactor feeder piping

Energy Technology Data Exchange (ETDEWEB)

Machowski, C. [Babcock & Wilcox Canada Ltd., Cambridge, Ontario (Canada)

2012-07-01

CANDU® PHT Feeder Piping is generally constructed from SA106 Grade B carbon steel, which is known to be susceptible to flow accelerated corrosion when exposed to certain environmental conditions. The configuration of the CANDU reactor promotes thinning of the inside surface of the pipe walls, predominantly at the outlet feeders. Inspection of this piping is currently conducted using ultrasonic techniques and is governed by the requirements established by the CANDU Owners Group (COG). There are many challenges associated with these inspections as a result of the complexity of the reactor piping configuration. Geometrical anomalies on the surface of the pipe and non-circular geometries at the tight radius bends hinder the performance of conventional ultrasonic techniques. This can cause lost signals in areas of interest, which in turn often results in rework in order to satisfy the inspection requirements and justify fitness for service of these components. There are also many inspection sites which have limited access due to physical restrictions on the reactor face; therefore in order to maximize the performance of an inspection campaign, it is paramount that the inspection personnel and the inspection technology be well integrated through training simulations prior to execution. These inspection challenges increase the complexity of the analysis process as ultrasonic signals get distorted and lost as a result of non-circular pipe geometries. In order to ensure a high level of integrity in the analysis results, a conservative process is utilized in which two analysts independently examine the data, and a third analyst reviews their results and submits the final call. A Data Management Software application (DMS) is used to input and store the three analysis results. Another important function of the DMS is to provide a communication link between the different work-groups associated with the inspection activities. The focus of this presentation discusses: �

Advanced nuclear fuel production by using fission-fusion hybrid reactor

International Nuclear Information System (INIS)

Al-Kusayer, T.A.; Sahin, S.; Abdulraoof, M.

1993-01-01

Efforts are made at the College of Engineering, King Saud University, Riyadh to lay out the main structure of a prototype experimental fusion and fusion-fission (hybrid) reactor blanket in cylindrical geometry. The geometry is consistent with most of the current fusion and hybrid reactor design concepts in respect of the neutronic considerations. Characteristics of the fusion chamber, fusion neutrons and the blanket are provided. The studies have further shown that 1 GWe fission-fusion reactor can produce up to 957 kg/year which is enough to fuel five light water reactors of comparable power. Fuel production can be increased further. 29 refs

Observations of the behaviour of gas in the wake behind a corner blockage in fast breeder reactor subassembly geometry

International Nuclear Information System (INIS)

Fukuzawa, Y.

1979-07-01

Observations were made of gas behaviour in the wake behind a 21% corner blockage in the subassembly geometry of a liquid metal fast breeder reactor. The test section used represented one half of the reactor fuel subassembly, divided along the vertical plane of symmetry through the blockage. A glass wall occupied the position of this plane. Water was allowed to flow between glass rods simulating fuel pins, the velocity being changed from 1.2 to 4.5 m/s. Argon was injected into the wake or into the flow upstream of the blockage, the injection rate being changed from 1 to 230 Ncm 3 /s (standard temperature and pressure). From the present experiment, the following is evident: The gas is accumulated in the wake behind the blockage, forming a gas cavity. The flow patterns of the two-phase mixture in the wake are classified into three types, depending on the liquid velocity. In the lower velocity range, a gas cavity cannot be present at rest, rising up through the wake as a single bubble due to buoyancy. In the higher velocity range, the gas cavity is broken up by the liquid flow forces, only small gas bubbles circulating in the wake. In the velocity range in between, the gas cavity is present in the wake. The cavity size depends on the gas injection rate and on the liquid velocity. From the results, the possibility of fuel failure caused by fission gas release at a blockage in the fast breeder reactor can be considered to depend on the operating conditions of the reactor, specially on the coolant velocity. (orig.) [de

Mixing In Jet-Stirred Reactors With Different Geometries

KAUST Repository

Ayass, Wassim W.

2013-01-01

analysis emerges with determining the experimental residence time distribution (RTD) curves of each reactor. Comparing these RTD curves with the ideal curve helped in eliminating two cases. Finally, the CFD simulations predict the RTD curves as well

A nuclear desalination complex with a VK-300 boiling type reactor facility

International Nuclear Information System (INIS)

Kuznetzov, Y.N.; Mishanina, Y.A.; Romenkov, A.A.

2004-01-01

RDIPE has developed a detailed design of an enhanced safety nuclear steam supply system (NSSS) with a VK-300 boiling water reactor for combined heat and power generation. The thermal power of the reactor is 750 MW. The maximum electrical power in the condensation mode is 250 MWe. The maximum heat generation capacity of 400 Gcal/h is reached at 150 MWe. This report describes, in brief, the basic technical concepts for the VK-300 NSSS and the power unit, with an emphasis on enhanced safety and good economic performance. With relatively small power, good technical and economic performance of the VK-300 reactor that is a base for the desalination complex is attained through: reduced capital costs of the nuclear plant construction thanks to technical approaches ensuring maximum simplicity of the reactor design and the NSSS layout; a single-circuit power unit configuration (reactor-turbine) excluding expensive equipment with a lot of metal, less pipelines and valves; reduced construction costs of the basic buildings thanks to reduced construction volumes due to rational arrangement concepts; higher reliability of equipment and reduced maintenance and repair costs; longer reactor design service life of up to 60 years; selection of the best reactor and desalination equipment interface pattern. The report considers the potential application of the VK-300 reactor as a source of energy for distillation desalination units. The heat from the reactor is transferred to the desalination unit via an intermediate circuit. Comparison is made between variants of the reactor integration with desalination units of the following types: multi-stage flash (MSF technology); multi-effect distillation horizontal-tube film units of the DOU GTPA type (MED technology). The NDC capacity with the VK-300 reactor, in terms of distillate, will be more than 200,000 m 3 /day, with the simultaneous output of electric power from the turbine generator buses of around 150 MWe. The variants of the

Standard Guide for Benchmark Testing of Light Water Reactor Calculations

CERN Document Server

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 ...

Complex resistivity spectra in relation to multiscale pore geometry in carbonates and mixed-siliciclastic rocks

Science.gov (United States)

Norbisrath, Jan Henrik

Carbonate rocks are known to have complex and heterogeneous pore structures, which result from their biogenic origin and strong affinity for diagenetic processes that change their pore structure after burial. The combination of sheer endless variations of precursor biogenic material, depositional environments, and diagenetic effects results in rocks that are interesting to study but intricate to understand. Many schemes to categorize the diversity of carbonate rocks are in use today; most are based on the macropore structure and qualitative thin-section analysis. Many studies, however, acknowledge that micropores have a significant influence on the macroscopic petrophysical rock properties, which are essential to determine reservoir quality. Micropores are, by definition, smaller than the thickness of a thin-section (four major carbonate microporosity types: (1) small intercrystalline, (2) large inter-crystalline, (3) intercement, and (4) micromoldic. Each microporosity type shows a distinct capacity to conduct electrical charge, which largely controls the magnitude and range of cementation factors (m) in rocks with such microporosity type. The BIB-SEM method is also used on a dataset of mixed carbonate-siliciclastic (mudrock) samples with high kerogen and pyrite content. Results show that the nanopore geometry here has little influence on cementation factors, and instead porosity is the main control on m in mudrocks. Cementation factors are crucial for estimates of oil-in-place and water saturation in a wireline application, and a slight change of (assumed) cementation factor can change the interpreter's evaluation from dry hole to discovery. Therefore, accurate determination of cementation factors is a critical task in formation evaluation, similar to accurate estimates of permeability. To achieve this goal, this dissertation utilizes a new approach of using complex resistivity spectra (CRS) to assess the pore geometry and its resulting electrical and fluid flow

Study on modeling technology in digital reactor system

International Nuclear Information System (INIS)

Liu Xiaoping; Luo Yuetong; Tong Lili

2004-01-01

Modeling is the kernel part of a digital reactor system. As an extensible platform for reactor conceptual design, it is very important to study modeling technology and develop some kind of tools to speed up preparation of all classical computing models. This paper introduces the background of the project and basic conception of digital reactor. MCAM is taken as an example for modeling and its related technologies used are given. It is an interface program for MCNP geometry model developed by FDS team (ASIPP and HUT), and designed to run on windows system. MCAM aims at utilizing CAD technology to facilitate creation of MCNP geometry model. There have been two ways for MCAM to utilize CAD technology: (1) Making use of user interface technology in aid of generation of MCNP geometry model; (2) Making use of existing 3D CAD model to accelerate creation of MCNP geometry model. This paper gives an overview of MCAM's major function. At last, several examples are given to demonstrate MCAM's various capabilities. (authors)

Beam propagation through a gaseous reactor: classical transport

International Nuclear Information System (INIS)

Yu, S.S.; Buchanan, H.L.; Lee, E.P.; Chambers, F.W.

1979-01-01

The present calculations are applicable to any beam geometry with cylindrical symmetry, including the converging beam geometry (large entrance port with radius > or approx. = 10 cm), as well as the pencil-shaped beam (small porthole with radius approx. mm). The small porthole is clearly advantageous from the reactor vessel design point of view. While the physics of the latter mode of propagation may be more complex, analyses up to this point have not revealed any detrimental instability effects that will inhibit propagation. In fact, the large perpendicular velocity v/sub perpendicular/ that the pinched mode can accommodate provides a mechanism for the quenching of filamentary instability. Furthermore, this mode of propagation can withstand more ion scattering and is not subject to the upper bound on pressure (p < 10 torr) which is imposed on the converging beam mode

Longitudinal Variation in Paleo-channel Complex Geometry and Associated Fill: Offshore South Carolina

Science.gov (United States)

Long, A. M.; Hill, J. C.

2017-12-01

In northeastern South Carolina, several shallow (migration of the ancestral Pee Dee River system along the southern limb of the Cape Fear Arch since the Pliocene. These paleo-channel complexes can be traced 80 km across the continental shelf via Boomer and Chirp subbottom data. The Murrells Inlet paleo-channel complex is the most well imaged offshore; and this data coverage provides an opportunity for a detailed seismic stratigraphic interpretation and analysis of downstream variability. Initial observations from this case study indicate that inner shelf incisions, where bedrock is folded and faulted, tend to be shallow with numerous channels, while the incisions across the middle shelf appear to be deeper and contains larger, more sinuous channels that are cut into broadly tilted strata with a gentle south-southeastward dip. This suggests the geometry and spatial distribution of the incisions were a function of the inherited fabric of the underlying basement, which created local deflection and areas of aggradation and degradation. The inner shelf paleo-channel complex fill is dominated by fluvial cut and fill seismic facies, while the middle shelf contains a wide variety of seismic facies (i.e. transparent, layered, chaotic, etc). This overall longitudinal fill pattern is most likely due to each location's general proximity to base level. The variation in the cut and fill seismic facies may be driven by substantial changes in discharge, driven locally by the joining of another major river or by climatic changes in the drainage basin. There also appears to be preferential reoccupation of previously filled paleo-channels, as the basement in this region is Tertiary and Cretaceous carbonates and siliciclastic rocks that are more resistant to erosion. The most recent occupation in any given paleo-channel tends to be on the southern margin, which may imply tectonic forcing from the uplift of the Cape Fear Arch. Preliminary results from this case study suggest that first

Reactor refurbishment in an outage environment

International Nuclear Information System (INIS)

Gowthorpe, P.; Hoare, R.

2012-01-01

Reactor life extension has typically been performed during specific refurbishment outages. These outages are long and costly due to the sheer complexity of the scope, not to mention the ever present discovery work. A scope of this size requires a huge labour force to execute, which poses significant challenges. The work is difficult to staff with qualified people able to execute the work smoothly and managing the required labour pool problematic. Cost and time overruns are inevitable in that environment. Reducing the cost and schedule is critical to the long term viability of reactor refurbishment projects. With planning, the total cost of the refurbishment can be reduced by managing the inspection and repairs during normal outages. Identifying what activities need to be done each outage for the life of the reactor and bringing the latest technology can make this viable. Tightly planned outages with a small well trained labour force will go a long way to reducing costs. The suite of services and tooling available to the utilities to manage their reactor integrity has improved significantly in recent years and continues to evolve. New feeder inspection technologies can provide improved inspection results for the complex feeder geometry. These improvements lead to more accurate wear rates and better predictions of component life. Feeders that need replacement based on improved inspection techniques can be replaced systematically during regular outages rather than specific refurbishment outages. Targeting areas rather than entire feeders reduces time, dose and cost. In cases where feeder replacement isn't feasible or where unpredicted wear is found, a feeder weld overlay process can be used. To manage the reactor work, new data systems are under development that allow for effective tracking of each activity performed and outcomes in a single package. (author)

The accuracy of geometries for iron porphyrin complexes from density functional theory

DEFF Research Database (Denmark)

Rydberg, Patrik Åke Anders; Olsen, Lars

2009-01-01

functionals is evaluated with regard to how they reproduce experimental structures. Seven different functionals (BP86, PBE, PBE0, TPSS, TPSSH, B3LYP, and B97-D) are used to study eight different iron porphyrin complexes. The results show that the TPSSH, PBE0, and TPSS functionals give the best results...... (absolute bond distance deviations of 0.015-0.016 A), but the geometries are well-reproduced by all functionals except B3LYP. We also test four different basis sets of double-zeta quality, and we find that a combination of double-zeta basis set of Schafer et al. on the iron atom and the 6-31G* basis set...

Geometry and analysis on manifolds in memory of professor Shoshichi Kobayashi

CERN Document Server

Mabuchi, Toshiki; Maeda, Yoshiaki; Noguchi, Junjiro; Weinstein, Alan

2015-01-01

This volume is dedicated to the memory of Shoshichi Kobayashi, and gathers contributions from distinguished researchers working on topics close to his research areas. The book is organized into three parts, with the first part presenting an overview of Professor Shoshichi Kobayashi’s career. This is followed by two expository course lectures (the second part) on recent topics in extremal Kähler metrics and value distribution theory, which will be helpful for graduate students in mathematics interested in new topics in complex geometry and complex analysis. Lastly, the third part of the volume collects authoritative research papers on differential geometry and complex analysis. Professor Shoshichi Kobayashi was a recognized international leader in the areas of differential and complex geometry. He contributed crucial ideas that are still considered fundamental in these fields. The book will be of interest to researchers in the fields of differential geometry, complex geometry, and several complex variables ...

Special metrics and group actions in geometry

CERN Document Server

Fino, Anna; Musso, Emilio; Podestà, Fabio; Vezzoni, Luigi

2017-01-01

The volume is a follow-up to the INdAM meeting “Special metrics and quaternionic geometry” held in Rome in November 2015. It offers a panoramic view of a selection of cutting-edge topics in differential geometry, including 4-manifolds, quaternionic and octonionic geometry, twistor spaces, harmonic maps, spinors, complex and conformal geometry, homogeneous spaces and nilmanifolds, special geometries in dimensions 5–8, gauge theory, symplectic and toric manifolds, exceptional holonomy and integrable systems. The workshop was held in honor of Simon Salamon, a leading international scholar at the forefront of academic research who has made significant contributions to all these subjects. The articles published here represent a compelling testimony to Salamon’s profound and longstanding impact on the mathematical community. Target readership includes graduate students and researchers working in Riemannian and complex geometry, Lie theory and mathematical physics.

A Framework for the Interactive Handling of High-Dimensional Simulation Data in Complex Geometries

KAUST Repository

Benzina, Amal; Buse, Gerrit; Butnaru, Daniel; Murarasu, Alin; Treib, Marc; Varduhn, Vasco; Mundani, Ralf-Peter

2013-01-01

Flow simulations around building infrastructure models involve large scale complex geometries, which when discretized in adequate detail entail high computational cost. Moreover, tasks such as simulation insight by steering or optimization require many such costly simulations. In this paper, we illustrate the whole pipeline of an integrated solution for interactive computational steering, developed for complex flow simulation scenarios that depend on a moderate number of both geometric and physical parameters. A mesh generator takes building information model input data and outputs a valid cartesian discretization. A sparse-grids-based surrogate model—a less costly substitute for the parameterized simulation—uses precomputed data to deliver approximated simulation results at interactive rates. Furthermore, a distributed multi-display visualization environment shows building infrastructure together with flow data. The focus is set on scalability and intuitive user interaction.

A methodology for modeling photocatalytic reactors for indoor pollution control using previously estimated kinetic parameters

Energy Technology Data Exchange (ETDEWEB)

Passalia, Claudio; Alfano, Orlando M. [INTEC - Instituto de Desarrollo Tecnologico para la Industria Quimica, CONICET - UNL, Gueemes 3450, 3000 Santa Fe (Argentina); FICH - Departamento de Medio Ambiente, Facultad de Ingenieria y Ciencias Hidricas, Universidad Nacional del Litoral, Ciudad Universitaria, 3000 Santa Fe (Argentina); Brandi, Rodolfo J., E-mail: rbrandi@santafe-conicet.gov.ar [INTEC - Instituto de Desarrollo Tecnologico para la Industria Quimica, CONICET - UNL, Gueemes 3450, 3000 Santa Fe (Argentina); FICH - Departamento de Medio Ambiente, Facultad de Ingenieria y Ciencias Hidricas, Universidad Nacional del Litoral, Ciudad Universitaria, 3000 Santa Fe (Argentina)

2012-04-15

Highlights: Black-Right-Pointing-Pointer Indoor pollution control via photocatalytic reactors. Black-Right-Pointing-Pointer Scaling-up methodology based on previously determined mechanistic kinetics. Black-Right-Pointing-Pointer Radiation interchange model between catalytic walls using configuration factors. Black-Right-Pointing-Pointer Modeling and experimental validation of a complex geometry photocatalytic reactor. - Abstract: A methodology for modeling photocatalytic reactors for their application in indoor air pollution control is carried out. The methodology implies, firstly, the determination of intrinsic reaction kinetics for the removal of formaldehyde. This is achieved by means of a simple geometry, continuous reactor operating under kinetic control regime and steady state. The kinetic parameters were estimated from experimental data by means of a nonlinear optimization algorithm. The second step was the application of the obtained kinetic parameters to a very different photoreactor configuration. In this case, the reactor is a corrugated wall type using nanosize TiO{sub 2} as catalyst irradiated by UV lamps that provided a spatially uniform radiation field. The radiative transfer within the reactor was modeled through a superficial emission model for the lamps, the ray tracing method and the computation of view factors. The velocity and concentration fields were evaluated by means of a commercial CFD tool (Fluent 12) where the radiation model was introduced externally. The results of the model were compared experimentally in a corrugated wall, bench scale reactor constructed in the laboratory. The overall pollutant conversion showed good agreement between model predictions and experiments, with a root mean square error less than 4%.

Colony geometry and structural complexity of the endangered species Acropora cervicornis partly explains the structure of their associated fish assemblage.

Science.gov (United States)

Agudo-Adriani, Esteban A; Cappelletto, Jose; Cavada-Blanco, Francoise; Croquer, Aldo

2016-01-01

In the past decade, significant efforts have been made to describe fish-habitat associations. However, most studies have oversimplified actual connections between fish assemblages and their habitats by using univariate correlations. The purpose of this study was to identify the features of habitat forming corals that facilitate and influences assemblages of associated species such as fishes. For this we developed three-dimensional models of colonies of Acropora cervicornis to estimate geometry (length and height), structural complexity (i.e., volume, density of branches, etc.) and biological features of the colonies (i.e., live coral tissue, algae). We then correlated these colony characteristics with the associated fish assemblage using multivariate analyses. We found that geometry and complexity were better predictors of the structure of fish community, compared to other variables such as percentage of live coral tissue or algae. Combined, the geometry of each colony explained 40% of the variability of the fish assemblage structure associated with this coral species; 61% of the abundance and 69% of fish richness, respectively. Our study shows that three-dimensional reconstructions of discrete colonies of Acropora cervicornis provides a useful description of the colonial structural complexity and may explain a great deal of the variance in the structure of the associated coral reef fish community. This demonstration of the strongly trait-dependent ecosystem role of this threatened species has important implications for restoration and conservation efforts.

Baikal-1 stand complex. Preparation and carrying out of the first energy start-up of the IVG-1 reactor

International Nuclear Information System (INIS)

Tikhomirov, L.N.

1995-01-01

The IVG-1 reactor was a first ground prototype of nuclear rocket engine. The reactor was built on the site 10 of the Semipalatinsk test site. Since the first energy start-up in 1975 the reactor was exploited 14 years till its modernization in 1989. The Bajkal-1 stand complex was designed and built for the carrying out of tests for fuel assemblies of different modifications. The energy start-up has been sum of long creative work of different research and constructive staffs on creation of high-temperature gas-cooled IVG-1 reactor. The history of construction, project and assembling of the stand complex is presented. Complex start and put works were carried out in the December 1974. Control physical start-up was carried out in the January 1975. Cold start-up by hydrogen was in the February 1975. Hot start-up was in the March 1975. The result of the hot start-up was experimental confirmation of metodics of thermohydrovlical estimations. 2 figs., 3 tabs

Molten salt reactors: reactor cores

International Nuclear Information System (INIS)

1983-01-01

In this critical analysis of the MSBR I project are examined the problems concerning the reactor core. Advantages of breeding depend essentially upon solutions to technological problems like continuous reprocessing or graphite behavior under neutron irradiation. Graphite deformation, moderator unloading, control rods and core instrumentation require more studies. Neutronics of the core, influence of core geometry and salt composition, fuel evolution, and thermohydraulics are reviewed [fr

Broad-Application Test Reactor

International Nuclear Information System (INIS)

Motloch, C.G.

1992-05-01

This report is about a new, safe, and operationally efficient DOE reactor of nuclear research and testing proposed for the early to mid- 21st Century. Dubbed the Broad-Application Test Reactor (BATR), the proposed facility incorporates a multiple-application, multiple-mission design to support DOE programs such as naval reactors and space power and propulsion, as well as research in medical, science, isotope, and electronics arenas. DOE research reactors are aging, and implementing major replacement projects requires long lead times. Primary design drivers include safety, low risk, minimum operation cost, mission flexibility, waste minimization, and long life. Scientists and engineers at the Idaho National Engineering Laboratory are evaluating possible fuel forms, structural materials, reactor geometries, coolants, and moderators

High-Order Finite-Difference Solution of the Poisson Equation Involving Complex Geometries in Embedded Meshes

Science.gov (United States)

Marques, Alexandre; Nave, Jean-Christophe; Rosales, Ruben

2011-11-01

The Poisson equation is of central importance in the description of fluid flows and other physical phenomena. In prior work, Marques, Nave, and Rosales introduced the Correction Function Method (CFM) to obtain fourth-order accurate solutions for the constant coefficient Poisson problem with prescribed jump conditions for the solution and its normal derivative across arbitrary interfaces. Here we combine this method with the ideas introduced by Mayo to solve other Poisson problems involving complex geometries. In summary, we are able to rewrite the problem as a boundary integral equation in terms of a potential distribution over the boundary or interface. The solution of this integral equation is discontinuous across the boundary or interface. Hence, after this integral equation is solved using standard techniques, the potential distribution can be used to determine the jump discontinuities. We are then able to use the CFM to solve the resulting Poisson equation with jump discontinuities. The outcome is a fourth-order accurate scheme to solve general Poisson problems which, over arbitrary geometries, has a cost that is approximately twice that of a fast Poisson solver using FFT on a rectangular geometry of the same size. Details of the method and applications will be presented.

3-DB, 3-D Multigroup Diffusion, X-Y-Z, R-Theta-Z, Triangular-Z Geometry, Fast Reactor Burnup

International Nuclear Information System (INIS)

Hardie, R.W.; Little, W.W. Jr.; Mroz, W.

1974-01-01

1 - Description of problem or function: 3DB is a three-dimensional (x-y-z, r-theta-z, triangular-z) multigroup diffusion code for use in detailed fast-reactor criticality and burnup analysis. The code can be used to - (a) compute k eff and perform criticality searches on time absorption, reactor composition, and reactor dimensions by means of either a flux or an adjoint model, (b) compute material burnup using a flexible material shuffling scheme, and (c) compute flux distributions for an arbitrary extraneous source. 2 - Method of solution: Eigenvalues are computed by standard source- iteration techniques. Group re-balancing and successive over-relaxation with line inversion are used to accelerate convergence. Adjoint solutions are obtained by inverting the input data and redefining the source terms. Material burnup is by reactor zone. The burnup rate is determined by the zone and energy-averaged cross sections which are recomputed after each time-step. The isotopic chains, which can contain any number of isotopes are formed by the user. The code does not contain built- in or internal chains. 3 - Restrictions on the complexity of the problem: Since variable dimensioning is employed, no simple bounds can be stated

FURNACE; a toroidal geometry neutronic program system method description and users manual

International Nuclear Information System (INIS)

Verschuur, K.A.

1984-12-01

The FURNACE program system performs neutronic and photonic calculations in 3D toroidal geometry for application to fusion reactors. The geometry description is quite general, allowing any torus cross section and any neutron source density distribution for the plasma, as well as simple parametric representations of circular, elliptic and D-shaped tori and plasmas. The numerical method is based on an approximate transport model that produces results with sufficient accuracy for reactor-design purposes, at acceptable calculational costs. A short description is given of the numerical method, and a user manual for the programs of the system: FURNACE, ANISN-PT, LIBRA, TAPEMA and DRAWER is presented

04 - Sodium cooled fast breeder fourth-generation reactors - The experimental reactor ALLEGRO, the other ways for fast breeder fourth-generation reactors

International Nuclear Information System (INIS)

2012-12-01

The authors first present the technology of gas-cooled fast breeder reactors (basic principles, specific innovations, feasibility studies, fuel element, safety) and notably the ALLEGRO project (design options and expected performances, preliminary safety demonstration). Then, they present the lead-cooled fast-breeder reactor technology: interests and obstacles, return on experience, the issue of lead density, neutron assessment, transmutation potential, dosimetry, safety chemical properties and compatibility with the fuel, water, air and steels. The next part addresses the technology of molten-salt fast-breeder reactors: choice of the liquid fuel and geometry, reactor concept (difficulties, lack of past R and D), demonstration and demonstrators, international context

Combinatorial geometry domain decomposition strategies for Monte Carlo simulations

Energy Technology Data Exchange (ETDEWEB)

Li, G.; Zhang, B.; Deng, L.; Mo, Z.; Liu, Z.; Shangguan, D.; Ma, Y.; Li, S.; Hu, Z. [Institute of Applied Physics and Computational Mathematics, Beijing, 100094 (China)

2013-07-01

Analysis and modeling of nuclear reactors can lead to memory overload for a single core processor when it comes to refined modeling. A method to solve this problem is called 'domain decomposition'. In the current work, domain decomposition algorithms for a combinatorial geometry Monte Carlo transport code are developed on the JCOGIN (J Combinatorial Geometry Monte Carlo transport INfrastructure). Tree-based decomposition and asynchronous communication of particle information between domains are described in the paper. Combination of domain decomposition and domain replication (particle parallelism) is demonstrated and compared with that of MERCURY code. A full-core reactor model is simulated to verify the domain decomposition algorithms using the Monte Carlo particle transport code JMCT (J Monte Carlo Transport Code), which has being developed on the JCOGIN infrastructure. Besides, influences of the domain decomposition algorithms to tally variances are discussed. (authors)

Combinatorial geometry domain decomposition strategies for Monte Carlo simulations

International Nuclear Information System (INIS)

Li, G.; Zhang, B.; Deng, L.; Mo, Z.; Liu, Z.; Shangguan, D.; Ma, Y.; Li, S.; Hu, Z.

2013-01-01

Analysis and modeling of nuclear reactors can lead to memory overload for a single core processor when it comes to refined modeling. A method to solve this problem is called 'domain decomposition'. In the current work, domain decomposition algorithms for a combinatorial geometry Monte Carlo transport code are developed on the JCOGIN (J Combinatorial Geometry Monte Carlo transport INfrastructure). Tree-based decomposition and asynchronous communication of particle information between domains are described in the paper. Combination of domain decomposition and domain replication (particle parallelism) is demonstrated and compared with that of MERCURY code. A full-core reactor model is simulated to verify the domain decomposition algorithms using the Monte Carlo particle transport code JMCT (J Monte Carlo Transport Code), which has being developed on the JCOGIN infrastructure. Besides, influences of the domain decomposition algorithms to tally variances are discussed. (authors)

Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1; Simulacion de un reactor FBR con geometria hexagonal-Z usando el codigo PARCS 3.1

Energy Technology Data Exchange (ETDEWEB)

Reyes F, M. C.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, U.P. Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Filio L, C., E-mail: rf.melisa@gmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

2013-10-15

The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S{sub 2} and P{sub 1}. Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

Comparative analysis of a fusion reactor blanket in cylindrical and toroidal geometry using Monte Carlo

International Nuclear Information System (INIS)

Chapin, D.L.

1976-03-01

Differences in neutron fluxes and nuclear reaction rates in a noncircular fusion reactor blanket when analyzed in cylindrical and toroidal geometry are studied using Monte Carlo. The investigation consists of three phases--a one-dimensional calculation using a circular approximation to a hexagonal shaped blanket; a two-dimensional calculation of a hexagonal blanket in an infinite cylinder; and a three-dimensional calculation of the blanket in tori of aspect ratios 3 and 5. The total blanket reaction rate in the two-dimensional model is found to be in good agreement with the circular model. The toroidal calculations reveal large variations in reaction rates at different blanket locations as compared to the hexagonal cylinder model, although the total reaction rate is nearly the same for both models. It is shown that the local perturbations in the toroidal blanket are due mainly to volumetric effects, and can be predicted by modifying the results of the infinite cylinder calculation by simple volume factors dependent on the blanket location and the torus major radius

Startup of Torrey Pines Mark III and Puerto Rico Nuclear Center reactors with TRIGA-FLIP fuel

Energy Technology Data Exchange (ETDEWEB)

Chesworth, R. H. [Gulf E and ES, San Diego, CA (United States)

1972-07-01

This paper discusses the characteristics of TRIGA FLIP cores in two different geometries: the normal TRIGA single-rod geometry as typified by the installation in the Torrey Pines Mark III reactor; and the four-rod cluster geometry as typified by the conversion core installed in the Puerto Rico Nuclear Center reactor at Mayaguez. In both reactors the fuel is 8-1/2 wt % uranium, 70% enriched in U-235. The hydrogen to zirconium atom ratio is 1.5 to 1.65 and the cladding material is stainless steel. The basic neutronic characteristics of the fuel in both reactor installations are briefly discussed.

The Y-12 National Security Complex Foreign Research Reactor Uranium Supply Production

Energy Technology Data Exchange (ETDEWEB)

Nelson, T. [Nuclear Technology and Nonproliferation Programs, B and W Y-12, L.L.C., Y-12 National Security Complex, Oak Ridge, Tennessee (United States); Keller, A.P. [Disposition and Supply Programs, B and W Y-12, L.L.C., Y-12 National Security Complex, Oak Ridge, Tennessee (United States)

2011-07-01

The Foreign Research Reactor (FRR) Uranium Supply Program at the Y-12 National Security Complex supports the nonproliferation objectives of the National Nuclear Security Administration (NNSA) HEU Disposition, the Reduced Enrichment Research and Test Reactors (RERTR), and the United States (U.S.) FRR Spent Nuclear Fuel (SNF) Acceptance Programs. The FRR Supply Program supports the important U.S. government nuclear nonproliferation commitment to serve as a reliable and cost-effective uranium supplier for those foreign research reactors that are converting or have converted to Low-Enriched Uranium (LEU) fuel under the RERTR Program. The NNSA Y-12 Site Office maintains the prime contracts with foreign government agencies for the supply of LEU for their research reactors. The LEU is produced by down blending Highly Enriched Uranium (HEU) that has been declared surplus to the U.S. national defense needs. The down blending and sale of the LEU supports the Surplus HEU Disposition Program Record of Decision to make the HEU non-weapons usable and to recover the economic value of the uranium to the extent feasible. In addition to uranium metal feedstock for fuel fabrication, Y-12 can produce LEU in different forms to support new fuel development or target fabrication for medical isotope production. With production improvements and efficient delivery preparations, Y-12 continues to successfully support the global research reactor community. (author)

ALGOL geometrical module for reactor and reactor cell calculations in the R-Z geometry with the Monte Carlo method

International Nuclear Information System (INIS)

Usikov, D.A.

1975-01-01

A description of a geometrical module used in a program of the ARMONT complex of the Monte Carlo calculations is given. The geometrical module is designed to simulate the particle trajectory in the R-Z geometry. The geometrical module follows the particle trajectory from the start point to the next collision or flight-out points. The flight direction at the scattering point is assumed isotropic in the laboratory coordinate system. In the module the angle between the flight direction before and after collision is not determined. The principles for the module construction are presented alongside with the text-module in the ALGOL language. The module is optimumized as to the counting rate and it is rather compact not to cause difficulties due to the translator limitations in common translation with other program blocks based on the use of the Monte Carlo calculations

The analytic nodal method in cylindrical geometry

International Nuclear Information System (INIS)

Prinsloo, Rian H.; Tomasevic, Djordje I.

2008-01-01

Nodal diffusion methods have been used extensively in nuclear reactor calculations, specifically for their performance advantage, but also for their superior accuracy. More specifically, the Analytic Nodal Method (ANM), utilising the transverse integration principle, has been applied to numerous reactor problems with much success. In this work, a nodal diffusion method is developed for cylindrical geometry. Application of this method to three-dimensional (3D) cylindrical geometry has never been satisfactorily addressed and we propose a solution which entails the use of conformal mapping. A set of 1D-equations with an adjusted, geometrically dependent, inhomogeneous source, is obtained. This work describes the development of the method and associated test code, as well as its application to realistic reactor problems. Numerical results are given for the PBMR-400 MW benchmark problem, as well as for a 'cylindrisized' version of the well-known 3D LWR IAEA benchmark. Results highlight the improved accuracy and performance over finite-difference core solutions and investigate the applicability of nodal methods to 3D PBMR type problems. Results indicate that cylindrical nodal methods definitely have a place within PBMR applications, yielding performance advantage factors of 10 and 20 for 2D and 3D calculations, respectively, and advantage factors of the order of 1000 in the case of the LWR problem

Colony geometry and structural complexity of the endangered species Acropora cervicornis partly explains the structure of their associated fish assemblage

Directory of Open Access Journals (Sweden)

Esteban A. Agudo-Adriani

2016-04-01

Full Text Available In the past decade, significant efforts have been made to describe fish-habitat associations. However, most studies have oversimplified actual connections between fish assemblages and their habitats by using univariate correlations. The purpose of this study was to identify the features of habitat forming corals that facilitate and influences assemblages of associated species such as fishes. For this we developed three-dimensional models of colonies of Acropora cervicornis to estimate geometry (length and height, structural complexity (i.e., volume, density of branches, etc. and biological features of the colonies (i.e., live coral tissue, algae. We then correlated these colony characteristics with the associated fish assemblage using multivariate analyses. We found that geometry and complexity were better predictors of the structure of fish community, compared to other variables such as percentage of live coral tissue or algae. Combined, the geometry of each colony explained 40% of the variability of the fish assemblage structure associated with this coral species; 61% of the abundance and 69% of fish richness, respectively. Our study shows that three-dimensional reconstructions of discrete colonies of Acropora cervicornis provides a useful description of the colonial structural complexity and may explain a great deal of the variance in the structure of the associated coral reef fish community. This demonstration of the strongly trait-dependent ecosystem role of this threatened species has important implications for restoration and conservation efforts.

Shape optimization of a sodium cooled fast reactor

International Nuclear Information System (INIS)

Schmitt, D.; Allaire, G.; Pantz, O.; Pozin, N.

2013-01-01

Traditional designs of sodium cooled fast reactors have a positive sodium expansion feedback. During a loss of flow transient without scram, sodium heating and boiling thus insert a positive reactivity and prevents the power from decreasing. Recent studies led at CEA, AREVA and EDF show that cores with complex geometries can feature a very low or even a negative sodium void worth. Usual optimization methods for core conception are based on a parametric description of a given core design. New core concepts and shapes can then only be found by hand. Shape optimization methods have proven very efficient in the conception of optimal structures under thermal or mechanical constraints. First studies show that these methods could be applied to sodium cooled core conception. In this paper, a shape optimization method is applied to the conception of a sodium cooled fast reactor core with low sodium void worth. An objective function to be minimized is defined. It includes the reactivity change induced by a 1% sodium density decrease. The optimization variable is a displacement field changing the core geometry from one shape to another. Additionally, a parametric optimization of the plutonium content distribution of the core is made, so as to ensure that the core is kept critical, and that the power shape is flat enough. The final shape obtained must then be adjusted to a given realistic core layout. Its characteristics can be checked with reference neutronic codes such as ERANOS. Thanks to this method, new shapes of reactor cores could be inferred, and lead to new design ideas. (authors)

Calculation of power density with MCNP in TRIGA reactor

International Nuclear Information System (INIS)

Snoj, L.; Ravnik, M.

2006-01-01

Modern Monte Carlo codes (e.g. MCNP) allow calculation of power density distribution in 3-D geometry assuming detailed geometry without unit-cell homogenization. To normalize MCNP calculation by the steady-state thermal power of a reactor, one must use appropriate scaling factors. The description of the scaling factors is not adequately described in the MCNP manual and requires detailed knowledge of the code model. As the application of MCNP for power density calculation in TRIGA reactors has not been reported in open literature, the procedure of calculating power density with MCNP and its normalization to the power level of a reactor is described in the paper. (author)

A ghost-cell immersed boundary method for flow in complex geometry

International Nuclear Information System (INIS)

Tseng, Y.-H.; Ferziger, Joel H.

2003-01-01

An efficient ghost-cell immersed boundary method (GCIBM) for simulating turbulent flows in complex geometries is presented. A boundary condition is enforced through a ghost cell method. The reconstruction procedure allows systematic development of numerical schemes for treating the immersed boundary while preserving the overall second-order accuracy of the base solver. Both Dirichlet and Neumann boundary conditions can be treated. The current ghost cell treatment is both suitable for staggered and non-staggered Cartesian grids. The accuracy of the current method is validated using flow past a circular cylinder and large eddy simulation of turbulent flow over a wavy surface. Numerical results are compared with experimental data and boundary-fitted grid results. The method is further extended to an existing ocean model (MITGCM) to simulate geophysical flow over a three-dimensional bump. The method is easily implemented as evidenced by our use of several existing codes

Experimental Approach for the Uncertainty Assessment of 3D Complex Geometry Dimensional Measurements Using Computed Tomography at the mm and Sub-mm Scales.

Science.gov (United States)

Jiménez, Roberto; Torralba, Marta; Yagüe-Fabra, José A; Ontiveros, Sinué; Tosello, Guido

2017-05-16

The dimensional verification of miniaturized components with 3D complex geometries is particularly challenging. Computed Tomography (CT) can represent a suitable alternative solution to micro metrology tools based on optical and tactile techniques. However, the establishment of CT systems' traceability when measuring 3D complex geometries is still an open issue. In this work, an alternative method for the measurement uncertainty assessment of 3D complex geometries by using CT is presented. The method is based on the micro-CT system Maximum Permissible Error (MPE) estimation, determined experimentally by using several calibrated reference artefacts. The main advantage of the presented method is that a previous calibration of the component by a more accurate Coordinate Measuring System (CMS) is not needed. In fact, such CMS would still hold all the typical limitations of optical and tactile techniques, particularly when measuring miniaturized components with complex 3D geometries and their inability to measure inner parts. To validate the presented method, the most accepted standard currently available for CT sensors, the Verein Deutscher Ingenieure/Verband Deutscher Elektrotechniker (VDI/VDE) guideline 2630-2.1 is applied. Considering the high number of influence factors in CT and their impact on the measuring result, two different techniques for surface extraction are also considered to obtain a realistic determination of the influence of data processing on uncertainty. The uncertainty assessment of a workpiece used for micro mechanical material testing is firstly used to confirm the method, due to its feasible calibration by an optical CMS. Secondly, the measurement of a miniaturized dental file with 3D complex geometry is carried out. The estimated uncertainties are eventually compared with the component's calibration and the micro manufacturing tolerances to demonstrate the suitability of the presented CT calibration procedure. The 2U/T ratios resulting from the

Energy deposition in STARFIRE reactor components

International Nuclear Information System (INIS)

Gohar, Y.; Brooks, J.N.

1985-04-01

The energy deposition in the STARFIRE commercial tokamak reactor was calculated based on detailed models for the different reactor components. The heat deposition and the 14 MeV neutron flux poloidal distributions in the first wall were obtained. The poloidal surface heat load distribution in the first wall was calculated from the plasma radiation. The Monte Carlo method was used for the calculation to allow an accurate modeling for the reactor geometry

Latest developments in prestressed concrete vessels for gas-cooled reactors

International Nuclear Information System (INIS)

Ople, F.S. Jr.

1979-01-01

This paper is an update of the design development of prestressed concrete vessels, commonly referred to as 'PCRVs' starting with the first single-cavity PCRV for the Fort St. Vrain Nuclear Generating Station to the latest multi-cavity PCRV configurations being utilized as the primary reactor vessels for both the High Temperature Gas-Cooled Reactor (HTGR) and the Gas-Cooled Fast Breeder Reactor (GCFR) in the U.S.A. The complexity of PCRV design varies not only due to the type of vessel configuration (single versus multi-cavity) but also on the application to the specific type of reactor concept. PCRV technology as applied to the Steam Cycle HTGR is fairly well established; however, some significant technical complexities are associated with PCRV design for the Gas Turbine HTGR and the GCFR. For the Gas Turbine HTGR, for instance, the fluid dynamics of the turbo-machinery cause multi-pressure conditions to exist in various portions of the power conversion loops during operation. This condition complicates the design approach and the proof test specification for the PCRV. The geometric configuration of the multi-cavity PCRV is also more complex due to the introduction of large horizontal cylindrical cavities (housing the turbo/machines for the Gas Turbine HTGR and circulators for the GCFR) in addition to the vertical cylindrical cavities for the core and heat exchangers. Because of this complex geometry, it becomes difficult to achieve an optimum prestressing arrangement for the PCRV. Other novel features of the multi-cavity PCRV resulting from the continuing design optimization effort are the incorporation of an asymmetric (offset core) configuration and the use of large vessel cavity/penetration concrete closures directly held down by prestressing tendons for both economic and safety reasons. (orig.)

Development of a numerical methodology for flowforming process simulation of complex geometry tubes

Science.gov (United States)

Varela, Sonia; Santos, Maite; Arroyo, Amaia; Pérez, Iñaki; Puigjaner, Joan Francesc; Puigjaner, Blanca

2017-10-01

Nowadays, the incremental flowforming process is widely explored because of the usage of complex tubular products is increasing due to the light-weighting trend and the use of expensive materials. The enhanced mechanical properties of finished parts combined with the process efficiency in terms of raw material and energy consumption are the key factors for its competitiveness and sustainability, which is consistent with EU industry policy. As a promising technology, additional steps for extending the existing flowforming limits in the production of tubular products are required. The objective of the present research is to further expand the current state of the art regarding limitations on tube thickness and diameter, exploring the feasibility to flowform complex geometries as tubes of elevated thickness of up to 60 mm. In this study, the analysis of the backward flowforming process of 7075 aluminum tubular preform is carried out to define the optimum process parameters, machine requirements and tooling geometry as demonstration case. Numerical simulation studies on flowforming of thin walled tubular components have been considered to increase the knowledge of the technology. The calculation of the rotational movement of the mesh preform, the high ratio thickness/length and the thermomechanical condition increase significantly the computation time of the numerical simulation model. This means that efficient and reliable tools able to predict the forming loads and the quality of flowformed thick tubes are not available. This paper aims to overcome this situation by developing a simulation methodology based on FEM simulation code including new strategies. Material characterization has also been performed through tensile test to able to design the process. Finally, to check the reliability of the model, flowforming tests at industrial environment have been developed.

Investigation of the applicability of MCNP code to complicated geometries

International Nuclear Information System (INIS)

Higuchi, Kenji; Yamaguchi, Yukichi

1994-03-01

Applicability of MCNP code, which is a general purpose Monte Carlo code for particle transport problems, to complicated geometries, has been investigated as a study in Human Acts Simulation Program (HASP), in which basic studies for intelligent robot for patrol and inspection of nuclear facilities are being performed. In HASP, basic software systems simulating the behavior of intelligent robot of human shape working in Japan Research Reactor No.3 are being developed. The aim of Dose Evaluation system in HASP is to establish the methodology to evaluate irradiation damage of the LSI/VLSI circuits embedded within a robot body and to give design criteria of intelligent robot. Monte Carlo method is used to solve particle transport problem in a complicated geometry such as robot body. Preliminary evaluation to establish the methodology has been conducted using continuous energy Monte Carlo code, MCNP with the anthropomorphic phantom. The phantom has the same degree of geometric complexity as robot body and is widely used for the calculation of the effective dose equivalent for radiological protection. It allowed us to verify the validity of the methodology by comparison of calculation results with the data in ICRP Pub. 51. In this report, the method used in the calculation of effective dose equivalent, visualization system supporting visualization of input data for complicated geometry and the results in the evaluation of validity of the method by the comparison of the calculated results with the data in the ICRP publication are described. (author)

PENTrack-a simulation tool for ultracold neutrons, protons, and electrons in complex electromagnetic fields and geometries

Science.gov (United States)

Schreyer, W.; Kikawa, T.; Losekamm, M. J.; Paul, S.; Picker, R.

2017-06-01

Modern precision experiments trapping low-energy particles require detailed simulations of particle trajectories and spin precession to determine systematic measurement limitations and apparatus deficiencies. We developed PENTrack, a tool that allows to simulate trajectories of ultracold neutrons and their decay products-protons and electrons-and the precession of their spins in complex geometries and electromagnetic fields. The interaction of ultracold neutrons with matter is implemented with the Fermi-potential formalism and diffuse scattering using Lambert and microroughness models. The results of several benchmark simulations agree with STARucn v1.2, uncovered several flaws in Geant4 v10.2.2, and agree with experimental data. Experiment geometry and electromagnetic fields can be imported from commercial computer-aided-design and finite-element software. All simulation parameters are defined in simple text files allowing quick changes. The simulation code is written in C++ and is freely available at github.com/wschreyer/PENTrack.git.

A Qualitative Investigation of Deposition Velocities of a Non-Newtonian Slurry in Complex Pipeline Geometries

Energy Technology Data Exchange (ETDEWEB)

Yokuda, Satoru T.; Poloski, Adam P.; Adkins, Harold E.; Casella, Andrew M.; Hohimer, Ryan E.; Karri, Naveen K.; Luna, Maria; Minette, Michael J.; Tingey, Joel M.

2009-05-11

The External Flowsheet Review Team (EFRT) has identified the issues relating to the Waste Treatment and Immobilization Plant (WTP) pipe plugging. Per the review’s executive summary, “Piping that transports slurries will plug unless it is properly designed to minimize this risk. This design approach has not been followed consistently, which will lead to frequent shutdowns due to line plugging.” To evaluate the potential for plugging, testing was performed to determine critical velocities for the complex WTP piping layout. Critical velocity is defined as the point at which a moving bed of particles begins to form on the pipe bottom during slurry-transport operations. Pressure drops across the fittings of the test pipeline were measured with differential pressure transducers, from which the critical velocities were determined. A WTP prototype flush system was installed and tested upon the completion of the pressure-drop measurements. We also provide the data for the overflow relief system represented by a WTP complex piping geometry with a non-Newtonian slurry. A waste simulant composed of alumina (nominally 50 μm in diameter) suspended in a kaolin clay slurry was used for this testing. The target composition of the simulant was 10 vol% alumina in a suspending medium with a yield stress of 3 Pa. No publications or reports are available to confirm the critical velocities for the complex geometry evaluated in this testing; therefore, for this assessment, the results were compared to those reported by Poloski et al. (2008) for which testing was performed for a straight horizontal pipe. The results of the flush test are compared to the WTP design guide 24590-WTP-GPG-M-0058, Rev. 0 (Hall 2006) in an effort to confirm flushing-velocity requirements.

A Qualitative Investigation of Deposition Velocities of a Non-Newtonian Slurry in Complex Pipeline Geometries

International Nuclear Information System (INIS)

Yokuda, Satoru T.; Poloski, Adam P.; Adkins, Harold E.; Casella, Andrew M.; Hohimer, Ryan E.; Karri, Naveen K.; Luna, Maria; Minette, Michael J.; Tingey, Joel M.

2009-01-01

The External Flowsheet Review Team (EFRT) has identified the issues relating to the Waste Treatment and Immobilization Plant (WTP) pipe plugging. Per the review's executive summary, ''Piping that transports slurries will plug unless it is properly designed to minimize this risk. This design approach has not been followed consistently, which will lead to frequent shutdowns due to line plugging.'' To evaluate the potential for plugging, testing was performed to determine critical velocities for the complex WTP piping layout. Critical velocity is defined as the point at which a moving bed of particles begins to form on the pipe bottom during slurry-transport operations. Pressure drops across the fittings of the test pipeline were measured with differential pressure transducers, from which the critical velocities were determined. A WTP prototype flush system was installed and tested upon the completion of the pressure-drop measurements. We also provide the data for the overflow relief system represented by a WTP complex piping geometry with a non-Newtonian slurry. A waste simulant composed of alumina (nominally 50 (micro)m in diameter) suspended in a kaolin clay slurry was used for this testing. The target composition of the simulant was 10 vol% alumina in a suspending medium with a yield stress of 3 Pa. No publications or reports are available to confirm the critical velocities for the complex geometry evaluated in this testing; therefore, for this assessment, the results were compared to those reported by Poloski et al. (2008) for which testing was performed for a straight horizontal pipe. The results of the flush test are compared to the WTP design guide 24590-WTP-GPG-M-0058, Rev. 0 (Hall 2006) in an effort to confirm flushing-velocity requirements.

Thermal-hydraulic limitations on water-cooled fusion reactor components

International Nuclear Information System (INIS)

Cha, Y.S.; Misra, B.

1986-01-01

An assessment of the cooling requirements for fusion reactor components, such as the first wall and limiter/divertor, was carried out using pressurized water as the coolant. In order to establish the coolant operating conditions, a survey of the literature on departure from nucleate boiling, critical heat flux, asymmetrical heating and heat transfer augmentation techniques was carried out. The experimental data and the empirical correlations indicate that thermal protection for the fusion reactor components based on conventional design concepts can be provided with an adequate margin of safety without resorting to either high coolant velocities, excessive coolant pressures, or heat transfer augmentation techniques. If, however, the future designs require unconventional shapes or heat transfer enhancement techniques, experimental verification would be necessary since no data on heat transfer augmentation techniques exist for complex geometries, especially under asymmetrically heated conditions. Since the data presented herein are concerned primarily with thermal protection of the reactor components, the final design should consider other factors such as thermal stresses, temperature limits, and fatigue

The geometry description markup language

International Nuclear Information System (INIS)

Chytracek, R.

2001-01-01

Currently, a lot of effort is being put on designing complex detectors. A number of simulation and reconstruction frameworks and applications have been developed with the aim to make this job easier. A very important role in this activity is played by the geometry description of the detector apparatus layout and its working environment. However, no real common approach to represent geometry data is available and such data can be found in various forms starting from custom semi-structured text files, source code (C/C++/FORTRAN), to XML and database solutions. The XML (Extensible Markup Language) has proven to provide an interesting approach for describing detector geometries, with several different but incompatible XML-based solutions existing. Therefore, interoperability and geometry data exchange among different frameworks is not possible at present. The author introduces a markup language for geometry descriptions. Its aim is to define a common approach for sharing and exchanging of geometry description data. Its requirements and design have been driven by experience and user feedback from existing projects which have their geometry description in XML

Fracture mechanics and residual fatigue life analysis for complex stress fields. Technical report

International Nuclear Information System (INIS)

Besuner, P.M.

1975-07-01

This report reviews the development and application of an influence function method for calculating stress intensity factors and residual fatigue life for two- and three-dimensional structures with complex stress fields and geometries. Through elastic superposition, the method properly accounts for redistribution of stress as the crack grows through the structure. The analytical methods used and the computer programs necessary for computation and application of load independent influence functions are presented. A new exact solution is obtained for the buried elliptical crack, under an arbitrary Mode I stress field, for stress intensity factors at four positions around the crack front. The IF method is then applied to two fracture mechanics problems with complex stress fields and geometries. These problems are of current interest to the electric power generating industry and include (1) the fatigue analysis of a crack in a pipe weld under nominal and residual stresses and (2) fatigue analysis of a reactor pressure vessel nozzle corner crack under a complex bivariate stress field

Statistical properties of reactor antineutrinos

CERN Document Server

Rusov, V D; Tarasov, V O; Shaaban, Y

2002-01-01

Based on the properties of the cascade statistics of reactor antineutrinos, the efficient method of searching for neutrino oscillations is offered. The determination of physical parameters of this statistics, i.e. the average number of fissions and the overage number of antineutrinos per fission, requires no a priori knowledge of the geometry and characteristics of the detector, the reactor power, and composition of nuclear fuel.

Modeling a Packed Bed Reactor Utilizing the Sabatier Process

Science.gov (United States)

Shah, Malay G.; Meier, Anne J.; Hintze, Paul E.

2017-01-01

A numerical model is being developed using Python which characterizes the conversion and temperature profiles of a packed bed reactor (PBR) that utilizes the Sabatier process; the reaction produces methane and water from carbon dioxide and hydrogen. While the specific kinetics of the Sabatier reaction on the RuAl2O3 catalyst pellets are unknown, an empirical reaction rate equation1 is used for the overall reaction. As this reaction is highly exothermic, proper thermal control is of the utmost importance to ensure maximum conversion and to avoid reactor runaway. It is therefore necessary to determine what wall temperature profile will ensure safe and efficient operation of the reactor. This wall temperature will be maintained by active thermal controls on the outer surface of the reactor. Two cylindrical PBRs are currently being tested experimentally and will be used for validation of the Python model. They are similar in design except one of them is larger and incorporates a preheat loop by feeding the reactant gas through a pipe along the center of the catalyst bed. The further complexity of adding a preheat pipe to the model to mimic the larger reactor is yet to be implemented and validated; preliminary validation is done using the smaller PBR with no reactant preheating. When mapping experimental values of the wall temperature from the smaller PBR into the Python model, a good approximation of the total conversion and temperature profile has been achieved. A separate CFD model incorporates more complex three-dimensional effects by including the solid catalyst pellets within the domain. The goal is to improve the Python model to the point where the results of other reactor geometry can be reasonably predicted relatively quickly when compared to the much more computationally expensive CFD approach. Once a reactor size is narrowed down using the Python approach, CFD will be used to generate a more thorough prediction of the reactors performance.

Spinning geometry = Twisted geometry

International Nuclear Information System (INIS)

Freidel, Laurent; Ziprick, Jonathan

2014-01-01

It is well known that the SU(2)-gauge invariant phase space of loop gravity can be represented in terms of twisted geometries. These are piecewise-linear-flat geometries obtained by gluing together polyhedra, but the resulting geometries are not continuous across the faces. Here we show that this phase space can also be represented by continuous, piecewise-flat three-geometries called spinning geometries. These are composed of metric-flat three-cells glued together consistently. The geometry of each cell and the manner in which they are glued is compatible with the choice of fluxes and holonomies. We first remark that the fluxes provide each edge with an angular momentum. By studying the piecewise-flat geometries which minimize edge lengths, we show that these angular momenta can be literally interpreted as the spin of the edges: the geometries of all edges are necessarily helices. We also show that the compatibility of the gluing maps with the holonomy data results in the same conclusion. This shows that a spinning geometry represents a way to glue together the three-cells of a twisted geometry to form a continuous geometry which represents a point in the loop gravity phase space. (paper)

Oklo reactors and implications for nuclear science

OpenAIRE

Davis, E. D.; Gould, C. R.; Sharapov, E. I.

2014-01-01

We summarize the nuclear physics interests in the Oklo natural nuclear reactors, focusing particularly on developments over the past two decades. Modeling of the reactors has become increasingly sophisticated, employing Monte Carlo simulations with realistic geometries and materials that can generate both the thermal and epithermal fractions. The water content and the temperatures of the reactors have been uncertain parameters. We discuss recent work pointing to lower temperatures than earlie...

Lectures on Symplectic Geometry

CERN Document Server

Silva, Ana Cannas

2001-01-01

The goal of these notes is to provide a fast introduction to symplectic geometry for graduate students with some knowledge of differential geometry, de Rham theory and classical Lie groups. This text addresses symplectomorphisms, local forms, contact manifolds, compatible almost complex structures, Kaehler manifolds, hamiltonian mechanics, moment maps, symplectic reduction and symplectic toric manifolds. It contains guided problems, called homework, designed to complement the exposition or extend the reader's understanding. There are by now excellent references on symplectic geometry, a subset of which is in the bibliography of this book. However, the most efficient introduction to a subject is often a short elementary treatment, and these notes attempt to serve that purpose. This text provides a taste of areas of current research and will prepare the reader to explore recent papers and extensive books on symplectic geometry where the pace is much faster. For this reprint numerous corrections and cl...

Investigation of flow stabilization in a compact reactor vessel of a FBR. Flow visualization in a reactor vessel

International Nuclear Information System (INIS)

Sato, Hiroyuki; Igarashi, Minoru; Kimura, Nobuyuki; Kamide, Hideki

2002-01-01

In the feasibility studies of Commercialized Fast Breeder Reactor Cycle System, a compact reactor vessel is considered from economical improvement point of a sodium cooled loop type fast reactor. The flow field was visualized by water experiment for a reactor vessel with 'a column type UIS (Upper Internal Structure)', which has a slit for fuel handling mechanism and is useful for a compact fast reactor. In this research, the 1/20 scale test equipment using water was made to understand coolant flow through a slit of a column type UIS' and fundamental behavior of reactor upper plenum flow. In the flow visualization tests, tracer particles were added in the water, and illuminated by the slit-shaped pulse laser. The flow visualization image was taken with a CCD camera. We obtained fluid velocity vectors from the visualization image using the Particle Imaging Velocimetry (PIV). The results are as follows. 1. Most of coolant flow through a slit of 'column type UIS' arrived the dip plate directly. In the opposite side of a slit, most of coolant flowed toward reactor vessel wall before it arrived the dip plate. 2. The PIV was useful to measure the flow field in the reactor vessel. The obtained velocity field was consistent with the flow visualization result. 3. The jet through the UIS slit was dependent on the UIS geometry. There is a possibility to control the jet by the UIS geometry. (author)

A novel porous Ffowcs-Williams and Hawkings acoustic methodology for complex geometries

Science.gov (United States)

Nitzkorski, Zane Lloyd

Predictive noise calculations from high Reynolds number flows in complex engineering geometry are becoming a possibility with the high performance computing resources that have become available in recent years. Increasing the applicability and reliability of solution methodologies have been two key challenges toward this goal. This dissertation develops a porous Ffowcs-Williams and Hawkings methodology that uses a novel endcap methodology, and can be applied to unstructured grids. The use of unstructured grids allows complex geometry to be represented while porous formulation eliminates difficulties with the choice of acoustic Green's function. Specifically, this dissertation (1) proposes and examines a novel endcap procedure to account for spurious noise, (2) uses the proposed methodology to investigate noise production from a range of subcritical Reynolds number circular cylinders, and (3) investigates a trailing edge geometry for noise production and to illustrate the generality of the Green's function. Porous acoustic analogies need an endcap scheme in order to prevent spurious noise due to truncation errors. A dynamic end cap methodology is proposed to account for spurious contributions to the far--field sound within the context of the Ffowcs--Williams and Hawkings (FW--H) acoustic analogy. The quadrupole source terms are correlated over multiple planes to obtain a convection velocity which is then used to determine a corrective convective flux at the FW--H porous surface. The proposed approach is first demonstrated for a convecting potential vortex. The correlation is investigated by examining it pass through multiple exit planes. It is then evaluated by computing the sound emitted by flow over a circular cylinder at Reynolds number of 150 and compared to other endcap methods, such as Shur et al. [1]. Insensitivity to end plane location and spacing and the effect of the dynamic convection velocity are computed. Subcritical Reynolds number circular cylinder

Special geometry

International Nuclear Information System (INIS)

Strominger, A.

1990-01-01

A special manifold is an allowed target manifold for the vector multiplets of D=4, N=2 supergravity. These manifolds are of interest for string theory because the moduli spaces of Calabi-Yau threefolds and c=9, (2,2) conformal field theories are special. Previous work has given a local, coordinate-dependent characterization of special geometry. A global description of special geometries is given herein, and their properties are studied. A special manifold M of complex dimension n is characterized by the existence of a holomorphic Sp(2n+2,R)xGL(1,C) vector bundle over M with a nowhere-vanishing holomorphic section Ω. The Kaehler potential on M is the logarithm of the Sp(2n+2,R) invariant norm of Ω. (orig.)

PENTrack—a simulation tool for ultracold neutrons, protons, and electrons in complex electromagnetic fields and geometries

Energy Technology Data Exchange (ETDEWEB)

Schreyer, W., E-mail: w.schreyer@tum.de [Technical University of Munich, James-Franck-Str. 1, 85748 Garching (Germany); Kikawa, T. [TRIUMF, 4004 Wesbrook Mall, Vancouver (Canada); Losekamm, M.J.; Paul, S. [Technical University of Munich, James-Franck-Str. 1, 85748 Garching (Germany); Picker, R. [TRIUMF, 4004 Wesbrook Mall, Vancouver (Canada); Simon Fraser University, 8888 University Drive, Burnaby (Canada)

2017-06-21

Modern precision experiments trapping low-energy particles require detailed simulations of particle trajectories and spin precession to determine systematic measurement limitations and apparatus deficiencies. We developed PENTrack, a tool that allows to simulate trajectories of ultracold neutrons and their decay products—protons and electrons—and the precession of their spins in complex geometries and electromagnetic fields. The interaction of ultracold neutrons with matter is implemented with the Fermi-potential formalism and diffuse scattering using Lambert and microroughness models. The results of several benchmark simulations agree with STARucn v1.2, uncovered several flaws in Geant4 v10.2.2, and agree with experimental data. Experiment geometry and electromagnetic fields can be imported from commercial computer-aided-design and finite-element software. All simulation parameters are defined in simple text files allowing quick changes. The simulation code is written in C++ and is freely available at (github.com/wschreyer/PENTrack.git).

The Hitchin model, Poisson-quasi-Nijenhuis, geometry and symmetry reduction

International Nuclear Information System (INIS)

Zucchini, Roberto

2007-01-01

We revisit our earlier work on the AKSZ-like formulation of topological sigma model on generalized complex manifolds, or Hitchin model, [20]. We show that the target space geometry geometry implied by the BV master equations is Poisson-quasi-Nijenhuis geometry recently introduced and studied by Stienon and Xu (in the untwisted case) in [44]. Poisson-quasi-Nijenhuis geometry is more general than generalized complex geometry and comprises it as a particular case. Next, we show how gauging and reduction can be implemented in the Hitchin model. We find that the geometry resulting form the BV master equation is closely related to but more general than that recently described by Lin and Tolman in [40, 41], suggesting a natural framework for the study of reduction of Poisson-quasi-Nijenhuis manifolds

Complexity characterization in a probabilistic approach to dynamical systems through information geometry and inductive inference

International Nuclear Information System (INIS)

Ali, S A; Kim, D-H; Cafaro, C; Giffin, A

2012-01-01

Information geometric techniques and inductive inference methods hold great promise for solving computational problems of interest in classical and quantum physics, especially with regard to complexity characterization of dynamical systems in terms of their probabilistic description on curved statistical manifolds. In this paper, we investigate the possibility of describing the macroscopic behavior of complex systems in terms of the underlying statistical structure of their microscopic degrees of freedom by the use of statistical inductive inference and information geometry. We review the maximum relative entropy formalism and the theoretical structure of the information geometrodynamical approach to chaos on statistical manifolds M S . Special focus is devoted to a description of the roles played by the sectional curvature K M S , the Jacobi field intensity J M S and the information geometrodynamical entropy S M S . These quantities serve as powerful information-geometric complexity measures of information-constrained dynamics associated with arbitrary chaotic and regular systems defined on M S . Finally, the application of such information-geometric techniques to several theoretical models is presented.

GEOMETRY – AN IMPORTANT MEANS OF EDUCATION IN THE CIVILISATION SCOPE

OpenAIRE

Liliana TOCARIU, PhD

2017-01-01

Geometry (from the Greek: γεωμετρία; geo = earth, metria = measure) is a genuine science, rooted in mathematics, which studies the plane and spatial forms of bodies from the objective or conceptual reality and the nature of the relationship that exists between them. Due to its complexity, geometry is divided into: Euclidian geometry, analytical geometry, descriptive geometry, projective geometry, kinematic geometry, surface and curve differential geometry, axiomatic geometry,...

Multiphase flows in complex geometries: a UQ perspective

KAUST Repository

Icardi, Matteo

2015-01-01

Nowadays computer simulations are widely used in many multiphase flow applications involving interphases, dispersed particles, and complex geometries. Most of these problems are solved with mixed models composed of fundamental physical laws, rigorous mathematical upscaling, and empirical correlations/closures. This means that classical inference techniques or forward parametric studies, for example, becomes computationally prohibitive and must take into account the physical meaning and constraints of the equations. However mathematical techniques commonly used in Uncertainty Quantification can come to the aid for the (i) modeling, (ii) simulation, and (iii) validation steps. Two relevant applications for environmental, petroleum, and chemical engineering will be presented to highlight these aspects and the importance of bridging the gaps between engineering applications, computational physics and mathematical methods. The first example is related to the mathematical modeling of sub-grid/sub-scale information with Probability Density Function (PDF) models in problems involving flow, mixing, and reaction in random environment. After a short overview of the research field, some connections and similarities with Polynomial Chaos techniques, will be investigated. In the second example, averaged correlations laws and effective parameters for multiphase flow and their statistical fluctuations, will be considered and efficient computational techniques, borrowed from high-dimensional stochastic PDE problems, will be applied. In presence of interfacial flow, where small spatial scales and fast time scales are neglected, the assessment of robustness and predictive capabilities are studied. These illustrative examples are inspired by common problems arising, for example, from the modeling and simulation of turbulent and porous media flows.

Multiphase flows in complex geometries: a UQ perspective

KAUST Repository

Icardi, Matteo

2015-01-07

Nowadays computer simulations are widely used in many multiphase flow applications involving interphases, dispersed particles, and complex geometries. Most of these problems are solved with mixed models composed of fundamental physical laws, rigorous mathematical upscaling, and empirical correlations/closures. This means that classical inference techniques or forward parametric studies, for example, becomes computationally prohibitive and must take into account the physical meaning and constraints of the equations. However mathematical techniques commonly used in Uncertainty Quantification can come to the aid for the (i) modeling, (ii) simulation, and (iii) validation steps. Two relevant applications for environmental, petroleum, and chemical engineering will be presented to highlight these aspects and the importance of bridging the gaps between engineering applications, computational physics and mathematical methods. The first example is related to the mathematical modeling of sub-grid/sub-scale information with Probability Density Function (PDF) models in problems involving flow, mixing, and reaction in random environment. After a short overview of the research field, some connections and similarities with Polynomial Chaos techniques, will be investigated. In the second example, averaged correlations laws and effective parameters for multiphase flow and their statistical fluctuations, will be considered and efficient computational techniques, borrowed from high-dimensional stochastic PDE problems, will be applied. In presence of interfacial flow, where small spatial scales and fast time scales are neglected, the assessment of robustness and predictive capabilities are studied. These illustrative examples are inspired by common problems arising, for example, from the modeling and simulation of turbulent and porous media flows.

Neutron shielding studies on an advanced molten salt fast reactor design

International Nuclear Information System (INIS)

Merk, Bruno; Konheiser, Jörg

2014-01-01

Highlights: • Material damage due to irradiation has already been discovered at the MSRE. • Neutronic analysis of MSFR with curved blanket wall geometry. • Neutron fluence limit at the wall of the outer vessel can be kept for 80 years. • Shielded MSFR core will be of same dimension than a SFR core. - Abstract: The molten salt reactor technology has gained some new interest. In contrast to the historic molten salt reactors, the current projects are based on designing a molten salt fast reactor. Thus the shielding becomes significantly more challenging than in historic concepts. One very interesting and innovative result of the most recent EURATOM project on molten salt reactors – EVOL – is the fluid flow optimized design of the inner reactor vessel using curved blanket walls. The developed structure leads to a very uniform flow distribution. The design avoids all internal structures. Based on this new geometry a model for neutron physics calculation is presented. The major steps are: the modeling of the curved geometry in the unstructured mesh neutron transport code HELIOS and the determination of the real neutron flux and power distribution for this new geometry. The developed model is then used for the determination of the neutron fluence distribution in the inner and outer wall of the system. Based on these results an optimized shielding strategy is developed for the molten salt fast reactor to keep the fluence in the safety related outer vessel below expected limit values. A lifetime of 80 years can be assured, but the size of the core/blanket system will be comparable to a sodium cooled fast reactor. The HELIOS results are verified against Monte-Carlo calculations with very satisfactory agreement for a deep penetration problem

Introduction to tropical geometry

CERN Document Server

Maclagan, Diane

2015-01-01

Tropical geometry is a combinatorial shadow of algebraic geometry, offering new polyhedral tools to compute invariants of algebraic varieties. It is based on tropical algebra, where the sum of two numbers is their minimum and the product is their sum. This turns polynomials into piecewise-linear functions, and their zero sets into polyhedral complexes. These tropical varieties retain a surprising amount of information about their classical counterparts. Tropical geometry is a young subject that has undergone a rapid development since the beginning of the 21st century. While establishing itself as an area in its own right, deep connections have been made to many branches of pure and applied mathematics. This book offers a self-contained introduction to tropical geometry, suitable as a course text for beginning graduate students. Proofs are provided for the main results, such as the Fundamental Theorem and the Structure Theorem. Numerous examples and explicit computations illustrate the main concepts. Each of t...

Progress of conversion system from CAD data to MCNP geometry data in Japan

International Nuclear Information System (INIS)

Sato, S.; Nashif, H.; Masuda, F.; Morota, H.; Iida, H.; Konno, C.

2010-01-01

Automatic conversion systems from CAD data to MCNP geometry input data have been developed to convert the CAD data of the fusion reactor with very complicated structure. So far, two conversion systems (GEOMIT-1 and ARCMCP) have been developed and the third system (GEOMIT-2) is under developing. The void data can be created in these systems. GEOMIT-1 was developed in 2007, but a lot of manual shape splitting work for the CAD data was required to convert the complicated geometry. ARCMCP was developed in 2008. The algorithm has been drastically improved on automatic creation of ambiguous surface in ARCMCP, but it still required a little manual shape splitting work. The latest system, GEOMIT-2, does not require additional commercial software packages, though the previous systems require them. It also has functions of the CAD data healing and the automatic shape splitting. Geometrical errors of CAD data can be automatically revised by the healing function, and complicated geometries can be automatically split into simple geometries by the shape splitting function. Any manual works for CAD data are not required in GEOMIT-2. GEOMIT-2 is very useful for nuclear analyses of fusion reactors.

Calculation models for a nuclear reactor

International Nuclear Information System (INIS)

Tashanii, Ahmed Ali

2010-01-01

Determination of different parameters of nuclear reactors requires neutron transport calculations. Due to complicity of geometry and material composition of the reactor core, neutron calculations were performed for simplified models of the real arrangement. In frame of the present work two models were used for calculations. First, an elementary cell model was used to prepare cross section data set for a homogenized-core reactor model. The homogenized-core reactor model was then used to perform neutron transport calculation. The nuclear reactor is a tank-shaped thermal reactor. The semi-cylindrical core arrangement consists of aluminum made fuel bundles immersed in water which acts as a moderator as well as a coolant. Each fuel bundle consists of aluminum cladded fuel rods arranged in square lattices. (author)

Design of experiments and springback prediction for AHSS automotive components with complex geometry

International Nuclear Information System (INIS)

Asgari, A.; Pereira, M.; Rolfe, B.; Dingle, M.; Hodgson, P.

2005-01-01

With the drive towards implementing Advanced High Strength Steels (AHSS) in the automotive industry; stamping engineers need to quickly answer questions about forming these strong materials into elaborate shapes. Commercially available codes have been successfully used to accurately predict formability, thickness and strains in complex parts. However, springback and twisting are still challenging subjects in numerical simulations of AHSS components. Design of Experiments (DOE) has been used in this paper to study the sensitivity of the implicit and explicit numerical results with respect to certain arrays of user input parameters in the forming of an AHSS component. Numerical results were compared to experimental measurements of the parts stamped in an industrial production line. The forming predictions of the implicit and explicit codes were in good agreement with the experimental measurements for the conventional steel grade, while lower accuracies were observed for the springback predictions. The forming predictions of the complex component with an AHSS material were also in good correlation with the respective experimental measurements. However, much lower accuracies were observed in its springback predictions. The number of integration points through the thickness and tool offset were found to be of significant importance, while coefficient of friction and Young's modulus (modeling input parameters) have no significant effect on the accuracy of the predictions for the complex geometry

Basic algebraic geometry, v.2

CERN Document Server

Shafarevich, Igor Rostislavovich

1994-01-01

Shafarevich Basic Algebraic Geometry 2 The second edition of Shafarevich's introduction to algebraic geometry is in two volumes. The second volume covers schemes and complex manifolds, generalisations in two different directions of the affine and projective varieties that form the material of the first volume. Two notable additions in this second edition are the section on moduli spaces and representable functors, motivated by a discussion of the Hilbert scheme, and the section on Kähler geometry. The book ends with a historical sketch discussing the origins of algebraic geometry. From the Zentralblatt review of this volume: "... one can only respectfully repeat what has been said about the first part of the book (...): a great textbook, written by one of the leading algebraic geometers and teachers himself, has been reworked and updated. As a result the author's standard textbook on algebraic geometry has become even more important and valuable. Students, teachers, and active researchers using methods of al...

Mispairs with Watson-Crick base-pair geometry observed in ternary complexes of an RB69 DNA polymerase variant.

Science.gov (United States)

Xia, Shuangluo; Konigsberg, William H

2014-04-01

Recent structures of DNA polymerase complexes with dGMPCPP/dT and dCTP/dA mispairs at the insertion site have shown that they adopt Watson-Crick geometry in the presence of Mn(2+) indicating that the tautomeric or ionization state of the base has changed. To see whether the tautomeric or ionization state of base-pair could be affected by its microenvironment, we determined 10 structures of an RB69 DNA polymerase quadruple mutant with dG/dT or dT/dG mispairs at position n-1 to n-5 of the Primer/Template duplex. Different shapes of the mispairs, including Watson-Crick geometry, have been observed, strongly suggesting that the local environment of base-pairs plays an important role in their tautomeric or ionization states. © 2014 The Protein Society.

Analysis and Geometry : MIMS-GGTM, in Honour of Mohammed Salah Baouendi

CERN Document Server

Kacimi, Aziz; Kallel, Sadok; Mir, Nordine

2015-01-01

This book includes selected papers presented at the MIMS (Mediterranean Institute for the Mathematical Sciences) - GGTM (Geometry and Topology Grouping for the Maghreb) conference, held in memory of Mohammed Salah Baouendi, a most renowned figure in the field of several complex variables, who passed away in 2011. All research articles were written by leading experts, some of whom are prize winners in the fields of complex geometry, algebraic geometry and analysis. The book offers a valuable resource for all researchers interested in recent developments in analysis and geometry.

Radiation absorption and optimization of solar photocatalytic reactors for environmental applications.

Science.gov (United States)

Colina-Márquez, Jose; Machuca-Martínez, Fiderman; Li Puma, Gianluca

2010-07-01

This study provides a systematic and quantitative approach to the analysis and optimization of solar photocatalytic reactors utilized in environmental applications such as pollutant remediation and conversion of biomass (waste) to hydrogen. Ray tracing technique was coupled with the six-flux absorption scattering model (SFM) to analyze the complex radiation field in solar compound parabolic collectors (CPC) and tubular photoreactors. The absorption of solar radiation represented by the spatial distribution of the local volumetric rate of photon absorption (LVRPA) depends strongly on catalyst loading and geometry. The total radiation absorbed in the reactors, the volumetric rate of absorption (VRPA), was analyzed as a function of the optical properties (scattering albedo) of the photocatalyst. The VRPA reached maxima at specific catalyst concentrations in close agreement with literature experimental studies. The CPC has on average 70% higher photon absorption efficiency than a tubular reactor and requires 39% less catalyst to operate under optimum conditions. The "apparent optical thickness" is proposed as a new dimensionless parameter for optimization of CPC and tubular reactors. It removes the dependence of the optimum catalyst concentration on tube diameter and photocatalyst scattering albedo. For titanium dioxide (TiO(2)) Degussa P25, maximum photon absorption occurs at apparent optical thicknesses of 7.78 for CPC and 12.97 for tubular reactors.

HERESY, 2-D Few-Group Static Eigenvalues Calculation for Thermal Reactor

International Nuclear Information System (INIS)

Finch, D.R.

1965-01-01

1 - Description of problem or function: HERESY3 solves the two- dimensional, few-group, static reactor eigenvalue problem using the heterogeneous (source-sink or Feinburg-Galanin) formalism. The solution yields the reactor k-effective and absorption reaction rates for each rod normalized to the most absorptive rod in the thermal level. Epithermal fissions are allowed at each resonance level, and lattice-averaged values of thermal utilization, resonance escape probability, thermal and resonance eta values, and the fast fission factor are calculated. Kernels in the calculation are based on age-diffusion theory. Both finite reactor lattices and infinitely repeating reactor super-cells may be calculated. Rod parameters may be calculated by several internal options, and a direct interface is provided to a HAMMER system (NESC Abstract 277) lattice library tape to obtain cell parameters. Criticality searches are provided on thermal utilization, thermal eta, and axial leakage buckling. 2 - Method of solution: Direct power iteration on matrix form of the heterogeneous critical equation is used. 3 - Restrictions on the complexity of the problem: Maxima of - 50 flux/geometry symmetry positions; 20 physically different assemblies; 9 resonance levels; 5000 rod coordinate positions

Methods in nuclear reactors calculations; Metodos de calculo en reactores nucleares

Energy Technology Data Exchange (ETDEWEB)

Velarde, G

1966-07-01

Studies are made of the neutron transport equation corresponding to the the real and virtual reactors, as well as the starting hypotheses. Methods are developed to solve the transport equation in slab geometry, and P{sub l}; B{sub l}; M{sub l}; S{sub n} and discrete ordinates approximations. (Author)

Finite element method solution of simplified P3 equation for flexible geometry handling

International Nuclear Information System (INIS)

Ryu, Eun Hyun; Joo, Han Gyu

2011-01-01

In order to obtain efficiently core flux solutions which would be much closer to the transport solution than the diffusion solution is, not being limited by the geometry of the core, the simplified P 3 (SP 3 ) equation is solved with the finite element method (FEM). A generic mesh generator, GMSH, is used to generate linear and quadratic mesh data. The linear system resulting from the SP 3 FEM discretization is solved by Krylov subspace methods (KSM). A symmetric form of the SP 3 equation is derived to apply the conjugate gradient method rather than the KSMs for nonsymmetric linear systems. An optional iso-parametric quadratic mapping scheme, which is to selectively model nonlinear shapes with a quadratic mapping to prevent significant mismatch in local domain volume, is also implemented for efficient application of arbitrary geometry handling. The gain in the accuracy attainable by the SP 3 solution over the diffusion solution is assessed by solving numerous benchmark problems having various core geometries including the IAEA PWR problems involving rectangular fuels and the Takeda fast reactor problems involving hexagonal fuels. The reference transport solution is produced by the McCARD Monte Carlo code and the multiplication factor and power distribution errors are assessed. In addition, the effect of quadratic mapping is examined for circular cell problems. It is shown that significant accuracy gain is possible with the SP 3 solution for the fast reactor problems whereas only marginal improvement is noted for thermal reactor problems. The quadratic mapping is also quite effective handling geometries with curvature. (author)

Practice in development and utilization of program-technical complex (PTK) in in-reactor control systems

International Nuclear Information System (INIS)

Gribov, A.A.; Kuzil, A.C.; Padun, S.P.; Surnachov, S.I.; Jakovlev, G.V.

2001-01-01

Experience with the development and utilization of the program-technical complex PTK 'KRUIZ' is analyzed in the paper. A peculiarity of PTK is the orientation on acquisition, processing and diagnostics of signals from in-reactor sensors (thermocouples and SPD). The PTK 'KRUIZ' represents a new generation of tools open for further development, oriented specifically on the use in in-reactor control systems in modernized and built power units of the WWER type. In the PTK 'KRUIZ', methods, models and algorithms proved in nuclear power plants are used accounting for the utilization of up to date technical tools and systematic technical solutions. Experience with the use of basic elements of the PTK 'KRUIZ' at existing WWER reactors including peculiarities of temperature control in nuclear power plants are also dealt within the paper. (Authors)

Experiences in the D ampersand D of the EBWR reactor complex at Argonne National Laboratory

International Nuclear Information System (INIS)

Bhattacharyya, S.K.; Boing, L.E.; Fellhauer, C.R.

1995-02-01

EBWR went critical in Dec 1957 at 20 MW(t), was upgraded to 100 MW(t) operation. EBWR was shut down July 1967 and placed in dry lay-up. In 1986, the D ampersand D work was planned in 4 phases: final planning and preparations for D ampersand D, removal of reactor systems, removal of reactor vessel complex, and final decontamination and project closeout. Despite precautions, there was an uptake of 241 Am by D ampersand D workers following underwater plasma arc cutting within the pool; the cause was traced to an experimental 241 Pu foil (200 μg) that was lost in the mid-1960s in the reactor vessel. Several major lessons were learned from this episode, among which is the fact that research facilities often involve unusual experiments which may not be recorded. Safety analysis and review procedure for D ampersand D operations need to be carefully considered since they represent considerably different situations than reactor operations. EBWR is one of the very few cases of a prototypic reactor facility designed, operated, tested and now D ampersand D'd by one organization

Application of advanced model of radiative heat transfer in a rod geometry to QUENCH and PARAMETER tests

International Nuclear Information System (INIS)

Vasiliev, A.D.; Kobelev, G.V.; Astafieva, V.O.

2007-01-01

Radiative heat transfer is very important in different fields of mechanical engineering and related technologies including nuclear reactors, heat transfer in furnaces, aerospace, different high-temperature assemblies. In particular, in the course of a hypothetical severe accident at PWR-type nuclear reactor the temperatures inside the reactor vessel reach high values at which taking into account of radiative heat exchange between the structures of reactor (including core and other reactor vessel elements) gets important. Radiative heat transfer dominates the late phase of severe accident because radiative heat fluxes (proportional to T4, where T is the temperature) are generally considerably higher than convective and conductive heat fluxes in a system. In particular, heat transfer due to radiation determines the heating and degradation of the core and surrounding steel in-vessel structures and finally influences the composition, temperature and mass of materials pouring out of the reactor vessel after its loss of integrity. Existing models of radiative heat exchange use many limitations and approximations: approximate estimation of view factors and beam lengths; the geometry change in the course of the accident is neglected; the database for emissivities of materials is not complete; absorption/emission by steam-noncondensable medium is taken into account approximately. The module MRAD was developed in this paper to model the radiative heat exchange in rod-like geometry typical of PWR-type reactor. Radiative heat exchange is computed using dividing on zones (zonal method) as in existing radiation models implemented to severe accident numerical codes such as ICARE, SCDAP/RELAP, MELCOR but improved in following aspects: new approach to evaluation of view factors and mean beam length; detailed evaluation of gas absorptivity and emissivity; account of effective radiative thermal conductivity for the large core; account of geometry modification in the course of severe

Modeling of the radiative field in complex geometries using computerized graphical tools. Application to comfort characterization in environments equipped with important radiative sources; Modelisation du champ radiatif dans des geometries complexes a l`aide d`outils infographiques. Application a la caracterisation du confort dans les ambiances munies de sources radiatives importantes

Energy Technology Data Exchange (ETDEWEB)

Manolescu, M; Sperandio, M; Allard, F [La Rochelle Universite, 17 - La Rochelle, LEPTAB (France)

1997-12-31

Bibliographic studies in the domain of radiant heat transfers in complex geometries demonstrate the impossibility of resolving such problems using classical analytical methods. The numerical analysis can theoretically be performed successfully but requires enormous computer means. The contribution of this study consists in using computerized graphical techniques to treat general problems of radiant heat transfers in complex geometries. This paper presents the model used, the calculation technique and the optimizations that allow to greatly reduce the computer memory required and the calculation time. The code developed uses evocative images for the synthetic presentation of results which facilitate the searcher`s and conceiver`s choices. Finally, an application to the characterization of thermal comfort in residential environments is developed to illustrate the potentialities of this method. (J.S.) 19 refs.

Modeling of the radiative field in complex geometries using computerized graphical tools. Application to comfort characterization in environments equipped with important radiative sources; Modelisation du champ radiatif dans des geometries complexes a l`aide d`outils infographiques. Application a la caracterisation du confort dans les ambiances munies de sources radiatives importantes

Energy Technology Data Exchange (ETDEWEB)

Manolescu, M.; Sperandio, M.; Allard, F. [La Rochelle Universite, 17 - La Rochelle, LEPTAB (France)

1996-12-31

Bibliographic studies in the domain of radiant heat transfers in complex geometries demonstrate the impossibility of resolving such problems using classical analytical methods. The numerical analysis can theoretically be performed successfully but requires enormous computer means. The contribution of this study consists in using computerized graphical techniques to treat general problems of radiant heat transfers in complex geometries. This paper presents the model used, the calculation technique and the optimizations that allow to greatly reduce the computer memory required and the calculation time. The code developed uses evocative images for the synthetic presentation of results which facilitate the searcher`s and conceiver`s choices. Finally, an application to the characterization of thermal comfort in residential environments is developed to illustrate the potentialities of this method. (J.S.) 19 refs.

Criteria for the assessment of reactor potential

International Nuclear Information System (INIS)

Carruthers, R.

1982-01-01

This article outlines some of the more general criteria to be used in assessing reactor potential. The interdependence of plasma and engineering parameters is considered. This demonstrated how it is the first wall power loading which is the critical parameter in assessing economic prospects. Taking some of the current conceptual designs of fusion reactors and raising the wall loading to the value needed to approach a competitive cost leads to a very challenging set of parameters. Although developed in terms of a tokamak they are figures which are applicable more generally to fusion reactors which are toroidal in form. It is not at all obvious that the tokamak could ever satisfy this criterion of economic viability, so we should not be using the parameters of existing tokamak reactor designs as the basis for assessing alternative approaches. We need to see whether there is an alternative sufficiently different as to offer a better chance of reaching these more onerous parameters. Unfortunately, so many of the alternatives differ only in magnetic geometry and their physical geometry leads to the same problems as faced by the tokamak. The traditional approach -- devising intriguing ''boxes'' for studying the confinement of plasma and then speculating on their reactor potential -- should give way to new initiatives. What we need in the fusion program is more ''reactor relevance pull'' and less ''plasma physics push'' when planning future activities

Holographic complexity for time-dependent backgrounds

Energy Technology Data Exchange (ETDEWEB)

Momeni, Davood, E-mail: davoodmomeni78@gmail.com [Eurasian International Center for Theoretical Physics and Department of General Theoretical Physics, Eurasian National University, Astana 010008 (Kazakhstan); Faizal, Mir, E-mail: mirfaizalmir@googlemail.com [Irving K. Barber School of Arts and Sciences, University of British Columbia, Okanagan, 3333 University Way, Kelowna, British Columbia V1V 1V7 (Canada); Department of Physics and Astronomy, University of Lethbridge, Lethbridge, Alberta, T1K 3M4 (Canada); Bahamonde, Sebastian, E-mail: sebastian.beltran.14@ucl.ac.uk [Department of Mathematics, University College London, Gower Street, London, WC1E 6BT (United Kingdom); Myrzakulov, Ratbay [Eurasian International Center for Theoretical Physics and Department of General Theoretical Physics, Eurasian National University, Astana 010008 (Kazakhstan)

2016-11-10

In this paper, we will analyze the holographic complexity for time-dependent asymptotically AdS geometries. We will first use a covariant zero mean curvature slicing of the time-dependent bulk geometries, and then use this co-dimension one spacelike slice of the bulk spacetime to define a co-dimension two minimal surface. The time-dependent holographic complexity will be defined using the volume enclosed by this minimal surface. This time-dependent holographic complexity will reduce to the usual holographic complexity for static geometries. We will analyze the time-dependence as a perturbation of the asymptotically AdS geometries. Thus, we will obtain time-dependent asymptotically AdS geometries, and we will calculate the holographic complexity for such time-dependent geometries.

Reactive flow simulation in complex 3D geometries using the COM3D code

International Nuclear Information System (INIS)

Breitung, W.; Kotchourko, A.; Veser, A.; Scholtyssek, W.

1999-01-01

The COM3D code, under development at the Forschungszentrum Karlsruhe (FZK), is a 3-d CFD code to describe turbulent combustion phenomena in complex geometries. It is intended to be part of the advanced integral code system for containment analysis (INCA) which includes in addition GASFLOW for distribution calculations, V3D for slow combustion and DET3D for detonation analysis. COM3D uses a TVD-solver and optional models for turbulence, chemistry and thermodynamics. The hydrodynamic model considers mass, momentum and energy conservation. Advanced procedures were provided to facilitate grid-development for complex 3-d structures. COM3D was validated on experiments performed on different scales with generally good agreement for important physical quantities. The code was applied to combustion analysis of a large PWR. The initial conditions were obtained from a GASFLOW distribution analysis for a LOOP scenario. Results are presented concerning flame propagation and pressure evolution in the containment which clearly demonstrate the effects of internal structures, their influence on turbulence formation and consequences for local loads. (author)

SABRINA, Geometry Plot Program for MCNP

International Nuclear Information System (INIS)

SEIDL, Marcus

2003-01-01

1 - Description of program or function: SABRINA is an interactive, three-dimensional, geometry-modeling code system, primarily for use with CCC-200/MCNP. SABRINA's capabilities include creation, visualization, and verification of three-dimensional geometries specified by either surface- or body-base combinatorial geometry; display of particle tracks are calculated by MCNP; and volume fraction generation. 2 - Method of solution: Rendering is performed by ray tracing or an edge and intersection algorithm. Volume fraction calculations are made by ray tracing. 3 - Restrictions on the complexity of the problem: A graphics display with X Window capability is required

Laminar simulation of intersubchannel mixing in a triangular nuclear fuel bundle geometry

International Nuclear Information System (INIS)

Zaretsky, A.; Lightstone, M.F.; Tullis, S.

2015-01-01

Highlights: • Quasi-periodic flow was observed through rod-to-wall gaps. • Triangular subchannel flows were fundamentally irregular. • Cross-gap flow was influenced both by local and adjacent cross-gap intensity. • Phase-linking between gaps induced cross-plane peripheral circulation through rod–wall gaps. • Cross-gap flow structure was dependent on subchannel geometry. - Abstract: Predicting temperature distributions in fuel rod bundles is an important component of nuclear reactor safety analysis. Intersubchannel mixing acts to homogenize coolant temperatures thus reducing the likelihood of localized regions of high fuel temperature. Previous research has shown that intersubchannel mixing in nuclear fuel rod bundles is enhanced by a large-scale quasi-periodic energetic fluid motion, which transports fluid on the cross-plane between the narrow gaps connecting subchannels. This phenomenon has also been observed in laminar flows. Unsteady laminar flow simulations were performed in a simplified bundle of three rods with a pipe. Three similar geometries of varying gap width were examined, and a thermal trace was implemented on the first geometry. Thermal mixing was driven by the advection of energy between subchannels by the cross-plane flow. Flow through the rod-to-wall gaps in the wall subchannels alternated with a dominant frequency, particularly when rod-to-wall gaps were smaller than rod-to-rod gaps. Significant phase-linking between rod-to-wall gaps was also observed such that a peripheral circulation occurred through each gap simultaneously. Cross-plane flow through the rod-to-rod gaps in the triangular subchannel was irregular in each case. This was due to the fundamental irregularity of the triangular subchannel geometry. Vortices were continually broken up by cross-plane flow from other gaps due to the odd number of fluid pathways within the central subchannel. Cross-plane flow in subchannel geometries is highly interconnected between gaps. The

Local analytic geometry

CERN Document Server

Abhyankar, Shreeram Shankar

1964-01-01

This book provides, for use in a graduate course or for self-study by graduate students, a well-motivated treatment of several topics, especially the following: (1) algebraic treatment of several complex variables; (2) geometric approach to algebraic geometry via analytic sets; (3) survey of local algebra; (4) survey of sheaf theory. The book has been written in the spirit of Weierstrass. Power series play the dominant role. The treatment, being algebraic, is not restricted to complex numbers, but remains valid over any complete-valued field. This makes it applicable to situations arising from

General Geometry and Geometry of Electromagnetism

OpenAIRE

Shahverdiyev, Shervgi S.

2002-01-01

It is shown that Electromagnetism creates geometry different from Riemannian geometry. General geometry including Riemannian geometry as a special case is constructed. It is proven that the most simplest special case of General Geometry is geometry underlying Electromagnetism. Action for electromagnetic field and Maxwell equations are derived from curvature function of geometry underlying Electromagnetism. And it is shown that equation of motion for a particle interacting with electromagnetic...

A simplified presentation of the multigroup analytic nodal method in 2-D Cartesian geometry

International Nuclear Information System (INIS)

Hebert, Alain

2008-01-01

The nodal diffusion algorithms used in many production reactor simulation codes are originating from a common ancestry developed in the 1970s, the analytic nodal method (ANM) of the QUANDRY code. However, this original presentation of the ANM is complex and makes difficult the calculation of the nodal coupling matrices. Moreover, QUANDRY is limited to two-energy groups and its generalization to more groups appears laborious. We are presenting a simplified implementation of the ANM requiring only limited programming work. This formulation is consistent with the initial QUANDRY implementation and is easily generalizable to arbitrary G-group problems. A Matlab script is provided to highlight the simplicity of our presentation. For the sake of clarity, our implementation is limited to G-group, 2-D Cartesian geometry

Lectures on discrete geometry

CERN Document Server

2002-01-01

Discrete geometry investigates combinatorial properties of configurations of geometric objects. To a working mathematician or computer scientist, it offers sophisticated results and techniques of great diversity and it is a foundation for fields such as computational geometry or combinatorial optimization. This book is primarily a textbook introduction to various areas of discrete geometry. In each area, it explains several key results and methods, in an accessible and concrete manner. It also contains more advanced material in separate sections and thus it can serve as a collection of surveys in several narrower subfields. The main topics include: basics on convex sets, convex polytopes, and hyperplane arrangements; combinatorial complexity of geometric configurations; intersection patterns and transversals of convex sets; geometric Ramsey-type results; polyhedral combinatorics and high-dimensional convexity; and lastly, embeddings of finite metric spaces into normed spaces. Jiri Matousek is Professor of Com...

Foreign research reactor uranium supply program: The Y-12 national security complex process

International Nuclear Information System (INIS)

Nelson, T.; Eddy, B.G.

2010-01-01

The Foreign Research Reactor (FRR) Uranium Supply Program at the Y-12 National Security Complex supports the nonproliferation objectives of the HEU Disposition Program, the Reduced Enrichment Research and Test Reactors (RERTR) Program, and the United States FRR Spent Nuclear Fuel (SNF) Acceptance Program. The Y-12 National Nuclear Security Administration (NNSA) Y-12 Site Office maintains the prime contracts with foreign governments for the supply of Low-Enriched Uranium (LEU) for their research reactors. The LEU is produced by down blending Highly Enriched Uranium (HEU) that has been declared surplus to the U.S. national defense needs. The down blending and sale of the LEU supports the Surplus HEU Disposition Program Record of Decision to make the HEU non-weapons usable and to recover the economic value of the uranium to the extent feasible. This program supports the important U.S. government and nuclear nonproliferation commitment to serve as a reliable and cost-effective uranium supplier for those foreign research reactors that are converting or have converted to LEU fuel under the guidance of the NNSA RERTR Program. In conjunction with the FRR SNF Acceptance Program which supports the global nonproliferation efforts to disposition U.S.-origin HEU, the Y-12 FRR Uranium Supply Program can provide the LEU for the replacement fuel fabrication. In addition to feedstock for fuel fabrication, Y-12 supplies LEU for target fabrication for medical isotope production. The Y-12 process uses supply forecasting tools, production improvements and efficient delivery preparations to successfully support the global research reactor community

Advances in discrete differential geometry

CERN Document Server

2016-01-01

This is one of the first books on a newly emerging field of discrete differential geometry and an excellent way to access this exciting area. It surveys the fascinating connections between discrete models in differential geometry and complex analysis, integrable systems and applications in computer graphics. The authors take a closer look at discrete models in differential geometry and dynamical systems. Their curves are polygonal, surfaces are made from triangles and quadrilaterals, and time is discrete. Nevertheless, the difference between the corresponding smooth curves, surfaces and classical dynamical systems with continuous time can hardly be seen. This is the paradigm of structure-preserving discretizations. Current advances in this field are stimulated to a large extent by its relevance for computer graphics and mathematical physics. This book is written by specialists working together on a common research project. It is about differential geometry and dynamical systems, smooth and discrete theories, ...

Development and experimental qualification of a calculation scheme for the evaluation of gamma heating in experimental reactors. Application to MARIA and Jules Horowitz (JHR) MTR Reactors

Energy Technology Data Exchange (ETDEWEB)

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

Using SpaceClaimTD Direct for Modeling Components with Complex Geometries for the Thermal Desktop-Based Advanced Stirling Radioisotope Generator Model

Science.gov (United States)

Fabanich, William A., Jr.

2014-01-01

SpaceClaim/TD Direct has been used extensively in the development of the Advanced Stirling Radioisotope Generator (ASRG) thermal model. This paper outlines the workflow for that aspect of the task and includes proposed best practices and lessons learned. The ASRG thermal model was developed to predict component temperatures and power output and to provide insight into the prime contractor's thermal modeling efforts. The insulation blocks, heat collectors, and cold side adapter flanges (CSAFs) were modeled with this approach. The model was constructed using mostly TD finite difference (FD) surfaces/solids. However, some complex geometry could not be reproduced with TD primitives while maintaining the desired degree of geometric fidelity. Using SpaceClaim permitted the import of original CAD files and enabled the defeaturing/repair of those geometries. TD Direct (a SpaceClaim add-on from CRTech) adds features that allowed the "mark-up" of that geometry. These so-called "mark-ups" control how finite element (FE) meshes are to be generated through the "tagging" of features (e.g. edges, solids, surfaces). These tags represent parameters that include: submodels, material properties, material orienters, optical properties, and radiation analysis groups. TD aliases were used for most tags to allow analysis to be performed with a variety of parameter values. "Domain-tags" were also attached to individual and groups of surfaces and solids to allow them to be used later within TD to populate objects like, for example, heaters and contactors. These tools allow the user to make changes to the geometry in SpaceClaim and then easily synchronize the mesh in TD without having to redefine the objects each time as one would if using TDMesher. The use of SpaceClaim/TD Direct helps simplify the process for importing existing geometries and in the creation of high fidelity FE meshes to represent complex parts. It also saves time and effort in the subsequent analysis.

Using SpaceClaim/TD Direct for Modeling Components with Complex Geometries for the Thermal Desktop-Based Advanced Stirling Radioisotope Generator Model

Science.gov (United States)

Fabanich, William

2014-01-01

SpaceClaim/TD Direct has been used extensively in the development of the Advanced Stirling Radioisotope Generator (ASRG) thermal model. This paper outlines the workflow for that aspect of the task and includes proposed best practices and lessons learned. The ASRG thermal model was developed to predict component temperatures and power output and to provide insight into the prime contractors thermal modeling efforts. The insulation blocks, heat collectors, and cold side adapter flanges (CSAFs) were modeled with this approach. The model was constructed using mostly TD finite difference (FD) surfaces solids. However, some complex geometry could not be reproduced with TD primitives while maintaining the desired degree of geometric fidelity. Using SpaceClaim permitted the import of original CAD files and enabled the defeaturing repair of those geometries. TD Direct (a SpaceClaim add-on from CRTech) adds features that allowed the mark-up of that geometry. These so-called mark-ups control how finite element (FE) meshes were generated and allowed the tagging of features (e.g. edges, solids, surfaces). These tags represent parameters that include: submodels, material properties, material orienters, optical properties, and radiation analysis groups. TD aliases were used for most tags to allow analysis to be performed with a variety of parameter values. Domain-tags were also attached to individual and groups of surfaces and solids to allow them to be used later within TD to populate objects like, for example, heaters and contactors. These tools allow the user to make changes to the geometry in SpaceClaim and then easily synchronize the mesh in TD without having to redefine these objects each time as one would if using TD Mesher.The use of SpaceClaim/TD Direct has helped simplify the process for importing existing geometries and in the creation of high fidelity FE meshes to represent complex parts. It has also saved time and effort in the subsequent analysis.

Synergism of the method of characteristic, R-functions and diffusion solution for accurate representation of 3D neutron interactions in research reactors using the AGENT code system

International Nuclear Information System (INIS)

Hursin, Mathieu; Xiao Shanjie; Jevremovic, Tatjana

2006-01-01

This paper summarizes the theoretical and numerical aspects of the AGENT code methodology accurately applied for detailed three-dimensional (3D) multigroup steady-state modeling of neutron interactions in complex heterogeneous reactor domains. For the first time we show the fine-mesh neutron scalar flux distribution in Purdue research reactor (that was built over forty years ago). The AGENT methodology is based on the unique combination of the three theories: the method of characteristics (MOC) used to simulate the neutron transport in two-dimensional (2D) whole core heterogeneous calculation, the theory of R-functions used as a mathematical tool to describe the true geometry and fuse with the MOC equations, and one-dimensional (1D) higher-order diffusion correction of 2D transport model to account for full 3D heterogeneous whole core representation. The synergism between the radial 2D transport and the 1D axial transport (to take into account the axial neutron interactions and leakage), called the 2D/1D method (used in DeCART and CHAPLET codes), provides a 3D computational solution. The unique synergism between the AGENT geometrical algorithm capable of modeling any current or future reactor core geometry and 3D neutron transport methodology is described in details. The 3D AGENT accuracy and its efficiency are demonstrated showing the eigenvalues, point-wise flux and reaction rate distributions in representative reactor geometries. The AGENT code, comprising this synergism, represents a building block of the computational system, called the virtual reactor. Its main purpose is to perform 'virtual' experiments and demonstrations of various mainly university research reactor experiments

Automatic generation of 3D fine mesh geometries for the analysis of the venus-3 shielding benchmark experiment with the Tort code

International Nuclear Information System (INIS)

Pescarini, M.; Orsi, R.; Martinelli, T.

2003-01-01

In many practical radiation transport applications today the cost for solving refined, large size and complex multi-dimensional problems is not so much computing but is linked to the cumbersome effort required by an expert to prepare a detailed geometrical model, verify and validate that it is correct and represents, to a specified tolerance, the real design or facility. This situation is, in particular, relevant and frequent in reactor core criticality and shielding calculations, with three-dimensional (3D) general purpose radiation transport codes, requiring a very large number of meshes and high performance computers. The need for developing tools that make easier the task to the physicist or engineer, by reducing the time required, by facilitating through effective graphical display the verification of correctness and, finally, that help the interpretation of the results obtained, has clearly emerged. The paper shows the results of efforts in this field through detailed simulations of a complex shielding benchmark experiment. In the context of the activities proposed by the OECD/NEA Nuclear Science Committee (NSC) Task Force on Computing Radiation Dose and Modelling of Radiation-Induced Degradation of Reactor Components (TFRDD), the ENEA-Bologna Nuclear Data Centre contributed with an analysis of the VENUS-3 low-flux neutron shielding benchmark experiment (SCK/CEN-Mol, Belgium). One of the targets of the work was to test the BOT3P system, originally developed at the Nuclear Data Centre in ENEA-Bologna and actually released to OECD/NEA Data Bank for free distribution. BOT3P, ancillary system of the DORT (2D) and TORT (3D) SN codes, permits a flexible automatic generation of spatial mesh grids in Cartesian or cylindrical geometry, through combinatorial geometry algorithms, following a simplified user-friendly approach. This system demonstrated its validity also in core criticality analyses, as for example the Lewis MOX fuel benchmark, permitting to easily

Modeling photonic crystal waveguides with noncircular geometry using green function method

International Nuclear Information System (INIS)

Uvarovaa, I.; Tsyganok, B.; Bashkatov, Y.; Khomenko, V.

2012-01-01

Currently in the field of photonics is an acute problem fast and accurate simulation photonic crystal waveguides with complex geometry. This paper describes an improved method of Green's functions for non-circular geometries. Based on comparison of selected efficient numerical method for finding the eigenvalues for the Green's function method for non-circular holes chosen effective method for our purposes. Simulation is realized in Maple environment. The simulation results confirmed experimentally. Key words: photonic crystal, waveguide, modeling, Green function, complex geometry

Development of CAD-Based Geometry Processing Module for a Monte Carlo Particle Transport Analysis Code

International Nuclear Information System (INIS)

Choi, Sung Hoon; Kwark, Min Su; Shim, Hyung Jin

2012-01-01

As The Monte Carlo (MC) particle transport analysis for a complex system such as research reactor, accelerator, and fusion facility may require accurate modeling of the complicated geometry. Its manual modeling by using the text interface of a MC code to define the geometrical objects is tedious, lengthy and error-prone. This problem can be overcome by taking advantage of modeling capability of the computer aided design (CAD) system. There have been two kinds of approaches to develop MC code systems utilizing the CAD data: the external format conversion and the CAD kernel imbedded MC simulation. The first approach includes several interfacing programs such as McCAD, MCAM, GEOMIT etc. which were developed to automatically convert the CAD data into the MCNP geometry input data. This approach makes the most of the existing MC codes without any modifications, but implies latent data inconsistency due to the difference of the geometry modeling system. In the second approach, a MC code utilizes the CAD data for the direct particle tracking or the conversion to an internal data structure of the constructive solid geometry (CSG) and/or boundary representation (B-rep) modeling with help of a CAD kernel. MCNP-BRL and OiNC have demonstrated their capabilities of the CAD-based MC simulations. Recently we have developed a CAD-based geometry processing module for the MC particle simulation by using the OpenCASCADE (OCC) library. In the developed module, CAD data can be used for the particle tracking through primitive CAD surfaces (hereafter the CAD-based tracking) or the internal conversion to the CSG data structure. In this paper, the performances of the text-based model, the CAD-based tracking, and the internal CSG conversion are compared by using an in-house MC code, McSIM, equipped with the developed CAD-based geometry processing module

Very High Efficiency Reactor (VHER) Concepts for Electrical Power Generation and Hydrogen Production

International Nuclear Information System (INIS)

PARMA JR, EDWARD J.; PICKARD, PAUL S.; SUO-ANTTILA, AHTI JORMA

2003-01-01

The goal of the Very High Efficiency Reactor study was to develop and analyze concepts for the next generation of nuclear power reactors. The next generation power reactor should be cost effective compared to current power generation plant, passively safe, and proliferation-resistant. High-temperature reactor systems allow higher electrical generating efficiencies and high-temperature process heat applications, such as thermo-chemical hydrogen production. The study focused on three concepts; one using molten salt coolant with a prismatic fuel-element geometry, the other two using high-pressure helium coolant with a prismatic fuel-element geometry and a fuel-pebble element design. Peak operating temperatures, passive-safety, decay heat removal, criticality, burnup, reactivity coefficients, and material issues were analyzed to determine the technical feasibility of each concept

Analytical solution of neutron transport equation in an annular reactor with a rotating pulsed source; Resolucao analitica da equacao de transporte de neutrons em um reator anelar com fonte pulsada rotativa

Energy Technology Data Exchange (ETDEWEB)

Teixeira, Paulo Cleber Mendonca

2002-12-01

In this study, an analytical solution of the neutron transport equation in an annular reactor is presented with a short and rotating neutron source of the type S(x) {delta} (x- Vt), where V is the speed of annular pulsed reactor. The study is an extension of a previous study by Williams [12] carried out with a pulsed source of the type S(x) {delta} (t). In the new concept of annular pulsed reactor designed to produce continuous high flux, the core consists of a subcritical annular geometry pulsed by a rotating modulator, producing local super prompt critical condition, thereby giving origin to a rotating neutron pulse. An analytical solution is obtained by opening up of the annular geometry and applying one energy group transport theory in one dimension using applied mathematical techniques of Laplace transform and Complex Variables. The general solution for the flux consists of a fundamental mode, a finite number of harmonics and a transient integral. A condition which limits the number of harmonics depending upon the circumference of the annular geometry has been obtained. Inverse Laplace transform technique is used to analyse instability condition in annular reactor core. A regenerator parameter in conjunction with perimeter of the ring and nuclear properties is used to obtain stable and unstable harmonics and to verify if these exist. It is found that the solution does not present instability in the conditions stated in the new concept of annular pulsed reactor. (author)

Geometry of quantum computation with qutrits.

Science.gov (United States)

Li, Bin; Yu, Zu-Huan; Fei, Shao-Ming

2013-01-01

Determining the quantum circuit complexity of a unitary operation is an important problem in quantum computation. By using the mathematical techniques of Riemannian geometry, we investigate the efficient quantum circuits in quantum computation with n qutrits. We show that the optimal quantum circuits are essentially equivalent to the shortest path between two points in a certain curved geometry of SU(3(n)). As an example, three-qutrit systems are investigated in detail.

Neutronics comparative analysis of plate-type research reactor using deterministic and stochastic methods

International Nuclear Information System (INIS)

Liu, Shichang; Wang, Guanbo; Wu, Gaochen; Wang, Kan

2015-01-01

Highlights: • DRAGON and DONJON are applied and verified in calculations of research reactors. • Continuous-energy Monte Carlo calculations by RMC are chosen as the references. • “ECCO” option of DRAGON is suitable for the calculations of research reactors. • Manual modifications of cross-sections are not necessary with DRAGON and DONJON. • DRAGON and DONJON agree well with RMC if appropriate treatments are applied. - Abstract: Simulation of the behavior of the plate-type research reactors such as JRR-3M and CARR poses a challenge for traditional neutronics calculation tools and schemes for power reactors, due to the characteristics of complex geometry, highly heterogeneity and large leakage of the research reactors. Two different theoretical approaches, the deterministic and the stochastic methods, are used for the neutronics analysis of the JRR-3M plate-type research reactor in this paper. For the deterministic method the neutronics codes DRAGON and DONJON are used, while the continuous-energy Monte Carlo code RMC (Reactor Monte Carlo code) is employed for the stochastic approach. The goal of this research is to examine the capability of the deterministic code system DRAGON and DONJON to reliably simulate the research reactors. The results indicate that the DRAGON and DONJON code system agrees well with the continuous-energy Monte Carlo simulation on both k eff and flux distributions if the appropriate treatments (such as the ECCO option) are applied

Chooz-B1, the new Electricite de France PWR: calculation scheme of neutron leakages from the reactor cavity

International Nuclear Information System (INIS)

Champion, G.; Thiriet, A.; Vergnaud, T.; Bourdet, L.; Nimal, J.C.; Brandicourt, G.

1987-04-01

A new calculation scheme has been set up to assess the neutron field characteristics inside French PWR. In order to take into account multiple neutron scattering and the complexity of the reactor geometry, the use of Monte-Carlo methods have been heavily increased. They are coupled with classical SN.-methods. The main goal aimed at was to find out the neutron field characteristics at the level of the reactor pit openings. These radiation reference sources will be used to check the neutron shielding efficiencies. The new calculation scheme has been applied to CHOOZ-B1, the first unit of the new N4 program. The former results have been compared with the measurement results related to PALUEL-I and II PWR, two units of the previous P4 program. Although the core and the geometry are not entirely similar, it is possible to check with confidence the calculation results along the vessel and at the core midplane level with the measurement results at the same locations. It appears that they are in good agreement. Consequently, the new calculation scheme appears reliable

Usage of burnable poison on research reactors

International Nuclear Information System (INIS)

Villarino, Eduardo Anibal

2002-01-01

The fuel assemblies with burnable poison are widely used on power reactors, but there are not commonly used on research reactors. This paper shows a neutronic analysis of the advantages and disadvantages of the burnable poison usage on research reactors. This paper analyses both burnable poison design used on research reactors: Boron on the lateral wall and Cadmium wires. Both designs include a parametric study on the design parameters like the amount and geometry of the burnable poison. This paper presents the design flexibility using burnable poisons, it does not find an optimal or final design, which it will strongly depend on the core characteristics and fuel management strategy. (author)

Influence of Irradiance, Flow Rate, Reactor Geometry, and Photopromoter Concentration in Mineralization Kinetics of Methane in Air and in Aqueous Solutions by Photocatalytic Membranes Immobilizing Titanium Dioxide

Directory of Open Access Journals (Sweden)

Ignazio Renato Bellobono

2008-01-01

Full Text Available Photomineralization of methane in air (10.0–1000 ppm (mass/volume of C at 100% relative humidity (dioxygen as oxygen donor was systematically studied at 318±3 K in an annular laboratory-scale reactor by photocatalytic membranes immobilizing titanium dioxide as a function of substrate concentration, absorbed power per unit length of membrane, reactor geometry, and concentration of a proprietary vanadium alkoxide as photopromoter. Kinetics of both substrate disappearance, to yield intermediates, and total organic carbon (TOC disappearance, to yield carbon dioxide, were followed. At a fixed value of irradiance (0.30 W⋅cm-1, the mineralization experiments in gaseous phase were repeated as a function of flow rate (4–400 m3⋅h−1. Moreover, at a standard flow rate of 300 m3⋅h−1, the ratio between the overall reaction volume and the length of the membrane was varied, substantially by varying the volume of reservoir, from and to which circulation of gaseous stream took place. Photomineralization of methane in aqueous solutions was also studied, in the same annular reactor and in the same conditions, but in a concentration range of 0.8–2.0 ppm of C, and by using stoichiometric hydrogen peroxide as an oxygen donor. A kinetic model was employed, from which, by a set of differential equations, four final optimised parameters, k1 and K1, k2 and K2, were calculated, which is able to fit the whole kinetic profile adequately. The influence of irradiance on k1 and k2, as well as of flow rate on K1 and K2, is rationalized. The influence of reactor geometry on k values is discussed in view of standardization procedures of photocatalytic experiments. Modeling of quantum yields, as a function of substrate concentration and irradiance, as well as of concentration of photopromoter, was carried out very satisfactorily. Kinetics of hydroxyl radicals reacting between themselves, leading to hydrogen peroxide, other than with substrate or

The code DYN3DR for steady-state and transient analyses of light water reactor cores with Cartesian geometry

International Nuclear Information System (INIS)

Grundmann, U.

1995-11-01

The code DYN3D/M2 was developed for 3-dimensional steady-state and transient analyses of reactor cores with hexagonal fuel assemblies. The neutron kinetics of the new version DYN3DR is based on a nodal method for the solution of the 3-dimensional 2-group neutron diffusion equation for Cartesian geometry. The thermal-hydraulic model FLOCAL simulating the two phase flow of coolant and the fuel rod behaviour is used in the two versions. The fundamentals for the solution of the neutron diffusion equations in DYN3DR are described. The 3-dimensional NEACRP benchmarks for rod ejections in LWR with quadratic fuel assemblies were calculated and the results were compared with the published solutions. The developed algorithm for neutron kinetics are suitable for using parallel processing. The behaviour of speed-up versus the number of processors is demonstrated for calculations of a static neutron flux distribution using a workstation with 4 processors. (orig.) [de

Application of the nodal method RTN-0 for the solution of the neutron diffusion equation dependent of time in hexagonal-Z geometry

International Nuclear Information System (INIS)

Esquivel E, J.; Alonso V, G.; Del Valle G, E.

2015-09-01

The solution of the neutron diffusion equation either for reactors in steady state or time dependent, is obtained through approximations generated by implementing of nodal methods such as RTN-0 (Raviart-Thomas-Nedelec of zero index), which is used in this study. Since the nodal methods are applied in quadrangular geometries, in this paper a technique in which the hexagonal geometry through the transfinite interpolation of Gordon-Hall becomes the appropriate geometry to make use of the nodal method RTN-0 is presented. As a result, a computer program was developed, whereby is possible to obtain among other results the neutron multiplication effective factor (k eff ), and the distribution of radial and/or axial power. To verify the operation of the code, was applied to three benchmark problems: in the first two reactors VVER and FBR, results k eff and power distribution are obtained, considering the steady state case of reactor; while the third problem a type VVER is analyzed, in its case dependent of time, which qualitative results are presented on the behavior of the reactor power. (Author)

Quasi-dimensional modeling of a fast-burn combustion dual-plug spark-ignition engine with complex combustion chamber geometries

International Nuclear Information System (INIS)

Altın, İsmail; Bilgin, Atilla

2015-01-01

This study builds on a previous parametric investigation using a thermodynamic-based quasi-dimensional (QD) cycle simulation of a spark-ignition (SI) engine with dual-spark plugs. The previous work examined the effects of plug-number and location on some performance parameters considering an engine with a simple cylindrical disc-shaped combustion chamber. In order to provide QD thermodynamic models applicable to complex combustion chamber geometries, a novel approach is considered here: flame-maps, which utilizes a computer aided design (CAD) software (SolidWorks). Flame maps are produced by the CAD software, which comprise all the possible flame radiuses with an increment of one-mm between them, according to the spark plug positions, spark timing, and piston position near the top dead center. The data are tabulated and stored as matrices. Then, these tabulated data are adapted to the previously reported cycle simulation. After testing for simple disc-shaped chamber geometries, the simulation is applied to a real production automobile (Honda-Fit) engine to perform the parametric study. - Highlights: • QD model was applied in dual plug engine with complex realistic combustion chamber. • This method successfully modeled the combustion in the dual-plug Honda-Fit engine. • The same combustion chamber is tested for various spark plug(s) locations. • The centrally located single spark-plug results in the fastest combustion

Reactor Structural Materials: Reactor Pressure Vessel Steels

International Nuclear Information System (INIS)

Chaouadi, R.

2000-01-01

The objectives of SCK-CEN's R and D programme on Rector Pressure Vessel (RPV) Steels are:(1) to complete the fracture toughness data bank of various reactor pressure vessel steels by using precracked Charpy specimens that were tested statically as well as dynamically; (2) to implement the enhanced surveillance approach in a user-friendly software; (3) to improve the existing reconstitution technology by reducing the input energy (short cycle welding) and modifying the stud geometry. Progress and achievements in 1999 are reported

XBWR, 1-D Xe Transients for BWR in Axial Geometry

International Nuclear Information System (INIS)

Forti, G.

1980-01-01

1 - Nature of the physical problem solved: 1-D xenon transients for BWRs in axial geometry. 2 - Method of solution: XBWR couples a two group neutron diffusion calculation in plane geometry with a two phase flow cooling channel calculation and the heat conduction in the typical fuel rod. The program allows following any given power time schedule, such as shut-down and restart, day-night power variation etc., while the reactor is being kept critical by control rod movement, variable poisoning of the core, or coolant flow recirculation rate. The xenon and iodine concentrations variation is evaluated pointwise (up to 100 points) by analytical solution for successive fixed time steps. At the end of each time step a new distribution of fluxes, power, voids and temperatures is obtained, which is consistent with the reactor critical condition as it is got by variation of the control parameter taking into account the feedbacks. The new flux distribution is used as input for xenon and iodine concentrations evolution in the next time step

Improvements in or relating to nuclear reactors

International Nuclear Information System (INIS)

Timofeev, A.V.; Batjukov, V.I.; Fadeev, A.I.; Shapkin, A.F.; Shikhiyan, T.G.; Ordynsky, G.V.; Drachev, V.P.; Pogodin, E.N.

1980-01-01

A refuelling installation for nuclear reactor complexes is described for recharging the reactor vessels of such complexes with new fuel assemblies and for removing spent fuel assemblies from the reactor vessel. (U.K.)

Flux compactifications and generalized geometries

International Nuclear Information System (INIS)

Grana, Mariana

2006-01-01

Following the lectures given at CERN Winter School 2006, we present a pedagogical overview of flux compactifications and generalized geometries, concentrating on closed string fluxes in type II theories. We start by reviewing the supersymmetric flux configurations with maximally symmetric four-dimensional spaces. We then discuss the no-go theorems (and their evasion) for compactifications with fluxes. We analyse the resulting four-dimensional effective theories for Calabi-Yau and Calabi-Yau orientifold compactifications, concentrating on the flux-induced superpotentials. We discuss the generic mechanism of moduli stabilization and illustrate with two examples: the conifold in IIB and a T 6 /(Z 3 x Z 3 ) torus in IIA. We finish by studying the effective action and flux vacua for generalized geometries in the context of generalized complex geometry

Flux compactifications and generalized geometries

Energy Technology Data Exchange (ETDEWEB)

Grana, Mariana [Service de Physique Theorique, CEA/Saclay, 91191 Gif-sur-Yvette Cedex (France)

2006-11-07

Following the lectures given at CERN Winter School 2006, we present a pedagogical overview of flux compactifications and generalized geometries, concentrating on closed string fluxes in type II theories. We start by reviewing the supersymmetric flux configurations with maximally symmetric four-dimensional spaces. We then discuss the no-go theorems (and their evasion) for compactifications with fluxes. We analyse the resulting four-dimensional effective theories for Calabi-Yau and Calabi-Yau orientifold compactifications, concentrating on the flux-induced superpotentials. We discuss the generic mechanism of moduli stabilization and illustrate with two examples: the conifold in IIB and a T{sup 6} /(Z{sub 3} x Z{sub 3}) torus in IIA. We finish by studying the effective action and flux vacua for generalized geometries in the context of generalized complex geometry.

Methods in nuclear reactors calculations

International Nuclear Information System (INIS)

Velarde, G.

1966-01-01

Studies are made of the neutron transport equation corresponding to the the real and virtual reactors, as well as the starting hypotheses. Methods are developed to solve the transport equation in slab geometry, and P l ; B l ; M l ; S n and discrete ordinates approximations. (Author)

CBM RICH geometry optimization

Energy Technology Data Exchange (ETDEWEB)

Mahmoud, Tariq; Hoehne, Claudia [II. Physikalisches Institut, Giessen Univ. (Germany); Collaboration: CBM-Collaboration

2016-07-01

The Compressed Baryonic Matter (CBM) experiment at the future FAIR complex will investigate the phase diagram of strongly interacting matter at high baryon density and moderate temperatures in A+A collisions from 2-11 AGeV (SIS100) beam energy. The main electron identification detector in the CBM experiment will be a RICH detector with a CO{sub 2} gaseous-radiator, focusing spherical glass mirrors, and MAPMT photo-detectors being placed on a PMT-plane. The RICH detector is located directly behind the CBM dipole magnet. As the final magnet geometry is now available, some changes in the RICH geometry become necessary. In order to guarantee a magnetic field of 1 mT at maximum in the PMT plane for effective operation of the MAPMTs, two measures have to be taken: The PMT plane is moved outwards of the stray field by tilting the mirrors by 10 degrees and shielding boxes have been designed. In this contribution the results of the geometry optimization procedure are presented.

Creating and using a type of free-form geometry in Monte Carlo particle transport

International Nuclear Information System (INIS)

Wessol, D.E.; Wheeler, F.J.

1993-01-01

While the reactor physicists were fine-tuning the Monte Carlo paradigm for particle transport in regular geometries, the computer scientists were developing rendering algorithms to display extremely realistic renditions of irregular objects ranging from the ubiquitous teakettle to dynamic Jell-O. Even though the modeling methods share a common basis, the initial strategies each discipline developed for variance reduction were remarkably different. Initially, the reactor physicist used Russian roulette, importance sampling, particle splitting, and rejection techniques. In the early stages of development, the computer scientist relied primarily on rejection techniques, including a very elegant hierarchical construction and sampling method. This sampling method allowed the computer scientist to viably track particles through irregular geometries in three-dimensional space, while the initial methods developed by the reactor physicists would only allow for efficient searches through analytical surfaces or objects. As time goes by, it appears there has been some merging of the variance reduction strategies between the two disciplines. This is an early (possibly first) incorporation of geometric hierarchical construction and sampling into the reactor physicists' Monte Carlo transport model that permits efficient tracking through nonuniform rational B-spline surfaces in three-dimensional space. After some discussion, the results from this model are compared with experiments and the model employing implicit (analytical) geometric representation

Industrial complex in organizing the high-speed in-line construction of reactor compartments at the Balakovo NPP

International Nuclear Information System (INIS)

Maksakov, A.I.; Kovrigin, Yu.K.; Zhila, V.P.

1986-01-01

Qualitatively new technology of reactor compartment construction presupposing organizing of an industrial-mounting in-line complex is described. Maximum level of construction industrialization and noticeable reduction of construction duration are noted to be ensured by means of this technology

Development of a code in three-dimensional cylindrical geometry based on analytic function expansion nodal (AFEN) method

International Nuclear Information System (INIS)

Lee, Joo Hee

2006-02-01

There is growing interest in developing pebble bed reactors (PBRs) as a candidate of very high temperature gas-cooled reactors (VHTRs). Until now, most existing methods of nuclear design analysis for this type of reactors are base on old finite-difference solvers or on statistical methods. But for realistic analysis of PBRs, there is strong desire of making available high fidelity nodal codes in three-dimensional (r,θ,z) cylindrical geometry. Recently, the Analytic Function Expansion Nodal (AFEN) method developed quite extensively in Cartesian (x,y,z) geometry and in hexagonal-z geometry was extended to two-group (r,z) cylindrical geometry, and gave very accurate results. In this thesis, we develop a method for the full three-dimensional cylindrical (r,θ,z) geometry and implement the method into a code named TOPS. The AFEN methodology in this geometry as in hexagonal geometry is 'robus' (e.g., no occurrence of singularity), due to the unique feature of the AFEN method that it does not use the transverse integration. The transverse integration in the usual nodal methods, however, leads to an impasse, that is, failure of the azimuthal term to be transverse-integrated over r-z surface. We use 13 nodal unknowns in an outer node and 7 nodal unknowns in an innermost node. The general solution of the node can be expressed in terms of that nodal unknowns, and can be updated using the nodal balance equation and the current continuity condition. For more realistic analysis of PBRs, we implemented em Marshak boundary condition to treat the incoming current zero boundary condition and the partial current translation (PCT) method to treat voids in the core. The TOPS code was verified in the various numerical tests derived from Dodds problem and PBMR-400 benchmark problem. The results of the TOPS code show high accuracy and fast computing time than the VENTURE code that is based on finite difference method (FDM)

Perspectives in Analysis, Geometry, and Topology

CERN Document Server

Itenberg, I V; Passare, Mikael

2012-01-01

The articles in this volume are invited papers from the Marcus Wallenberg symposium and focus on research topics that bridge the gap between analysis, geometry, and topology. The encounters between these three fields are widespread and often provide impetus for major breakthroughs in applications. Topics include new developments in low dimensional topology related to invariants of links and three and four manifolds; Perelman's spectacular proof of the Poincare conjecture; and the recent advances made in algebraic, complex, symplectic, and tropical geometry.

Resonance integral calculations for high temperature reactors

International Nuclear Information System (INIS)

Blake, J.P.H.

1960-02-01

Methods of calculation of resonance integrals of finite dilution and temperature are given for both, homogeneous and heterogeneous geometries, together with results obtained from these methods as applied to the design of high temperature reactors. (author)

Coupled fast-thermal core 'HERBE', as the benchmark experiment at the RB reactor

International Nuclear Information System (INIS)

Pesic, M.

2003-10-01

Validation of the well-known Monte Carlo code MCNP TM against measured criticality data for the coupled fast-thermal HERBE. System at the RB research reactor is shown in this paper. Experimental data are obtained for regular HERBE core and for the cases of controlled flooding of the neutron converter zone by heavy water. Earlier calculations of these criticality parameters, done by combination of transport and diffusion codes using 2D geometry model are also compared to new calculations carried out by the MCNP code in 3D geometry, applying new detailed 3D model of the HEU fuel slug, developed recently. Satisfactory agreements in comparison of the HERBE criticality calculation results with experimental data, in spite complex heterogeneous composition of the HERBE core, are obtained and confirmed that HERBE core could be used as a criticality benchmark for coupled fast-thermal core. (author)

Reactor BR2. Introduction

International Nuclear Information System (INIS)

Gubel, P.

2002-01-01

The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system

Reactor BR2. Introduction

Energy Technology Data Exchange (ETDEWEB)

Gubel, P

2002-04-01

The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system.

Estimate of the induced activity in the fixed structure of the Pluto Reactor

International Nuclear Information System (INIS)

Goodill, D.R.; Moore, D.C.; Tymons, B.J.

1984-11-01

This report presents an inventory of neutron-induced activity in the main components of a Materials Testing Reactor at the end of reactor life. The calculations were carried out for the PLUTO reactor at Harwell which is taken to be typical of all MTRs. The results were derived by using the FISPIN computer code, taking into account the geometry and construction of the reactor components. (author)

The Persistification of the ATLAS Geometry

CERN Document Server

AUTHOR|(INSPIRE)INSPIRE-00068562; The ATLAS collaboration; Bianchi, Riccardo-Maria

2016-01-01

The complex geometry of the whole detector of the ATLAS experiment at LHC is currently stored only in custom online databases, from which it is built on-the- y on request. Accessing the online geometry guarantees accessing the latest version of the detector description, but requires the setup of the full ATLAS so ware framework “Athena”, which provides the online services and the tools to retrieve the data from the database. is operation is cumbersome and slows down the applications that need to access the geometry. Moreover, all applications that need to access the detector geom- etry need to be built and run on the same platform as the ATLAS framework, preventing the usage of the actual detector geometry in stand-alone applications. Here we propose a new mechanism to persistify and serve the geometry of HEP experiments. e new mechanism is composed by a new le format and a REST API. e new le format allows to store the whole detector description locally in a at le, and it is especially optimized to descri...

SCORE-4, 2-D Removal Diffusion in X-Y or R-Z Geometry for Rectangular Shields

International Nuclear Information System (INIS)

Richardson, B.L.

1974-01-01

1 - Nature of physical problem solved: The neutron flux is calculated for a shield made up of rectangular regions. The geometry is either x-y or r-z. 2 - Method of solution: Removal fluxes and sources throughout the shield regions are calculated from a given reactor core power distribution using a point kernel method. The diffusion neutron fluxes are obtained from the removal source distribution using an iterative Method of solution. 3 - Restrictions on the complexity of the problem: The amount of fast core required for the program depends on the size of shield being calculated. For example, a 100 by 100 mesh shielding calculation would require approximately 300 k bytes. Larger problems could be solved by increasing the fast storage requirements

General solution of the multigroup spherical harmonics equations in R-Z geometry

International Nuclear Information System (INIS)

Matausek, M.

1983-01-01

In the present paper the generalization is performed of the procedure to solve multigroup spherical harmonics equations, which has originally been proposed and developed foe one-dimensional systems in cylindrical or spherical geometry, and later extended for special case of a two-dimensional system in r-z geometry. The expressions are derived for the axial and the radial dependence of the group values of the neutron flux moments, in the P-3 approximation of the spherical harmonics method, in a cylindrically symmetrical system with an arbitrary number of material regions in both r and z directions. In the special case of an axially homogeneous system, these expressions reduce to the relations derived previously. The analysis is performed of the possibilities to satisfy the boundary conditions in the case when the system considered represents an elementary reactor lattice cell and in the case when the system represents a reactor as a whole. The computational effort is estimated for system of a given configuration. (author)

Automated Design and Optimization of Pebble-bed Reactor Cores

International Nuclear Information System (INIS)

Gougar, Hans D.; Ougouag, Abderrafi M.; Terry, William K.

2010-01-01

We present a conceptual design approach for high-temperature gas-cooled reactors using recirculating pebble-bed cores. The design approach employs PEBBED, a reactor physics code specifically designed to solve for and analyze the asymptotic burnup state of pebble-bed reactors, in conjunction with a genetic algorithm to obtain a core that maximizes a fitness value that is a function of user-specified parameters. The uniqueness of the asymptotic core state and the small number of independent parameters that define it suggest that core geometry and fuel cycle can be efficiently optimized toward a specified objective. PEBBED exploits a novel representation of the distribution of pebbles that enables efficient coupling of the burnup and neutron diffusion solvers. With this method, even complex pebble recirculation schemes can be expressed in terms of a few parameters that are amenable to modern optimization techniques. With PEBBED, the user chooses the type and range of core physics parameters that represent the design space. A set of traits, each with acceptable and preferred values expressed by a simple fitness function, is used to evaluate the candidate reactor cores. The stochastic search algorithm automatically drives the generation of core parameters toward the optimal core as defined by the user. The optimized design can then be modeled and analyzed in greater detail using higher resolution and more computationally demanding tools to confirm the desired characteristics. For this study, the design of pebble-bed high temperature reactor concepts subjected to demanding physical constraints demonstrated the efficacy of the PEBBED algorithm.

Validation and Analysis of Forward Osmosis CFD Model in Complex 3D Geometries

Science.gov (United States)

Gruber, Mathias F.; Johnson, Carl J.; Tang, Chuyang; Jensen, Mogens H.; Yde, Lars; Hélix-Nielsen, Claus

2012-01-01

In forward osmosis (FO), an osmotic pressure gradient generated across a semi-permeable membrane is used to generate water transport from a dilute feed solution into a concentrated draw solution. This principle has shown great promise in the areas of water purification, wastewater treatment, seawater desalination and power generation. To ease optimization and increase understanding of membrane systems, it is desirable to have a comprehensive model that allows for easy investigation of all the major parameters in the separation process. Here we present experimental validation of a computational fluid dynamics (CFD) model developed to simulate FO experiments with asymmetric membranes. Simulations are compared with experimental results obtained from using two distinctly different complex three-dimensional membrane chambers. It is found that the CFD model accurately describes the solute separation process and water permeation through membranes under various flow conditions. It is furthermore demonstrated how the CFD model can be used to optimize membrane geometry in such as way as to promote the mass transfer. PMID:24958428

Validation and Analysis of Forward Osmosis CFD Model in Complex 3D Geometries

Directory of Open Access Journals (Sweden)

Lars Yde

2012-11-01

Full Text Available In forward osmosis (FO, an osmotic pressure gradient generated across a semi-permeable membrane is used to generate water transport from a dilute feed solution into a concentrated draw solution. This principle has shown great promise in the areas of water purification, wastewater treatment, seawater desalination and power generation. To ease optimization and increase understanding of membrane systems, it is desirable to have a comprehensive model that allows for easy investigation of all the major parameters in the separation process. Here we present experimental validation of a computational fluid dynamics (CFD model developed to simulate FO experiments with asymmetric membranes. Simulations are compared with experimental results obtained from using two distinctly different complex three-dimensional membrane chambers. It is found that the CFD model accurately describes the solute separation process and water permeation through membranes under various flow conditions. It is furthermore demonstrated how the CFD model can be used to optimize membrane geometry in such as way as to promote the mass transfer.

Control rod studies in small and medium sized fast reactors

International Nuclear Information System (INIS)

John, T.M.; Mohanakrishnan, P.; Mahalakshmi, B.; Singh, R.S.

1988-01-01

Control rods are the primary safety mechanism in the operation of fast reactors. Neutronic parameters associated with the control rods have to be evaluated precisely for studying the behaviour of the reactor under various operating conditions. Control rods are strong neutron absorbers discretely distributed in the reactor core. Accurate estimation of control rod parameters demand, in principle transport theory solutions in exact geometry. But computer codes for such evaluations usually consume exorbitantly large computer time and memory for even a single parameter evaluation. During the design of reactors, evaluation of these parameters will be required for many configurations of control rods. In this paper, the method used at Indira Gandhi Centre for Atomic Research for estimating the parameters associated with control rods is presented. Diffusion theory solutions were used for computations. A scheme using three dimensional geometry represented by triangular meshes and diffusion theory solutions in few energy groups for control rod parameter evaluation is presented. This scheme was employed in estimating the control rod parameters in a 500 Mw(e) fast reactor. Error due to group collapsing is estimated by comparing with 25 group calculations in three dimensions for typical cases. (author). 5 refs, 4 figs, 3 tabs

Laplacians on discrete and quantum geometries

International Nuclear Information System (INIS)

Calcagni, Gianluca; Oriti, Daniele; Thürigen, Johannes

2013-01-01

We extend discrete calculus for arbitrary (p-form) fields on embedded lattices to abstract discrete geometries based on combinatorial complexes. We then provide a general definition of discrete Laplacian using both the primal cellular complex and its combinatorial dual. The precise implementation of geometric volume factors is not unique and, comparing the definition with a circumcentric and a barycentric dual, we argue that the latter is, in general, more appropriate because it induces a Laplacian with more desirable properties. We give the expression of the discrete Laplacian in several different sets of geometric variables, suitable for computations in different quantum gravity formalisms. Furthermore, we investigate the possibility of transforming from position to momentum space for scalar fields, thus setting the stage for the calculation of heat kernel and spectral dimension in discrete quantum geometries. (paper)

Parallelization of the unstructured Navier-stoke solver LILAC for the aero-thermal analysis of a gas-cooled reactor

International Nuclear Information System (INIS)

Kim, J. T.; Kim, S. B.; Lee, W. J.

2004-01-01

Currently lilac code is under development to analyse thermo-hydraulics of the gas-cooled reactor(GCR) especially high-temperature GCR which is one of the gen IV nuclear reactors. The lilac code was originally developed for the analysis of thermo-hydraulics in a molten pool. And now it is modified to resolve the compressible gas flows in the GCR. The more complexities in the internal flow geometries of the GCR reactor and aero-thermal flows, the number of computational cells are increased and finally exceeds the current computing powers of the desktop computers. To overcome the problem and well resolve the interesting physics in the GCR it is conducted to parallels the lilac code by the decomposition of a computational domain or grid. Some benchmark problems are solved with the parallelized lilac code and its speed-up characteristics by the parallel computation is evaluated and described in the article

A study of reactor vessel integrity assessment

Energy Technology Data Exchange (ETDEWEB)

Lee, Jae Hoon [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Kim, Jong Kyung; Shin, Chang Ho; Seo, Bo Kyun [Hanyang Univ., Seoul (Korea, Republic of)

1999-02-15

The fast neutron fluence at the Reactor Pressure Vessel(RPV) of KNGR designed for 60 years lifetime was calculated by full-scope Monte Carlo simulation for reactor vessel integrity assessment. KNGR core geometry was modeled on a three-dimensional representation of the one-sixteenth of the reactor in-vessel component. Each fuel assemblies were modeled explicitly, and each fuel pins were axially divided into 5 segments. The maximum flux of 4.3 x 10{sup 10} neutrons/cm{sup 2}. sec at the RPV was obtained by tallying neutrons crossing the beltline of inner surface of the RPV.

HETERO code, heterogeneous procedure for reactor calculation

International Nuclear Information System (INIS)

Jovanovic, S.M.; Raisic, N.M.

1966-11-01

This report describes the procedure for calculating the parameters of heterogeneous reactor system taking into account the interaction between fuel elements related to established geometry. First part contains the analysis of single fuel element in a diffusion medium, and criticality condition of the reactor system described by superposition of elements interactions. the possibility of performing such analysis by determination of heterogeneous system lattice is described in the second part. Computer code HETERO with the code KETAP (calculation of criticality factor η n and flux distribution) is part of this report together with the example of RB reactor square lattice

Geometries

CERN Document Server

Sossinsky, A B

2012-01-01

The book is an innovative modern exposition of geometry, or rather, of geometries; it is the first textbook in which Felix Klein's Erlangen Program (the action of transformation groups) is systematically used as the basis for defining various geometries. The course of study presented is dedicated to the proposition that all geometries are created equal--although some, of course, remain more equal than others. The author concentrates on several of the more distinguished and beautiful ones, which include what he terms "toy geometries", the geometries of Platonic bodies, discrete geometries, and classical continuous geometries. The text is based on first-year semester course lectures delivered at the Independent University of Moscow in 2003 and 2006. It is by no means a formal algebraic or analytic treatment of geometric topics, but rather, a highly visual exposition containing upwards of 200 illustrations. The reader is expected to possess a familiarity with elementary Euclidean geometry, albeit those lacking t...

Reactor assessments of advanced bumpy torus configurations

International Nuclear Information System (INIS)

Uckan, N.A.; Owen, L.W.; Spong, D.A.; Miller, R.L.; Ard, W.B.; Pipkins, J.F.; Schmitt, R.J.

1984-02-01

Recently, several innovative approaches were introduced for enhancing the performance of the basic ELMO Bumpy Torus (EBT) concept and for improving its reactor potential. These include planar racetrack and square geometries, Andreoletti coil systems, and bumpy torus-stellarator hybrids (which include twisted racetrack and helical axis stellarator - snakey torus). Preliminary evaluations of reactor implications of each approach have been carried out based on magnetics (vacuum) calculations, transport and scaling relationships, and stability properties deduced from provisional configurations that implement the approach but are not necessarily optimized. Further optimization is needed in all cases to evaluate the full potential of each approach. Results of these studies indicate favorable reactor projections with a significant reduction in reactor physical size as compared to conventional EBT reactor designs carried out in the past

High-order discrete ordinate transport in hexagonal geometry: A new capability in ERANOS

International Nuclear Information System (INIS)

Le Tellier, R.; Suteau, C.; Fournier, D.; Ruggieri, J.M.

2010-01-01

This paper presents the implementation of an arbitrary order discontinuous Galerkin scheme within the framework of a discrete ordinate solver of the neutron transport equation for nuclear reactor calculations. More precisely, it deals with non-conforming spatial meshes for the 2 D and 3 D modeling of core geometries based on hexagonal assemblies. This work aims at improving the capabilities of the ERANOS code system dedicated to fast reactor analysis and design. Both the angular quadrature and spatial scheme peculiarities for hexagonal geometries are presented. A particular focus is set on the spatial non-conforming mesh and variable order capabilities of this scheme in anticipation to the development of spatial adaptiveness algorithms. These features are illustrated on a 3 D numerical benchmark with comparison to a Monte Carlo reference and a 2 D benchmark that shows the potential of this scheme for both h-and p-adaptation.

Computer realization of an algorithm for determining the optimal arrangement of a fast power reactor core with hexagonal assemblies

International Nuclear Information System (INIS)

Karpov, V.A.; Rybnikov, A.F.

1983-01-01

An algorithm for solving the problems associated with fast nuclear reactor computer-aided design is suggested. Formulation of the discrete optimization problem dealing with chosing of the first loading arrangement, determination of the control element functional purpose and the order of their rearrangement during reactor operation as well as the choice of operations for core reloading is given. An algorithm for computerized solutions of the mentioned optimization problem based on variational methods relized in the form of the DESIGN program complex written in FORTRAN for the BEhSM-6 computer is proposed. A fast-response program for solving the diffusion equations of two-dimensional reactor permitting to obtain the optimization problem solution at reasonable period of time is developed to conduct necessary neutron-physical calculations for the reactor in hexagonal geometry. The DESIGN program can be included into a computer-aided design system for automation of the procedure of determining the fast power reactor core arrangement. Application of the DESIGN program permits to avoid the routine calculations on substantiation of neutron-physical and thermal-hydraulic characteristics of the reactor core that releases operators from essential waste of time and increases efficiency of their work

Radiation shielding for fusion reactors

International Nuclear Information System (INIS)

Santoro, R.T.

2000-01-01

Radiation shielding requirements for fusion reactors present different problems than those for fission reactors and accelerators. Fusion devices, particularly tokamak reactors, are complicated by geometry constraints that complicate disposition of fully effective shielding. This paper reviews some of these shielding issues and suggested solutions for optimizing the machine and biological shielding. Radiation transport calculations are essential for predicting and confirming the nuclear performance of the reactor and, as such, must be an essential part of the reactor design process. Development and optimization of reactor components from the first wall and primary shielding to the penetrations and containment shielding must be carried out in a sensible progression. Initial results from one-dimensional transport calculations are used for scoping studies and are followed by detailed two- and three-dimensional analyses to effectively characterize the overall radiation environment. These detail model calculations are essential for accounting for the radiation leakage through ports and other penetrations in the bulk shield. Careful analysis of component activation and radiation damage is cardinal for defining remote handling requirements, in-situ replacement of components, and personnel access at specific locations inside the reactor containment vessel. (author)

Operation and utilization of low power research reactor critical facility for Advanced Heavy Water Reactor (AHWR)

International Nuclear Information System (INIS)

De, S.K.; Karhadkar, C.G.

2017-01-01

An Advanced Heavy Water Reactor (AHWR) has been designed and developed for maximum power generation from thorium considering large reserves of thorium. The design envisages using 54 pin MOX cluster with different enrichment of "2"3"3U and Pu in Thoria fuel pins. Theoretical models developed to neutron transport and the geometrical details of the reactor including all reactivity devices involve approximations in modelling, resulting in uncertainties. With a view to minimize these uncertainties, a low power research reactor Critical Facility was built in which cold clean fuel can be arranged in a desired and precise geometry. Different experiments conducted in this facility greatly contribute to understand and validate the physics design parameters

Creep/fatigue damage prediction of fast reactor components using shakedown methods

International Nuclear Information System (INIS)

Buckthorpe, D.E.

1997-01-01

The present status of the shakedown method is reviewed, the application of the shakedown based principles to complex hardening and creep behaviour is described and justified and the prediction of damage against design criteria outlined. Comparisons are made with full inelastic analysis solutions where these are available and against damage assessments using elastic and inelastic design code methods. Current and future developments of the method are described including a summary of the advances made in the development of the post process ADAPT, which has enabled the method to be applied to complex geometry features and loading cases. The paper includes a review of applications of the method to typical Fast Reactor structural example cases within the primary and secondary circuits. For the primary circuit this includes structures such as the large diameter internal shells which are surrounded by hot sodium and subject to slow and rapid thermal transient loadings. One specific case is the damage assessment associated with thermal stratifications within sodium and the effects of moving sodium surfaces arising from reactor trip conditions. Other structures covered are geometric features within components such as the Above Core structure and Intermediate Heat Exchanger. For the secondary circuit the method has been applied to alternative and more complex forms of geometry namely thick section tubeplates of the Steam Generator and a typical secondary circuit piping run. Both of these applications are in an early stage of development but are expected to show significant advantages with respect to creep and fatigue damage estimation compared with existing code methods. The principle application of the method to design has so far been focused on Austenitic Stainless steel components however current work shows some significant benefits may be possible from the application of the method to structures made from Ferritic steels such as Modified 9Cr 1Mo. This aspect is briefly

The geometry of some natural conjugacies in ℂn dynamics

Directory of Open Access Journals (Sweden)

John W. Robertson

2004-01-01

Full Text Available We show that under some simple conditions a topological conjugacy h between two holomorphic self-maps f1 and f2 of complex n-dimensional projective space ℙn lifts canonically to a topological conjugacy H between the two corresponding polynomial self-maps of ℂn+1, and this conjugacy relates the two Green functions of f1 and f2. These conjugacies are interesting because their geometry is not inherited entirely from the geometry of the conjugacy on ℙn. Part of the geometry of such a conjugacy is given (locally by a complex-valued function whose absolute value is determined by the Green functions for the two maps, but whose argument seems to appear out of thin air. We work out the local geometry of such conjugacies over the Fatou set and over Fatou varieties of the original map.

Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report

International Nuclear Information System (INIS)

Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.

1998-03-01

The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [de

Solving the neutron diffusion equation on combinatorial geometry computational cells for reactor physics calculations

International Nuclear Information System (INIS)

Azmy, Y. Y.

2004-01-01

An approach is developed for solving the neutron diffusion equation on combinatorial geometry computational cells, that is computational cells composed by combinatorial operations involving simple-shaped component cells. The only constraint on the component cells from which the combinatorial cells are assembled is that they possess a legitimate discretization of the underlying diffusion equation. We use the Finite Difference (FD) approximation of the x, y-geometry diffusion equation in this work. Performing the same combinatorial operations involved in composing the combinatorial cell on these discrete-variable equations yields equations that employ new discrete variables defined only on the combinatorial cell's volume and faces. The only approximation involved in this process, beyond the truncation error committed in discretizing the diffusion equation over each component cell, is a consistent-order Legendre series expansion. Preliminary results for simple configurations establish the accuracy of the solution to the combinatorial geometry solution compared to straight FD as the system dimensions decrease. Furthermore numerical results validate the consistent Legendre-series expansion order by illustrating the second order accuracy of the combinatorial geometry solution, the same as standard FD. Nevertheless the magnitude of the error for the new approach is larger than FD's since it incorporates the additional truncated series approximation. (authors)

Microinstabilities in complex magnetic field geometries and high-β sheared sheath structure. Progress report, June 1, 1975--February 27, 1976

International Nuclear Information System (INIS)

Bakshi, P.; Kalman, G.

1976-02-01

A new approach for the solution of the Vlasov equation for complex magnetic field geometries has been developed using operator techniques. The general approach is illustrated by determining the perturbed distribution function and density operator for the problem of shear stabilization of drift waves for transverse and arbitrary directions of propagation. The ensuing corrections to stability criteria of current theories are obtained for certain domains of physical parameters. Preliminary work on the integral equation approach to the dispersion relation has been initiated. As a prelude to the study of particle orbits in complex mirror geometries, the adiabatic and non-adiabatic behavior of a harmonic oscillator has been studied using operator methods. High-β, high shear plasma sheath configurations have been studied with the full ion dynamics taken into account and electrons treated in the zero and first order approximation, in the ratio of the electron Larmor radius to the scale length. The resulting sheath structure equation in the lowest order approximation has been solved for certain entering ion distributions, and prepared for computer analysis for others. In this approximation the electron current parallel to magnetic field lines has to be assumed suppressed or predetermined. Equations in the next order approximation include the finite Larmor radius stress tensor. This equation is under study

Spectral dimension of quantum geometries

International Nuclear Information System (INIS)

Calcagni, Gianluca; Oriti, Daniele; Thürigen, Johannes

2014-01-01

The spectral dimension is an indicator of geometry and topology of spacetime and a tool to compare the description of quantum geometry in various approaches to quantum gravity. This is possible because it can be defined not only on smooth geometries but also on discrete (e.g., simplicial) ones. In this paper, we consider the spectral dimension of quantum states of spatial geometry defined on combinatorial complexes endowed with additional algebraic data: the kinematical quantum states of loop quantum gravity (LQG). Preliminarily, the effects of topology and discreteness of classical discrete geometries are studied in a systematic manner. We look for states reproducing the spectral dimension of a classical space in the appropriate regime. We also test the hypothesis that in LQG, as in other approaches, there is a scale dependence of the spectral dimension, which runs from the topological dimension at large scales to a smaller one at short distances. While our results do not give any strong support to this hypothesis, we can however pinpoint when the topological dimension is reproduced by LQG quantum states. Overall, by exploring the interplay of combinatorial, topological and geometrical effects, and by considering various kinds of quantum states such as coherent states and their superpositions, we find that the spectral dimension of discrete quantum geometries is more sensitive to the underlying combinatorial structures than to the details of the additional data associated with them. (paper)

Reactor vessel nozzle cracks: a photoelastic study

International Nuclear Information System (INIS)

Smith, C.W.

1979-01-01

A method consisting of a marriage between the ''frozen stress'' photoelastic approach and the local stress field equations of linear elastic fracture mechanics for estimating stress intensity factor distributions in three dimensional, finite cracked body problems is reviewed and extensions of the method are indicated. The method is then applied to the nuclear reactor vessel nozzle corner crack problem for both Intermediate Test Vessel and Boiling Water Reactor geometries. Results are compared with those of other investigators. 35 refs

Development of a 3-D flow analysis computer program for integral reactor

International Nuclear Information System (INIS)

Youn, H. Y.; Lee, K. H.; Kim, H. K.; Whang, Y. D.; Kim, H. C.

2003-01-01

A 3-D computational fluid dynamics program TASS-3D is being developed for the flow analysis of primary coolant system consists of complex geometries such as SMART. A pre/post processor also is being developed to reduce the pre/post processing works such as a computational grid generation, set-up the analysis conditions and analysis of the calculated results. TASS-3D solver employs a non-orthogonal coordinate system and FVM based on the non-staggered grid system. The program includes the various models to simulate the physical phenomena expected to be occurred in the integral reactor and will be coupled with core dynamics code, core T/H code and the secondary system code modules. Currently, the application of TASS-3D is limited to the single phase of liquid, but the code will be further developed including 2-phase phenomena expected for the normal operation and the various transients of the integrator reactor in the next stage

NF-6 program complex for BESM-6 computation of the basic neutron-physical characteristics of nuclear reactors

International Nuclear Information System (INIS)

Zizin, M.N.; Savochkina, O.A.; Chukhlova, O.P.

1978-01-01

A structure of standard designations is described and semantics of a number of standard values used in a NF-6 program complex is given. Main source data and results of neutron-physical reactor calculation are standard values, the peculiarities of FORTRAN and ALGOL-GDR algorithm languages in the DUBNA monitoring system were taken account of. As a base of standard values list the FIHAR system list, supplemented with new standard designations for integral reactor characteristics, is used. Developed is also a list of standard values to organize the exchange with external memory in the process of task solution and long-range storage

High Order Finite Element Method for the Lambda modes problem on hexagonal geometry

International Nuclear Information System (INIS)

Gonzalez-Pintor, S.; Ginestar, D.; Verdu, G.

2009-01-01

A High Order Finite Element Method to approximate the Lambda modes problem for reactors with hexagonal geometry has been developed. This method is based on the expansion of the neutron flux in terms of the modified Dubiner's polynomials on a triangular mesh. This mesh is fixed and the accuracy of the method is improved increasing the degree of the polynomial expansions without the necessity of remeshing. The performance of method has been tested obtaining the dominant Lambda modes of different 2D reactor benchmark problems.

Random geometry capability in RMC code for explicit analysis of polytype particle/pebble and applications to HTR-10 benchmark

International Nuclear Information System (INIS)

Liu, Shichang; Li, Zeguang; Wang, Kan; Cheng, Quan; She, Ding

2018-01-01

Highlights: •A new random geometry was developed in RMC for mixed and polytype particle/pebble. •This capability was applied to the full core calculations of HTR-10 benchmark. •Reactivity, temperature coefficient and control rod worth of HTR-10 were compared. •This method can explicitly model different packing fraction of different pebbles. •Monte Carlo code with this method can simulate polytype particle/pebble type reactor. -- Abstract: With the increasing demands of high fidelity neutronics analysis and the development of computer technology, Monte Carlo method is becoming more and more attractive in accurate simulation of pebble bed High Temperature gas-cooled Reactor (HTR), owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. For the double-heterogeneous geometry of pebble bed, traditional Monte Carlo codes can treat it by explicit geometry description. However, packing methods such as Random Sequential Addition (RSA) can only produce a sphere packing up to 38% volume packing fraction, while Discrete Element Method (DEM) is troublesome and also time consuming. Moreover, traditional Monte Carlo codes are difficult and inconvenient to simulate the mixed and polytype particles or pebbles. A new random geometry method was developed in Monte Carlo code RMC to simulate the particle transport in polytype particle/pebble in double heterogeneous geometry systems. This method was verified by some test cases, and applied to the full core calculations of HTR-10 benchmark. The reactivity, temperature coefficient and control rod worth of HTR-10 were compared for full core and initial core in helium and air atmosphere respectively, and the results agree well with the benchmark results and experimental results. This work would provide an efficient tool for the innovative design of pebble bed, prism HTRs and molten salt reactors with polytype particles or pebbles using Monte Carlo method.

Coolant mixing in pressurized water reactors

Energy Technology Data Exchange (ETDEWEB)

Hoehne, T; Grunwald, G

1998-10-01

The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

Fractal geometry and number theory complex dimensions of fractal strings and zeros of zeta functions

CERN Document Server

Lapidus, Michael L

1999-01-01

A fractal drum is a bounded open subset of R. m with a fractal boundary. A difficult problem is to describe the relationship between the shape (geo­ metry) of the drum and its sound (its spectrum). In this book, we restrict ourselves to the one-dimensional case of fractal strings, and their higher dimensional analogues, fractal sprays. We develop a theory of complex di­ mensions of a fractal string, and we study how these complex dimensions relate the geometry with the spectrum of the fractal string. We refer the reader to [Berrl-2, Lapl-4, LapPol-3, LapMal-2, HeLapl-2] and the ref­ erences therein for further physical and mathematical motivations of this work. (Also see, in particular, Sections 7. 1, 10. 3 and 10. 4, along with Ap­ pendix B. ) In Chapter 1, we introduce the basic object of our research, fractal strings (see [Lapl-3, LapPol-3, LapMal-2, HeLapl-2]). A 'standard fractal string' is a bounded open subset of the real line. Such a set is a disjoint union of open intervals, the lengths of which ...

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

Energy Technology Data Exchange (ETDEWEB)

Rauck, St

2000-10-01

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

PITR: Princeton Ignition Test Reactor

International Nuclear Information System (INIS)

1978-12-01

The principal objectives of the PITR - Princeton Ignition Test Reactor - are to demonstrate the attainment of thermonuclear ignition in deuterium-tritium, and to develop optimal start-up techniques for plasma heating and current induction, in order to determine the most favorable means of reducing the size and cost of tokamak power reactors. This report describes the status of the plasma and engineering design features of the PITR. The PITR geometry is chosen to provide the highest MHD-stable values of beta in a D-shaped plasma, as well as ease of access for remote handling and neutral-beam injection

Geometry

CERN Document Server

Prasolov, V V

2015-01-01

This book provides a systematic introduction to various geometries, including Euclidean, affine, projective, spherical, and hyperbolic geometries. Also included is a chapter on infinite-dimensional generalizations of Euclidean and affine geometries. A uniform approach to different geometries, based on Klein's Erlangen Program is suggested, and similarities of various phenomena in all geometries are traced. An important notion of duality of geometric objects is highlighted throughout the book. The authors also include a detailed presentation of the theory of conics and quadrics, including the theory of conics for non-Euclidean geometries. The book contains many beautiful geometric facts and has plenty of problems, most of them with solutions, which nicely supplement the main text. With more than 150 figures illustrating the arguments, the book can be recommended as a textbook for undergraduate and graduate-level courses in geometry.

DRAGON, Reactor Cell Calculation System with Burnup

International Nuclear Information System (INIS)

2007-01-01

1 - Description of program or function: DRAGON is a collection of models to simulate the neutronic behavior of a unit cell or a fuel assembly in a nuclear reactor. It includes all of the functions that characterize a lattice cell code, namely: interpolation of microscopic cross sections supplied by means of standard libraries; resonance self-shielding calculations in multidimensional geometries; multigroup and multidimensional neutron flux calculations which can take into account neutron leakage; transport-transport or transport-diffusion equivalence calculations as well as editing of condensed and homogenized nuclear properties for reactor calculations; and finally isotopic depletion calculations. The user must supply cross sections. DRAGON can access directly standard microscopic cross-section libraries in the following formats: DRAGON, MATXS (TRANSX-CTR), WIMSD4, WIMS-AECL, and APOLLO. It has the capability of exchanging macroscopic and microscopic cross-section libraries with a code such as PSR-0206/TRANSX-CTR or PSR-0317/TRANSX-2 by the use of the GOXS and ISOTXS format files. Macroscopic cross sections can also be read in DRAGON via the input data stream. 2 - Method of solution: DRAGON contains a multigroup iterator conceived to control a number of different algorithms for the solution of the neutron transport equation. Each of these algorithms is presented in the form of a one-group solution procedure where the contributions from other energy groups are included in a source term. The current version, DRAGON 9 71124 (Release 3.02), which was released in January 1998, contains three such algorithms. The JPM option solves the integral transport equation using the interface current method applied to homogeneous blocks; the SYBIL option solves the integral transport equation using the collision probability method for simple one-dimensional (1-D) or two-dimensional (2-D) geometries and the interface current method for 2-D Cartesian or hexagonal assemblies; and the

Validation of Reactor Physics-Thermal hydraulics Calculations for Research Reactors Cooled by the Laminar Flow of Water

Energy Technology Data Exchange (ETDEWEB)

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

A collaboration between the University of Florida and the Swiss Federal Institute of Technology, Lausanne (EPFL) has been formed to develop and validate detailed coupled multiphysics models of the zero-power (100 W) CROCUS reactor at EPFL and the 100 kW University of Florida Training Reactor, for the comprehensive analysis of the reactor behavior under transient (neutronic or thermal-hydraulic induced) conditions. These two reactors differ significantly in the core design and thermal power output, but share unique heat transfer and flow characteristics. They are characterized by single-phase laminar water flow at near-atmospheric pressures in complex geometries with the possibility of mechanically entrained air bubbles. Validation experiments will be designed to expand the validation domain of these existing models, computational codes and techniques. In this process, emphasis will be placed on validation of the coupled models developed to gain confidence in their applicability for safety analysis. EPFL is responsible for the design and implementation of transient experiments to generate a database of reactor parameters (flow distribution, power profile, and power evolution) to be used to validate against code predictions. The transient experiments performed at EPFL will be simulated on the basis of developed models for these tasks. Comparative analysis will be performed with SERPENT and MCNPX reference core models. UF focuses on the generation of the coupled neutron kinetics and thermal-hydraulic models, including implementation of a TRACE/PARCS reactor simulator model, a PARET model, and development of full-field computational fluid dynamics models (using OpenFOAM) for refined thermal-hydraulics physics treatments. In this subtask of the project, the aim is to verify by means of CFD the validity of TRACE predictions for near-atmospheric pressure water flow in the presence of mechanically entrained air bubbles. The scientific understanding of these multiphysics

Validation of Reactor Physics-Thermal hydraulics Calculations for Research Reactors Cooled by the Laminar Flow of Water

International Nuclear Information System (INIS)

Jordan, K. A.; Schubring, D.; Girardin, G.; Pautz, A.

2013-01-01

A collaboration between the University of Florida and the Swiss Federal Institute of Technology, Lausanne (EPFL) has been formed to develop and validate detailed coupled multiphysics models of the zero-power (100 W) CROCUS reactor at EPFL and the 100 kW University of Florida Training Reactor, for the comprehensive analysis of the reactor behavior under transient (neutronic or thermal-hydraulic induced) conditions. These two reactors differ significantly in the core design and thermal power output, but share unique heat transfer and flow characteristics. They are characterized by single-phase laminar water flow at near-atmospheric pressures in complex geometries with the possibility of mechanically entrained air bubbles. Validation experiments will be designed to expand the validation domain of these existing models, computational codes and techniques. In this process, emphasis will be placed on validation of the coupled models developed to gain confidence in their applicability for safety analysis. EPFL is responsible for the design and implementation of transient experiments to generate a database of reactor parameters (flow distribution, power profile, and power evolution) to be used to validate against code predictions. The transient experiments performed at EPFL will be simulated on the basis of developed models for these tasks. Comparative analysis will be performed with SERPENT and MCNPX reference core models. UF focuses on the generation of the coupled neutron kinetics and thermal-hydraulic models, including implementation of a TRACE/PARCS reactor simulator model, a PARET model, and development of full-field computational fluid dynamics models (using OpenFOAM) for refined thermal-hydraulics physics treatments. In this subtask of the project, the aim is to verify by means of CFD the validity of TRACE predictions for near-atmospheric pressure water flow in the presence of mechanically entrained air bubbles. The scientific understanding of these multiphysics

Determination of the transfer function of a reactor

International Nuclear Information System (INIS)

Dencs, B.

1976-01-01

The theoretical and experimental methods of the determination of reactor transfer functions are reviewed. Preliminary measurements were made on the experimental and final core of the training reactor of the Budapest Technical University. The rod-drop curves, the hole effect of the reactor and the control rod worths were determined. The effect of Cd ring and Cd profile was studied, too. The neutron flux distribution in the core was determined in several geometries. The oscillatory method is treated in detail. After the zero measurements of the core the oscillatory determination of the transfer function has been made on some frequency. The simplified model of the reactor transfer function was reconstructed from the measurement data. (R.J.)

Beryllium reflectors for research reactors. Review and preliminary finite element analysis

Energy Technology Data Exchange (ETDEWEB)

Bejarano, Pablo S; Cocco, Roxana G., E-mail: rcocco@invap.com.ar [INVAP S.E., Rio Negro (Argentina)

2012-03-15

Beryllium is used in numerous research reactors to moderate neutron energy and to reflect neutrons back into the core, thus intensifying the thermal neutron flux. However, beryllium is degraded by radiation damage, as a result of both displacement and transmutation. Displacement damage leads to point defect clustering, irradiation hardening and embrittlement. Transmutation produces helium, which results in high levels of gas and swelling, even at low temperatures. A brief state-of-the-art review on the use of reflector assemblies reveals that each user has adopted a different method for overcoming problems related to swelling: strengthening, cracking and distortion. In the present work a preliminary study about the geometry influence on the reflector assembly behavior was performed by a Finite Element Analysis (FEA). A simplified study was made varying its geometry in height, thickness and width. The results showed that the most influencing parameter in avoiding distortion due to swelling is firstly the reflector's assembly height, H; secondly its thickness, L, and lastly its angle/width, {theta}. These results contribute to the understanding of distortion behavior and the stresses generated in a simple geometry Be bar subjected to radiation, which can be a useful tool for mechanical design of more complex components. (author)

Flow inversion and natural convection in a MTR (Materials Testing Reactor)

International Nuclear Information System (INIS)

Gimenez, M.O.; Clausse, A.

1990-01-01

The thermohydraulic evolution of a refrigerating channel of the MTR (Materials Testing Reactors) RA-6 reactor's core, at the Bariloche Atomic Center, has been studied during the transient caused by the primary system's pump decommissioning. This transient constitutes one of the reactor's operating power boundaries due to the maximum temperature permissible in fuel plates. The problem regarding the thermohydraulic code altered for the rectangular geometry calculation characteristic of the MTR design is analyzed. (Author) [es

Effect of hydraulic retention time on the biodegradation of complex phenolic mixture from simulated coal wastewater in hybrid UASB reactors

Energy Technology Data Exchange (ETDEWEB)

Ramakrishnan, Anushuya [Centre for Environmental Science and Engineering, Indian Institute of Technology Bombay, Powai, Mumbai 400076 (India); Gupta, Sudhir Kumar [Centre for Environmental Science and Engineering, Indian Institute of Technology Bombay, Powai, Mumbai 400076 (India)], E-mail: skgupta@iitb.ac.in

2008-05-01

This study describes the feasibility of anaerobic treatment of complex phenolics mixture from a simulated synthetic coal wastewater using four identical 13.5 L (effective volume) bench scale hybrid up-flow anaerobic sludge blanket (HUASB) (combining UASB + anaerobic filter) reactors at four different hydraulic retention times (HRT) under mesophilic (27 {+-} 5 {sup o}C) conditions. Synthetic coal wastewater with an average chemical oxygen demand (COD) of 2240 mg/L and phenolics concentration of 752 mg/L was used as substrate. The phenolics contained phenol (490 mg/L); m-, o-, p-cresols (123.0, 58.6, 42 mg/L); 2,4-, 2,5-, 3,4- and 3,5-dimethyl phenols (6.3, 6.3, 4.4 and 21.3 mg/L) as major phenolic compounds. The study demonstrated that at optimum HRT, 24 h, and phenolic loading rate of 0.75 g COD/(m{sup 3}-d), the phenolics and COD removal efficiency of the reactors were 96% and 86%, respectively. Bio-kinetic models were applied to data obtained from experimental studies in hybrid UASB reactor. Grau second-order multi-component substrate removal model was best fitted to the hybrid UASB reactor. The second-order substrate removal rate constant (k{sub 2(s)}) was found as 1.72 h{sup -1} for the hybrid reactor treating complex phenolic mixture. Morphological examination of the sludge revealed rod-type Methanothrix-like, cells to be dominant on the surface.

Effect of hydraulic retention time on the biodegradation of complex phenolic mixture from simulated coal wastewater in hybrid UASB reactors

International Nuclear Information System (INIS)

Ramakrishnan, Anushuya; Gupta, Sudhir Kumar

2008-01-01

This study describes the feasibility of anaerobic treatment of complex phenolics mixture from a simulated synthetic coal wastewater using four identical 13.5 L (effective volume) bench scale hybrid up-flow anaerobic sludge blanket (HUASB) (combining UASB + anaerobic filter) reactors at four different hydraulic retention times (HRT) under mesophilic (27 ± 5 o C) conditions. Synthetic coal wastewater with an average chemical oxygen demand (COD) of 2240 mg/L and phenolics concentration of 752 mg/L was used as substrate. The phenolics contained phenol (490 mg/L); m-, o-, p-cresols (123.0, 58.6, 42 mg/L); 2,4-, 2,5-, 3,4- and 3,5-dimethyl phenols (6.3, 6.3, 4.4 and 21.3 mg/L) as major phenolic compounds. The study demonstrated that at optimum HRT, 24 h, and phenolic loading rate of 0.75 g COD/(m 3 -d), the phenolics and COD removal efficiency of the reactors were 96% and 86%, respectively. Bio-kinetic models were applied to data obtained from experimental studies in hybrid UASB reactor. Grau second-order multi-component substrate removal model was best fitted to the hybrid UASB reactor. The second-order substrate removal rate constant (k 2(s) ) was found as 1.72 h -1 for the hybrid reactor treating complex phenolic mixture. Morphological examination of the sludge revealed rod-type Methanothrix-like, cells to be dominant on the surface

State of the art seismic analysis for CANDU reactor structure components using condensation method

Energy Technology Data Exchange (ETDEWEB)

Soliman, S A; Ibraham, A M; Hodgson, S [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

1996-12-31

The reactor structure assembly seismic analysis is a relatively complex process because of the intricate geometry with many different discontinuities, and due to the hydraulic attached mass which follows the structure during its vibration. In order to simulate reasonably accurate behaviour of the reactor structure assembly, detailed finite element models are generated and used for both modal and stress analysis. Guyan reduction condensation method was used in the analysis. The attached mass, which includes the fluid mass contained in the components plus the added mass which accounts for the inertia of the surrounding fluid entrained by the accelerating structure immersed in the fluid, was calculated and attached to the vibrating structures. The masses of the attached components, supported partly or totally by the assembly which includes piping, reactivity control units, end fittings, etc. are also considered in the analysis. (author). 4 refs., 6 tabs., 4 figs.

MORET: a Monte Carlo program for fast computation of the effective multiplying factors of fissile media within complex geometries

International Nuclear Information System (INIS)

Caizergues, Robert; Poullot, Gilles; Teillet, J.-R.

1976-06-01

The MORET code determines effective multiplying factors. It uses the Monte Carlo technique and the multigroup theory; a collision is taken as isotropic, but anisotropy is taken into account by means of the transport correction. Complex geometries can be rapidly treated: the array to be studied is divided in simple elementary volumes (spheres, cylinders, boxes, cones, half space planes...) to which are applied operators of the theory of sets. Some constant or differential (albedos) reflection coefficients simulate neighboring reflections on the outer volume [fr

Vectorising the detector geometry to optimize particle transport

CERN Document Server

Apostolakis, John; Carminati, Federico; Gheata, Andrei; Wenzel, Sandro

2014-01-01

Among the components contributing to particle transport, geometry navigation is an important consumer of CPU cycles. The tasks performed to get answers to "basic" queries such as locating a point within a geometry hierarchy or computing accurately the distance to the next boundary can become very computing intensive for complex detector setups. So far, the existing geometry algorithms employ mainly scalar optimisation strategies (voxelization, caching) to reduce their CPU consumption. In this paper, we would like to take a different approach and investigate how geometry navigation can benefit from the vector instruction set extensions that are one of the primary source of performance enhancements on current and future hardware. While on paper, this form of microparallelism promises increasing performance opportunities, applying this technology to the highly hierarchical and multiply branched geometry code is a difficult challenge. We refer to the current work done to vectorise an important part of the critica...

Differential geometry based multiscale models.

Science.gov (United States)

Wei, Guo-Wei

2010-08-01

Large chemical and biological systems such as fuel cells, ion channels, molecular motors, and viruses are of great importance to the scientific community and public health. Typically, these complex systems in conjunction with their aquatic environment pose a fabulous challenge to theoretical description, simulation, and prediction. In this work, we propose a differential geometry based multiscale paradigm to model complex macromolecular systems, and to put macroscopic and microscopic descriptions on an equal footing. In our approach, the differential geometry theory of surfaces and geometric measure theory are employed as a natural means to couple the macroscopic continuum mechanical description of the aquatic environment with the microscopic discrete atomistic description of the macromolecule. Multiscale free energy functionals, or multiscale action functionals are constructed as a unified framework to derive the governing equations for the dynamics of different scales and different descriptions. Two types of aqueous macromolecular complexes, ones that are near equilibrium and others that are far from equilibrium, are considered in our formulations. We show that generalized Navier-Stokes equations for the fluid dynamics, generalized Poisson equations or generalized Poisson-Boltzmann equations for electrostatic interactions, and Newton's equation for the molecular dynamics can be derived by the least action principle. These equations are coupled through the continuum-discrete interface whose dynamics is governed by potential driven geometric flows. Comparison is given to classical descriptions of the fluid and electrostatic interactions without geometric flow based micro-macro interfaces. The detailed balance of forces is emphasized in the present work. We further extend the proposed multiscale paradigm to micro-macro analysis of electrohydrodynamics, electrophoresis, fuel cells, and ion channels. We derive generalized Poisson-Nernst-Planck equations that are

Differential Geometry Based Multiscale Models

Science.gov (United States)

Wei, Guo-Wei

2010-01-01

Large chemical and biological systems such as fuel cells, ion channels, molecular motors, and viruses are of great importance to the scientific community and public health. Typically, these complex systems in conjunction with their aquatic environment pose a fabulous challenge to theoretical description, simulation, and prediction. In this work, we propose a differential geometry based multiscale paradigm to model complex macromolecular systems, and to put macroscopic and microscopic descriptions on an equal footing. In our approach, the differential geometry theory of surfaces and geometric measure theory are employed as a natural means to couple the macroscopic continuum mechanical description of the aquatic environment with the microscopic discrete atom-istic description of the macromolecule. Multiscale free energy functionals, or multiscale action functionals are constructed as a unified framework to derive the governing equations for the dynamics of different scales and different descriptions. Two types of aqueous macromolecular complexes, ones that are near equilibrium and others that are far from equilibrium, are considered in our formulations. We show that generalized Navier–Stokes equations for the fluid dynamics, generalized Poisson equations or generalized Poisson–Boltzmann equations for electrostatic interactions, and Newton's equation for the molecular dynamics can be derived by the least action principle. These equations are coupled through the continuum-discrete interface whose dynamics is governed by potential driven geometric flows. Comparison is given to classical descriptions of the fluid and electrostatic interactions without geometric flow based micro-macro interfaces. The detailed balance of forces is emphasized in the present work. We further extend the proposed multiscale paradigm to micro-macro analysis of electrohydrodynamics, electrophoresis, fuel cells, and ion channels. We derive generalized Poisson–Nernst–Planck equations that

Single-Phase Crossflow Mixing in a Vertical Tube Bundle Geometry : An Experimental Study

NARCIS (Netherlands)

Mahmood, A.

2011-01-01

The vertical rod/tube bundle geometry has a wide variety of industrial applications. Typical examples are the core of light water nuclear reactors (LWR) and vertical tube steam generators. In the core of a LWR, primarily coolant flows upward but their also exist a flow in lateral direction, called

A short course in computational geometry and topology

CERN Document Server

Edelsbrunner, Herbert

2014-01-01

With the aim to bring the subject of Computational Geometry and Topology closer to the scientific audience, this book is written in thirteen ready-to-teach sections organized in four parts: tessellations, complexes, homology, persistence. To speak to the non-specialist, detailed formalisms are often avoided in favor of lively 2- and 3-dimensional illustrations. The book is warmly recommended to everybody who loves geometry and the fascinating world of shapes.

Effect of cosine current approximation in lattice cell calculations in cylindrical geometry

International Nuclear Information System (INIS)

Mohanakrishnan, P.

1978-01-01

It is found that one-dimensional cylindrical geometry reactor lattice cell calculations using cosine angular current approximation at spatial mesh interfaces give results surprisingly close to the results of accurate neutron transport calculations as well as experimental measurements. This is especially true for tight light water moderated lattices. Reasons for this close agreement are investigated here. By re-examining the effects of reflective and white cell boundary conditions in these calculations it is concluded that one major reason is the use of white boundary condition necessitated by the approximation of the two-dimensional reactor lattice cell by a one-dimensional one. (orig.) [de

Studies on validation possibilities for computational codes for criticality and burnup calculations of boiling water reactor fuel; Untersuchungen zu Validierungsmoeglichkeiten von Rechencodes fuer Kritikalitaets- und Abbrandrechnungen von Siedewasserreaktor-Brennstoff

Energy Technology Data Exchange (ETDEWEB)

Behler, Matthais; Hannstein, Volker; Kilger, Robert; Sommer, Fabian; Stuke, Maik

2017-06-15

The Application of the method of Burn-up Credit on Boiling Water Reactor fuel is much more complex than in the case of Pressurized Water Reactors due to the increased heterogeneity and complexity of the fuel assemblies. Strongly varying enrichments, complex fuel assembly geometries, partial length fuel rods, and strong axial variations of the moderator density make the verification of conservative irradiation conditions difficult. In this Report, it was investigated whether it is possible to take into account the burn-up in criticality analyses for systems with irradiated Boiling Water Reactor fuel on the basis of freely available experimental data and by additionally applying stochastic methods. In order to achieve this goal, existing methods for stochastic analysis were adapted and further developed in order to being applicable to the specific conditions needed in Boiling Water Reactor analysis. The aim was to gain first insight whether a workable scheme for using burn-up credit in Boiling Water Reactor applications can be derived. Due to the fact that the different relevant quantities, like e.g. moderator density and the axial power profile, are strongly correlated, the GRS-tool SUnCISTT for Monte-Carlo uncertainty quantification was used in the analysis. This tool was coupled to a simplified, consistent model for the irradiation conditions. In contrast to conventional methods, this approach allows to simultaneously analyze all involved effects.

A versatile embedded boundary adaptive mesh method for compressible flow in complex geometry

KAUST Repository

Almarouf, Mohamad Abdulilah Alhusain Alali

2017-02-25

We present an embedded ghost-fluid method for numerical solutions of the compressible Navier Stokes (CNS) equations in arbitrary complex domains. A PDE multidimensional extrapolation approach is used to reconstruct the solution in the ghost-fluid regions and imposing boundary conditions on the fluid-solid interface, coupled with a multi-dimensional algebraic interpolation for freshly cleared cells. The CNS equations are numerically solved by the second order multidimensional upwind method. Block-structured adaptive mesh refinement, implemented with the Chombo framework, is utilized to reduce the computational cost while keeping high resolution mesh around the embedded boundary and regions of high gradient solutions. The versatility of the method is demonstrated via several numerical examples, in both static and moving geometry, ranging from low Mach number nearly incompressible flows to supersonic flows. Our simulation results are extensively verified against other numerical results and validated against available experimental results where applicable. The significance and advantages of our implementation, which revolve around balancing between the solution accuracy and implementation difficulties, are briefly discussed as well.

A versatile embedded boundary adaptive mesh method for compressible flow in complex geometry

KAUST Repository

Almarouf, Mohamad Abdulilah Alhusain Alali; Samtaney, Ravi

2017-01-01

We present an embedded ghost-fluid method for numerical solutions of the compressible Navier Stokes (CNS) equations in arbitrary complex domains. A PDE multidimensional extrapolation approach is used to reconstruct the solution in the ghost-fluid regions and imposing boundary conditions on the fluid-solid interface, coupled with a multi-dimensional algebraic interpolation for freshly cleared cells. The CNS equations are numerically solved by the second order multidimensional upwind method. Block-structured adaptive mesh refinement, implemented with the Chombo framework, is utilized to reduce the computational cost while keeping high resolution mesh around the embedded boundary and regions of high gradient solutions. The versatility of the method is demonstrated via several numerical examples, in both static and moving geometry, ranging from low Mach number nearly incompressible flows to supersonic flows. Our simulation results are extensively verified against other numerical results and validated against available experimental results where applicable. The significance and advantages of our implementation, which revolve around balancing between the solution accuracy and implementation difficulties, are briefly discussed as well.

The relation between geometry and function of the ankle joint complex: a biomechanical review.

Science.gov (United States)

Kleipool, Roeland P; Blankevoort, Leendert

2010-05-01

This review deals with the relation between the anatomy and function of the ankle joint complex. The questions addressed are how high do the forces in the ankle joint get, where can the joints go (range of motion) and where do they go during walking and running. Finally the role of the ligaments and the articular surfaces is discussed, i.e. how does it happen. The magnitude of the loads on the ankle joint complex are primarily determined by muscle activity and can be as high as four times the body weight during walking. For the maximal range of motion, plantar and dorsiflexion occurs in the talocrural joint and marginally at the subtalar joint. In-eversion takes place at both levels. The functional range of motion is well within the limits of the maximal range of motion. The ligaments do not contribute to the forces for the functional range of motion but determine the maximal range of motion together with the articular surfaces. The geometry of the articular surfaces primarily determines the kinematics. Clinical studies must include these anatomical aspects to better understand the mechanism of injury, recovery, and interventions. Models can elucidate the mechanism by which the anatomy relates to the function. The relation between the anatomy and mechanical properties of the joint structures and joint function should be considered for diagnosis and treatment of ankle joint pathology.

Experimental and numerical investigation of bubble column reactors

NARCIS (Netherlands)

Bai, W.

2010-01-01

Due to various advantages, such as simple geometry, ease of operation, low operating and maintenance costs, excellent heat and mass transfer characteristics, bubble column reactors are frequently used in chemical, petrochemical, biochemical, pharmaceutical, metallurgical industries for a variety of

Seismic responses of N-Reactor core. Independent review of Phase II work

International Nuclear Information System (INIS)

Chen, J.C.; Lo, T.; Chinn, D.J.; Murray, R.C.; Johnson, J.J.; Maslenikov, O.R.

1985-08-01

Seismic response of the N-Reactor core was independently analyzed to validate the results of Impell's analysis. The analysis procedure consists of two major stages: linear soil-structure interaction (SSI) analysis of the overall N-Reactor structure complex and nonlinear dynamic analysis of the reactor core. In the SSI analysis, CLASSI computer codes were used to calculate the SSI response of the structures and to generate the input motions for the nonlinear reactor core analysis. In addition, the response was compared to the response from the SASSI analysis under review. The impact of foundation modeling techniques and the effect of soil stiffness variation on SSI response were also investigated. In the core analysis, a nonlinear dynamic analysis model was developed. The stiffness representation of the model was calculated through a finite element analysis of several local core geometries. Finite element analyses were also used to study the block to block interaction characteristics. Using this nonlinear dynamic model along with the basemat time histories generated from CLASSI and SASSI, several dynamic analyses of the core were performed. A series of sensitivity studies was performed to investigate the discretization of the core, the effect of vertical acceleration, the effect of basemat rocking, and modeling assumptions. In general, our independent analysis of core response validates the order of magnitude of the displacement calculated by Impell. 11 refs., 110 figs., 12 tabs

Engineering graphics theoretical foundations of engineering geometry for design

CERN Document Server

Brailov, Aleksandr Yurievich

2016-01-01

This professional treatise on engineering graphics emphasizes engineering geometry as the theoretical foundation for communication of design ideas with real world structures and products. It considers each theoretical notion of engineering geometry as a complex solution of direct- and inverse-problems of descriptive geometry and each solution of basic engineering problems presented is accompanied by construction of biunique two- and three-dimension models of geometrical images. The book explains the universal structure of formal algorithms of the solutions of positional, metric, and axonometric problems, as well as the solutions of problems of construction in developing a curvilinear surface. The book further characterizes and explains the added laws of projective connections to facilitate construction of geometrical images in any of eight octants. Laws of projective connections allow constructing the complex drawing of a geometrical image in the American system of measurement and the European system of measu...

Advancing liquid metal reactor technology with nitride fuels

International Nuclear Information System (INIS)

Lyon, W.F.; Baker, R.B.; Leggett, R.D.; Matthews, R.B.

1991-08-01

A review of the use of nitride fuels in liquid metal fast reactors is presented. Past studies indicate that both uranium nitride and uranium/plutonium nitride possess characteristics that may offer enhanced performance, particularly in the area of passive safety. To further quantify these effects, the analysis of a mixed-nitride fuel system utilizing the geometry and power level of the US Advanced Liquid Metal Reactor as a reference is described. 18 refs., 2 figs., 2 tabs

Optimization of a fuel bundle within a CANDU supercritical water reactor

International Nuclear Information System (INIS)

Schofield, M.E.

2009-01-01

The supercritical water reactor is one of six nuclear reactor concepts being studied under the Generation IV International Forum. Generation IV nuclear reactors will improve the metrics of economics, sustainability, safety and reliability, and physical protection and proliferation resistance over current nuclear reactor designs. The supercritical water reactor has specific benefits in the areas of economics, safety and reliability, and physical protection. This work optimizes the fuel composition and bundle geometry to maximize the fuel burnup, and minimize the surface heat flux and the form factor. In optimizing these factors, improvements can be achieved in the areas of economics, safety and reliability of the supercritical water reactor. The WIMS-AECL software was used to model a fuel bundle within a CANDU supercritical water reactor. The Gauss' steepest descent method was used to optimize the above mentioned factors. Initially the fresh fuel composition was optimized within a 43-rod CANFLEX bundle and a 61-rod bundle. In both the 43-rod and 61-rod bundle scenarios an online refuelling scheme and non-refuelling scheme were studied. The geometry of the fuel bundles was then optimized. Finally, a homogeneous mixture of thorium and uranium fuel was studied in a 60-rod bundle. Each optimization process showed definitive improvements in the factors being studied, with the most significant improvement being an increase in the fuel burnup. The 43-rod CANFLEX bundle was the most successful at being optimized. There was little difference in the final fresh fuel content when comparing an online refuelling scheme and non-refuelling scheme. Through each optimization scenario the ratio of the fresh fuel content between the annuli was a significant determining cause in the improvements in the factors being optimized. The geometry optimization showed that improvement in the design of a fuel bundle is indeed possible, although it would be more advantageous to pursue it

Fixed-bed Reactor Dynamics and Control - A Review

DEFF Research Database (Denmark)

Jørgensen, S. B.

1986-01-01

The industrial diversity of fixed bed reactors offers a challenging and relevant set of control problems. These intricate problems arise due to the rather complex dynamics of fixed bed reactors and to the complexity of actual reactor configurations. Many of these control problems are nonlinear...... and multi-variable. During the last decade fixed bed reactor control strategies have been proposed and investigated experimentally. This paper reviews research on these complex control problems with an emphasis upon solutions which have been demon-strated to work in the laboratory and hold promise...

Current status of the reactor physics code WIMS and recent developments

International Nuclear Information System (INIS)

Lindley, B.A.; Hosking, J.G.; Smith, P.J.; Powney, D.J.; Tollit, B.S.; Newton, T.D.; Perry, R.; Ware, T.C.; Smith, P.N.

2017-01-01

Highlights: • The current status of the WIMS reactor physics code is presented. • Applications range from 2D lattice calculations up to 3D whole core geometries. • Gamma transport and thermal-hydraulic feedback models added. • Calculations methodologies described for several Gen II, III and IV reactor types. - Abstract: The WIMS modular reactor physics code has been under continuous development for over fifty years. This paper discusses the current status of WIMS and recent developments, in particular developments to the resonance shielding methodology and 3D transport solvers. Traditionally, WIMS is used to perform 2D lattice calculations, typically to generate homogenized reactor physics parameters for a whole core code such as PANTHER. However, with increasing computational resources there has been a growing trend for performing transport calculations on larger problems, up to and including 3D full core models. To this end, a number of the WIMS modules have been parallelised to allow efficient performance for whole core calculations, and WIMS includes a 3D method of characteristics solver with reflective and once-through tracking methods, which can be used to analyse problems of varying size and complexity. A time-dependent flux solver has been incorporated and thermal-hydraulic modelling capability is also being added to allow steady-state and transient coupled calculations to be performed. WIMS has been validated against a range of experimental data and other codes, in particular for water and graphite moderated thermal reactors. Future developments will include improved parallelization, enhancing the thermal-hydraulic feedback models and validating the WIMS/PANTHER code system for BWRs and fast reactors.

Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

Energy Technology Data Exchange (ETDEWEB)

Afifah, Maryam, E-mail: maryam.afifah210692@gmail.com; Su’ud, Zaki [Nuclear Research Group, FMIPA, Bandung Institute of Technology Jl. Ganesha 10, Bandung 40132 (Indonesia); Miura, Ryosuke; Takaki, Naoyuki [Department of Nuclear Safety Engineering, Tokyo City University 1-28-1 Tamazutsumi, Setagaya, Tokyo 158-8557 (Japan); Sekimoto, H. [Emerritus Prof. of Research Laboratory for Nuclear Reactors, Tokyo Inst. of Technology (Japan)

2015-09-30

Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don’t need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.

Space nuclear reactor shields for manned and unmanned applications

International Nuclear Information System (INIS)

McKissock, B.I.; Bloomfield, H.S.

1990-01-01

Missions which use nuclear reactor power systems require radiation shielding of payload and/or crew areas to predetermined dose rates. Since shielding can become a significant fraction of the total mass of the system, it is of interest to show the effect of various parameters on shield thickness and mass for manned and unmanned applications. Algorithms were developed to give the thicknesses needed if reactor thermal power, separation distances and dose rates are given as input. The thickness algorithms were combined with models for four different shield geometries to allow tradeoff studies of shield volume and mass for a variety of manned and unmanned missions. The shield design tradeoffs presented in this study include the effects of: higher allowable dose rates; radiation hardened electronics; shorter crew exposure times; shield geometry; distance of the payload and/or crew from the reactor; and changes in the size of the shielded area. Specific NASA missions that were considered in this study include unmanned outer planetary exploration, manned advanced/evolutionary space station and advanced manned lunar base. (author)

Space nuclear reactor shields for manned and unmanned applications

International Nuclear Information System (INIS)

Mckissock, B.I.; Bloomfield, H.S.

1989-01-01

Missions which use nuclear reactor power systems require radiation shielding of payload and/or crew areas to predetermined dose rates. Since shielding can become a significant fraction of the total mass of the system, it is of interest to show the effect of various parameters on shield thickness and mass for manned and unmanned applications. Algorithms were developed to give the thicknesses needed if reactor thermal power, separation distances, and dose rates are given as input. The thickness algorithms were combined with models for four different shield geometries to allow tradeoff studies of shield volume and mass for a variety of manned and unmanned missions. Shield design tradeoffs presented in this study include the effects of: higher allowable dose rates; radiation hardened electronics; shorter crew exposure times; shield geometry; distance of the payload and/or crew from the reactor; and changes in the size of the shielded area. Specific NASA missions that were considered in this study include unmanned outer planetary exploration, manned advanced/evolutionary space station, and advanced manned lunar base

Comparison of PANTHER nodal solutions in hexagonal-z geometry

International Nuclear Information System (INIS)

Knight, M.; Hutt, P.; Lewis, I.

1995-01-01

The reactor physics code PANTHER has been extended to hexagonal geometries. Steady-state, depletion, and transient calculations with feedback can all be performed. Two hexagonal nodal flux solutions have been developed. In the first method, transverse integration is performed exactly as in the rectangular case. The resulting transverse integrated equation has singular terms, which are simply ignored. The second approach applies a conformal mapping that transforms the hexagon onto a rectangle. Pin power reconstruction has also been developed with both methods. For a benchmark VVER-1000 reactor depletion problem, both methods give accurate results for standard depletion calculations. In the more extreme situation with all rods inserted, the simpler method breaks down. However, the accuracy of the conformal solution was found to be excellent in all cases studied

Physical properties corresponding to vortical flow geometry

Energy Technology Data Exchange (ETDEWEB)

Nakayama, K, E-mail: nakayama@aitech.ac.jp [Department of Mechanical Engineering, Aichi Institute of Technology, Toyota, Aichi 470-0392 (Japan)

2014-10-01

We examine a vortical flow geometry specified by the velocity gradient tensor ∇v, and derive properties representing the symmetry (axisymmetry or skewness) of the vortical flow in the swirl plane and a property specifying inflowing (outflowing) motion in all directions around the point. We focus on the radial and azimuthal velocities in a plane nonparallel to the eigenvector corresponding to the real eigenvalue of ∇v and show that these components are expressed as specific quadratic forms. The real and imaginary parts of the complex eigenvalues of ∇v represent averages of these eigenvalues of the quadratic forms, and are inadequate to specify the detailed flow geometry uniquely. The new properties complement specifying the precise flow geometry of the vortical flow.

Use of sup(233)U for high flux reactors

International Nuclear Information System (INIS)

Sekimoto, Hiroshi; Liem, P.H.

1991-01-01

The feasibility design study on the graphite moderated gas cooled reactor as a high flux reactor has been performed. The core of the reactor is equipped with two graphite reflectors, i.e., the inner reflector and the outer reflector. The highest value of the thermal neutron flux and moderately high thermal neutron flux are expected to be achieved in the inner reflector region and in the outer reflector region respectively. This reactor has many merits comparing to the conventional high flux reactors. It has the inherent safety features associated with the modular high temperature reactors. Since the core is composed with pebble bed, the on-power refueling can be performed and the experiment time can be chosen as long as necessary. Since the thermal-to-fast flux ratio is large, the background neutron level is low and material damage induced by fast neutrons are small. The calculation was performed using a four groups diffusion approximation in a one-dimensional spherical geometry and a two-dimensional cylindrical geometry. By choosing the optimal values of the core-reflector geometrical parameters and moderator-to-fuel atomic density, high thermal neutron flux can be obtained. Because of the thermal neutron flux can be obtained. Because of the thermal design constraint, however, this design will produce a relatively large core volume (about 10 7 cc) and consequently a higher reactor power (100 MWth). Preliminary calculational results show that with an average power density of only 10 W/cc, maximum thermal neutron flux of 10 15 cm -2 s -1 can be achieved in the inner reflector. The eta value of 233 U is larger than 235 U. By introducing 233 U as the fissile material for this reactor, the thermal neutron flux level can be increased by about 15%. (author). 3 refs., 2 figs., 4 tabs

Planning for Evolution in a Production Environment: Migration from a Legacy Geometry Code to an Abstract Geometry Modeling Language in STAR

Science.gov (United States)

Webb, Jason C.; Lauret, Jerome; Perevoztchikov, Victor

2012-12-01

Increasingly detailed descriptions of complex detector geometries are required for the simulation and analysis of today's high-energy and nuclear physics experiments. As new tools for the representation of geometry models become available during the course of an experiment, a fundamental challenge arises: how best to migrate from legacy geometry codes developed over many runs to the new technologies, such as the ROOT/TGeo [1] framework, without losing touch with years of development, tuning and validation. One approach, which has been discussed within the community for a number of years, is to represent the geometry model in a higher-level language independent of the concrete implementation of the geometry. The STAR experiment has used this approach to successfully migrate its legacy GEANT 3-era geometry to an Abstract geometry Modelling Language (AgML), which allows us to create both native GEANT 3 and ROOT/TGeo implementations. The language is supported by parsers and a C++ class library which enables the automated conversion of the original source code to AgML, supports export back to the original AgSTAR[5] representation, and creates the concrete ROOT/TGeo geometry implementation used by our track reconstruction software. In this paper we present our approach, design and experience and will demonstrate physical consistency between the original AgSTAR and new AgML geometry representations.

Nitritation performance and biofilm development of co- and counter-diffusion biofilm reactors: Modeling and experimental comparison

DEFF Research Database (Denmark)

Wang, Rongchang; Terada, Akihiko; Lackner, Susanne

2009-01-01

A comparative study was conducted on the start-up performance and biofilm development in two different biofilm reactors with aim of obtaining partial nitritation. The reactors were both operated under oxygen limited conditions, but differed in geometry. While substrates (O-2, NH3) co......-diffused in one geometry, they counter-diffused in the other. Mathematical simulations of these two geometries were implemented in two 1-D multispecies biofilm models using the AQUASIM software. Sensitivity analysis results showed that the oxygen mass transfer coefficient (K-i) and maximum specific growth rate...... results showed that the counter-diffusion biofilms developed faster and attained a larger maximum biofilm thickness than the co-diffusion biofilms. Under oxygen limited condition (DO

Numerical investigation on complex target geometries in the context of laser-accelerated proton beams

Energy Technology Data Exchange (ETDEWEB)

Deppert, O.; Harres, K.; Busold, S.; Schaumann, G.; Roth, M. [IKP, Technische Universitaet Darmstadt (Germany); Brabetz, C. [IAP, Goethe Universitaet Frankfurt (Germany); Schollmeier, M.; Geissel, M. [Sandia National Laboratories, NM (United States); Bagnoud, V. [GSI - Helmholtzzentrum fuer Schwerionenforschung, Darmstadt (Germany); Neely, D. [Rutherford Appleton Laboratory (United Kingdom); McKenna, P. [University of Strathclyde (United Kingdom)

2012-07-01

The irradiation of thin metal foils by an ultra-intense laser pulse leads to the generation of a highly laminar, intense proton beam accelerated from the target rear side by a mechanism called TNSA. This acceleration mechanism strongly depends on the geometry of the target. The acceleration originates from the formation of a Gaussian-like electron sheath leading to an electric field in the order of TV/m. This sheath field-ionizes the target rear side and is able to accelerate protons from a hydrogen contamination layer. The Gaussian-like sheath adds an energy dependent divergence to the spatial proton beam profile. For future applications it is essential to reduce the divergence already from the source of the acceleration process. Therefore different target geometries were studied numerically with the help of Particle-In-Cell (PIC) simulations. Both, the influence of the target geometry as well as the influence of the laser beam profile onto the proton trajectories are discussed. Furthermore, the first experimental results of a dedicated target geometry for laser-ion acceleration are presented.

Comparative assessment of different approaches for the use of CAD geometry in Monte Carlo transport calculations

International Nuclear Information System (INIS)

Weinhorst, Bastian; Fischer, Ulrich; Lu, Lei; Qiu, Yuefeng; Wilson, Paul

2015-01-01

Highlights: • Comparison of different approaches for the use of CAD geometry for Monte Carlo transport calculations. • Comparison with regard to user-friendliness and computation performance. • Three approaches, namely conversion with McCad, unstructured mesh feature of MCN6 and DAGMC. • Installation most complex for DAGMC, model preparation worst for McCad, computation performance worst for MCNP6. • Installation easiest for McCad, model preparation best for MCNP6, computation speed fastest for McCad. - Abstract: Computer aided design (CAD) is an important industrial way to produce high quality designs. Therefore, CAD geometries are in general used for engineering and the design of complex facilities like the ITER tokamak. Although Monte Carlo codes like MCNP are well suited to handle the complex 3D geometry of ITER for transport calculations, they rely on their own geometry description and are in general not able to directly use the CAD geometry. In this paper, three different approaches for the use of CAD geometries with MCNP calculations are investigated and assessed with regard to calculation performance and user-friendliness. The first method is the conversion of the CAD geometry into MCNP geometry employing the conversion software McCad developed by KIT. The second approach utilizes the MCNP6 mesh geometry feature for the particle tracking and relies on the conversion of the CAD geometry into a mesh model. The third method employs DAGMC, developed by the University of Wisconsin-Madison, for the direct particle tracking on the CAD geometry using a patched version of MCNP. The obtained results show that each method has its advantages depending on the complexity and size of the model, the calculation problem considered, and the expertise of the user.

Comparative assessment of different approaches for the use of CAD geometry in Monte Carlo transport calculations

Energy Technology Data Exchange (ETDEWEB)

Weinhorst, Bastian, E-mail: bastian.weinhorst@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology, Eggenstein-Leopoldshafen (Germany); Fischer, Ulrich; Lu, Lei; Qiu, Yuefeng [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology, Eggenstein-Leopoldshafen (Germany); Wilson, Paul [University of Wisconsin-Madison, Computational Nuclear Engineering Research Group, Madison, WI (United States)

2015-10-15

Highlights: • Comparison of different approaches for the use of CAD geometry for Monte Carlo transport calculations. • Comparison with regard to user-friendliness and computation performance. • Three approaches, namely conversion with McCad, unstructured mesh feature of MCN6 and DAGMC. • Installation most complex for DAGMC, model preparation worst for McCad, computation performance worst for MCNP6. • Installation easiest for McCad, model preparation best for MCNP6, computation speed fastest for McCad. - Abstract: Computer aided design (CAD) is an important industrial way to produce high quality designs. Therefore, CAD geometries are in general used for engineering and the design of complex facilities like the ITER tokamak. Although Monte Carlo codes like MCNP are well suited to handle the complex 3D geometry of ITER for transport calculations, they rely on their own geometry description and are in general not able to directly use the CAD geometry. In this paper, three different approaches for the use of CAD geometries with MCNP calculations are investigated and assessed with regard to calculation performance and user-friendliness. The first method is the conversion of the CAD geometry into MCNP geometry employing the conversion software McCad developed by KIT. The second approach utilizes the MCNP6 mesh geometry feature for the particle tracking and relies on the conversion of the CAD geometry into a mesh model. The third method employs DAGMC, developed by the University of Wisconsin-Madison, for the direct particle tracking on the CAD geometry using a patched version of MCNP. The obtained results show that each method has its advantages depending on the complexity and size of the model, the calculation problem considered, and the expertise of the user.

Geometry and Hamiltonian mechanics on discrete spaces

International Nuclear Information System (INIS)

Talasila, V; Clemente-Gallardo, J; Schaft, A J van der

2004-01-01

Numerical simulation is often crucial for analysing the behaviour of many complex systems which do not admit analytic solutions. To this end, one either converts a 'smooth' model into a discrete (in space and time) model, or models systems directly at a discrete level. The goal of this paper is to provide a discrete analogue of differential geometry, and to define on these discrete models a formal discrete Hamiltonian structure-in doing so we try to bring together various fundamental concepts from numerical analysis, differential geometry, algebraic geometry, simplicial homology and classical Hamiltonian mechanics. For example, the concept of a twisted derivation is borrowed from algebraic geometry for developing a discrete calculus. The theory is applied to a nonlinear pendulum and we compare the dynamics obtained through a discrete modelling approach with the dynamics obtained via the usual discretization procedures. Also an example of an energy-conserving algorithm on a simple harmonic oscillator is presented, and its effect on the Poisson structure is discussed

Study of an hypothetical reactor meltdown accident for a 50 MW sub(th) fast reactor

International Nuclear Information System (INIS)

Azevedo, E.M. de.

1983-01-01

A melhodology for determining the energy released in hypothetical reactor meltdown accidents is presented. A numerical code was developed based upon the Nicholson method for a uniform and homogeneous reactor with spherical geometry. A comparative study with other know programs in the literature which use better approximations for small energy released, shows that the methodology used were compatible with those under comparison. Besides the influence of some parameters on the energy released, such as the initial power level and the prompt neutron lifetime was studied under this metodology and its result exhibitted. The Doppler effect was also analyzed and its influence on the energy released has been emphasized. (Author) [pt

Reactive turbulent flow CFD study in supercritical water oxidation process: application to a stirred double shell reactor

International Nuclear Information System (INIS)

Moussiere, S.

2006-12-01

Supercritical water oxidation is an innovative process to treat organic liquid waste which uses supercritical water properties to mix efficiency the oxidant and the organic compounds. The reactor is a stirred double shell reactor. In the step of adaptation to nuclear constraints, the computational fluid dynamic modeling is a good tool to know required temperature field in the reactor for safety analysis. Firstly, the CFD modeling of tubular reactor confirms the hypothesis of an incompressible fluid and the use of k-w turbulence model to represent the hydrodynamic. Moreover, the EDC model is as efficiency as the kinetic to compute the reaction rate in this reactor. Secondly, the study of turbulent flow in the double shell reactor confirms the use of 2D axisymmetric geometry instead of 3D geometry to compute heat transfer. Moreover, this study reports that water-air mixing is not in single phase. The reactive turbulent flow is well represented by EDC model after adaptation of initial conditions. The reaction rate in supercritical water oxidation reactor is mainly controlled by the mixing. (author)

TOPIC: a debugging code for torus geometry input data of Monte Carlo transport code

International Nuclear Information System (INIS)

Iida, Hiromasa; Kawasaki, Hiromitsu.

1979-06-01

TOPIC has been developed for debugging geometry input data of the Monte Carlo transport code. the code has the following features: (1) It debugs the geometry input data of not only MORSE-GG but also MORSE-I capable of treating torus geometry. (2) Its calculation results are shown in figures drawn by Plotter or COM, and the regions not defined or doubly defined are easily detected. (3) It finds a multitude of input data errors in a single run. (4) The input data required in this code are few, so that it is readily usable in a time sharing system of FACOM 230-60/75 computer. Example TOPIC calculations in design study of tokamak fusion reactors (JXFR, INTOR-J) are presented. (author)

WIMSD5, Deterministic Multigroup Reactor Lattice Calculations

International Nuclear Information System (INIS)

2004-01-01

1 - Description of program or function: The Winfrith improved multigroup scheme (WIMS) is a general code for reactor lattice cell calculation on a wide range of reactor systems. In particular, the code will accept rod or plate fuel geometries in either regular arrays or in clusters and the energy group structure has been chosen primarily for thermal calculations. The basic library has been compiled with 14 fast groups, 13 resonance groups and 42 thermal groups, but the user is offered the choice of accurate solutions in many groups or rapid calculations in few groups. Temperature dependent thermal scattering matrices for a variety of scattering laws are included in the library for the principal moderators which include hydrogen, deuterium, graphite, beryllium and oxygen. WIMSD5 is a successor version of WIMS-D/4. 2 - Method of solution: The treatment of resonances is based on the use of equivalence theorems with a library of accurately evaluated resonance integrals for equivalent homogeneous systems at a variety of temperatures. The collision theory procedure gives accurate spectrum computations in the 69 groups of the library for the principal regions of the lattice using a simplified geometric representation of complicated lattice cells. The computed spectra are then used for the condensation of cross-sections to the number of groups selected for solution of the transport equation in detailed geometry. Solution of the transport equation is provided either by use of the Carlson DSN method or by collision probability methods. Leakage calculations including an allowance for streaming asymmetries may be made using either diffusion theory or the more elaborate B1-method. The output of the code provides Eigenvalues for the cases where a simple buckling mode is applicable or cell-averaged parameters for use in overall reactor calculations. Various reaction rate edits are provided for direct comparison with experimental measurements. 3 - Restrictions on the complexity of

Transmission probability method based on triangle meshes for solving unstructured geometry neutron transport problem

Energy Technology Data Exchange (ETDEWEB)

Wu Hongchun [Nuclear Engineering Department, Xi' an Jiaotong University, Xi' an 710049, Shaanxi (China)]. E-mail: hongchun@mail.xjtu.edu.cn; Liu Pingping [Nuclear Engineering Department, Xi' an Jiaotong University, Xi' an 710049, Shaanxi (China); Zhou Yongqiang [Nuclear Engineering Department, Xi' an Jiaotong University, Xi' an 710049, Shaanxi (China); Cao Liangzhi [Nuclear Engineering Department, Xi' an Jiaotong University, Xi' an 710049, Shaanxi (China)

2007-01-15

In the advanced reactor, the fuel assembly or core with unstructured geometry is frequently used and for calculating its fuel assembly, the transmission probability method (TPM) has been used widely. However, the rectangle or hexagon meshes are mainly used in the TPM codes for the normal core structure. The triangle meshes are most useful for expressing the complicated unstructured geometry. Even though finite element method and Monte Carlo method is very good at solving unstructured geometry problem, they are very time consuming. So we developed the TPM code based on the triangle meshes. The TPM code based on the triangle meshes was applied to the hybrid fuel geometry, and compared with the results of the MCNP code and other codes. The results of comparison were consistent with each other. The TPM with triangle meshes would thus be expected to be able to apply to the two-dimensional arbitrary fuel assembly.

Development of a 2-D Simplified P3 FEM Solver for Arbitrary Geometry Applications

Energy Technology Data Exchange (ETDEWEB)

Ryu, Eun Hyun; Joo, Han Gyu [Seoul National University, Seoul (Korea, Republic of)

2010-10-15

In the calculation of power distributions and multiplication factors in a nuclear reactor, the Finite Difference Method (FDM) and the nodal methods are primarily used. These methods are, however, limited to particular geometries and lack general application involving arbitrary geometries. The Finite Element Method (FEM) can be employed for arbitrary geometry application and there are numerous FEM codes to solve the neutron diffusion equation or the Sn transport equation. The diffusion based FEM codes have the drawback of inferior accuracy while the Sn based ones require a considerable computing time. This work here is to seek a compromise between these two by employing the simplified P3 (SP3) method for arbitrary geometry applications. Sufficient accuracy with affordable computing time and resources would be achieved with this choice of approximate transport solution when compared to full FEM based Pn or Sn solutions. For now only 2-D solver is considered

Study on dynamic behavior of fusion reactor materials and their response to variable and complex irradiation environment

International Nuclear Information System (INIS)

Abe, K.; Kohyama, A.; Namba, C.; Wiffen, F.W.; Jones, R.H.

2001-01-01

A Japan-USA Program of irradiation experiments for fusion research, 'JUPITER', has been established as a 6 year program from 1995 to 2000. The goal is to study the dynamic behavior of fusion reactor materials and their response to variable and complex irradiation environment using fission reactors. The irradiation experiments in this program include low activation structural materials, functional ceramics and other innovative materials. The experimental data are analyzed by theoretical modeling and computer simulation to integrate the above effects. The irradiation capsules for in-situ measurement and varying temperature were developed successfully. It was found that insulating ceramics were worked up to 3 dpa. The property changes and related issues in low activation structural materials were summarized. (author)

A voxelization approach to navigate through nested geometries

CERN Document Server

Harrison, Brent Andrew

2016-01-01

High energy physics experiment software typically implements a detailed description of the geometry of the relevant detector. As modern detectors increase in complexity, modelling them becomes more challenging. Typically such models are built as a nested hierarchy of O(10000) volumes reaching a depth of 10 - 20. It is desirable to develop data structures and algorithms which allow fast and efficient navigation though a given detector geometry model. We investigate the feasibility of voxelisation techniques to this end.

Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

Science.gov (United States)

Ilham, Muhammad; Su'ud, Zaki

2017-01-01

Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

Preliminary analysis of basic reactor physics of the Dual Fluid Reactor - 15270

International Nuclear Information System (INIS)

Wang, X.; Macian-Juan, R.; Seidl, M.

2015-01-01

The Dual Fluid Reactor (DFR) is a novel fast nuclear reactor concept invented by the IFK based on the Generation IV Molten Salt Reactor and the Liquid Metal Cooled Reactor. The DFR uses a chloride based molten fuel salt in order to harden the neutron spectrum. The molten fuel salt is cooled with a separated liquid lead loop, which in principle allows for higher power densities and better breeding performance. The DFR does not combine heat removal and breeding into a single circuit but separates the two functions into two independent circuits. Since there are attractive features mentioned in this design, the main task of this paper is to verify the model of the whole reactor based on this concept. For this purpose several calculations are presented, including steady state calculations, sensitivity calculations with regard to the nuclide cross sections, the temperature and geometry coefficient of k eff as well as the burnup calculation. The Monte Carlo calculation codes MCNP, SERPENT and SCALE are used for the analysis. As expected the study shows a significant negative reactivity feedback with temperature in the overall fission zone. For the coupled coolant and reflector design the temperature feedback is rather small for practical purposes such as reactor control during normal operation. In the view of these results the DFR in principle can be self-regulated totally by the temperature change of its own fuel salt and consequently can rely on fully passive safety systems for accident management

Architectural Geometry and Fabrication-Aware Design

KAUST Repository

Pottmann, Helmut

2013-04-27

Freeform shapes and structures with a high geometric complexity play an increasingly important role in contemporary architecture. While digital models are easily created, the actual fabrication and construction remains a challenge. This is the source of numerous research problems many of which fall into the area of Geometric Computing and form part of a recently emerging research area, called "Architectural Geometry". The present paper provides a short survey of research in Architectural Geometry and shows how this field moves towards a new direction in Geometric Modeling which aims at combining shape design with important aspects of function and fabrication. © 2013 Kim Williams Books, Turin.

Comparison of BEACON and COMPARE reactor cavity subcompartment analyses

International Nuclear Information System (INIS)

Burkett, M.W.; Idar, E.S.; Gido, R.G.; Lime, J.F.; Koestel, A.

1984-04-01

In this study, a more advanced best-estimate containment code, BEACON-MOD3A, was ued to calculate force and moment loads resulting from a high-energy blowdown for two reactor cavity geometries previously analyzed with the licensing computer code COMPARE-MOD1A. The BEACON force and moment loads were compared with the COMPARE results to determine the safety margins provided by the COMPARE code. The forces and moments calculated by the codes were found to be different, although not in any consistent manner, for the two reactor cavity geometries studied. Therefore, generic summary statements regarding margins cannot be made because of the effects of the detailed physical configuration. However, differences in the BEACON and COMPARE calculated forces and moments can be attributed to differences in the modeling assumptions used in the codes and the analyses

Progress on RMC: a Monte Carlo neutron transport code for reactor analysis

International Nuclear Information System (INIS)

Wang, Kan; Li, Zeguang; She, Ding; Liu, Yuxuan; Xu, Qi; Shen, Huayun; Yu, Ganglin

2011-01-01

This paper presents a new 3-D Monte Carlo neutron transport code named RMC (Reactor Monte Carlo code), specifically intended for reactor physics analysis. This code is being developed by Department of Engineering Physics in Tsinghua University and written in C++ and Fortran 90 language with the latest version of RMC 2.5.0. The RMC code uses the method known as the delta-tracking method to simulate neutron transport, the advantages of which include fast simulation in complex geometries and relatively simple handling of complicated geometrical objects. Some other techniques such as computational-expense oriented method and hash-table method have been developed and implemented in RMC to speedup the calculation. To meet the requirements of reactor analysis, the RMC code has the calculational functions including criticality calculation, burnup calculation and also kinetics simulation. In this paper, comparison calculations of criticality problems, burnup problems and transient problems are carried out using RMC code and other Monte Carlo codes, and the results show that RMC performs quite well in these kinds of problems. Based on MPI, RMC succeeds in parallel computation and represents a high speed-up. This code is still under intensive development and the further work directions are mentioned at the end of this paper. (author)

Linking computer-aided design (CAD) to Geant4-based Monte Carlo simulations for precise implementation of complex treatment head geometries

International Nuclear Information System (INIS)

Constantin, Magdalena; Constantin, Dragos E; Keall, Paul J; Narula, Anisha; Svatos, Michelle; Perl, Joseph

2010-01-01

Most of the treatment head components of medical linear accelerators used in radiation therapy have complex geometrical shapes. They are typically designed using computer-aided design (CAD) applications. In Monte Carlo simulations of radiotherapy beam transport through the treatment head components, the relevant beam-generating and beam-modifying devices are inserted in the simulation toolkit using geometrical approximations of these components. Depending on their complexity, such approximations may introduce errors that can be propagated throughout the simulation. This drawback can be minimized by exporting a more precise geometry of the linac components from CAD and importing it into the Monte Carlo simulation environment. We present a technique that links three-dimensional CAD drawings of the treatment head components to Geant4 Monte Carlo simulations of dose deposition. (note)

Constrained Geometry Organotitanium Catalysts Supported on Nanosized Silica for Ethylene (co)Polymerization.

Science.gov (United States)

Li, Kuo-Tseng; Wu, Ling-Huey

2017-05-05

Supported olefin polymerization catalysts can prevent reactor-fouling problems and produce uniform polymer particles. Constrained geometry complexes (CGCs) have less sterically hindered active sites than bis-cyclopentadienyl metallocene catalysts. In the literature, micrometer-sized silica particles were used for supporting CGC catalysts, which might have strong mass transfer limitations. This study aims to improve the activity of supported CGC catalysts by using nanometer-sized silica. Ti[(C₅Me₄)SiMe₂(N t Bu)]Cl₂, a "constrained-geometry" titanium catalyst, was supported on MAO-treated silicas (nano-sized and micro-sized) by an impregnation method. Ethylene homo-polymerization and co-polymerization with 1-octene were carried out in a temperature range of 80-120 °C using toluene as the solvent. Catalysts prepared and polymers produced were characterized. For both catalysts and for both reactions, the maximum activities occurred at 100 °C, which is significantly higher than that (60 °C) reported before for supported bis-cyclopentadienyl metallocene catalysts containing zirconium, and is lower than that (≥140 °C) used for unsupported Ti[(C₅Me₄)SiMe₂(N t Bu)]Me₂ catalyst. Activities of nano-sized catalyst were 2.6 and 1.6 times those of micro-sized catalyst for homopolymerization and copolymerization, respectively. The former produced polymers with higher crystallinity and melting point than the latter. In addition, copolymer produced with nanosized catalyst contained more 1-octene than that produced with microsized catalyst.

Immobilization of metal-humic acid complexes in anaerobic granular sludge for their application as solid-phase redox mediators in the biotransformation of iopromide in UASB reactors.

Science.gov (United States)

Cruz-Zavala, Aracely S; Pat-Espadas, Aurora M; Rangel-Mendez, J Rene; Chazaro-Ruiz, Luis F; Ascacio-Valdes, Juan A; Aguilar, Cristobal N; Cervantes, Francisco J

2016-05-01

Metal-humic acid complexes were synthesized and immobilized by a granulation process in anaerobic sludge for their application as solid-phase redox mediators (RM) in the biotransformation of iopromide. Characterization of Ca- and Fe-humic acid complexes revealed electron accepting capacities of 0.472 and 0.556milli-equivalentsg(-1), respectively. Once immobilized, metal-humic acid complexes significantly increased the biotransformation of iopromide in upflow anaerobic sludge blanket (UASB) reactors. Control UASB reactor (without humic material) achieved 31.6% of iopromide removal, while 80% was removed in UASB reactors supplied with each metal-humic acid complex. Further analyses indicated multiple transformation reactions taking place in iopromide including deiodination, N-dealkylation, decarboxylation and deacetylation. This is the first successful application of immobilized RM, which does not require a supporting material to maintain the solid-phase RM in long term operation of bioreactors. The proposed redox catalyst could be suitable for enhancing the redox conversion of different recalcitrant pollutants present in industrial effluents. Copyright © 2016 Elsevier Ltd. All rights reserved.

Elementary algebraic geometry

CERN Document Server

Kendig, Keith

2015-01-01

Designed to make learning introductory algebraic geometry as easy as possible, this text is intended for advanced undergraduates and graduate students who have taken a one-year course in algebra and are familiar with complex analysis. This newly updated second edition enhances the original treatment's extensive use of concrete examples and exercises with numerous figures that have been specially redrawn in Adobe Illustrator. An introductory chapter that focuses on examples of curves is followed by a more rigorous and careful look at plane curves. Subsequent chapters explore commutative ring th

Subgroup complexes

CERN Document Server

Smith, Stephen D

2011-01-01

This book is intended as an overview of a research area that combines geometries for groups (such as Tits buildings and generalizations), topological aspects of simplicial complexes from p-subgroups of a group (in the spirit of Brown, Quillen, and Webb), and combinatorics of partially ordered sets. The material is intended to serve as an advanced graduate-level text and partly as a general reference on the research area. The treatment offers optional tracks for the reader interested in buildings, geometries for sporadic simple groups, and G-equivariant equivalences and homology for subgroup complexes.

Electron beam solenoid reactor concept

International Nuclear Information System (INIS)

Bailey, V.; Benford, J.; Cooper, R.; Dakin, D.; Ecker, B.; Lopez, O.; Putman, S.; Young, T.S.T.

1977-01-01

The electron Beam Heated Solenoid (EBHS) reactor is a linear magnetically confined fusion device in which the bulk or all of the heating is provided by a relativistic electron beam (REB). The high efficiency and established technology of the REB generator and the ability to vary the coupling length make this heating technique compatible with several radial and axial enery loss reduction options including multiple-mirrors, electrostatic and gas end-plug techniques. This paper addresses several of the fundamental technical issues and provides a current evaluation of the concept. The enhanced confinement of the high energy plasma ions due to nonadiabatic scattering in the multiple mirror geometry indicates the possibility of reactors of the 150 to 300 meter length operating at temperatures > 10 keV. A 275 meter EBHS reactor with a plasma Q of 11.3 requiring 33 MJ of beam eneergy is presented

Design Process Control for Improved Surface Finish of Metal Additive Manufactured Parts of Complex Build Geometry

Directory of Open Access Journals (Sweden)

Mikdam Jamal

2017-12-01

Full Text Available Metal additive manufacturing (AM is increasingly used to create complex 3D components at near net shape. However, the surface finish (SF of the metal AM part is uneven, with surface roughness being variable over the facets of the design. Standard post-processing methods such as grinding and linishing often meet with major challenges in finishing parts of complex shape. This paper reports on research that demonstrated that mass finishing (MF processes are able to deliver high-quality surface finishes (Ra and Sa on AM-generated parts of a relatively complex geometry (both internal features and external facets under select conditions. Four processes were studied in this work: stream finishing, high-energy (HE centrifuge, drag finishing and disc finishing. Optimisation of the drag finishing process was then studied using a structured design of experiments (DOE. The effects of a range of finishing parameters were evaluated and optimal parameters and conditions were determined. The study established that the proposed method can be successfully applied in drag finishing to optimise the surface roughness in an industrial application and that it is an economical way of obtaining the maximum amount of information in a short period of time with a small number of tests. The study has also provided an important step in helping understand the requirements of MF to deliver AM-generated parts to a target quality finish and cycle time.

Neutron-photon energy deposition in CANDU reactor fuel channels: a comparison of modelling techniques using ANISN and MCNP computer codes

International Nuclear Information System (INIS)

Bilanovic, Z.; McCracken, D.R.

1994-12-01

In order to assess irradiation-induced corrosion effects, coolant radiolysis and the degradation of the physical properties of reactor materials and components, it is necessary to determine the neutron, photon, and electron energy deposition profiles in the fuel channels of the reactor core. At present, several different computer codes must be used to do this. The most recent, advanced and versatile of these is the latest version of MCNP, which may be capable of replacing all the others. Different codes have different assumptions and different restrictions on the way they can model the core physics and geometry. This report presents the results of ANISN and MCNP models of neutron and photon energy deposition. The results validate the use of MCNP for simplified geometrical modelling of energy deposition by neutrons and photons in the complex geometry of the CANDU reactor fuel channel. Discrete ordinates codes such as ANISN were the benchmark codes used in previous work. The results of calculations using various models are presented, and they show very good agreement for fast-neutron energy deposition. In the case of photon energy deposition, however, some modifications to the modelling procedures had to be incorporated. Problems with the use of reflective boundaries were solved by either including the eight surrounding fuel channels in the model, or using a boundary source at the bounding surface of the problem. Once these modifications were incorporated, consistent results between the computer codes were achieved. Historically, simple annular representations of the core were used, because of the difficulty of doing detailed modelling with older codes. It is demonstrated that modelling by MCNP, using more accurate and more detailed geometry, gives significantly different and improved results. (author). 9 refs., 12 tabs., 20 figs

Sandia Pulse Reactor-IV Project

International Nuclear Information System (INIS)

Reuscher, J.A.

1983-01-01

Sandia National Laboratories has developed, designed and operated fast burst reactors for over 20 years. These reactors have been used for a variety of radiation effects programs. During this period, programs have required larger irradiation volumes primarily to expose complex electronic systems to postulated threat environments. As experiment volumes increased, a new reactor was built so that these components could be tested. The Sandia Pulse Reactor-IV is a logical evolution of the two decades of fast burst reactor development at Sandia

Optical geometry

International Nuclear Information System (INIS)

Robinson, I.; Trautman, A.

1988-01-01

The geometry of classical physics is Lorentzian; but weaker geometries are often more appropriate: null geodesics and electromagnetic fields, for example, are well known to be objects of conformal geometry. To deal with a single null congruence, or with the radiative electromagnetic fields associated with it, even less is needed: flag geometry for the first, optical geometry, with which this paper is chiefly concerned, for the second. The authors establish a natural one-to-one correspondence between optical geometries, considered locally, and three-dimensional Cauchy-Riemann structures. A number of Lorentzian geometries are shown to be equivalent from the optical point of view. For example the Goedel universe, the Taub-NUT metric and Hauser's twisting null solution have an optical geometry isomorphic to the one underlying the Robinson congruence in Minkowski space. The authors present general results on the problem of lifting a CR structure to a Lorentz manifold and, in particular, to Minkowski space; and exhibit the relevance of the deviation form to this problem

Fuel enrichment reduction for heavy water moderated research reactors

International Nuclear Information System (INIS)

McCulloch, D.B.

1984-01-01

Twelve heavy-water-moderated research reactors of significant power level (5 MW to 125 MW) currently operate in a number of countries, and use highly enriched uranium (HEU) fuel. Most of these reactors could in principle be converted to use uranium of lower enrichment, subject in some cases to the successful development and demonstration of new fuel materials and/or fuel element designs. It is, however, generally accepted as desirable that existing fuel element geometry be retained unaltered to minimise the capital costs and licensing difficulties associated with enrichment conversion. The high flux Australian reactor, HIFAR, at Lucas Heights, Sydney is one of 5 Dido-class reactors in the above group. It operates at 10 MW using 80% 235 U HEU fuel. Theoretical studies of neutronic, thermohydraulic and operational aspects of converting HIFAR to use fuels of reduced enrichment have been made over a period. It is concluded that with no change of fuel element geometry and no penalty in the present HEU fuel cycle burn-up performance, conversion to MEU (nominally 45% 235 U) would be feasible within the limits of current fully qualified U-Al fuel materials technology. There would be no significant, adverse effects on safety-related parameters (e.g. reactivity coefficients) and only small penalties in reactor flux. Conversion to LEU (nominally 20% 235 U) a similar basis would require that fuel materials of about 2.3 g U cm -3 be fully qualified, and would depress the in-core thermal neutron flux by about 15 per cent relative to HEU fuelling. In qualitative terms, similar conclusions would be expected to hold for a majority of the above heavy water moderated reactors. (author)

Principle of an operational complexity index for the characterization of the human factor relevance of future reactors concepts

International Nuclear Information System (INIS)

Papin, Bernard

2004-01-01

With the increasing reliability of the modern technological systems, the human contribution to the global risk in the operation of industrial systems is becoming more and more significant : in the nuclear reactor operation for example, a recent PSA estimation of this contribution is about 25% of the risk of core melting, all situations considered. This urges the designers of future nuclear reactors to consider the minimisation of this Human Factor (HF) contribution, at the very early stage of their design : the experience feedback shows that this is indeed at this stage that the fundamental design options, impacting the most the human reliability in operation, are fixed. The problem is that at these early design stages, it is also quite impossible to apply formal human reliability methods to support this HF optimisation, while the precise operating conditions of the reactor are not yet known in enough details. In this paper, another approach of the HF evaluation during the design, based on the functional and operational complexity assessment, is proposed. As an illustration, this approach is used to compare various concepts of Pressurized Water Reactors from the point of view of the Human Factor relevance. (Author)

Discrete quantum geometries and their effective dimension

International Nuclear Information System (INIS)

Thuerigen, Johannes

2015-01-01

In several approaches towards a quantum theory of gravity, such as group field theory and loop quantum gravity, quantum states and histories of the geometric degrees of freedom turn out to be based on discrete spacetime. The most pressing issue is then how the smooth geometries of general relativity, expressed in terms of suitable geometric observables, arise from such discrete quantum geometries in some semiclassical and continuum limit. In this thesis I tackle the question of suitable observables focusing on the effective dimension of discrete quantum geometries. For this purpose I give a purely combinatorial description of the discrete structures which these geometries have support on. As a side topic, this allows to present an extension of group field theory to cover the combinatorially larger kinematical state space of loop quantum gravity. Moreover, I introduce a discrete calculus for fields on such fundamentally discrete geometries with a particular focus on the Laplacian. This permits to define the effective-dimension observables for quantum geometries. Analysing various classes of quantum geometries, I find as a general result that the spectral dimension is more sensitive to the underlying combinatorial structure than to the details of the additional geometric data thereon. Semiclassical states in loop quantum gravity approximate the classical geometries they are peaking on rather well and there are no indications for stronger quantum effects. On the other hand, in the context of a more general model of states which are superposition over a large number of complexes, based on analytic solutions, there is a flow of the spectral dimension from the topological dimension d on low energy scales to a real number between 0 and d on high energy scales. In the particular case of 1 these results allow to understand the quantum geometry as effectively fractal.

Numerical simulation of vortex pyrolysis reactors for condensable tar production from biomass

Energy Technology Data Exchange (ETDEWEB)

Miller, R.S.; Bellan, J. [California Inst. of Tech., Pasadena, CA (United States). Jet Propulsion Lab.

1998-08-01

A numerical study is performed in order to evaluate the performance and optimal operating conditions of vortex pyrolysis reactors used for condensable tar production from biomass. A detailed mathematical model of porous biomass particle pyrolysis is coupled with a compressible Reynolds stress transport model for the turbulent reactor swirling flow. An initial evaluation of particle dimensionality effects is made through comparisons of single- (1D) and multi-dimensional particle simulations and reveals that the 1D particle model results in conservative estimates for total pyrolysis conversion times and tar collection. The observed deviations are due predominantly to geometry effects while directional effects from thermal conductivity and permeability variations are relatively small. Rapid ablative particle heating rates are attributed to a mechanical fragmentation of the biomass particles that is modeled using a critical porosity for matrix breakup. Optimal thermal conditions for tar production are observed for 900 K. Effects of biomass identity, particle size distribution, and reactor geometry and scale are discussed.

On Monte Carlo estimation of radiation damage in light water reactor systems

International Nuclear Information System (INIS)

Read, Edward A.; Oliveira, Cassiano R.E. de

2010-01-01

There has been a growing need in recent years for the development of methodologies to calculate damage factors, namely displacements per atom (dpa), of structural components for Light Water Reactors (LWRs). The aim of this paper is discuss and highlight the main issues associated with the calculation of radiation damage factors utilizing the Monte Carlo method. Among these issues are: particle tracking and tallying in complex geometries, dpa calculation methodology, coupled fuel depletion and uncertainty propagation. The capabilities of the Monte Carlo code Serpent such as Woodcock tracking and burnup are assessed for radiation damage calculations and its capability demonstrated and compared to those of the MCNP code for dpa calculations of a typical LWR configuration involving the core vessel and the downcomer. (author)

Introducing geometry concept based on history of Islamic geometry

Science.gov (United States)

Maarif, S.; Wahyudin; Raditya, A.; Perbowo, K. S.

2018-01-01

Geometry is one of the areas of mathematics interesting to discuss. Geometry also has a long history in mathematical developments. Therefore, it is important integrated historical development of geometry in the classroom to increase’ knowledge of how mathematicians earlier finding and constructing a geometric concept. Introduction geometrical concept can be started by introducing the Muslim mathematician who invented these concepts so that students can understand in detail how a concept of geometry can be found. However, the history of mathematics development, especially history of Islamic geometry today is less popular in the world of education in Indonesia. There are several concepts discovered by Muslim mathematicians that should be appreciated by the students in learning geometry. Great ideas of mathematicians Muslim can be used as study materials to supplement religious character values taught by Muslim mathematicians. Additionally, by integrating the history of geometry in teaching geometry are expected to improve motivation and geometrical understanding concept.

CIMPA Summer School on Arithmetic and Geometry Around Hypergeometric Functions

CERN Document Server

Uludağ, A; Yoshida, Masaaki; Arithmetic and Geometry Around Hypergeometric Functions

2007-01-01

This volume comprises the Lecture Notes of the CIMPA Summer School "Arithmetic and Geometry around Hypergeometric Functions" held at Galatasaray University, Istanbul in 2005. It contains lecture notes, a survey article, research articles, and the results of a problem session. Key topics are moduli spaces of points on P1 and Picard-Terada-Deligne-Mostow theory, moduli spaces of K3 surfaces, complex hyperbolic geometry, ball quotients, GKZ hypergeometric structures, Hilbert and Picard modular surfaces, uniformizations of complex orbifolds, algebraicity of values of Schwartz triangle functions, and Thakur's hypergeometric function. The book provides a background, gives detailed expositions and indicates new research directions. It is directed to postgraduate students and researchers.

Study of the distributions of flow rate and enthalpy in the sub-channels of a bundle geometry of nuclear reactors in one and two-phase flow

International Nuclear Information System (INIS)

Bayoumi, M.A.A.

1976-10-01

A bibliographic study shows that the experimental studies examined, have been developed to understand the phenomenon acting on the mixing between the sub-channels of which geometries are such these of rod bundles used in some nuclear reactors. Experimental devices and tests have been developed to study the influence of the following parameters, operating conditions, pressure, flow rate, power brought to the bundle and inlet temperature on the distribution of flow rates and vapor content among the different sub-channels. By means of non isokinetic sampling, one has determined the enthalpy of the fluid participating to the mixing between the communicating sub-channels and it has been shown that the value of this enthalpy depends strongly on the type of fluid flow and that this enthalpy cannot be either the enthalpy of one of the two sub-channels, nor (always) an average of these two enthalpies. The experimental results have been compared with calculations developed with the code FLICA, concerning the mass velocity distribution, the exchange term of linear momentum, and the variation of the transversal enthalpy with regard to the type of fluid flow. A study of local void ratio measurement, by means of optical probes, has been proposed. The present study has been carried out with a smooth geometry [fr

Applicability of AWJ technique for dismantling reactor of the Fukushima Daiichi Nuclear Power Station. Cutting test of imitation of fuel debris and optimization of the cutting condition

International Nuclear Information System (INIS)

Maruyama, Shin-ichiro; Watatani, Satoshi

2016-01-01

Based on findings during recovery works that followed the accident at Three Mile Island Station 2, it is assumed that the reactor internals at the Fukushima Daiichi Nuclear Power Station (1F) have complex geometries intermixed with melted fuel and confined in limited spaces. Accordingly, abrasive water jet (AWJ) cutting method is considered to be a promising technique that can be safely and reasonably used for cutting and removing reactor internals. The authors conducted tests to examine the possibility of application and to solve the problems of this technique. In the tests imitation of fuel debris and optimization of the cutting condition is used. The test result made the measures for some of the associated issues clear, and demonstrated that AWJ cutting method is assumed as one of the promising techniques for removing reactor internals. (author)

Transient neutrons flux behaviour in a spherical reactor core

International Nuclear Information System (INIS)

Souza, A.W.A. de.

1978-11-01

This work studies the transient neutron flux in a fast reactor of spherical geometry. The burning of U 235 nuclei is equated and two kinds of reflector were studied. The numeric solutions are then compared with the results for those reflectors. (author) [pt

4d quantum geometry from 3d supersymmetric gauge theory and holomorphic block

International Nuclear Information System (INIS)

Han, Muxin

2016-01-01

A class of 3d N=2 supersymmetric gauge theories are constructed and shown to encode the simplicial geometries in 4-dimensions. The gauge theories are defined by applying the Dimofte-Gaiotto-Gukov construction http://dx.doi.org/10.1007/s00220-013-1863-2 in 3d-3d correspondence to certain graph complement 3-manifolds. Given a gauge theory in this class, the massive supersymmetric vacua of the theory contain the classical geometries on a 4d simplicial complex. The corresponding 4d simplicial geometries are locally constant curvature (either dS or AdS), in the sense that they are made by gluing geometrical 4-simplices of the same constant curvature. When the simplicial complex is sufficiently refined, the simplicial geometries can approximate all possible smooth geometries on 4-manifold. At the quantum level, we propose that a class of holomorphic blocks defined in http://dx.doi.org/10.1007/JHEP12(2014)177 from the 3d N=2 gauge theories are wave functions of quantum 4d simplicial geometries. In the semiclassical limit, the asymptotic behavior of holomorphic block reproduces the classical action of 4d Einstein-Hilbert gravity in the simplicial context.

Reactivity transient calculatios in research reactor

International Nuclear Information System (INIS)

Santos, R.S. dos

1986-01-01

A digital program for reactivity transient analysis in research reactor and cylindrical geometry was showed quite efficient when compared with methods and programs of the literature, as much in the solution of the neutron kinetics equation as in the thermohydraulic. An improvement in the representation of the feedback reactivity adopted on the program reduced markedly the computation time, with some accuracy. (Author) [pt

Low-Rynolds number k-ε turbulence model for calculation of fast-reactor-channel flows

International Nuclear Information System (INIS)

Mikhin, V.I.

2000-01-01

For calculating the turbulent flows in the complex geometry channels typical for the nuclear reactor installation elements the low-Reynolds-number k-ε turbulence model with the model functions not containing the spatial coordinate like y + is proposed. Such spatial coordinate is usually used for modeling the turbulence near the wall correctly. The model completed on the developed flow of the non-viscous incompressible liquid in the plane channel correctly describes the transition from the laminar regime to the turbulent one. The calculated skin friction coefficients obey the well-known Dean and Zarbi - Reynolds laws. The mean velocity distributions are close to that obtained from the empirical three-layer Karman model. (author)

Resistive toroidal-field coils for tokamak reactors

International Nuclear Information System (INIS)

Kalnavarns, J.; Jassby, D.L.

1980-11-01

This paper analyzes the optimization of the geometry of resistive TF coils of rectangular bore for tokamak fusion test reactors and practical neutron generators. In examining the trade-offs between geometric parameters and magnetic field for reactors giving a specified neutron wall loading, either the resistive power loss or the lifetime coil cost can be minimized. Aspects of cooling, magnetic stress, and construction are addressed for several reference designs. Bending moment distributions in closed form have been derived for rectangular coils on the basis of the theory of rigid frames. Candidate methods of fabrication and of implementing demountable joints are summarized

Hilbert's Nullstellensatz and the Beginning of Algebraic Geometry

Indian Academy of Sciences (India)

The objects of study in algebraic geometry are the loci, or zero sets of polynomials. ... of polynomials in n-variables with real coefficients, indexed by some (finite or ... complex, can defined by only finitely many polynomials.) _ _ _ _ _ _ _ _ ...

Ratcheting problems for ITER [International Thermonuclear Experimental Reactor

International Nuclear Information System (INIS)

Majumdar, S.

1991-01-01

Because of the presence of high cyclic thermal stress, pressure-induced primary stress, and disruption-induced high cyclic primary stress, ratcheting of the first wall poses a serious challenge to the designers of ITER (International Thermonuclear Experimental Reactor). Existing design tools such as the Bree diagram in the ASME Boiler and Pressure Vessels Code, are not directly applicable to ITER, because of important differences in geometry and loading modes. Available alternative models for ratcheting are discussed and new Bree diagrams, that are more relevant for fusion reactor applications, are proposed. 9 refs., 17 figs

TiO2 Solar Photocatalytic Reactor Systems: Selection of Reactor Design for Scale-up and Commercialization—Analytical Review

Directory of Open Access Journals (Sweden)

Yasmine Abdel-Maksoud

2016-09-01

Full Text Available For the last four decades, viability of photocatalytic degradation of organic compounds in water streams has been demonstrated. Different configurations for solar TiO2 photocatalytic reactors have been used, however pilot and demonstration plants are still countable. Degradation efficiency reported as a function of treatment time does not answer the question: which of these reactor configurations is the most suitable for photocatalytic process and optimum for scale-up and commercialization? Degradation efficiency expressed as a function of the reactor throughput and ease of catalyst removal from treated effluent are used for comparing performance of different reactor configurations to select the optimum for scale-up. Comparison included parabolic trough, flat plate, double skin sheet, shallow ponds, shallow tanks, thin-film fixed-bed, thin film cascade, step, compound parabolic concentrators, fountain, slurry bubble column, pebble bed and packed bed reactors. Degradation efficiency as a function of system throughput is a powerful indicator for comparing the performance of photocatalytic reactors of different types and geometries, at different development scales. Shallow ponds, shallow tanks and fountain reactors have the potential of meeting all the process requirements and a relatively high throughput are suitable for developing into continuous industrial-scale treatment units given that an efficient immobilized or supported photocatalyst is used.

Geometry system used in the General Monte Carlo transport code SPARTAN

International Nuclear Information System (INIS)

Bending, R.C.; Easter, P.G.

1974-01-01

The geometry routines used in the general-purpose, three-dimensional particle transport code SPARTAN are described. The code is designed to deal with the very complex geometries encountered in lattice cell and fuel handling calculations, health physics, and shielding problems. Regions of the system being studied may be represented by simple shapes (spheres, cylinders, and so on) or by multinomial surfaces of any order, and many simple shapes may be combined to make up a complex layout. The geometry routines are designed to allow the program to carry out a number of tasks (such as sampling for a random point or tracking a path through several regions) in any order, so that the use of the routines is not restricted to a particular tracking or scoring method. Routines for reading, checking, and printing the data are included. (U.S.)

A spectral nodal method for eigenvalue SN transport problems in two-dimensional rectangular geometry for energy multigroup nuclear reactor global calculations

International Nuclear Information System (INIS)

Silva, Davi Jose M.; Alves Filho, Hermes; Barros, Ricardo C.

2015-01-01

A spectral nodal method is developed for multigroup x,y-geometry discrete ordinates (S N ) eigenvalue problems for nuclear reactor global calculations. This method uses the conventional multigroup SN discretized spatial balance nodal equations with two non-standard auxiliary equations: the spectral diamond (SD) auxiliary equations for the discretization nodes inside the fuel regions, and the spectral Green's function (SGF) auxiliary equations for the non-multiplying regions, such as the baffle and the reactor. This spectral nodal method is derived from the analytical general solution of the SN transverse integrated nodal equations with constant approximations for the transverse leakage terms within each discretization node. The SD and SGF auxiliary equations have parameters, which are determined to preserve the homogeneous and the particular components of these local general solutions. Therefore, we refer to the offered method as the hybrid SD-SGF-Constant Nodal (SD-SGF-CN) method. The S N discretized spatial balance equations, together with the SD and the SGF auxiliary equations form the SD-SGF-CN equations. We solve the SD-SGF-CN equations by using the one-node block inversion inner iterations (NBI), wherein the most recent estimates for the incoming group node-edge average or prescribed boundary conditions are used to evaluate the outgoing group node-edge average fluxes in the directions of the S N transport sweeps, for each estimate of the dominant eigenvalue in the conventional Power outer iterations. We show in numerical calculations that the SD-SGF-CN method is very accurate for coarse-mesh multigroup S N eigenvalue problems, even though the transverse leakage terms are approximated rather simply. (author)

Beam model for seismic analysis of complex shear wall structure based on the strain energy equivalence

International Nuclear Information System (INIS)

Reddy, G.R.; Mahajan, S.C.; Suzuki, Kohei

1997-01-01

A nuclear reactor building structure consists of shear walls with complex geometry, beams and columns. The complexity of the structure is explained in the section Introduction. Seismic analysis of the complex reactor building structure using the continuum mechanics approach may produce good results but this method is very difficult to apply. Hence, the finite element approach is found to be an useful technique for solving the dynamic equations of the reactor building structure. In this approach, the model which uses finite elements such as brick, plate and shell elements may produce accurate results. However, this model also poses some difficulties which are explained in the section Modeling Techniques. Therefore, seismic analysis of complex structures is generally carried out using a lumped mass beam model. This model is preferred because of its simplicity and economy. Nevertheless, mathematical modeling of a shear wall structure as a beam requires specialized skill and a thorough understanding of the structure. For accurate seismic analysis, it is necessary to model more realistically the stiffness, mass and damping. In linear seismic analysis, modeling of the mass and damping may pose few problems compared to modeling the stiffness. When used to represent a complex structure, the stiffness of the beam is directly related to the shear wall section properties such as area, shear area and moment of inertia. Various beam models which are classified based on the method of stiffness evaluation are also explained under the section Modeling Techniques. In the section Case Studies the accuracy and simplicity of the beam models are explained. Among various beam models, the one which evaluates the stiffness using strain energy equivalence proves to be the simplest and most accurate method for modeling the complex shear wall structure. (author)

Application of MCNPX 2.7.D for reactor core management at the research reactor BR2

International Nuclear Information System (INIS)

Kalcheva, Silva; Koonen, Edgar

2011-01-01

The paper discusses application of the Monte Carlo burn up code MCNPX 2.7.D for whole core criticality and depletion analysis of the Material Testing Research Reactor BR2 at SCK-CEN in Mol, Belgium. Two different approaches in the use of MCNPX 2.7.D are presented. The first methodology couples the evolution of fuel depletion, evaluated by MCNPX 2.7.D in an infinite lattice with a steady-state 3-D power distribution in the full core model. The second method represents fully automatic whole core depletion and criticality calculations in the detailed 3-D heterogeneous geometry model of the BR2 reactor. The accuracy of the method and computational time as function of the number of used unique burn up materials in the model are being studied. The depletion capabilities of MCNPX 2.7.D are compared vs. the developed at the BR2 reactor department MCNPX & ORIGEN-S combined method. Testing of MCNPX 2.7.D on the criticality measurements at the BR2 reactor is presented. (author)

Geometry simulation and physics with the CMS forward pixel detector

Energy Technology Data Exchange (ETDEWEB)

Parashar, N [Purdue University Calumet, Hammond, Indiana (United States)], E-mail: Neeti@fnal.gov

2008-06-15

The Forward Pixel Detector of CMS is an integral part of the Tracking system, which will play a key role in addressing the full physics potential of the collected data. It has a very complex geometry that encompasses multilayer structure of its detector modules. This presentation describes the development of geometry simulation for the Forward Pixel Detector. A new geometry package has been developed, which uses the detector description database (DDD) interface for the XML (eXtensive Markup Language) to GEANT simulation. This is necessary for digitization and GEANT4 reconstruction software for tracking. The expected physics performance is also discussed.

Geometry simulation and physics with the CMS forward pixel detector

International Nuclear Information System (INIS)

Parashar, N

2008-01-01

The Forward Pixel Detector of CMS is an integral part of the Tracking system, which will play a key role in addressing the full physics potential of the collected data. It has a very complex geometry that encompasses multilayer structure of its detector modules. This presentation describes the development of geometry simulation for the Forward Pixel Detector. A new geometry package has been developed, which uses the detector description database (DDD) interface for the XML (eXtensive Markup Language) to GEANT simulation. This is necessary for digitization and GEANT4 reconstruction software for tracking. The expected physics performance is also discussed

CFD analysis of the VHTR prismatic core with variation of geometry parameters

Energy Technology Data Exchange (ETDEWEB)

Lira, Carlos A.B.O.; Paiva, Pedro P.D.S., E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear

2017-11-01

The Very High Temperature Reactor is a thermal, graphite moderated and helium cooled nuclear reactor. The purpose of this work is to study the behavior of the VHTR by means of parametric analysis, altering the energy generation profile in the fuel blocks and the influence of modifications in the geometry itself. The coolant flow through the coolant channels and by-pass channels were analyzed in a 1/12{sup th} section of a fuel block column. Geometry was used with by-pass channels of different dimensions, besides one that had only the cooling channels, without by-pass channel. It has been found that the existence of a by-pass flow induces an increase in the temperature gradient in the fuel block. Comparative studies were performed between the results obtained in simulations carried out with different profiles of thermal energy generation (uniform and sinusoidal) in the fuel channels. It was verified that when there is the same total thermal energy generation in the fuel block, the maximum temperature observed in each of the materials is smaller for the generation with sinusoidal profile. Computer simulations were performed using a geometry with a central channel with the same diameter as the others to verify the hypothesis that the existence of a temperature gradient in the fuel block, with the highest temperature at the center and the lowest temperature being at the periphery of this block, is due to the smaller dimension of the coolant channel located in the center of this block. The results obtained confirm the hypothesis. (author)

Topology, ergodic theory, real algebraic geometry Rokhlin's memorial

CERN Document Server

Turaev, V

2001-01-01

This book is dedicated to the memory of the outstanding Russian mathematician, V. A. Rokhlin (1919-1984). It is a collection of research papers written by his former students and followers, who are now experts in their fields. The topics in this volume include topology (the Morse-Novikov theory, spin bordisms in dimension 6, and skein modules of links), real algebraic geometry (real algebraic curves, plane algebraic surfaces, algebraic links, and complex orientations), dynamics (ergodicity, amenability, and random bundle transformations), geometry of Riemannian manifolds, theory of Teichmüller

Geometry through history Euclidean, hyperbolic, and projective geometries

CERN Document Server

Dillon, Meighan I

2018-01-01

Presented as an engaging discourse, this textbook invites readers to delve into the historical origins and uses of geometry. The narrative traces the influence of Euclid’s system of geometry, as developed in his classic text The Elements, through the Arabic period, the modern era in the West, and up to twentieth century mathematics. Axioms and proof methods used by mathematicians from those periods are explored alongside the problems in Euclidean geometry that lead to their work. Students cultivate skills applicable to much of modern mathematics through sections that integrate concepts like projective and hyperbolic geometry with representative proof-based exercises. For its sophisticated account of ancient to modern geometries, this text assumes only a year of college mathematics as it builds towards its conclusion with algebraic curves and quaternions. Euclid’s work has affected geometry for thousands of years, so this text has something to offer to anyone who wants to broaden their appreciation for the...

CFD analysis of multiphase coolant flow through fuel rod bundles in advanced pressure tube nuclear reactors

International Nuclear Information System (INIS)

Catana, A.; Turcu, I.; Prisecaru, I.; Dupleac, D.; Danila, N.

2010-01-01

The key component of a pressure tube nuclear reactor core is pressure tube filled with a stream of fuel bundles. This feature makes them suitable for CFD thermal-hydraulic analysis. A methodology for CFD analysis applied to pressure tube nuclear reactors is presented in this paper, which is focused on advanced pressure tube nuclear reactors. The complex flow conditions inside pressure tube are analysed by using the Eulerian multiphase model implemented in FLUENT CFD computer code. Fuel rods in these channels are superheated but the liquid is under high pressure, so it is sub-cooled in normal operating conditions on most of pressure tube length. In the second half of pressure tube length, the onset of boiling occurs, so the flow consists of a gas liquid mixture, with the volume of gas increasing along the length of the channel in the direction of the flow. Limited computer resources enforced us to use CFD analysis for segments of pressure tube. Significant local geometries (junctions, spacers) were simulated. Main results of this work are: prediction of main thermal-hydraulic parameters along pressure tube including CHF evaluation through fuel assemblies. (authors)

Automated scoping methodology for liquid metal natural circulation small reactor

International Nuclear Information System (INIS)

Son, Hyung M.; Suh, Kune Y.

2014-01-01

Highlights: • Automated scoping methodology for natural circulation small modular reactor is developed. • In-house code is developed to carry out system analysis and core geometry generation during scoping. • Adjustment relations are obtained to correct the critical core geometry out of diffusion theory. • Optimized design specification is found using objective function value. • Convex hull volume is utilized to quantify the impact of different constraints on the scope range. - Abstract: A novel scoping method is proposed that can automatically generate design variable range of the natural circulation driven liquid metal cooled small reactor. From performance requirements based upon Generation IV system roadmap, appropriate structure materials are selected and engineering constraints are compiled based upon literature. Utilizing ASME codes and standards, appropriate geometric sizing criteria on constituting components are developed to ensure integrity of the system during its lifetime. In-house one dimensional thermo-hydraulic system analysis code is developed based upon momentum integral model and finite element methods to deal with non-uniform descritization of temperature nodes for convection and thermal diffusion equation of liquid metal coolant. In order to quickly generate critical core dimensions out of given unit cell information, an adjustment relation that relates the critical geometry estimated from one-group diffusion and that from MCNP code is constructed and utilized throughout the process. For the selected unit cell dimension ranges, burnup calculations are carried out to check the cores can generate energy over the reactor lifetime. Utilizing random method, sizing criteria, and in-house analysis codes, an automated scoping methodology is developed. The methodology is applied to nitride fueled integral type lead cooled natural circulation reactor concept to generate design scopes which satisfies given constraints. Three dimensional convex

Space, Geometry and the Imagination from Antiquity to the Early Modern Age

CERN Document Server

Mathematizing Space : The Objects of Geometry from Antiquity to the Early Modern Age

2015-01-01

This book brings together papers of the conference on 'Space, Geometry and the Imagination from Antiquity to the Modern Age' held in Berlin, Germany, 27-29 August 2012. Focusing on the interconnections between the history of geometry and the philosophy of space in the pre-Modern and Early Modern Age, the essays in this volume are particularly directed toward elucidating the complex epistemological revolution that transformed the classical geometry of figures into the modern geometry of space. Contributors: Graciela De Pierris Franco Farinelli Michael Friedman Daniel Garber Jeremy Gray Gary Hatfield Andrew Janiak Douglas Jesseph Alexander Jones Henry Mendell David Rabouin

A high temperature reactor for ship propulsion

International Nuclear Information System (INIS)

Lobet, P.; Seigel, R.; Thompson, A.C.; Beadnell, R.M.; Beeley, P.A.

2002-01-01

The initial thermal hydraulic and physics design of a high temperature gas cooled reactor for ship propulsion is described. The choice of thermodynamic cycle and thermal power is made to suit the marine application. Several configurations of a Helium cooled, Graphite moderated reactor are then analysed using the WIMS and MONK codes from AEA Technology. Two geometries of fuel elements formed using micro spheres in prismatic blocks, and various arrangements of control rods and poison rods are examined. Reactivity calculations through life are made and a pattern of rod insertion to flatten the flux is proposed and analysed. Thermal hydraulic calculations are made to find maximum fuel temperature under high power with optimized flow distribution. Maximum temperature after loss of flow and temperatures in the reactor vessel are also computed. The temperatures are significantly below the known limits for the type of fuel proposed. It is concluded that the reactor can provide the required power and lifetime between refueling within likely space and weight constraints. (author)

Comparison of THALES and VIPRE-01 Subchannel Codes for Loss of Flow and Single Reactor Coolant Pump Rotor Seizure Accidents using Lumped Channel APR1400 Geometry

Energy Technology Data Exchange (ETDEWEB)

Oezdemir, Erdal; Moon, Kang Hoon; Oh, Seung Jong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of); Kim, Yongdeog [KHNP-CRI, Daejeon (Korea, Republic of)

2014-10-15

Subchannel analysis plays important role to evaluate safety critical parameters like minimum departure from nucleate boiling ratio (MDNBR), peak clad temperature and fuel centerline temperature. In this study, two different subchannel codes, VIPRE-01 (Versatile Internals and Component Program for Reactors: EPRI) and THALES (Thermal Hydraulic AnaLyzer for Enhanced Simulation of core) are examined. In this study, two different transient cases for which MDNBR result play important role are selected to conduct analysis with THALES and VIPRE-01 subchannel codes. In order to get comparable results same core geometry, fuel parameters, correlations and models are selected for each code. MDNBR results from simulations by both code are agree with each other with negligible difference. Whereas, simulations conducted by enabling conduction model in VIPRE-01 shows significant difference from the results of THALES.

Use of MCNP for characterization of reactor vessel internals waste from decommissioned nuclear reactors

International Nuclear Information System (INIS)

Love, E.F.; Pauley, K.A.; Reid, B.D.

1995-09-01

This study describes the use of the Monte Carlo Neutron-Photon (MCNP) code for determining activation levels of irradiated reactor vessel internals hardware. The purpose of the analysis is to produce data for the Department of Energy's Greater-Than-Class C Low-Level Radioactive Waste Program. An MCNP model was developed to analyze the Yankee Rowe reactor facility. The model incorporates reactor geometry, material compositions, and operating history data acquired from Yankee Atomic Electric Company. In addition to the base activation analysis, parametric studies were performed to determine the sensitivity of activation to specific parameters. A component sampling plan was also developed to validate the model results, although the plan was not implemented. The calculations for the Yankee Rowe reactor predict that only the core baffle and the core support plates will be activated to levels above the Class C limits. The parametric calculations show, however, that the large uncertainties in the material compositions could cause errors in the estimates that could also increase the estimated activation level of the core barrel to above the Class C limits. Extrapolation of the results to other reactor facilities indicates that in addition to the baffle and support plates, core barrels may also be activated to above Class C limits; however the classification will depend on the specific operating conditions of the reactor and the specific material compositions of the metal, as well as the use of allowable concentration averaging practices in packaging and classifying the waste

An ensemble classifier to predict track geometry degradation

International Nuclear Information System (INIS)

Cárdenas-Gallo, Iván; Sarmiento, Carlos A.; Morales, Gilberto A.; Bolivar, Manuel A.; Akhavan-Tabatabaei, Raha

2017-01-01

Railway operations are inherently complex and source of several problems. In particular, track geometry defects are one of the leading causes of train accidents in the United States. This paper presents a solution approach which entails the construction of an ensemble classifier to forecast the degradation of track geometry. Our classifier is constructed by solving the problem from three different perspectives: deterioration, regression and classification. We considered a different model from each perspective and our results show that using an ensemble method improves the predictive performance. - Highlights: • We present an ensemble classifier to forecast the degradation of track geometry. • Our classifier considers three perspectives: deterioration, regression and classification. • We construct and test three models and our results show that using an ensemble method improves the predictive performance.

Preliminary neutronic design of TRIGA Mark II Reactor

International Nuclear Information System (INIS)

Sarikaya, B.; Tombakoglu, M.; Cecen, Y.; Kadiroglu, O. K.

2001-01-01

It is very important to analyse the behaviour of the research reactors, since, they play a key role in developing the power reactor technology and radiation applications such as isotope generation for medical treatments. In this study, the neutronic behaviour of the TRIGA MARK II reactor, owned and operated by Istanbul Technical University is analysed by using the SCALE code system. In the analysis, in order to overcome the disadvantages of special TRIGA codes, such as TRIGAP, the SCALE code system is chosen to perform the calculations. TRIGAP and similar codes have limited geometrical (one-dimensional geometry) and cross sectional options (two-group calculations), however, SCALE has the capability of wider range of geometrical modelling capability (three-dimensional modelling is possible) and multi-group calculations are possible

Interface tracking simulations of bubbly flows in PWR relevant geometries

Energy Technology Data Exchange (ETDEWEB)

Fang, Jun, E-mail: jfang3@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Rasquin, Michel, E-mail: michel.rasquin@colorado.edu [Aerospace Engineering Department, University of Colorado, Boulder, CO 80309 (United States); Bolotnov, Igor A., E-mail: igor_bolotnov@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States)

2017-02-15

Highlights: • Simulations were performed for turbulent bubbly flows in PWR subchannel geometry. • Liquid turbulence is fully resolved by direct numerical simulation approach. • Bubble behavior is captured using level-set interface tracking method. • Time-averaged single- and two-phase turbulent flow statistical quantities are obtained. - Abstract: The advances in high performance computing (HPC) have allowed direct numerical simulation (DNS) approach coupled with interface tracking methods (ITM) to perform high fidelity simulations of turbulent bubbly flows in various complex geometries. In this work, we have chosen the geometry of the pressurized water reactor (PWR) core subchannel to perform a set of interface tracking simulations (ITS) with fully resolved liquid turbulence. The presented research utilizes a massively parallel finite-element based code, PHASTA, for the subchannel geometry simulations of bubbly flow turbulence. The main objective for this research is to demonstrate the ITS capabilities in gaining new insight into bubble/turbulence interactions and assisting the development of improved closure laws for multiphase computational fluid dynamics (M-CFD). Both single- and two-phase turbulent flows were studied within a single PWR subchannel. The analysis of numerical results includes the mean gas and liquid velocity profiles, void fraction distribution and turbulent kinetic energy profiles. Two sets of flow rates and bubble sizes were used in the simulations. The chosen flow rates corresponded to the Reynolds numbers of 29,079 and 80,775 based on channel hydraulic diameter (D{sub h}) and mean velocity. The finite element unstructured grids utilized for these simulations include 53.8 million and 1.11 billion elements, respectively. This has allowed to fully resolve all the turbulence scales and the deformable interfaces of individual bubbles. For the two-phase flow simulations, a 1% bubble volume fraction was used which resulted in 17 bubbles in

FURNACE-J, 2-D Diffusion Burnup for Fast Reactors from JAERI Fast-Set

International Nuclear Information System (INIS)

Ikawa, Koji

1984-01-01

1 - Nature of physical problem solved: FURNACEJ is a two-dimensional diffusion-burnup code for use in the detailed burnup analysis of fast reactors. The code is an extension code of the FURNACE. There exists no essential difference between FURNACE and FURNACEJ. However, the latter can deal with JAERI-Fast-Set as its cross section library, while the former is designed to use ABBN set. Additionally, in FURNACEJ, group-dependent and -independent transverse buckling of each region can be computed and punched on cards, if desired. This is prepared for users so as to use them as input data for detailed two-dimensional x-y calculations. 2 - Restrictions on the complexity of the problem: Only r-z geometry is available

Architectural geometry

KAUST Repository

Pottmann, Helmut

2014-11-26

Around 2005 it became apparent in the geometry processing community that freeform architecture contains many problems of a geometric nature to be solved, and many opportunities for optimization which however require geometric understanding. This area of research, which has been called architectural geometry, meanwhile contains a great wealth of individual contributions which are relevant in various fields. For mathematicians, the relation to discrete differential geometry is significant, in particular the integrable system viewpoint. Besides, new application contexts have become available for quite some old-established concepts. Regarding graphics and geometry processing, architectural geometry yields interesting new questions but also new objects, e.g. replacing meshes by other combinatorial arrangements. Numerical optimization plays a major role but in itself would be powerless without geometric understanding. Summing up, architectural geometry has become a rewarding field of study. We here survey the main directions which have been pursued, we show real projects where geometric considerations have played a role, and we outline open problems which we think are significant for the future development of both theory and practice of architectural geometry.

Architectural geometry

KAUST Repository

Pottmann, Helmut; Eigensatz, Michael; Vaxman, Amir; Wallner, Johannes

2014-01-01

Around 2005 it became apparent in the geometry processing community that freeform architecture contains many problems of a geometric nature to be solved, and many opportunities for optimization which however require geometric understanding. This area of research, which has been called architectural geometry, meanwhile contains a great wealth of individual contributions which are relevant in various fields. For mathematicians, the relation to discrete differential geometry is significant, in particular the integrable system viewpoint. Besides, new application contexts have become available for quite some old-established concepts. Regarding graphics and geometry processing, architectural geometry yields interesting new questions but also new objects, e.g. replacing meshes by other combinatorial arrangements. Numerical optimization plays a major role but in itself would be powerless without geometric understanding. Summing up, architectural geometry has become a rewarding field of study. We here survey the main directions which have been pursued, we show real projects where geometric considerations have played a role, and we outline open problems which we think are significant for the future development of both theory and practice of architectural geometry.

RCPO1 - A Monte Carlo program for solving neutron and photon transport problems in three dimensional geometry with detailed energy description and depletion capability

International Nuclear Information System (INIS)

Ondis, L.A. II; Tyburski, L.J.; Moskowitz, B.S.

2000-01-01

The RCP01 Monte Carlo program is used to analyze many geometries of interest in nuclear design and analysis of light water moderated reactors such as the core in its pressure vessel with complex piping arrangement, fuel storage arrays, shipping and container arrangements, and neutron detector configurations. Written in FORTRAN and in use on a variety of computers, it is capable of estimating steady state neutron or photon reaction rates and neutron multiplication factors. The energy range covered in neutron calculations is that relevant to the fission process and subsequent slowing-down and thermalization, i.e., 20 MeV to 0 eV. The same energy range is covered for photon calculations

RCPO1 - A Monte Carlo program for solving neutron and photon transport problems in three dimensional geometry with detailed energy description and depletion capability

Energy Technology Data Exchange (ETDEWEB)

Ondis, L.A., II; Tyburski, L.J.; Moskowitz, B.S.

2000-03-01

The RCP01 Monte Carlo program is used to analyze many geometries of interest in nuclear design and analysis of light water moderated reactors such as the core in its pressure vessel with complex piping arrangement, fuel storage arrays, shipping and container arrangements, and neutron detector configurations. Written in FORTRAN and in use on a variety of computers, it is capable of estimating steady state neutron or photon reaction rates and neutron multiplication factors. The energy range covered in neutron calculations is that relevant to the fission process and subsequent slowing-down and thermalization, i.e., 20 MeV to 0 eV. The same energy range is covered for photon calculations.

Investigation on the influence of electrode geometry on characteristics of coaxial dielectric barrier discharge reactor driven by an oscillating microsecond pulsed power supply

Science.gov (United States)

Miao, Chuanrun; Liu, Feng; Wang, Qian; Cai, Meiling; Fang, Zhi

2018-03-01

In this paper, an oscillating microsecond pulsed power supply with rise time of several tens of nanosecond (ns) is used to excite a coaxial DBD with double layer dielectric barriers. The effects of various electrode geometries by changing the size of inner quartz tube (different electrode gaps) on the discharge uniformity, power deposition, energy efficiency, and operation temperature are investigated by electrical, optical, and temperature diagnostics. The electrical parameters of the coaxial DBD are obtained from the measured applied voltage and current using an equivalent electrical model. The energy efficiency and the power deposition in air gap of coaxial DBD with various electrode geometries are also obtained with the obtained electrical parameters, and the heat loss and operation temperature are analyzed by a heat conduction model. It is found that at the same applied voltage, with the increasing of the air gap, the discharge uniformity becomes worse and the discharge power deposition and the energy efficiency decrease. At 2.5 mm air gap and 24 kV applied voltage, the energy efficiency of the coaxial DBD reaches the maximum value of 68.4%, and the power deposition in air gap is 23.6 W and the discharge uniformity is the best at this case. The corresponding operation temperature of the coaxial DBD reaches 64.3 °C after 900 s operation and the temperature of the inner dielectric barrier is 114.4 °C under thermal balance. The experimental results provide important experimental references and are important to optimize the design and the performance of coaxial DBD reactor.

Mathematical efficiency calibration with uncertain source geometries using smart optimization

International Nuclear Information System (INIS)

Menaa, N.; Bosko, A.; Bronson, F.; Venkataraman, R.; Russ, W. R.; Mueller, W.; Nizhnik, V.; Mirolo, L.

2011-01-01

The In Situ Object Counting Software (ISOCS), a mathematical method developed by CANBERRA, is a well established technique for computing High Purity Germanium (HPGe) detector efficiencies for a wide variety of source shapes and sizes. In the ISOCS method, the user needs to input the geometry related parameters such as: the source dimensions, matrix composition and density, along with the source-to-detector distance. In many applications, the source dimensions, the matrix material and density may not be well known. Under such circumstances, the efficiencies may not be very accurate since the modeled source geometry may not be very representative of the measured geometry. CANBERRA developed an efficiency optimization software known as 'Advanced ISOCS' that varies the not well known parameters within user specified intervals and determines the optimal efficiency shape and magnitude based on available benchmarks in the measured spectra. The benchmarks could be results from isotopic codes such as MGAU, MGA, IGA, or FRAM, activities from multi-line nuclides, and multiple counts of the same item taken in different geometries (from the side, bottom, top etc). The efficiency optimization is carried out using either a random search based on standard probability distributions, or using numerical techniques that carry out a more directed (referred to as 'smart' in this paper) search. Measurements were carried out using representative source geometries and radionuclide distributions. The radionuclide activities were determined using the optimum efficiency and compared against the true activities. The 'Advanced ISOCS' method has many applications among which are: Safeguards, Decommissioning and Decontamination, Non-Destructive Assay systems and Nuclear reactor outages maintenance. (authors)

Two lectures on D-geometry and noncommutative geometry

International Nuclear Information System (INIS)

Douglas, M.R.

1999-01-01

This is a write-up of lectures given at the 1998 Spring School at the Abdus Salam ICTP. We give a conceptual introduction to D-geometry, the study of geometry as seen by D-branes in string theory, and to noncommutative geometry as it has appeared in D-brane and Matrix theory physics. (author)

Burnup-dependent core neutronics analysis of plate-type research reactor using deterministic and stochastic methods

International Nuclear Information System (INIS)

Liu, Shichang; Wang, Guanbo; Liang, Jingang; Wu, Gaochen; Wang, Kan

2015-01-01

Highlights: • DRAGON & DONJON were applied in burnup calculations of plate-type research reactors. • Continuous-energy Monte Carlo burnup calculations by RMC were chosen as references. • Comparisons of keff, isotopic densities and power distribution were performed. • Reasons leading to discrepancies between two different approaches were analyzed. • DRAGON & DONJON is capable of burnup calculations with appropriate treatments. - Abstract: The burnup-dependent core neutronics analysis of the plate-type research reactors such as JRR-3M poses a challenge for traditional neutronics calculational tools and schemes for power reactors, due to the characteristics of complex geometry, highly heterogeneity, large leakage and the particular neutron spectrum of the research reactors. Two different theoretical approaches, the deterministic and the stochastic methods, are used for the burnup-dependent core neutronics analysis of the JRR-3M plate-type research reactor in this paper. For the deterministic method the neutronics codes DRAGON & DONJON are used, while the continuous-energy Monte Carlo code RMC (Reactor Monte Carlo code) is employed for the stochastic one. In the first stage, the homogenizations of few-group cross sections by DRAGON and the full core diffusion calculations by DONJON have been verified by comparing with the detailed Monte Carlo simulations. In the second stage, the burnup-dependent calculations of both assembly level and the full core level were carried out, to examine the capability of the deterministic code system DRAGON & DONJON to reliably simulate the burnup-dependent behavior of research reactors. The results indicate that both RMC and DRAGON & DONJON code system are capable of burnup-dependent neutronics analysis of research reactors, provided that appropriate treatments are applied in both assembly and core levels for the deterministic codes

Fracture network modelling: an integrated approach for realisation of complex fracture network geometries

International Nuclear Information System (INIS)