Energy Technology Data Exchange (ETDEWEB)
Lieberoth, J.
1975-06-15
The numerical solution of the neutron diffusion equation plays a very important role in the analysis of nuclear reactors. A wide variety of numerical procedures has been proposed, at which most of the frequently used numerical methods are fundamentally based on the finite- difference approximation where the partial derivatives are approximated by the finite difference. For complex geometries, typical of the practical reactor problems, the computational accuracy of the finite-difference method is seriously affected by the size of the mesh width relative to the neutron diffusion length and by the heterogeneity of the medium. Thus, a very large number of mesh points are generally required to obtain a reasonably accurate approximate solution of the multi-dimensional diffusion equation. Since the computation time is approximately proportional to the number of mesh points, a detailed multidimensional analysis, based on the conventional finite-difference method, is still expensive even with modern large-scale computers. Accordingly, there is a strong incentive to develop alternatives that can reduce the number of mesh-points and still retain accuracy. One of the promising alternatives is the finite element method, which consists of the expansion of the neutron flux by piecewise polynomials. One of the advantages of this procedure is its flexibility in selecting the locations of the mesh points and the degree of the expansion polynomial. The small number of mesh points of the coarse grid enables to store the results of several of the least outer iterations and to calculate well extrapolated values of them by comfortable formalisms. This holds especially if only one energy distribution of fission neutrons is assumed for all fission processes in the reactor, because the whole information of an outer iteration is contained in a field of fission rates which has the size of all mesh points of the coarse grid.
Symmetries and the coarse-mesh method
International Nuclear Information System (INIS)
Makai, M.
1980-10-01
This report approaches the basic problem of the coarse-mesh method from a new side. Group theory is used for the determination of the space dependency of the flux. The result is a method called ANANAS after the analytic-analytic solution. This method was tested on two benchmark problems: one given by Melice and the IAEA benchmark. The ANANAS program is an experimental one. The method was intended for use in hexagonal geometry. (Auth.)
Coarse-mesh rebalancing acceleration for eigenvalue problems
International Nuclear Information System (INIS)
Asaoka, T.; Nakahara, Y.; Miyasaka, S.
1974-01-01
The coarse-mesh rebalance method is adopted for Monte Carlo schemes for aiming at accelerating the convergence of a source iteration process. At every completion of the Monte Carlo game for one batch of neutron histories, the scaling factor for the neutron flux is calculated to achieve the neutron balance in each coarse-mesh zone into which the total system is divided. This rebalance factor is multiplied to the weight of each fission source neutron in the coarse-mesh zone for playing the next Monte Carlo game. The numerical examples have shown that the coarse-mesh rebalance Monte Carlo calculation gives a good estimate of the eigenvalue already after several batches with a negligible extra computer time compared to the standard Monte Carlo. 5 references. (U.S.)
Application of coarse-mesh methods to fluid dynamics equations
International Nuclear Information System (INIS)
Romstedt, P.; Werner, W.
1977-01-01
An Asymmetric Weighted Residual (ASWR) method for fluid dynamics equations is described. It leads to local operators with a 7-point Finite Difference (FD) structure, which is independent of the degree of the approximating polynomials. An 1-dimensional problem was solved by both this ASWR-method and a commonly used FD-method. The numerical results demonstrate that the ASWR-method combines high accuracy on a coarse computational mesh with short computing time per space point. The posibility of using fewer space points consequently brings about a considerable reduction in total running time for the ASWR-method as compared with conventional FD-methods. (orig.) [de
Reactor calculation in coarse mesh by finite element method applied to matrix response method
International Nuclear Information System (INIS)
Nakata, H.
1982-01-01
The finite element method is applied to the solution of the modified formulation of the matrix-response method aiming to do reactor calculations in coarse mesh. Good results are obtained with a short running time. The method is applicable to problems where the heterogeneity is predominant and to problems of evolution in coarse meshes where the burnup is variable in one same coarse mesh, making the cross section vary spatially with the evolution. (E.G.) [pt
New procedure for criticality search using coarse mesh nodal methods
International Nuclear Information System (INIS)
Pereira, Wanderson F.; Silva, Fernando C. da; Martinez, Aquilino S.
2011-01-01
The coarse mesh nodal methods have as their primary goal to calculate the neutron flux inside the reactor core. Many computer systems use a specific form of calculation, which is called nodal method. In classical computing systems that use the criticality search is made after the complete convergence of the iterative process of calculating the neutron flux. In this paper, we proposed a new method for the calculation of criticality, condition which will be over very iterative process of calculating the neutron flux. Thus, the processing time for calculating the neutron flux was reduced by half compared with the procedure developed by the Nuclear Engineering Program of COPPE/UFRJ (PEN/COPPE/UFRJ). (author)
New procedure for criticality search using coarse mesh nodal methods
Energy Technology Data Exchange (ETDEWEB)
Pereira, Wanderson F.; Silva, Fernando C. da; Martinez, Aquilino S., E-mail: wneto@con.ufrj.b, E-mail: fernando@con.ufrj.b, E-mail: Aquilino@lmp.ufrj.b [Coordenacao dos Programas de Pos-Graduacao de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear
2011-07-01
The coarse mesh nodal methods have as their primary goal to calculate the neutron flux inside the reactor core. Many computer systems use a specific form of calculation, which is called nodal method. In classical computing systems that use the criticality search is made after the complete convergence of the iterative process of calculating the neutron flux. In this paper, we proposed a new method for the calculation of criticality, condition which will be over very iterative process of calculating the neutron flux. Thus, the processing time for calculating the neutron flux was reduced by half compared with the procedure developed by the Nuclear Engineering Program of COPPE/UFRJ (PEN/COPPE/UFRJ). (author)
A coarse-mesh nodal method-diffusive-mesh finite difference method
International Nuclear Information System (INIS)
Joo, H.; Nichols, W.R.
1994-01-01
Modern nodal methods have been successfully used for conventional light water reactor core analyses where the homogenized, node average cross sections (XSs) and the flux discontinuity factors (DFs) based on equivalence theory can reliably predict core behavior. For other types of cores and other geometries characterized by tightly-coupled, heterogeneous core configurations, the intranodal flux shapes obtained from a homogenized nodal problem may not accurately portray steep flux gradients near fuel assembly interfaces or various reactivity control elements. This may require extreme values of DFs (either very large, very small, or even negative) to achieve a desired solution accuracy. Extreme values of DFs, however, can disrupt the convergence of the iterative methods used to solve for the node average fluxes, and can lead to a difficulty in interpolating adjacent DF values. Several attempts to remedy the problem have been made, but nothing has been satisfactory. A new coarse-mesh nodal scheme called the Diffusive-Mesh Finite Difference (DMFD) technique, as contrasted with the coarse-mesh finite difference (CMFD) technique, has been developed to resolve this problem. This new technique and the development of a few-group, multidimensional kinetics computer program are described in this paper
Variationally derived coarse mesh methods using an alternative flux representation
International Nuclear Information System (INIS)
Wojtowicz, G.; Holloway, J.P.
1995-01-01
Investigation of a previously reported variational technique for the solution of the 1-D, 1-group neutron transport equation in reactor lattices has inspired the development of a finite element formulation of the method. Compared to conventional homogenization methods in which node homogenized cross sections are used, the coefficients describing this system take on greater spatial dependence. However, the methods employ an alternative flux representation which allows the transport equation to be cast into a form whose solution has only a slow spatial variation and, hence, requires relatively few variables to describe. This alternative flux representation and the stationary property of a variational principle define a class of coarse mesh discretizations of transport theory capable of achieving order of magnitude reductions of eigenvalue and pointwise scalar flux errors as compared with diffusion theory while retaining diffusion theory's relatively low cost. Initial results of a 1-D spectral element approach are reviewed and used to motivate the finite element implementation which is more efficient and almost as accurate; one and two group results of this method are described
International Nuclear Information System (INIS)
Nishida, Takahiko; Horikami, Kunihiko; Suzuki, Tadakazu; Nakahara, Yasuaki; Taji, Yukichi
1975-09-01
The coarse-mesh rebalancing technique is introduced into the general-purpose neutron and gamma-ray Monte Carlo transport code MORSE, to accelerate the convergence rate of the iteration process for eigenvalue calculation in a nuclear reactor system. Two subroutines are thus attached to the code. One is bookkeeping routine 'COARSE' for obtaining the quantities related with the neutron balance in each coarse mesh cell, such as the number of neutrons absorbed in the cell, from random walks of neutrons in a batch. The other is rebalance factor calculation routine 'REBAL' for obtaining the scaling factor whereby the neutron flux in the cell is multiplied to attain the neutron balance. The two subroutines and algorithm of the coarse mesh rebalancing acceleration in a Monte Carlo game are described. (auth.)
Coarse-mesh discretized low-order quasi-diffusion equations for subregion averaged scalar fluxes
International Nuclear Information System (INIS)
Anistratov, D. Y.
2004-01-01
In this paper we develop homogenization procedure and discretization for the low-order quasi-diffusion equations on coarse grids for core-level reactor calculations. The system of discretized equations of the proposed method is formulated in terms of the subregion averaged group scalar fluxes. The coarse-mesh solution is consistent with a given fine-mesh discretization of the transport equation in the sense that it preserves a set of average values of the fine-mesh transport scalar flux over subregions of coarse-mesh cells as well as the surface currents, and eigenvalue. The developed method generates numerical solution that mimics the large-scale behavior of the transport solution within assemblies. (authors)
Formulation of coarse mesh finite difference to calculate mathematical adjoint flux
International Nuclear Information System (INIS)
Pereira, Valmir; Martinez, Aquilino Senra; Silva, Fernando Carvalho da
2002-01-01
The objective of this work is the obtention of the mathematical adjoint flux, having as its support the nodal expansion method (NEM) for coarse mesh problems. Since there are difficulties to evaluate this flux by using NEM. directly, a coarse mesh finite difference program was developed to obtain this adjoint flux. The coarse mesh finite difference formulation (DFMG) adopted uses results of the direct calculation (node average flux and node face averaged currents) obtained by NEM. These quantities (flux and currents) are used to obtain the correction factors which modify the classical finite differences formulation . Since the DFMG formulation is also capable of calculating the direct flux it was also tested to obtain this flux and it was verified that it was able to reproduce with good accuracy both the flux and the currents obtained via NEM. In this way, only matrix transposition is needed to calculate the mathematical adjoint flux. (author)
Coarse-mesh method for multidimensional, mixed-lattice diffusion calculations
International Nuclear Information System (INIS)
Dodds, H.L. Jr.; Honeck, H.C.; Hostetler, D.E.
1977-01-01
A coarse-mesh finite difference method has been developed for multidimensional, mixed-lattice reactor diffusion calculations, both statics and kinetics, in hexagonal geometry. Results obtained with the coarse-mesh (CM) method have been compared with a conventional mesh-centered finite difference method and with experiment. The results of this comparison indicate that the accuracy of the CM method for highly heterogeneous (mixed) lattices using one point per hexagonal mesh element (''hex'') is about the same as the conventional method with six points per hex. Furthermore, the computing costs (i.e., central processor unit time and core storage requirements) of the CM method with one point per hex are about the same as the conventional method with one point per hex
Network topology exploration of mesh-based coarse-grain reconfigurable architectures
Bansal, N.; Gupta, S.; Dutt, N.D.; Nicolau, A.; Gupta, R.
2004-01-01
Several coarse-grain reconfigurable architectures proposed recently consist of a large number of processing elements (PEs) connected in a mesh-like network topology. We study the effects of three aspects of network topology exploration on the performance of applications on these architectures: (a)
A general coarse and fine mesh solution scheme for fluid flow modeling in VHTRS
International Nuclear Information System (INIS)
Clifford, I; Ivanov, K; Avramova, M.
2011-01-01
Coarse mesh Computational Fluid Dynamics (CFD) methods offer several advantages over traditional coarse mesh methods for the safety analysis of helium-cooled graphite-moderated Very High Temperature Reactors (VHTRs). This relatively new approach opens up the possibility for system-wide calculations to be carried out using a consistent set of field equations throughout the calculation, and subsequently the possibility for hybrid coarse/fine mesh or hierarchical multi scale CFD simulations. To date, a consistent methodology for hierarchical multi-scale CFD has not been developed. This paper describes work carried out in the initial development of a multi scale CFD solver intended to be used for the safety analysis of VHTRs. The VHTR is considered on any scale to consist of a homogenized two-phase mixture of fluid and stationary solid material of varying void fraction. A consistent set of conservation equations was selected such that they reduce to the single-phase conservation equations for the case where void fraction is unity. The discretization of the conservation equations uses a new pressure interpolation scheme capable of capturing the discontinuity in pressure across relatively large changes in void fraction. Based on this, a test solver was developed which supports fully unstructured meshes for three-dimensional time-dependent compressible flow problems, including buoyancy effects. For typical VHTR flow phenomena the new solver shows promise as an effective candidate for predicting the flow behavior on multiple scales, as it is capable of modeling both fine mesh single phase flows as well as coarse mesh flows in homogenized regions containing both fluid and solid materials. (author)
Simulation of transients with space-dependent feedback by coarse mesh flux expansion method
International Nuclear Information System (INIS)
Langenbuch, S.; Maurer, W.; Werner, W.
1975-01-01
For the simulation of the time-dependent behaviour of large LWR-cores, even the most efficient Finite-Difference (FD) methods require a prohibitive amount of computing time in order to achieve results of acceptable accuracy. Static CM-solutions computed with a mesh-size corresponding to the fuel element structure (about 20 cm) are at least as accurate as FD-solutions computed with about 5 cm mesh-size. For 3d-calculations this results in a reduction of storage requirements by a factor 60 and of computing costs by a factor 40, relative to FD-methods. These results have been obtained for pure neutronic calculations, where feedback is not taken into account. In this paper it is demonstrated that the method retains its accuracy also in kinetic calculations, even in the presence of strong space dependent feedback. (orig./RW) [de
Coarse mesh finite element method for boiling water reactor physics analysis
International Nuclear Information System (INIS)
Ellison, P.G.
1983-01-01
A coarse mesh method is formulated for the solution of Boiling Water Reactor physics problems using two group diffusion theory. No fuel assembly cross-section homogenization is required; water gaps, control blades and fuel pins of varying enrichments are treated explicitly. The method combines constrained finite element discretization with infinite lattice super cell trial functions to obtain coarse mesh solutions for which the only approximations are along the boundaries between fuel assemblies. The method is applied to bench mark Boiling Water Reactor problems to obtain both the eigenvalue and detailed flux distributions. The solutions to these problems indicate the method is useful in predicting detailed power distributions and eigenvalues for Boiling Water Reactor physics problems
A study of coarse mesh collision probability correction factors in slab lattices
International Nuclear Information System (INIS)
Buckler, A.N.
1975-07-01
Calculations of collision probability leakage estimates are performed in one dimensional slab geometry with one neutron group to gain some insight into methods of correction for the coarseness of the mesh H. The chief result is that the correction factor, beta, can be written as CD/H where C → 4 for the diffusion limit. An explicit expression for C is derived in terms of the E 3 function, for a linear flux variation across the slabs. (author)
Generalized Coarse-Mesh Rebalance Method for Acceleration of Neutron Transport Calculations
International Nuclear Information System (INIS)
Yamamoto, Akio
2005-01-01
This paper proposes a new acceleration method for neutron transport calculations: the generalized coarse-mesh rebalance (GCMR) method. The GCMR method is a unified scheme of the traditional coarse-mesh rebalance (CMR) and the coarse-mesh finite difference (CMFD) acceleration methods. Namely, by using an appropriate acceleration factor, formulation of the GCMR method becomes identical to that of the CMR or CMFD method. This also indicates that the convergence property of the GCMR method can be controlled by the acceleration factor since the convergence properties of the CMR and CMFD methods are generally different. In order to evaluate the convergence property of the GCMR method, a linearized Fourier analysis was carried out for a one-group homogeneous medium, and the results clarified the relationship between the acceleration factor and the spectral radius. It was also shown that the spectral radius of the GCMR method is smaller than those of the CMR and CMFD methods. Furthermore, the Fourier analysis showed that when an appropriate acceleration factor was used, the spectral radius of the GCMR method did not exceed unity in this study, which was in contrast to the results of the CMR or the CMFD method. Application of the GCMR method to practical calculations will be easy when the CMFD acceleration is already adopted in a transport code. By multiplying a suitable acceleration factor to a coefficient (D FD ) of a finite difference formulation, one can improve the numerical instability of the CMFD acceleration method
Coarse mesh and one-cell block inversion based diffusion synthetic acceleration
Kim, Kang-Seog
DSA (Diffusion Synthetic Acceleration) has been developed to accelerate the SN transport iteration. We have developed solution techniques for the diffusion equations of FLBLD (Fully Lumped Bilinear Discontinuous), SCB (Simple Comer Balance) and UCB (Upstream Corner Balance) modified 4-step DSA in x-y geometry. Our first multi-level method includes a block Gauss-Seidel iteration for the discontinuous diffusion equation, uses the continuous diffusion equation derived from the asymptotic analysis, and avoids void cell calculation. We implemented this multi-level procedure and performed model problem calculations. The results showed that the FLBLD, SCB and UCB modified 4-step DSA schemes with this multi-level technique are unconditionally stable and rapidly convergent. We suggested a simplified multi-level technique for FLBLD, SCB and UCB modified 4-step DSA. This new procedure does not include iterations on the diffusion calculation or the residual calculation. Fourier analysis results showed that this new procedure was as rapidly convergent as conventional modified 4-step DSA. We developed new DSA procedures coupled with 1-CI (Cell Block Inversion) transport which can be easily parallelized. We showed that 1-CI based DSA schemes preceded by SI (Source Iteration) are efficient and rapidly convergent for LD (Linear Discontinuous) and LLD (Lumped Linear Discontinuous) in slab geometry and for BLD (Bilinear Discontinuous) and FLBLD in x-y geometry. For 1-CI based DSA without SI in slab geometry, the results showed that this procedure is very efficient and effective for all cases. We also showed that 1-CI based DSA in x-y geometry was not effective for thin mesh spacings, but is effective and rapidly convergent for intermediate and thick mesh spacings. We demonstrated that the diffusion equation discretized on a coarse mesh could be employed to accelerate the transport equation. Our results showed that coarse mesh DSA is unconditionally stable and is as rapidly convergent
International Nuclear Information System (INIS)
Watson, F.V.
1982-01-01
An adaptation of the alternate direction method for coarse mesh calculation, is presented. The algorithm is applicable to two-and three dimensional problems, the last being the more interesting one. (E.G.) [pt
Development of a coarse mesh code for the solution of two group static diffusion problems
International Nuclear Information System (INIS)
Barros, R.C. de.
1985-01-01
This new coarse mesh code designed for the solution of 2 and 3 dimensional static diffusion problems, is based on an alternating direction method which consists in the solution of one dimensional problem along each coordinate direction with leakage terms for the remaining directions estimated from previous interactions. Four versions of this code have been developed: AD21 - 2D - 1/4, AD21 - 2D - 4/4, AD21 - 3D - 1/4 and AD21 - 3D - 4/4; these versions have been designed for 2 and 3 dimensional problems with or without 1/4 symmetry. (Author) [pt
Splitting Method for Solving the Coarse-Mesh Discretized Low-Order Quasi-Diffusion Equations
International Nuclear Information System (INIS)
Hiruta, Hikaru; Anistratov, Dmitriy Y.; Adams, Marvin L.
2005-01-01
In this paper, the development is presented of a splitting method that can efficiently solve coarse-mesh discretized low-order quasi-diffusion (LOQD) equations. The LOQD problem can reproduce exactly the transport scalar flux and current. To solve the LOQD equations efficiently, a splitting method is proposed. The presented method splits the LOQD problem into two parts: (a) the D problem that captures a significant part of the transport solution in the central parts of assemblies and can be reduced to a diffusion-type equation and (b) the Q problem that accounts for the complicated behavior of the transport solution near assembly boundaries. Independent coarse-mesh discretizations are applied: the D problem equations are approximated by means of a finite element method, whereas the Q problem equations are discretized using a finite volume method. Numerical results demonstrate the efficiency of the methodology presented. This methodology can be used to modify existing diffusion codes for full-core calculations (which already solve a version of the D problem) to account for transport effects
International Nuclear Information System (INIS)
Miller, W.F. Jr.
1975-10-01
The coarse-mesh rebalance method, based on neutron conservation, is used in discrete ordinates neutron transport codes to accelerate convergence of the within-group scattering source. Though very powerful for this application, the method is ineffective in accelerating the iteration on the discrete-ordinates-to-spherical-harmonics fictitious sources used for ray-effect elimination. This is largely because this source makes a minimum contribution to the neutron balance equation. The traditional rebalance approach is derived in a variational framework and compared with new rebalance approaches tailored to be compatible with the fictitious source. The new approaches are compared numerically to determine their relative advantages. It is concluded that there is little incentive to use the new methods. (3 tables, 5 figures)
International Nuclear Information System (INIS)
Grill, S.F.; Jonsson, A.; Crump, M.W.
1983-01-01
The inclusion of 3-D effects in PWR analysis is necessary for accurate predictions of reactivity, power distributions, and reactivity coefficients. The ROCS/MC code system has been developed by Combustion Engineering to provide 3-D coarse mesh analysis (ROCS) with the capability to retrieve local information on flux, power and burnup (MC). A review of the finite difference representation of the MC diffusion equation, along with recent improvements to the ROCS/MC system are presented. These improvements include the implementation if fine mesh radial boundary conditions and internal calculation of coarse mesh boundary conditions, generalization of the imbedded calculation to account for the local neighboring environment, and the automation of ROCS/MC links to C-E's code system for in-core power distribution monitoring and core-follow analysis. The results of the ROCS/MC verification program are described and show good agreement with C-E's ROCS/PDQ based methodologies
International Nuclear Information System (INIS)
Shen, W.
2012-01-01
Recent assessment results indicate that the coarse-mesh finite-difference method (FDM) gives consistently smaller percent differences in channel powers than the fine-mesh FDM when compared to the reference MCNP solution for CANDU-type reactors. However, there is an impression that the fine-mesh FDM should always give more accurate results than the coarse-mesh FDM in theory. To answer the question if the better performance of the coarse-mesh FDM for CANDU-type reactors was just a coincidence (cancellation of errors) or caused by the use of heavy water or the use of lattice-homogenized cross sections for the cluster fuel geometry in the diffusion calculation, three benchmark problems were set up with three different fuel lattices: CANDU, HWR and PWR. These benchmark problems were then used to analyze the root cause of the better performance of the coarse-mesh FDM for CANDU-type reactors. The analyses confirm that the better performance of the coarse-mesh FDM for CANDU-type reactors is mainly caused by the use of lattice-homogenized cross sections for the sub-meshes of the cluster fuel geometry in the diffusion calculation. Based on the analyses, it is recommended to use 2 x 2 coarse-mesh FDM to analyze CANDU-type reactors when lattice-homogenized cross sections are used in the core analysis. (authors)
Energy Technology Data Exchange (ETDEWEB)
Shen, W. [Candu Energy Inc., 2285 Speakman Dr., Mississauga, ON L5B 1K (Canada)
2012-07-01
Recent assessment results indicate that the coarse-mesh finite-difference method (FDM) gives consistently smaller percent differences in channel powers than the fine-mesh FDM when compared to the reference MCNP solution for CANDU-type reactors. However, there is an impression that the fine-mesh FDM should always give more accurate results than the coarse-mesh FDM in theory. To answer the question if the better performance of the coarse-mesh FDM for CANDU-type reactors was just a coincidence (cancellation of errors) or caused by the use of heavy water or the use of lattice-homogenized cross sections for the cluster fuel geometry in the diffusion calculation, three benchmark problems were set up with three different fuel lattices: CANDU, HWR and PWR. These benchmark problems were then used to analyze the root cause of the better performance of the coarse-mesh FDM for CANDU-type reactors. The analyses confirm that the better performance of the coarse-mesh FDM for CANDU-type reactors is mainly caused by the use of lattice-homogenized cross sections for the sub-meshes of the cluster fuel geometry in the diffusion calculation. Based on the analyses, it is recommended to use 2 x 2 coarse-mesh FDM to analyze CANDU-type reactors when lattice-homogenized cross sections are used in the core analysis. (authors)
International Nuclear Information System (INIS)
Barros, R. C.; Filho, H. A.; Platt, G. M.; Oliveira, F. B. S.; Militao, D. S.
2009-01-01
Coarse-mesh numerical methods are very efficient in the sense that they generate accurate results in short computational time, as the number of floating point operations generally decrease, as a result of the reduced number of mesh points. On the other hand, they generate numerical solutions that do not give detailed information on the problem solution profile, as the grid points can be located considerably away from each other. In this paper we describe two analytical reconstruction schemes for the coarse-mesh solution generated by the spectral nodal method for neutral particle discrete ordinates (S N ) transport model in slab geometry. The first scheme we describe is based on the analytical reconstruction of the coarse-mesh solution within each discretization cell of the spatial grid set up on the slab. The second scheme is based on the angular reconstruction of the discrete ordinates solution between two contiguous ordinates of the angular quadrature set used in the S N model. Numerical results are given so we can illustrate the accuracy of the two reconstruction schemes, as described in this paper. (authors)
Development of a New core/reflector model for coarse-mesh nodal methods
International Nuclear Information System (INIS)
Pogosbekyan, Leonid; Cho, Jin Young; Kim, Young Il; Kim, Young Jin; Joo, Hyung Kuk; Chang, Moon Hee.
1997-10-01
This work presents two approaches for reflector simulation in coarse-mesh nodal methods. The first approach is called Interface Matrix Technique (IMT), which simulates the baffle as a banishingly thin layer having the property of reflection and transmission. We applied this technique within the frame of AFEN (Analytic Function Expansion Nodal) method, and developed the AFEN-IM (Interface Matrix) method. AFEN-IM method shows 1.24% and 0.42 % in maximum and RMS (Root Mean Square) assemblywise power error for ZION-1 benchmark problem. The second approach is L-shaped reflector homogenization method. This method is based on the integral response conservation along the L-shaped core-reflector interface. The reference reflector response is calculated from 2-dimensional spectral calculation and the response of the homogenized reflector is derived from the one-node 2-dimensional AFEN problem solution. This method shows 5 times better accuracy than the 1-dimensional homogenization technique in the assemblywise power. Also, the concept of shroud/reflector homogenization for hexagonal core have been developed. The 1-dimensional spectral calculation was used for the determination of 2 group cross sections. The essence of homogenization concept consists in the calculation of equivalent shroud width, which preserve albedo for the fast neutrons in 2-dimensional reflector. This method shows a relative error less than 0.42% in assemblywise power and a difference of 9x10 -5 in multiplication factor for full-core model. (author). 9 refs., 3 tabs., 28 figs
International Nuclear Information System (INIS)
Park, Beom Woo; Joo, Han Gyu
2015-01-01
Highlights: • The stiffness confinement method is combined with multigroup CMFD with SENM nodal kernel. • The systematic methods for determining the shape and amplitude frequencies are established. • Eigenvalue problems instead of fixed source problems are solved in the transient calculation. • It is demonstrated that much larger time step sizes can be used with the SCM–CMFD method. - Abstract: An improved Stiffness Confinement Method (SCM) is formulated within the framework of the coarse mesh finite difference (CMFD) formulation for efficient multigroup spatial kinetics calculation. The algorithm for searching for the amplitude frequency that makes the dynamic eigenvalue unity is developed in a systematic way along with the methods for determining the shape and precursor frequencies. A nodal calculation scheme is established within the CMFD framework to incorporate the cross section changes due to thermal feedback and dynamic frequency update. The conditional nodal update scheme is employed such that the transient calculation is performed mostly with the CMFD formulation and the CMFD parameters are conditionally updated by intermittent nodal calculations. A quadratic representation of amplitude frequency is introduced as another improvement. The performance of the improved SCM within the CMFD framework is assessed by comparing the solution accuracy and computing times for the NEACRP control rod ejection benchmark problems with those obtained with the Crank–Nicholson method with exponential transform (CNET). It is demonstrated that the improved SCM is beneficial for large time step size calculations with stability and accuracy enhancement
A simplified treatment of the boundary conditions of the k- ε model in coarse-mesh CFD-type codes
International Nuclear Information System (INIS)
Analytis, G.Th.; Andreani, M.
1999-01-01
In coarse-mesh, CFD-type codes such as the containment analysis code GOTHIC, one of the options that can be used for modelling of turbulence is the k - ε model. However, in contrast to most other CFD codes which are designed to perform detailed CFD calculations with a large number of spatial meshes, codes such as GOTHIC are primarily aimed at simplified calculation of transients in large spaces (e.g., reactor containments), and generally use coarse meshes. The solution of the two parabolic equations for the k - ε model requires the definition of boundary conditions at physical boundaries and this, in turn, requires very small spatial meshes near these boundaries. Hence, while in codes like CFX this is done in a rigorous and consistent manner, codes like GOTHIC adopt an indirect and heuristic approach, due to the fact that the spatial meshes are usually large. This can have adverse consequences during the calculation of a transient and in this work, we shall give some examples of this and outline a method by which this problem can be avoided. (author)
International Nuclear Information System (INIS)
Garcia-Herranz, Nuria; Cabellos, Oscar; Aragones, Jose M.; Ahnert, Carol
2003-01-01
In order to take into account in a more effective and accurate way the intranodal heterogeneities in coarse-mesh finite-difference (CMFD) methods, a new equivalent parameter generation methodology has been developed and tested. This methodology accounts for the dependence of the nodal homogeneized two-group cross sections and nodal coupling factors, with interface flux discontinuity (IFD) factors that account for heterogeneities on the flux-spectrum and burnup intranodal distributions as well as on neighbor effects.The methodology has been implemented in an analytic CMFD method, rigorously obtained for homogeneous nodes with transverse leakage and generalized now for heterogeneous nodes by including IFD heterogeneity factors. When intranodal mesh node heterogeneity vanishes, the heterogeneous solution tends to the analytic homogeneous nodal solution. On the other hand, when intranodal heterogeneity increases, a high accuracy is maintained since the linear and nonlinear feedbacks on equivalent parameters have been shown to be as a very effective way of accounting for heterogeneity effects in two-group multidimensional coarse-mesh diffusion calculations
A 3D coarse-mesh time dependent code for nuclear reactor kinetic calculations
International Nuclear Information System (INIS)
Montagnini, B.; Raffaelli, P.; Sumini, M.; Zardini, D.M.
1996-01-01
A course-mesh code for time-dependent multigroup neutron diffusion calculation based on a direct integration scheme for the time dependence and a low order nodal flux expansion approximation for the space variables has been implemented as a fast tool for transient analysis. (Author)
Energy Technology Data Exchange (ETDEWEB)
Pereira, Valmir; Martinez, Aquilino Senra; Silva, Fernando Carvalho da [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear
2002-07-01
The objective of this work is the obtention of the mathematical adjoint flux, having as its support the nodal expansion method (NEM) for coarse mesh problems. Since there are difficulties to evaluate this flux by using NEM. directly, a coarse mesh finite difference program was developed to obtain this adjoint flux. The coarse mesh finite difference formulation (DFMG) adopted uses results of the direct calculation (node average flux and node face averaged currents) obtained by NEM. These quantities (flux and currents) are used to obtain the correction factors which modify the classical finite differences formulation . Since the DFMG formulation is also capable of calculating the direct flux it was also tested to obtain this flux and it was verified that it was able to reproduce with good accuracy both the flux and the currents obtained via NEM. In this way, only matrix transposition is needed to calculate the mathematical adjoint flux. (author)
International Nuclear Information System (INIS)
Mayya, Y.S.; Mishra, Rosaline; Prajith, Rama; Sapra, B.K.; Kushwaha, H.S.
2010-01-01
Deposition-based 222 Rn and 220 Rn progeny sensors act as unique, passive tools for determining the long time-averaged progeny deposition fluxes in the environment. The use of these deposition sensors as progeny concentration monitors was demonstrated in typical indoor environments as conceptually superior alternatives to gas-based indirect monitoring methods. In the present work, the dependency of these deposition monitors on various environmental parameters is minimized by capping the deposition sensor with a suitable wire mesh. These wire-mesh capped deposition sensors measure the coarse fraction deposition flux, which is less dependent on the change in environmental parameters like ventilation rate and turbulence. The calibration of these wire-mesh capped coarse fraction progeny sensors was carried out by laboratory controlled experiments. These sensors were deployed both in indoor and in occupational environments having widely different ventilation rates. The obtained coarse fraction deposition velocities were fairly constant in these environments, which further confirmed that the signal on the wire-mesh capped sensors show the least dependency on the change in environmental parameters. This technique has the potential to serve as a passive particle sizer in the general context of nanoparticles using progeny species as surrogates. On the whole, there exists a strong case for developing a passive system that responds only to coarse fraction for providing alternative tools for dosimetry and environmental fine particle research. - Research highlights: → Wire-mesh capped deposition sensor measures the coarse fraction deposition flux → Coarse fraction deposition flux less dependent on environmental conditions → Wire-mesh capped deposition sensor as passive particle sizer
Performance Improvements for Coarse Mesh Finite Difference Acceleration L3:RTM.PRT.P13.02
International Nuclear Information System (INIS)
Collins, Benjamin S.; Hamilton, Steven P.; Stimpson, Shane; Yee, Ben; Larsen, Edward W.; Kochunas, Brendan
2016-01-01
The development of VERA-CS in recent years has focused on developing the capability to simulate multiple cycles of operating commercial nuclear power plants. Now that these capabilities have advanced to the point where it is being deployed to users, the focus is on improving the computational performance of various components in VERA-CS. In this work, the focus is on the Coarse Mesh Finite Difference (CMFD) solution in MPACT. CMFD serves multiple purposes in the 2D/1D solution methodology. First, it is a natural mechanism to tie together the radial MOC transport and the axial SP3 solution. Because the CMFD system solves the multigroup three-dimensional core in one system, it pulls together the global response of the system. In addition, the CMFD solution provides a framework to accelerate the convergence of the eigenvalue problem.
International Nuclear Information System (INIS)
Santos, Frederico P.; Xavier, Vinicius S.; Alves Filho, Hermes; Barros, Ricardo C.
2011-01-01
The scattering source iterative (SI) scheme is traditionally applied to converge fine-mesh numerical solutions to fixed-source discrete ordinates (S N ) neutron transport problems. The SI scheme is very simple to implement under a computational viewpoint. However, the SI scheme may show very slow convergence rate, mainly for diffusive media (low absorption) with several mean free paths in extent. In this work we describe an acceleration technique based on an improved initial guess for the scattering source distribution within the slab. In other words, we use as initial guess for the fine-mesh scattering source, the coarse-mesh solution of the neutron diffusion equation with special boundary conditions to account for the classical S N prescribed boundary conditions, including vacuum boundary conditions. Therefore, we first implement a spectral nodal method that generates coarse-mesh diffusion solution that is completely free from spatial truncation errors, then we reconstruct this coarse-mesh solution within each spatial cell of the discretization grid, to further yield the initial guess for the fine-mesh scattering source in the first S N transport sweep (μm > 0 and μm < 0, m = 1:N) across the spatial grid. We consider a number of numerical experiments to illustrate the efficiency of the offered diffusion synthetic acceleration (DSA) technique. (author)
International Nuclear Information System (INIS)
Santos, Frederico P.; Alves Filho, Hermes; Barros, Ricardo C.; Xavier, Vinicius S.
2011-01-01
The scattering source iterative (SI) scheme is traditionally applied to converge fine-mesh numerical solutions to fixed-source discrete ordinates (S N ) neutron transport problems. The SI scheme is very simple to implement under a computational viewpoint. However, the SI scheme may show very slow convergence rate, mainly for diffusive media (low absorption) with several mean free paths in extent. In this work we describe an acceleration technique based on an improved initial guess for the scattering source distribution within the slab. In other words, we use as initial guess for the fine-mesh scattering source, the coarse-mesh solution of the neutron diffusion equation with special boundary conditions to account for the classical S N prescribed boundary conditions, including vacuum boundary conditions. Therefore, we first implement a spectral nodal method that generates coarse-mesh diffusion solution that is completely free from spatial truncation errors, then we reconstruct this coarse-mesh solution within each spatial cell of the discretization grid, to further yield the initial guess for the fine-mesh scattering source in the first S N transport sweep (μm > 0 and μm < 0, m = 1:N) across the spatial grid. We consider a number of numerical experiments to illustrate the efficiency of the offered diffusion synthetic acceleration (DSA) technique. (author)
MUSIC: a mesh-unrestricted simulation code
International Nuclear Information System (INIS)
Bonalumi, R.A.; Rouben, B.; Dastur, A.R.; Dondale, C.S.; Li, H.Y.H.
1978-01-01
A general formalism to solve the G-group neutron diffusion equation is described. The G-group flux is represented by complementing an ''asymptotic'' mode with (G-1) ''transient'' modes. A particular reduction-to-one-group technique gives a high computational efficiency. MUSIC, a 2-group code using the above formalism, is presented. MUSIC is demonstrated on a fine-mesh calculation and on 2 coarse-mesh core calculations: a heavy-water reactor (HWR) problem and the 2-D lightwater reactor (LWR) IAEA benchmark. Comparison is made to finite-difference results
HEXNOD23, 2-D, 3-D Coarse Mesh Solution of Steady State Diffusion Equation in Hexagonal Geometry
International Nuclear Information System (INIS)
Grundmann, Ulrich
1986-01-01
1 - Description of program or function: Two- or three dimensional coarse mesh solution of steady state two group neutron diffusion equation in arrays of regular hexagons or hexagonal subassemblies. 2 - Method of solution: The neutron flux in a hexagonal node is expanded in a series of Bessel functions in the hexagonal plane. Polynomials up to the 4. order are used for the approximation of neutron flux in axial direction of three dimensional cases. Resulting relations between node averaged fluxes and mean partial currents of node faces in connection with the neutron balance of nodes are used to calculate the eigenvalue Keff, mean fluxes and mean powers of nodes. The iterations process is divided into inner and outer iterations. The iterations are accelerated by Ljusternik and Tschebyscheff extrapolation schemes. The power densities in the nodes and subassembly powers are computed for given reactor power in three dimensional cases. 30 degree reflectional, 60 and 120 degree rotational core symmetry and the whole core can be treated. 3 - Restrictions on the complexity of the problem: If the problem size designated by LIAR and LRAR exceeds 3000 and 50000 respectively, the lengths of the working array MIAR and MRAR in the main program can be increased. External sources are not permitted
International Nuclear Information System (INIS)
Andreani, Michele; Paladino, Domenico
2010-01-01
The recently concluded OECD SETH project included twenty-four experiments on basic flows and gas transport and mixing driven by jets and plumes in two, large, connected vessels of the PANDA facility. The experiments featured injection of saturated or superheated steam, or a mixture of steam and helium in one vessel and venting from the same vessel or from the connected one. These tests have been especially designed for providing an extensive data base for the assessment of three-dimensional codes, including CFD codes. In particular, one of the goals of the analytical activities associated with the experiments was to evaluate the detail of the model (mesh) necessary for capturing the various phenomena. This work reports an overview of the results obtained for these experimental data using the advanced containment code GOTHIC and relatively coarse meshes, which are coarser than the ones typically used for the simulation with commercial CFD codes, but are still representative of the models which are currently affordable for a full containment analysis. In general, the phenomena were correctly represented in the simulations with GOTHIC, and the agreement of the results with the data was in most cases pretty good, in some cases excellent. Only for a few tests (or particular phenomena occurring in some tests) the simulations showed noticeable discrepancies with the experimental data, which could be referred to either an insufficiently detailed mesh or to lack of specialized models for local effects.
Thermal-hydraulics verification of a coarse-mesh OpenFOAM-based solver for a Sodium Fast Reactor
Energy Technology Data Exchange (ETDEWEB)
Bonet López, M.
2015-07-01
Recently, in the Institute Swiss Paul Scherrer Institut, is has developed a platform Multiphysics, based in OpenFOAM, that is capable of performing an analysis multidimensional of a reactor nuclear. One of the main objectives of this project is to verify the part of the code responsible for the Thermo-hydraulic analysis of the reactor. To carry out simulations this part of the code uses the approximation of thick mesh based on the equations of a porous medium. Therefore, the other objective is demonstrate that this method is applicable to the analysis of a reactor nuclear fast of sodium, focusing is in his capacity of predict the transfer of heat between a subset and the space vacuum between subsets of the core of the reactor. (Author)
National Aeronautics and Space Administration — Unstructured HIRENASD mesh: - coarse size (5.7 million nodes, 14.4 million elements) - for node centered solvers - 01.06.2011 - caution: dimensions in mm
International Nuclear Information System (INIS)
Seregin, A.S.
2000-01-01
In the paper the formulae for perturbation theory functionals calculation are given and equations are based on improved coarse mesh discretization of diffusion problem in 3-dimensional geometry (Hex-Z). Expressions for the reactivity effect components and reactivity coefficients, written in the framework of the first order perturbation theory, are presented. On this basis the formulae for estimation of the sensitivity coefficients of different reactivity effects group cross-sections were derived. Expressions for the reactivity effect and its components obtained in the framework of the strict perturbation theory, are also presented in the paper. (author)
International Nuclear Information System (INIS)
Borresen, S.
1995-01-01
A simplified, finite-difference diffusion scheme for a three-dimensional calculation of the gross power distribution in the core of a boiling water reactor (BWR) is presented. Results obtained in a series of one- and two-dimensional test cases indicate that this method may be of sufficient accuracy and simplicity for implementation in BWR-simulator computer programs. Computer requirements are very modest; thus, only 3N memory locations are required for in-core treatment of the inner iteration in the solution of a problem with N mesh points. The mesh width may be chosen equal to the fuel assembly pitch. Input data are in the form of conventional 2-group diffusion parameters. It is concluded that the method presented has definite advantages in comparison with the nodal coupling method. (author)
Cleveland, Mathew A.
We investigate several aspects of the numerical solution of the radiative transfer equation in the context of coal combustion: the parallel efficiency of two commonly-used opacity models, the sensitivity of turbulent radiation interaction (TRI) effects to the presence of coal particulate, and an improvement of the order of temporal convergence using the coarse mesh finite difference (CMFD) method. There are four opacity models commonly employed to evaluate the radiative transfer equation in combustion applications; line-by-line (LBL), multigroup, band, and global. Most of these models have been rigorously evaluated for serial computations of a spectrum of problem types [1]. Studies of these models for parallel computations [2] are limited. We assessed the performance of the Spectral-Line-Based weighted sum of gray gasses (SLW) model, a global method related to K-distribution methods [1], and the LBL model. The LBL model directly interpolates opacity information from large data tables. The LBL model outperforms the SLW model in almost all cases, as suggested by Wang et al. [3]. The SLW model, however, shows superior parallel scaling performance and a decreased sensitivity to load imbalancing, suggesting that for some problems, global methods such as the SLW model, could outperform the LBL model. Turbulent radiation interaction (TRI) effects are associated with the differences in the time scales of the fluid dynamic equations and the radiative transfer equations. Solving on the fluid dynamic time step size produces large changes in the radiation field over the time step. We have modified the statistically homogeneous, non-premixed flame problem of Deshmukh et al. [4] to include coal-type particulate. The addition of low mass loadings of particulate minimally impacts the TRI effects. Observed differences in the TRI effects from variations in the packing fractions and Stokes numbers are difficult to analyze because of the significant effect of variations in problem
DEFF Research Database (Denmark)
2015-01-01
Mesh generation and visualization software based on the CGAL library. Folder content: drawmesh Visualize slices of the mesh (surface/volumetric) as wireframe on top of an image (3D). drawsurf Visualize surfaces of the mesh (surface/volumetric). img2mesh Convert isosurface in image to volumetric m...... mesh (medit format). img2off Convert isosurface in image to surface mesh (off format). off2mesh Convert surface mesh (off format) to volumetric mesh (medit format). reduce Crop and resize 3D and stacks of images. data Example data to test the library on...
Light Water Reactor (LWR) safety
International Nuclear Information System (INIS)
Sehgal, Bal Raj
2006-01-01
In this paper, a historical review of the developments in the safely of LWR power plants is presented. The paper reviews the developments prior to the TMI-2 accident, i.e. the concept of the defense in depth, the design basis, the large LOCA technical controversies and the LWR safety research programs. The TMI-2 accident, which became a turning point in the history of the development of nuclear power is described briefly. The Chernobyl accident, which terrified the world and almost completely curtailed the development of nuclear power is also described briefly. The great international effort of research in the LWR design-base and severe accidents, which was, respectively, conducted prior to and following the TMI-2 and Chernobyl accidents is described next. We conclude that with the knowledge gained and the improvements in plant organisation/management and in the training of the staff at the presently-installed nuclear power stations, the LWR plants have achieved very high standards of safety and performance. The Generation 3 + LWR power plants, next to be installed, may claim to have reached the goal of assuring the safety of the public to a very large extent. This review is based on the historical developments in LWR safety that occurred primarily in USA. however, they are valid for the rest of the Western World. This review can not do justice to the many many fine contributions that have been made over the last fifty years to the cause of LWR safety. We apologize if we have not mentioned them. We also apologize for not providing references to many of the fine investigations, which have contributed towards LWR safety earning the conclusions that we describe just above
RETRANS, Reactivity Transients in LWR
International Nuclear Information System (INIS)
Kamelander, G.
1989-01-01
1 - Description of program or function: RETRANS is appropriate to calculate power excursions in light water reactors initiated by reactivity insertions due to withdrawal of control elements. As in the code TWIGL, the neutron physics model is based on the time-dependent two-group neutron diffusion equations. The equation of state of the coolant is approximated by a table built into the code. RETRANS solves the heat conduction equation and calculates the heat transfer coefficient for representative fuel rods at each time-step. 2 - Method of solution: The time-dependent neutron diffusion equations are modified by an exponential transformation and solved by means of a finite difference method. There is an option accelerating the inner iterations of the difference scheme by a coarse-mesh re-balancing method. The heat balance equations of the thermo- hydraulic model are discretized and converted into a tri-diagonal system of linear equations which is solved recursively. 3 - Restrictions on the complexity of the problem: r-z-geometry, one- phase-flow
International Nuclear Information System (INIS)
Paratte, J.M.
1982-07-01
The LWR-Core behaviour project concerns the mathematical simulation of a light water reactor in normal operation (emergency situations excluded). Computational tools are assembled, i.e. programs and libraries of data. These computational tools can likewise be used in nuclear power applications, industry and control applications. The project is divided into three parts: the development and application of calculation methods for quantisation determination of LWR physics; investigation of the behaviour of nuclear fuels under radiation with special attention to higher burnup; simulation of the operating transients of nuclear power stations. (A.N.K.)
Predicting mesh density for adaptive modelling of the global atmosphere.
Weller, Hilary
2009-11-28
The shallow water equations are solved using a mesh of polygons on the sphere, which adapts infrequently to the predicted future solution. Infrequent mesh adaptation reduces the cost of adaptation and load-balancing and will thus allow for more accurate mapping on adaptation. We simulate the growth of a barotropically unstable jet adapting the mesh every 12 h. Using an adaptation criterion based largely on the gradient of the vorticity leads to a mesh with around 20 per cent of the cells of a uniform mesh that gives equivalent results. This is a similar proportion to previous studies of the same test case with mesh adaptation every 1-20 min. The prediction of the mesh density involves solving the shallow water equations on a coarse mesh in advance of the locally refined mesh in order to estimate where features requiring higher resolution will grow, decay or move to. The adaptation criterion consists of two parts: that resolved on the coarse mesh, and that which is not resolved and so is passively advected on the coarse mesh. This combination leads to a balance between resolving features controlled by the large-scale dynamics and maintaining fine-scale features.
ERDA LWR plant technology program: role of government/industry in improving LWR performance
International Nuclear Information System (INIS)
1975-01-01
Information is presented under the following chapter headings: executive summary; LWR plant outages; LWR plant construction delays and cancellations; programs addressing plant outages, construction delays, and cancellations; need for additional programs to remedy continuing problems; criteria for government role in LWR commercialization; and the proposed government program
Isotropic 2D quadrangle meshing with size and orientation control
Pellenard, Bertrand; Alliez, Pierre; Morvan, Jean-Marie
2011-01-01
We propose an approach for automatically generating isotropic 2D quadrangle meshes from arbitrary domains with a fine control over sizing and orientation of the elements. At the heart of our algorithm is an optimization procedure that, from a coarse
International Nuclear Information System (INIS)
Zheng Hualing.
1986-01-01
This article, from viewpoints of technical feasibility, safety evaluation and socioeconomic benefit-risk analysis, introduces and comments on history and status of recycling U and Pu in LWR, dealing with reactor, reprocessing, conversion and fuel element fabrication et al. Author has analysed LWR fuel cycle strategies in China and made a proposal
Mesh requirements for neutron transport calculations
International Nuclear Information System (INIS)
Askew, J.R.
1967-07-01
Fine-structure calculations are reported for a cylindrical natural uranium-graphite cell using different solution methods (discrete ordinate and collision probability codes) and varying the spatial mesh. It is suggested that of formulations assuming the source constant in a mesh interval the differential approach is generally to be preferred. Due to cancellation between approximations made in the derivation of the finite difference equations and the errors in neglecting source variation, the discrete ordinate code gave a more accurate estimate of fine structure for a given mesh even for unusually coarse representations. (author)
Nonlinear Multigrid solver exploiting AMGe Coarse Spaces with Approximation Properties
DEFF Research Database (Denmark)
Christensen, Max la Cour; Villa, Umberto; Engsig-Karup, Allan Peter
The paper introduces a nonlinear multigrid solver for mixed finite element discretizations based on the Full Approximation Scheme (FAS) and element-based Algebraic Multigrid (AMGe). The main motivation to use FAS for unstructured problems is the guaranteed approximation property of the AMGe coarse...... properties of the coarse spaces. With coarse spaces with approximation properties, our FAS approach on unstructured meshes has the ability to be as powerful/successful as FAS on geometrically refined meshes. For comparison, Newton’s method and Picard iterations with an inner state-of-the-art linear solver...... are compared to FAS on a nonlinear saddle point problem with applications to porous media flow. It is demonstrated that FAS is faster than Newton’s method and Picard iterations for the experiments considered here. Due to the guaranteed approximation properties of our AMGe, the coarse spaces are very accurate...
International Nuclear Information System (INIS)
Zbytovsky, A.; Lehmann, M.; Vyskocil, V.; Vacek, J.; Krysl, V.
1980-01-01
Computation of nuclear power reactors of the WWER-1000 type is described as are computer programs used by Skoda Works for the solution of neutron problems. The programs are analyzed for applicability in the unified program system of the CMEA countries which will be used in the preparation of safety reports, the evaluation of safety hazards, the design of fuel charges, economical studies etc. A detailed description is also presented of multigroup transport calculations and of the preparation of input data for macrocalculations of the heterogeneous lattices of LWR's. (author)
NONLINEAR MULTIGRID SOLVER EXPLOITING AMGe COARSE SPACES WITH APPROXIMATION PROPERTIES
Energy Technology Data Exchange (ETDEWEB)
Christensen, Max La Cour [Technical Univ. of Denmark, Lyngby (Denmark); Villa, Umberto E. [Univ. of Texas, Austin, TX (United States); Engsig-Karup, Allan P. [Technical Univ. of Denmark, Lyngby (Denmark); Vassilevski, Panayot S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)
2016-01-22
The paper introduces a nonlinear multigrid solver for mixed nite element discretizations based on the Full Approximation Scheme (FAS) and element-based Algebraic Multigrid (AMGe). The main motivation to use FAS for unstruc- tured problems is the guaranteed approximation property of the AMGe coarse spaces that were developed recently at Lawrence Livermore National Laboratory. These give the ability to derive stable and accurate coarse nonlinear discretization problems. The previous attempts (including ones with the original AMGe method, [5, 11]), were less successful due to lack of such good approximation properties of the coarse spaces. With coarse spaces with approximation properties, our FAS approach on un- structured meshes should be as powerful/successful as FAS on geometrically re ned meshes. For comparison, Newton's method and Picard iterations with an inner state-of-the-art linear solver is compared to FAS on a nonlinear saddle point problem with applications to porous media ow. It is demonstrated that FAS is faster than Newton's method and Picard iterations for the experiments considered here. Due to the guaranteed approximation properties of our AMGe, the coarse spaces are very accurate, providing a solver with the potential for mesh-independent convergence on general unstructured meshes.
Urogynecologic Surgical Mesh Implants
... procedures performed to treat pelvic floor disorders with surgical mesh: Transvaginal mesh to treat POP Transabdominal mesh to treat ... address safety risks Final Order for Reclassification of Surgical Mesh for Transvaginal Pelvic Organ Prolapse Repair Final Order for Effective ...
Roe, John
2003-01-01
Coarse geometry is the study of spaces (particularly metric spaces) from a 'large scale' point of view, so that two spaces that look the same from a great distance are actually equivalent. This point of view is effective because it is often true that the relevant geometric properties of metric spaces are determined by their coarse geometry. Two examples of important uses of coarse geometry are Gromov's beautiful notion of a hyperbolic group and Mostow's proof of his famous rigidity theorem. The first few chapters of the book provide a general perspective on coarse structures. Even when only metric coarse structures are in view, the abstract framework brings the same simplification as does the passage from epsilons and deltas to open sets when speaking of continuity. The middle section reviews notions of negative curvature and rigidity. Modern interest in large scale geometry derives in large part from Mostow's rigidity theorem and from Gromov's subsequent 'large scale' rendition of the crucial properties of n...
Nonlinear multigrid solvers exploiting AMGe coarse spaces with approximation properties
DEFF Research Database (Denmark)
Christensen, Max la Cour; Vassilevski, Panayot S.; Villa, Umberto
2017-01-01
discretizations on general unstructured grids for a large class of nonlinear partial differential equations, including saddle point problems. The approximation properties of the coarse spaces ensure that our FAS approach for general unstructured meshes leads to optimal mesh-independent convergence rates similar...... to those achieved by geometric FAS on a nested hierarchy of refined meshes. In the numerical results, Newton’s method and Picard iterations with state-of-the-art inner linear solvers are compared to our FAS algorithm for the solution of a nonlinear saddle point problem arising from porous media flow...
Outline of Swedish activities on LWR fuel
Energy Technology Data Exchange (ETDEWEB)
Grounes, M [Studsvik Nuclear, Nykoeping (Sweden); Roennberg, G [OKG AB (Sweden)
1997-12-01
The presentation outlines the Swedish activities on LWR fuel and considers the following issues: electricity production; performance of operating nuclear power plants; nuclear fuel cycle and waste management; research and development in nuclear field. 4 refs, 4 tabs.
Grid adaptation using chimera composite overlapping meshes
Kao, Kai-Hsiung; Liou, Meng-Sing; Chow, Chuen-Yen
1994-01-01
The objective of this paper is to perform grid adaptation using composite overlapping meshes in regions of large gradient to accurately capture the salient features during computation. The chimera grid scheme, a multiple overset mesh technique, is used in combination with a Navier-Stokes solver. The numerical solution is first converged to a steady state based on an initial coarse mesh. Solution-adaptive enhancement is then performed by using a secondary fine grid system which oversets on top of the base grid in the high-gradient region, but without requiring the mesh boundaries to join in any special way. Communications through boundary interfaces between those separated grids are carried out using trilinear interpolation. Application to the Euler equations for shock reflections and to shock wave/boundary layer interaction problem are tested. With the present method, the salient features are well-resolved.
Grid adaption using Chimera composite overlapping meshes
Kao, Kai-Hsiung; Liou, Meng-Sing; Chow, Chuen-Yen
1993-01-01
The objective of this paper is to perform grid adaptation using composite over-lapping meshes in regions of large gradient to capture the salient features accurately during computation. The Chimera grid scheme, a multiple overset mesh technique, is used in combination with a Navier-Stokes solver. The numerical solution is first converged to a steady state based on an initial coarse mesh. Solution-adaptive enhancement is then performed by using a secondary fine grid system which oversets on top of the base grid in the high-gradient region, but without requiring the mesh boundaries to join in any special way. Communications through boundary interfaces between those separated grids are carried out using tri-linear interpolation. Applications to the Euler equations for shock reflections and to a shock wave/boundary layer interaction problem are tested. With the present method, the salient features are well resolved.
Energy Technology Data Exchange (ETDEWEB)
Schöberl, Markus, E-mail: m.schoeberl@tum.de [Continuum Mechanics Group, Technical University of Munich, Boltzmannstraße 15, 85748 Garching (Germany); Zabaras, Nicholas [Institute for Advanced Study, Technical University of Munich, Lichtenbergstraße 2a, 85748 Garching (Germany); Department of Aerospace and Mechanical Engineering, University of Notre Dame, 365 Fitzpatrick Hall, Notre Dame, IN 46556 (United States); Koutsourelakis, Phaedon-Stelios [Continuum Mechanics Group, Technical University of Munich, Boltzmannstraße 15, 85748 Garching (Germany)
2017-03-15
We propose a data-driven, coarse-graining formulation in the context of equilibrium statistical mechanics. In contrast to existing techniques which are based on a fine-to-coarse map, we adopt the opposite strategy by prescribing a probabilistic coarse-to-fine map. This corresponds to a directed probabilistic model where the coarse variables play the role of latent generators of the fine scale (all-atom) data. From an information-theoretic perspective, the framework proposed provides an improvement upon the relative entropy method and is capable of quantifying the uncertainty due to the information loss that unavoidably takes place during the coarse-graining process. Furthermore, it can be readily extended to a fully Bayesian model where various sources of uncertainties are reflected in the posterior of the model parameters. The latter can be used to produce not only point estimates of fine-scale reconstructions or macroscopic observables, but more importantly, predictive posterior distributions on these quantities. Predictive posterior distributions reflect the confidence of the model as a function of the amount of data and the level of coarse-graining. The issues of model complexity and model selection are seamlessly addressed by employing a hierarchical prior that favors the discovery of sparse solutions, revealing the most prominent features in the coarse-grained model. A flexible and parallelizable Monte Carlo – Expectation–Maximization (MC-EM) scheme is proposed for carrying out inference and learning tasks. A comparative assessment of the proposed methodology is presented for a lattice spin system and the SPC/E water model.
Fatigue management considering LWR coolant environments
International Nuclear Information System (INIS)
Park, Heung Bae; Jin, Tae eun
2000-01-01
Design fatigue curve for structural material in the ASME Boiler and Pressure Vessel Code do not explicitly address the effects of reactor coolant environments on fatigue life. Environmentally assisted cracking (EAC) of low-alloy steels in light water reactor (LWR) coolant environments has been a concern ever since the early 1970's. And, recent fatigue test data indicate a significant decrease in fatigue lives of carbon steels, low-alloy steels and austenitic stainless steels in LWR coolant environments. For these reasons, fatigue of major components has been identified as a technical issue remaining to be resolved for life management and license renewal of nuclear power plants. In the present paper, results of recent investigations by many organizations are reviewed to provide technical justification to support the development of utility approach regarding the management of fatigue considering LWR coolant environments for the purpose of life management and license renewal of nuclear power plants. (author)
Generalized coarse-grained Becker-Doering equations
International Nuclear Information System (INIS)
Bolton, Colin D; Wattis, Jonathan A D
2003-01-01
We present and apply a generalized coarse-graining method of reducing the Becker-Doering model; originally formulated to describe the stepwise aggregation and fragmentation of clusters during nucleation. Previous formulations of the coarse-graining procedure have allowed a temporal rescaling of the coarse-grained reaction rates; this is generalized to allow the rescaling to depend on cluster size. The form of this factor is derived for general reaction rates and general mesh function so that the steady-state solution is preserved; in the case of an even mesh function the kinetics can also be accurately reproduced. With a size-dependent mesh function the equilibrium solution and the form of convergence to this state are matched for a specific example. Finally we consider reaction rates relevant to the classical nucleation theory of spherical cluster growth, and numerically compare solutions of the full system to the generalized coarse-grained system in both constant monomer and constant mass formulations, demonstrating the accuracy of the method
HFR irradiation testing of light water reactor (LWR) fuel
International Nuclear Information System (INIS)
Markgraf, J.F.W.
1985-01-01
For the materials testing reactor HFR some characteristic information with emphasis on LWR fuel rod testing capabilities and hot cell investigation is presented. Additionally a summary of LWR fuel irradiation programmes performed and forthcoming programmes are described. Project management information and a list of publications pertaining to LWR fuel rod test programmes is given
Texturing of continuous LOD meshes with the hierarchical texture atlas
Birkholz, Hermann
2006-02-01
For the rendering of detailed virtual environments, trade-offs have to be made between image quality and rendering time. An immersive experience of virtual reality always demands high frame-rates with the best reachable image qual-ity. Continuous Level of Detail (cLoD) triangle-meshes provide an continuous spectrum of detail for a triangle mesh that can be used to create view-dependent approximations of the environment in real-time. This enables the rendering with a constant number of triangles and thus with constant frame-rates. Normally the construction of such cLoD mesh representations leads to the loss of all texture information of the original mesh. To overcome this problem, a parameter domain can be created, in order to map the surface properties (colour, texture, normal) to it. This parameter domain can be used to map the surface properties back to arbitrary approximations of the original mesh. The parameter domain is often a simplified version of the mesh to be parameterised. This limits the reachable simplification to the domain mesh which has to map the surface of the original mesh with the least possible stretch. In this paper, a hierarchical domain mesh is presented, that scales between very coarse domain meshes and good property-mapping.
Challenges in coupled thermal-hydraulics and neutronics simulations for LWR safety analysis
International Nuclear Information System (INIS)
Ivanov, Kostadin; Avramova, Maria
2007-01-01
The simulation of nuclear power plant accident conditions requires three-dimensional (3D) modeling of the reactor core to ensure a realistic description of physical phenomena. The operational flexibility of Light Water Reactor (LWR) plants can be improved by utilizing accurate 3D coupled neutronics/thermal-hydraulics calculations for safety margins evaluations. There are certain requirements to the coupling of thermal-hydraulic system codes and neutron-kinetics codes that ought to be considered. The objective of these requirements is to provide accurate solutions in a reasonable amount of CPU time in coupled simulations of detailed operational transient and accident scenarios. These requirements are met by the development and implementation of six basic components of the coupling methodologies: ways of coupling (internal or external coupling); coupling approach (integration algorithm or parallel processing); spatial mesh overlays; coupled time-step algorithms; coupling numerics (explicit, semi-implicit and implicit schemes); and coupled convergence schemes. These principles of the coupled simulations are discussed in details along with the scientific issues associated with the development of appropriate neutron cross-section libraries for coupled code transient modeling. The current trends in LWR nuclear power generation and regulation as well as the design of next generation LWR reactor concepts along with the continuing computer technology progress stimulate further development of these coupled code systems. These efforts have been focused towards extending the analysis capabilities as well as refining the scale and level of detail of the coupling. This article analyses the coupled phenomena and modeling challenges on both global (assembly-wise) and local (pin-wise) levels. The issues related to the consistent qualification of coupled code systems as well as their application to different types of LWR transients are presented. Finally, the advances in numerical
Mesh Excision: Is Total Mesh Excision Necessary?
Wolff, Gillian F; Winters, J Christian; Krlin, Ryan M
2016-04-01
Nearly 29% of women will undergo a secondary, repeat operation for pelvic organ prolapse (POP) symptom recurrence following a primary repair, as reported by Abbott et al. (Am J Obstet Gynecol 210:163.e1-163.e1, 2014). In efforts to decrease the rates of failure, graft materials have been utilized to augment transvaginal repairs. Following the success of using polypropylene mesh (PPM) for stress urinary incontinence (SUI), the use of PPM in the transvaginal repair of POP increased. However, in recent years, significant concerns have been raised about the safety of PPM mesh. Complications, some specific to mesh, such as exposures, erosion, dyspareunia, and pelvic pain, have been reported with increased frequency. In the current literature, there is not substantive evidence to suggest that PPM has intrinsic properties that warrant total mesh removal in the absence of complications. There are a number of complications that can occur after transvaginal mesh placement that do warrant surgical intervention after failure of conservative therapy. In aggregate, there are no high-quality controlled studies that clearly demonstrate that total mesh removal is consistently more likely to achieve pain reduction. In the cases of obstruction and erosion, it seems clear that definitive removal of the offending mesh is associated with resolution of symptoms in the majority of cases and reasonable practice. There are a number of complications that can occur with removal of mesh, and patients should be informed of this as they formulate a choice of treatment. We will review these considerations as we examine the clinical question of whether total versus partial removal of mesh is necessary for the resolution of complications following transvaginal mesh placement.
Isotropic 2D quadrangle meshing with size and orientation control
Pellenard, Bertrand
2011-12-01
We propose an approach for automatically generating isotropic 2D quadrangle meshes from arbitrary domains with a fine control over sizing and orientation of the elements. At the heart of our algorithm is an optimization procedure that, from a coarse initial tiling of the 2D domain, enforces each of the desirable mesh quality criteria (size, shape, orientation, degree, regularity) one at a time, in an order designed not to undo previous enhancements. Our experiments demonstrate how well our resulting quadrangle meshes conform to a wide range of input sizing and orientation fields.
Development of training simulator for LWR
International Nuclear Information System (INIS)
Sureshbabu, R.M.
2009-01-01
A full-scope training simulator was developed for a light water reactor (LWR). This paper describes how the development evolved from a desktop simulator to the full-scope training simulator. It also describes the architecture and features of the simulator including the large number of failures that it simulates. The paper also explains the three-level validation tests that were used to qualify the training simulator. (author)
'CANDLE' burnup regime after LWR regime
International Nuclear Information System (INIS)
Sekimoto, Hiroshi; Nagata, Akito
2008-01-01
CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) burnup strategy can derive many merits. From safety point of view, the change of excess reactivity along burnup is theoretically zero, and the core characteristics, such as power feedback coefficients and power peaking factor, are not changed along burnup. Application of this burnup strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replacing fuels. About 40% of natural or depleted uranium undergoes fission without the conventional reprocessing and enrichment. If the LWR produced energy of X Joules, the CANDLE reactor can produce about 50X Joules from the depleted uranium left at the enrichment facility for the LWR fuel. If we can say LWRs have produced energy sufficient for full 20 years, we can produce the energy for 1000 years by using the CANDLE reactors with depleted uranium. We need not mine any uranium ore, and do not need reprocessing facility. The burnup of spent fuel becomes 10 times. Therefore, the spent fuel amount per produced energy is also reduced to one-tenth. The details of the scenario of CANDLE burnup regime after LWR regime will be presented at the symposium. (author)
Modeling the economic consequences of LWR accidents
International Nuclear Information System (INIS)
Burke, R.P.; Aldrich, D.C.; Rasmussen, N.C.
1984-01-01
Models to be used for analyses of economic risks from events which may occur during LWR plant operation are developed in this study. The models include capabilities to estimate both onsite and offsite costs of LWR events ranging from routine plant outages to severe core-melt accidents resulting in large releases of radioactive material to the environment. The models can be used by both the nuclear power industry and regulatory agencies in cost-benefit analyses for decisionmaking purposes. The newly developed economic consequence models are applied in an example to estimate the economic risks from operation of the Surry Unit 2 plant. The analyses indicate that economic risks from US LWR operation, in contrast to public health risks, are dominated by relatively high-frequency forced outage events. Even for severe (e.g., core-melt) accidents, expected offsite costs are less than expected onsite costs for the Surry site. The implications of these conclusions for nuclear power plant operation and regulation are discussed
Technical report on LWR design decision methodology. Phase I
International Nuclear Information System (INIS)
1980-03-01
Energy Incorporated (EI) was selected by Sandia Laboratories to develop and test on LWR design decision methodology. Contract Number 42-4229 provided funding for Phase I of this work. This technical report on LWR design decision methodology documents the activities performed under that contract. Phase I was a short-term effort to thoroughly review the curret LWR design decision process to assure complete understanding of current practices and to establish a well defined interface for development of initial quantitative design guidelines
Status of LWR fuel design and future usage of JENDL
International Nuclear Information System (INIS)
Ito, Takuya
2008-01-01
For all conventional LWR fuel design codes of LWR fuel manufactures in Japan, the cross section library are based on the ENDF/B. Recently we can see several movements for the utilization of JENDL library for the LWR fuel design. The latest version of NEUPHYS cross section library is based on the JENDL-3.2. To accelerate this movement of JENDL utilization in LWR fuel design, it is necessary to prepare a high quality JENDL document, systematic validation of JENDL and to appeal them abroad effectively. (author)
Implementation of static generalized perturbation theory for LWR design applications
International Nuclear Information System (INIS)
Byron, R.F.; White, J.R.
1987-01-01
A generalized perturbation theory (GPT) formulation is developed for application to light water reactor (LWR) design. The extensions made to standard generalized perturbation theory are the treatment of thermal-hydraulic and fission product poisoning feedbacks, and criticality reset. This formulation has been implemented into a standard LWR design code. The method is verified by comparing direct calculations with GPT calculations. Data are presented showing that feedback effects need to be considered when using GPT for LWR problems. Some specific potential applications of this theory to the field of LWR design are discussed
Recycled Coarse Aggregate Produced by Pulsed Discharge in Water
Namihira, Takao; Shigeishi, Mitsuhiro; Nakashima, Kazuyuki; Murakami, Akira; Kuroki, Kaori; Kiyan, Tsuyoshi; Tomoda, Yuichi; Sakugawa, Takashi; Katsuki, Sunao; Akiyama, Hidenori; Ohtsu, Masayasu
In Japan, the recycling ratio of concrete scraps has been kept over 98 % after the Law for the Recycling of Construction Materials was enforced in 2000. In the present, most of concrete scraps were recycled as the Lower Subbase Course Material. On the other hand, it is predicted to be difficult to keep this higher recycling ratio in the near future because concrete scraps increase rapidly and would reach to over 3 times of present situation in 2010. In addition, the demand of concrete scraps as the Lower Subbase Course Material has been decreased. Therefore, new way to reuse concrete scraps must be developed. Concrete scraps normally consist of 70 % of coarse aggregate, 19 % of water and 11 % of cement. To obtain the higher recycling ratio, the higher recycling ratio of coarse aggregate is desired. In this paper, a new method for recycling coarse aggregate from concrete scraps has been developed and demonstrated. The system includes a Marx generator and a point to hemisphere mesh electrode immersed in water. In the demonstration, the test piece of concrete scrap was located between the electrodes and was treated by the pulsed discharge. After discharge treatment of test piece, the recycling coarse aggregates were evaluated under JIS and TS and had enough quality for utilization as the coarse aggregate.
Creep damage in zircaloy-4 at LWR temperatures
International Nuclear Information System (INIS)
Keusseyan, R.L.; Hu, C.P.; Li, C.Y.
1978-08-01
The observation of creep damage in the form of grain boundary cavitation in Zircaloy-4 in the temperature range of interest to Light Water Reactor (LWR) applications is reported. The observed damage is shown to reduce the ductility of Zircaloy-4 in a tensile test at LWR temperatures
Convergence study of global meshing on enamel-cement-bracket finite element model
Samshuri, S. F.; Daud, R.; Rojan, M. A.; Basaruddin, K. S.; Abdullah, A. B.; Ariffin, A. K.
2017-09-01
This paper presents on meshing convergence analysis of finite element (FE) model to simulate enamel-cement-bracket fracture. Three different materials used in this study involving interface fracture are concerned. Complex behavior ofinterface fracture due to stress concentration is the reason to have a well-constructed meshing strategy. In FE analysis, meshing size is a critical factor that influenced the accuracy and computational time of analysis. The convergence study meshing scheme involving critical area (CA) and non-critical area (NCA) to ensure an optimum meshing sizes are acquired for this FE model. For NCA meshing, the area of interest are at the back of enamel, bracket ligature groove and bracket wing. For CA meshing, area of interest are enamel area close to cement layer, the cement layer and bracket base. The value of constant NCA meshing tested are meshing size 1 and 0.4. The value constant CA meshing tested are 0.4 and 0.1. Manipulative variables are randomly selected and must abide the rule of NCA must be higher than CA. This study employed first principle stresses due to brittle failure nature of the materials used. Best meshing size are selected according to convergence error analysis. Results show that, constant CA are more stable compare to constant NCA meshing. Then, 0.05 constant CA meshing are tested to test the accuracy of smaller meshing. However, unpromising result obtained as the errors are increasing. Thus, constant CA 0.1 with NCA mesh of 0.15 until 0.3 are the most stable meshing as the error in this region are lowest. Convergence test was conducted on three selected coarse, medium and fine meshes at the range of NCA mesh of 0.15 until 3 and CA mesh area stay constant at 0.1. The result shows that, at coarse mesh 0.3, the error are 0.0003% compare to 3% acceptable error. Hence, the global meshing are converge as the meshing size at CA 0.1 and NCA 0.15 for this model.
Is it the end of history for LWR safety?
International Nuclear Information System (INIS)
Sehgal, Bal Raj
2004-01-01
In this essay a parallel is drawn between the struggle for recognition, which is argued by Fukuyama as the 'motor' of human history and that waged by the LWR safety for the public to recognize the LWR plants as a source of safe nuclear power. The end of history for the ''human struggle for recognition'' as the capitalistic liberal democracy is equated with the ''end of history'' for the LWR safety to provide assurance to the public of termination of a severe accident it ever would occur. It is suggested that we are near ''the end of history'' of the LWR safety for the new-design LWR plants but fall short for the presently-installed plants. The essay bases these suggestions on an examination of the history of nuclear power development in U.S.A., but also considering the more recent regulatory and public acceptance developments in Europe and the rest of the World. (author)
... knitted mesh or non-knitted sheet forms. The synthetic materials used can be absorbable, non-absorbable or a combination of absorbable and non-absorbable materials. Animal-derived mesh are made of animal tissue, such as intestine or skin, that has been processed and disinfected to be ...
Hybrid continuum-coarse-grained modeling of erythrocytes
Lyu, Jinming; Chen, Paul G.; Boedec, Gwenn; Leonetti, Marc; Jaeger, Marc
2018-06-01
The red blood cell (RBC) membrane is a composite structure, consisting of a phospholipid bilayer and an underlying membrane-associated cytoskeleton. Both continuum and particle-based coarse-grained RBC models make use of a set of vertices connected by edges to represent the RBC membrane, which can be seen as a triangular surface mesh for the former and a spring network for the latter. Here, we present a modeling approach combining an existing continuum vesicle model with a coarse-grained model for the cytoskeleton. Compared to other two-component approaches, our method relies on only one mesh, representing the cytoskeleton, whose velocity in the tangential direction of the membrane may be different from that of the lipid bilayer. The finitely extensible nonlinear elastic (FENE) spring force law in combination with a repulsive force defined as a power function (POW), called FENE-POW, is used to describe the elastic properties of the RBC membrane. The mechanical interaction between the lipid bilayer and the cytoskeleton is explicitly computed and incorporated into the vesicle model. Our model includes the fundamental mechanical properties of the RBC membrane, namely fluidity and bending rigidity of the lipid bilayer, and shear elasticity of the cytoskeleton while maintaining surface-area and volume conservation constraint. We present three simulation examples to demonstrate the effectiveness of this hybrid continuum-coarse-grained model for the study of RBCs in fluid flows.
Milestoning with coarse memory
Hawk, Alexander T.
2013-04-01
Milestoning is a method used to calculate the kinetics of molecular processes occurring on timescales inaccessible to traditional molecular dynamics (MD) simulations. In the method, the phase space of the system is partitioned by milestones (hypersurfaces), trajectories are initialized on each milestone, and short MD simulations are performed to calculate transitions between neighboring milestones. Long trajectories of the system are then reconstructed with a semi-Markov process from the observed statistics of transition. The procedure is typically justified by the assumption that trajectories lose memory between crossing successive milestones. Here we present Milestoning with Coarse Memory (MCM), a generalization of Milestoning that relaxes the memory loss assumption of conventional Milestoning. In the method, milestones are defined and sample transitions are calculated in the standard Milestoning way. Then, after it is clear where trajectories sample milestones, the milestones are broken up into distinct neighborhoods (clusters), and each sample transition is associated with two clusters: the cluster containing the coordinates the trajectory was initialized in, and the cluster (on the terminal milestone) containing trajectory's final coordinates. Long trajectories of the system are then reconstructed with a semi-Markov process in an extended state space built from milestone and cluster indices. To test the method, we apply it to a process that is particularly ill suited for Milestoning: the dynamics of a polymer confined to a narrow cylinder. We show that Milestoning calculations of both the mean first passage time and the mean transit time of reversal—which occurs when the end-to-end vector reverses direction—are significantly improved when MCM is applied. Finally, we note the overhead of performing MCM on top of conventional Milestoning is negligible.
Recycle of LWR actinides to an IFR
International Nuclear Information System (INIS)
Pierce, R.D.; Ackerman, J.P.; Johnson, G.K.; Mulcahey, T.P.; Poa, D.S.
1991-01-01
Large quantities of actinide elements are present in irradiated light water reactor fuel that is stored throughout the world. Because of the high fission to capture ratio for the transuranium (TRU) elements with the high energy neutrons in metal-fueled integral fast reactors (IFR), that reactor can consume these elements effectively. The stored fuel may represent valuable resource for the expanding application of fast power reactors. In addition, the removal of TRU elements from spent LWR fuel has the potential for increasing the capacity of high level waste facilities by reducing the heat load and may increase the margin of safety in meeting licensing requirement. Argonne National Laboratory is developing a pyrochemical process, which is compatible with the IFR fuel cycle for the recovery of TRU elements from LWR fuel. The proposed product is a metallic actinide ingot, which can be introduced into the electrorefining step of the IFR process. Two pyrochemical processes, that is, salt transport process and blanket processing study, are discussed in this paper. Also the experimental studies are reported. (K.I.)
Vectorization of LWR transient analysis code RELAP5/MOD1 and its effect
International Nuclear Information System (INIS)
Ishiguro, Misako; Harada, Hiroo; Shinozawa, Naohisa; Naraoka, Ken-itsu
1985-03-01
The RELAP5/MOD1 is a large thermal-hydraulic code to analyze LWR LOCA and non-LOCA transients. The code originally was designed for use on a CDC Cyber-176. This report documents vectorization of the RELAP5/MOD1 code conducted for the purpose of efficient use of VP-100 (peak speed 250 MFLOPS, clock period 7.5 ns) at the JAERI. The code was vectorized using the junction and volume level parallelisms in the hydrodynamic calculations, and the heat-structure and heat-mesh level in the heat conduction calculations. The vectorized version runs as much as 2.4 to 2.8 times faster than the original scalar version, while the speedup ratio is dependent on the number of spactial cells included in the problem. (author)
LWR Spent Fuel Management for the Smooth Deployment of FBR
International Nuclear Information System (INIS)
Fukasawa, T.; Yamashita, J.; Hoshino, K.; Sasahira, A.; Inoue, T.; Minato, K.; Sato, S.
2015-01-01
Fast breeder reactors (FBR) and FBR fuel cycle are indispensable to prevent the global warming and to secure the long-term energy supply. Commercial FBR expects to be deployed from around 2050 until around 2110 in Japan by the replacement of light water reactors (LWR) after their 60 years life. The FBR deployment needs Pu (MOX) from the LWR-spent fuel (SF) reprocessing. As Japan can posses little excess Pu, its balance control is necessary between LWR-SF management (reprocessing) and FBR deployment. The fuel cycle systems were investigated for the smooth FBR deployment and the effectiveness of proposed flexible system was clarified in this work. (author)
Safety research for LWR type reactors
International Nuclear Information System (INIS)
1989-07-01
The current R and D activities are to be seen in connection with the LWR risk assessment studies. Two trends are emerging, of which the one concentrates more on BWR-specific problems, and the other on the efficiency or safety-related assessment of accident management activities. This annual report of 1988 reviews the progress of work done by the institutes and departments of the Karlsruhe Nuclear Research Center, (KfK), or on behalf of KfK by external institutions, in the field of safety research. The papers of this report present the state of work at the end of the year 1988. They are written in German, with an abstract in English. (orig./HP) [de
LWR nuclear power plant component failures
International Nuclear Information System (INIS)
Schmidt, W.H.
1980-10-01
An analysis of the most significant light water reactor (LWR) nuclear power plant component failures, from information in the computerized Nuclear Safety Information Center (NSIC) data bank, shows that for both pressurized water reactor (PWR) and boiling water reactor (BWR) plants the component category most responsible for reactor shutdowns is valves. Next in importance for PWR shutdowns is steam generators followed by seals of all kinds. For BWR plants, seals, and pipes and pipe fittings are the second and third most important component failure categories which lead to reactor shutdown. The data are for records extending from early 1972 through September 1978. A list of the most significant component categories and a breakdown of the number of component citations for both PWR and BWR reactor types are presented
Expert system for estimating LWR plutonium production
International Nuclear Information System (INIS)
Sandquist, G.M.
1988-01-01
An Artificial Intelligence-Expert System called APES (Analysis of Proliferation by Expert System) has been developed and tested to permit a non proliferation expert to evaluate the capability and capacity of a specified LWR reactor and PUREX reprocessing system for producing and separating plutonium even when system information may be limited and uncertain. APES employs an expert system coded in LISP and based upon an HP-RL (Hewlett Packard-Representational Language) Expert System Shell. The user I/O interface communicates with a blackboard and the knowledge base which contains the quantitative models required to describe the reactor, selected fission product production and radioactive decay processes, Purex reprocessing and ancillary knowledge
Problems associated with domestic LWR technology development
International Nuclear Information System (INIS)
Watamori, Tikara
1975-01-01
To cope with the future energy problem in Japan, the enhancement of her own technology is continuing in the nuclear power field. Developments in the past, current state, and problems for the future are described regarding LWR power plants. The technology introduced from overseas countries cannot be used as it is. The domestic technology thus consists of the conversion of nuclear power technology so as to meet Japan's own condition and the domestic manufacture of machinery. In the former category, there are the aspects of aseismatic design, waste disposal, software, etc. In the latter, there are the productions of reactor vessels, steam generators, large valves, piping, etc. As the problems for the future, there are reliability and safety and the associated standardization. (Mori, K.)
Geometrically Consistent Mesh Modification
Bonito, A.
2010-01-01
A new paradigm of adaptivity is to execute refinement, coarsening, and smoothing of meshes on manifolds with incomplete information about their geometry and yet preserve position and curvature accuracy. We refer to this collectively as geometrically consistent (GC) mesh modification. We discuss the concept of discrete GC, show the failure of naive approaches, and propose and analyze a simple algorithm that is GC and accuracy preserving. © 2010 Society for Industrial and Applied Mathematics.
Perspectives on the economic risks of LWR accidents
International Nuclear Information System (INIS)
Ritchie, L.T.; Burke, R.P.
1986-01-01
Models which can be used for the analysis of the economic risks from events which may occur during LWR operation have been developed. The models include capabilities to estimate both onsite and offsite costs of LWR events ranging from routine plant forced outages to severe core-melt accidents resulting in large releases of radioactive material to the environment. The economic consequence models have been applied in studies of the economic risks from the operation of US LWR plants. The results of the analyses provide some important perspectives regarding the economic risks of LWR accidents. The analyses indicate that economic risks, in contrast to public health risks, are dominated by the onsite costs of relatively high-frequency forced outage events. Even for severe (e.g., core-melt) accidents, expected offsite costs are less than expected onsite costs for a typical US plant
Status of the CONTAIN computer code for LWR containment analysis
International Nuclear Information System (INIS)
Bergeron, K.D.; Murata, K.K.; Rexroth, P.E.; Clauser, M.J.; Senglaub, M.E.; Sciacca, F.W.; Trebilcock, W.
1983-01-01
The current status of the CONTAIN code for LWR safety analysis is reviewed. Three example calculations are discussed as illustrations of the code's capabilities: (1) a demonstration of the spray model in a realistic PWR problem, and a comparison with CONTEMPT results; (2) a comparison of CONTAIN results for a major aerosol experiment against experimental results and predictions of the HAARM aerosol code; and (3) an LWR sample problem, involving a TMLB' sequence for the Zion reactor containment
Status of the CONTAIN computer code for LWR containment analysis
International Nuclear Information System (INIS)
Bergeron, K.D.; Murata, K.K.; Rexroth, P.E.; Clauser, M.J.; Senglaub, M.E.; Sciacca, F.W.; Trebilcock, W.
1982-01-01
The current status of the CONTAIN code for LWR safety analysis is reviewed. Three example calculations are discussed as illustrations of the code's capabilities: (1) a demonstration of the spray model in a realistic PWR problem, and a comparison with CONTEMPT results; (2) a comparison of CONTAIN results for a major aerosol experiment against experimental results and predictions of the HAARM aerosol code; and (3) an LWR sample problem, involving a TMLB' sequence for the Zion reactor containment
EURLIB-LWR-45/16 and - 15/5. Two board group libraries for LWR-shielding problems
Energy Technology Data Exchange (ETDEWEB)
Herrnberger, V
1982-04-01
Specifications of the broad group cross section libraries EURLIB-LWR-45/16 and -15/5 are given. They are based on EURLIB-III data and produced for LWR shielding problems. The elements considered are H, C{sub 12}, O, Na, Al, Si, Ca, Cr, Mn, Fe, Ni, Zr, U{sub 235}, U{sub 238}. The cross section libraries are available upon request from EIR, RSIC, NEA-CPL and IAEA-NDS. (author) Refs, figs, tabs
Parallel adaptation of general three-dimensional hybrid meshes
International Nuclear Information System (INIS)
Kavouklis, Christos; Kallinderis, Yannis
2010-01-01
A new parallel dynamic mesh adaptation and load balancing algorithm for general hybrid grids has been developed. The meshes considered in this work are composed of four kinds of elements; tetrahedra, prisms, hexahedra and pyramids, which poses a challenge to parallel mesh adaptation. Additional complexity imposed by the presence of multiple types of elements affects especially data migration, updates of local data structures and interpartition data structures. Efficient partition of hybrid meshes has been accomplished by transforming them to suitable graphs and using serial graph partitioning algorithms. Communication among processors is based on the faces of the interpartition boundary and the termination detection algorithm of Dijkstra is employed to ensure proper flagging of edges for refinement. An inexpensive dynamic load balancing strategy is introduced to redistribute work load among processors after adaptation. In particular, only the initial coarse mesh, with proper weighting, is balanced which yields savings in computation time and relatively simple implementation of mesh quality preservation rules, while facilitating coarsening of refined elements. Special algorithms are employed for (i) data migration and dynamic updates of the local data structures, (ii) determination of the resulting interpartition boundary and (iii) identification of the communication pattern of processors. Several representative applications are included to evaluate the method.
Documentation for MeshKit - Reactor Geometry (&mesh) Generator
Energy Technology Data Exchange (ETDEWEB)
Jain, Rajeev [Argonne National Lab. (ANL), Argonne, IL (United States); Mahadevan, Vijay [Argonne National Lab. (ANL), Argonne, IL (United States)
2015-09-30
This report gives documentation for using MeshKit’s Reactor Geometry (and mesh) Generator (RGG) GUI and also briefly documents other algorithms and tools available in MeshKit. RGG is a program designed to aid in modeling and meshing of complex/large hexagonal and rectilinear reactor cores. RGG uses Argonne’s SIGMA interfaces, Qt and VTK to produce an intuitive user interface. By integrating a 3D view of the reactor with the meshing tools and combining them into one user interface, RGG streamlines the task of preparing a simulation mesh and enables real-time feedback that reduces accidental scripting mistakes that could waste hours of meshing. RGG interfaces with MeshKit tools to consolidate the meshing process, meaning that going from model to mesh is as easy as a button click. This report is designed to explain RGG v 2.0 interface and provide users with the knowledge and skills to pilot RGG successfully. Brief documentation of MeshKit source code, tools and other algorithms available are also presented for developers to extend and add new algorithms to MeshKit. RGG tools work in serial and parallel and have been used to model complex reactor core models consisting of conical pins, load pads, several thousands of axially varying material properties of instrumentation pins and other interstices meshes.
Utility requirements for advanced LWR passive plants
International Nuclear Information System (INIS)
Yedidia, J.M.; Sugnet, W.R.
1992-01-01
LWR Passive Plants are becoming an increasingly attractive and prominent option for future electric generating capacity for U.S. utilities. Conceptual designs for ALWR Passive Plants are currently being developed by U.S. suppliers. EPRI-sponsored work beginning in 1985 developed preliminary conceptual designs for a passive BWR and PWR. DOE-sponsored work from 1986 to the present in conjunction with further EPRI-sponsored studies has continued this development to the point of mature conceptual designs. The success to date in developing the ALWR Passive Plant concepts has substantially increased utility interest. The EPRI ALWR Program has responded by augmenting its initial scope to develop a Utility Requirements Document for ALWR Passive Plants. These requirements will be largely based on the ALWR Utility Requirements Document for Evolutionary Plants, but with significant changes in areas related to the passive safety functions and system configurations. This work was begun in late 1988, and the thirteen-chapter Passive Plant Utility Requirements Document will be completed in 1990. This paper discusses the progress to date in developing the Passive Plant requirements, reviews the top-level requirements, and discusses key issues related to adaptation of the utility requirements to passive safety functions and system configurations. (orig.)
Criticality impacts on LWR fuel storage efficiency
International Nuclear Information System (INIS)
Napolitano, D.
1992-01-01
This presentation discusses the criticality impacts throughout storage of fuel onsite including new fuel storage, spent fuel storage, consolidation, and dry storage. The general principles for criticality safety are also be discussed. There is first an introduction which explains today's situation for criticality safety concerns. This is followed by a discussion of criticality safety Regulatory Guides, safety limits and fundamental principles. Design objectives for criticality safety in the 1990's include higher burnups, longer cycles, and higher enrichments which impact the criticality safety design. Criticality safety for new fuel storage, spent fuel storage, fuel consolidation, and dry storage are followed by conclusions. Today's situation is one in which the US does not reprocess, and does not have an operating MRS facility or repository. High density fuel storage rack designs of the 1980s, are filling up. Dry cask storage systems for spent fuel storage are being utilized. Enrichments continue to increase PWR fuel assemblies with enrichments of 4.5 to 5.0 weight percent U-235 and BWR fuel assemblies with enrichments of 3.25 to 3.5 weight percent U-235 are common. Criticality concerns affect the capacity and the economics of light water reactor (LWR) fuel storage arrays by dictating the spacing of fuel assemblies in a storage system, or the use of poisons or exotic materials in the storage system design
Economic analyses of LWR fuel cycles
International Nuclear Information System (INIS)
Field, F.R.
1977-05-01
An economic comparison was made of three options for handling irradiated light-water reactor (LWR) fuel. These options are reprocessing of spent reactor fuel and subsequent recycle of both uranium and plutonium, reprocessing and recycle of uranium only, and direct terminal storage of spent fuel not reprocessed. The comparison was based on a peak-installed nuclear capacity of 507 GWe by CY 2000 and retirement of reactors after 30 years of service. Results of the study indicate that: Through the year 2000, recycle of uranium and plutonium in LWRs saves about $12 billion (FY 1977 dollars) compared with the throwaway cycle, but this amounts to only about 1.3% of the total cost of generating electricity by nuclear power. If deferred costs are included for fuel that has been discharged from reactors but not reprocessed, the economic advantage increases to $17.7 billion. Recycle of uranium only (storage of plutonium) is approximately $7 billion more expensive than the throwaway fuel cycle and is, therefore, not considered an economically viable option. The throwaway fuel cycle ultimately requires >40% more uranium resources (U 3 O 8 ) than does reprocessing spent fuel where both uranium and plutonium are recycled
NUPEC proves reliability of LWR fuel assemblies
International Nuclear Information System (INIS)
Anon.
1987-01-01
It is very important in assuring the safety of nuclear reactors to confirm the reliability of fuel assemblies. The test program of the Nuclear Power Engineering Center on the reliability of fuel assemblies has verified the high performance and reliability of Japanese LWR fuels, and confirmed the propriety of their design and fabrication. This claim is based on the data obtained from the fuel assemblies irradiated in commercial reactors. The NUPEC program includes irradiation test which has been conducted for 11 years since fiscal 1976, and the maximum thermal loading test using the out of pile test facilities simulating a real reactor which has been continued since fiscal 1978. The irradiation test on BWR fuel assemblies in No.3 reactor in Fukushima No.1 Nuclear Power Station, Tokyo Electric Power Co., Inc., and on PWR fuel assemblies in No.3 reactor in Mihama Power Station, Kansai Electric Power Co., Inc., and the maximum thermal loading test on BWR and PWR fuel assemblies are reported. The series of postirradiation examination of the fuel assemblies used for commercial reactors was conducted for the first time in Japan, and the highly systematic data on 27 items were obtained. (Kako, I.)
Advanced LWR Nuclear Fuel Cladding Development
International Nuclear Information System (INIS)
Bragg-Sitton, S.; Griffith, G.
2012-01-01
The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R and D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental enhancements are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction to allow improved fuel economy via power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an 'accident tolerant' fuel system that would offer improved coping time under accident scenarios. In a staged development approach, the LWRS program will engage stakeholders throughout the development process to ensure commercial viability of the investigated technologies. Applying minimum performance criteria, several of the top-ranked materials and fabrication concepts will undergo a rigorous series of mechanical, thermal and chemical characterization tests to better define their properties and operating potential in a relatively low-cost, nonnuclear test series. A reduced number of options will be recommended for test rodlet fabrication and in-pile nuclear testing under steady-state, transient and accident conditions. (author)
LWR primary coolant pipe rupture test rig
International Nuclear Information System (INIS)
Yoshitoshi, Shyoji
1978-01-01
The rupture test rig for primary coolant pipes is constructed in the Japan Atomic Energy Research Institute to verify the reliability of the primary coolant pipes for both PWRs and BWRs. The planned test items consisted of reaction force test, restraint test, whip test, jet test and continuous release test. A pressure vessel of about 4 m 3 volume, a circulating pump, a pressurizer, a heater, an air cooler and the related instrumentation and control system are included in this test rig. The coolant test condition is 160 kg/cm 2 g, 325 deg C for PWR test, and 70 kg/cm 2 g, saturated water and steam for BWR test, 100 ton of test load for the ruptured pipe bore of 8B Schedule 160, and 20 lit/min. discharge during 20 h for continuous release of coolant. The maximum pit internal pressure was estimated for various pipe diameters and time under the PWR and BWR conditions. The spark rupturing device was adopted for the rupture mechanics in this test rig. The computer PANAFACOM U-300 is used for the data processing. This test rig is expected to operate in 1978 effectively for the improvement of reliability of LWR primary coolant pipes. (Nakai, Y.)
Results of LWR core transient benchmarks
International Nuclear Information System (INIS)
Finnemann, H.; Bauer, H.; Galati, A.; Martinelli, R.
1993-10-01
LWR core transient (LWRCT) benchmarks, based on well defined problems with a complete set of input data, are used to assess the discrepancies between three-dimensional space-time kinetics codes in transient calculations. The PWR problem chosen is the ejection of a control assembly from an initially critical core at hot zero power or at full power, each for three different geometrical configurations. The set of problems offers a variety of reactivity excursions which efficiently test the coupled neutronic/thermal - hydraulic models of the codes. The 63 sets of submitted solutions are analyzed by comparison with a nodal reference solution defined by using a finer spatial and temporal resolution than in standard calculations. The BWR problems considered are reactivity excursions caused by cold water injection and pressurization events. In the present paper, only the cold water injection event is discussed and evaluated in some detail. Lacking a reference solution the evaluation of the 8 sets of BWR contributions relies on a synthetic comparative discussion. The results of this first phase of LWRCT benchmark calculations are quite satisfactory, though there remain some unresolved issues. It is therefore concluded that even more challenging problems can be successfully tackled in a suggested second test phase. (authors). 46 figs., 21 tabs., 3 refs
Toward An Unstructured Mesh Database
Rezaei Mahdiraji, Alireza; Baumann, Peter Peter
2014-05-01
Unstructured meshes are used in several application domains such as earth sciences (e.g., seismology), medicine, oceanography, cli- mate modeling, GIS as approximate representations of physical objects. Meshes subdivide a domain into smaller geometric elements (called cells) which are glued together by incidence relationships. The subdivision of a domain allows computational manipulation of complicated physical structures. For instance, seismologists model earthquakes using elastic wave propagation solvers on hexahedral meshes. The hexahedral con- tains several hundred millions of grid points and millions of hexahedral cells. Each vertex node in the hexahedrals stores a multitude of data fields. To run simulation on such meshes, one needs to iterate over all the cells, iterate over incident cells to a given cell, retrieve coordinates of cells, assign data values to cells, etc. Although meshes are used in many application domains, to the best of our knowledge there is no database vendor that support unstructured mesh features. Currently, the main tool for querying and manipulating unstructured meshes are mesh libraries, e.g., CGAL and GRAL. Mesh li- braries are dedicated libraries which includes mesh algorithms and can be run on mesh representations. The libraries do not scale with dataset size, do not have declarative query language, and need deep C++ knowledge for query implementations. Furthermore, due to high coupling between the implementations and input file structure, the implementations are less reusable and costly to maintain. A dedicated mesh database offers the following advantages: 1) declarative querying, 2) ease of maintenance, 3) hiding mesh storage structure from applications, and 4) transparent query optimization. To design a mesh database, the first challenge is to define a suitable generic data model for unstructured meshes. We proposed ImG-Complexes data model as a generic topological mesh data model which extends incidence graph model to multi
Environmental development plan. LWR commercial waste management
International Nuclear Information System (INIS)
1980-08-01
This Environmental Development Plan (EDP) identifies the planning and managerial requirements and schedules needed to evaluate and assess the environmental, health and safety (EH and S) aspects of the Commercial Waste Management Program (CWM). Environment is defined in its broadest sense to include environmental, health (occupational and public), safety, socioeconomic, legal and institutional aspects. This plan addresses certain present and potential Federal responsibilities for the storage, treatment, transfer and disposal of radioactive waste materials produced by the nuclear power industry. The handling and disposal of LWR spent fuel and processed high-level waste (in the event reprocessing occurs) are included in this plan. Defense waste management activities, which are addressed in detail in a separate EDP, are considered only to the extent that such activities are common to the commercial waste management program. This EDP addresses three principal elements associated with the disposal of radioactive waste materials from the commercial nuclear power industry, namely Terminal Isolation Research and Development, Spent Fuel Storage and Waste Treatment Technology. The major specific concerns and requirements addressed are assurance that (1) radioactivity will be contained during waste transport, interim storage or while the waste is considered as retrievable from a repository facility, (2) the interim storage facilities will adequately isolate the radioactive material from the biosphere, (3) the terminal isolation facility will isolate the wastes from the biosphere over a time period allowing the radioactivity to decay to innocuous levels, (4) the terminal isolation mode for the waste will abbreviate the need for surveillance and institutional control by future generations, and (5) the public will accept the basic waste management strategy and geographical sites when needed
Design consideration on severe accident for future LWR
International Nuclear Information System (INIS)
Omoto, A.
1998-01-01
Utilities' Severe Accident Management strategies, selected based on Individual Plant Examination, are in the process of implementation for each operating plant. Activities for the next generation LWR design are going on by Utilities, NSSS vendors and Research Institutes. The proposed new designs vary from evolutionary design to revolutionary design such as the supercritical LWR. Discussion on the consideration of Severe Accident in the design of next generation LWR is being held to establish the industry's self-regulatory document on containment design and its performance, which ABWR-IER (Improved Evolutionary Reactor) on the part of BWR and Evolutionary APWR and New PWR21 on the part of PWR are expected to comply. Conceptual design study for ABWR-IER will illustrate an example of design approach for the prevention and mitigation of Severe Accident and its impact on capital cost
How coarse is too coarse for salmon spawning substrates?
Wooster, J. K.; Riebe, C. S.; Ligon, F. K.; Overstreet, B. T.
2009-12-01
Populations of Pacific salmon species have declined sharply in many rivers of the western US. Reversing these declines is a top priority and expense of many river restoration projects. To help restore salmon populations, managers often inject gravel into rivers, to supplement spawning habitat that has been depleted by gravel mining and the effects of dams—which block sediment and thus impair habitat downstream by coarsening the bed where salmon historically spawned. However, there is little quantitative understanding nor a methodology for determining when a river bed has become too coarse for salmon spawning. Hence there is little scientific basis for selecting sites that would optimize the restoration benefits of gravel injection (e.g., sites where flow velocities are suitable but bed materials are too coarse for spawning). To develop a quantitative understanding of what makes river beds too coarse for salmon spawning, we studied redds and spawning use in a series of California and Washington rivers where salmon spawning ability appears to be affected by coarse bed material. Our working hypothesis is that for a given flow condition, there is a maximum “threshold” particle size that a salmon of a given size is able to excavate and/or move as she builds her redd. A second, related hypothesis is that spawning use should decrease and eventually become impossible with increasing percent coverage by immovable particles. To test these hypotheses, we quantified the sizes and spatial distributions of immovably coarse particles in a series of salmon redds in each river during the peak of spawning. We also quantified spawning use and how it relates to percent coverage by immovable particles. Results from our studies of fall-run chinook salmon (Oncorhynchus tshawytsha) in the Feather River suggest that immovable particle size varies as a function of flow velocity over the redd, implying that faster water helps fish move bigger particles. Our Feather River study also
SUPERIMPOSED MESH PLOTTING IN MCNP
Energy Technology Data Exchange (ETDEWEB)
J. HENDRICKS
2001-02-01
The capability to plot superimposed meshes has been added to MCNP{trademark}. MCNP4C featured a superimposed mesh weight window generator which enabled users to set up geometries without having to subdivide geometric cells for variance reduction. The variance reduction was performed with weight windows on a rectangular or cylindrical mesh superimposed over the physical geometry. Experience with the new capability was favorable but also indicated that a number of enhancements would be very beneficial, particularly a means of visualizing the mesh and its values. The mathematics for plotting the mesh and its values is described here along with a description of other upgrades.
The need for LWR metrology standardization: the imec roughness protocol
Lorusso, Gian Francesco; Sutani, Takumichi; Rutigliani, Vito; van Roey, Frieda; Moussa, Alain; Charley, Anne-Laure; Mack, Chris; Naulleau, Patrick; Constantoudis, Vassilios; Ikota, Masami; Ishimoto, Toru; Koshihara, Shunsuke
2018-03-01
As semiconductor technology keeps moving forward, undeterred by the many challenges ahead, one specific deliverable is capturing the attention of many experts in the field: Line Width Roughness (LWR) specifications are expected to be less than 2nm in the near term, and to drop below 1nm in just a few years. This is a daunting challenge and engineers throughout the industry are trying to meet these targets using every means at their disposal. However, although current efforts are surely admirable, we believe they are not enough. The fact is that a specification has a meaning only if there is an agreed methodology to verify if the criterion is met or not. Such a standardization is critical in any field of science and technology and the question that we need to ask ourselves today is whether we have a standardized LWR metrology or not. In other words, if a single reference sample were provided, would everyone measuring it get reasonably comparable results? We came to realize that this is not the case and that the observed spread in the results throughout the industry is quite large. In our opinion, this makes the comparison of LWR data among institutions, or to a specification, very difficult. In this paper, we report the spread of measured LWR data across the semiconductor industry. We investigate the impact of image acquisition, measurement algorithm, and frequency analysis parameters on LWR metrology. We review critically some of the International Technology Roadmap for Semiconductors (ITRS) metrology guidelines (such as measurement box length larger than 2μm and the need to correct for SEM noise). We compare the SEM roughness results to AFM measurements. Finally, we propose a standardized LWR measurement protocol - the imec Roughness Protocol (iRP) - intended to ensure that every time LWR measurements are compared (from various sources or to specifications), the comparison is sensible and sound. We deeply believe that the industry is at a point where it is
The minimum attention plant inherent safety through LWR simplification
International Nuclear Information System (INIS)
Turk, R.S.; Matzie, R.A.
1987-01-01
The Minimum Attention Plant (MAP) is a unique small LWR that achieves greater inherent safety, improved operability, and reduced costs through design simplification. The MAP is a self-pressurized, indirect-cycle light water reactor with full natural circulation primary coolant flow and multiple once-through steam generators located within the reactor vessel. A fundamental tenent of the MAP design is its complete reliance on existing LWR technology. This reliance on conventional technology provides an extensive experience base which gives confidence in judging the safety and performance aspects of the design
LWR and HTGR coolant dynamics: the containment of severe accidents
International Nuclear Information System (INIS)
Theofanous, T.G.; Gherson, P.; Nourbakhsh, H.P.; Hu, K.; Iyer, K.; Viskanta, R.; Lommers, L.
1983-07-01
This is the final report of a project containing three major tasks. Task I deals with the fundamental aspects of energetic fuel/coolant interactions (steam explosions) as they pertain to LWR core melt accidents. Task II deals with the applied aspects of LWR core melt accident sequences and mechanisms important to containment response, and includes consideration of energetic fuel/coolant interaction events, as well as non-explosive ones, corium material disposition and eventual coolability, and containment pressurization phenomena. Finally, Task III is concerned with HTGR loss of forced circulation accidents. This report is organized into three major parts corresponding to these three tasks respectively
Metal Matrix Microencapsulated Fuel Technology for LWR Applications
International Nuclear Information System (INIS)
Terrani, Kurt A.; Bell, Gary L.; Kiggans, Jim; Snead, Lance Lewis
2012-01-01
An overview of the metal matrix microencapsulated (M3) fuel concept for the specific LWR application has been provided. Basic fuel properties and characteristics that aim to improve operational reliability, enlarge performance envelope, and enhance safety margins under design-basis accident scenarios are summarized. Fabrication of M3 rodlets with various coated fuel particles over a temperature range of 800-1300 C is discussed. Results from preliminary irradiation testing of LWR M3 rodlets with surrogate coated fuel particles are also reported.
Wang, Xinheng
2008-01-01
Wireless telemedicine using GSM and GPRS technologies can only provide low bandwidth connections, which makes it difficult to transmit images and video. Satellite or 3G wireless transmission provides greater bandwidth, but the running costs are high. Wireless networks (WLANs) appear promising, since they can supply high bandwidth at low cost. However, the WLAN technology has limitations, such as coverage. A new wireless networking technology named the wireless mesh network (WMN) overcomes some of the limitations of the WLAN. A WMN combines the characteristics of both a WLAN and ad hoc networks, thus forming an intelligent, large scale and broadband wireless network. These features are attractive for telemedicine and telecare because of the ability to provide data, voice and video communications over a large area. One successful wireless telemedicine project which uses wireless mesh technology is the Emergency Room Link (ER-LINK) in Tucson, Arizona, USA. There are three key characteristics of a WMN: self-organization, including self-management and self-healing; dynamic changes in network topology; and scalability. What we may now see is a shift from mobile communication and satellite systems for wireless telemedicine to the use of wireless networks based on mesh technology, since the latter are very attractive in terms of cost, reliability and speed.
Directory of Open Access Journals (Sweden)
Jorge Pérez Mañes
2014-01-01
Full Text Available The Institute for Neutron Physics and Reactor Technology (INR at the Karlsruhe Institute of Technology (KIT is investigating the application of the meso- and microscale analysis for the prediction of local safety parameters for light water reactors (LWR. By applying codes like CFD (computational fluid dynamics and SP3 (simplified transport reactor dynamics it is possible to describe the underlying phenomena in a more accurate manner than by the nodal/coarse 1D thermal hydraulic coupled codes. By coupling the transport (SP3 based neutron kinetics (NK code DYN3D with NEPTUNE-CFD, within a parallel MPI-environment, the NHESDYN platform is created. The newly developed system will allow high fidelity simulations of LWR fuel assemblies and cores. In NHESDYN, a heat conduction solver, SYRTHES, is coupled to NEPTUNE-CFD. The driver module of NHESDYN controls the sequence of execution of the solvers as well as the communication between the solvers based on MPI. In this paper, the main features of NHESDYN are discussed and the proof of the concept is done by solving a single pin problem. The prediction capability of NHESDYN is demonstrated by a code-to-code comparison with the DYNSUB code. Finally, the future developments and validation efforts are highlighted.
Hursin, Mathieu; Leray, Olivier; Perret, Gregory; Pautz, Andreas; Bostelmann, Friederike; Aures, Alexander; Zwermann, Winfried
2017-09-01
In the present work, PSI and GRS sensitivity analysis (SA) and uncertainty quantification (UQ) methods, SHARK-X and XSUSA respectively, are compared for reactivity coefficient calculation; for reference the results of the TSUNAMI and SAMPLER modules of the SCALE code package are also provided. The main objective of paper is to assess the impact of the implicit effect, e.g., considering the effect of cross section perturbation on the self-shielding calculation, on the Doppler coefficient SA and UQ. Analyses are done for a Light Water Reactor (LWR) pin cell based on Phase I of the UAM LWR benchmark. The negligence of implicit effects in XSUSA and TSUNAMI leads to deviations of a few percent between the sensitivity profiles compared to SAMPLER and TSUNAMI (incl. implicit effects) except for 238U elastic scattering. The implicit effect is much larger for the SHARK-X calculations because of its coarser energy group structure between 10 eV and 10 keV compared to the applied SCALE libraries. It is concluded that the influence of the implicit effect strongly depends on the energy mesh of the nuclear data library of the neutron transport solver involved in the UQ calculations and may be magnified by the response considered.
Mesh erosion after abdominal sacrocolpopexy.
Kohli, N; Walsh, P M; Roat, T W; Karram, M M
1998-12-01
To report our experience with erosion of permanent suture or mesh material after abdominal sacrocolpopexy. A retrospective chart review was performed to identify patients who underwent sacrocolpopexy by the same surgeon over 8 years. Demographic data, operative notes, hospital records, and office charts were reviewed after sacrocolpopexy. Patients with erosion of either suture or mesh were treated initially with conservative therapy followed by surgical intervention as required. Fifty-seven patients underwent sacrocolpopexy using synthetic mesh during the study period. The mean (range) postoperative follow-up was 19.9 (1.3-50) months. Seven patients (12%) had erosions after abdominal sacrocolpopexy with two suture erosions and five mesh erosions. Patients with suture erosion were asymptomatic compared with patients with mesh erosion, who presented with vaginal bleeding or discharge. The mean (+/-standard deviation) time to erosion was 14.0+/-7.7 (range 4-24) months. Both patients with suture erosion were treated conservatively with estrogen cream. All five patients with mesh erosion required transvaginal removal of the mesh. Mesh erosion can follow abdominal sacrocolpopexy over a long time, and usually presents as vaginal bleeding or discharge. Although patients with suture erosion can be managed successfully with conservative treatment, patients with mesh erosion require surgical intervention. Transvaginal removal of the mesh with vaginal advancement appears to be an effective treatment in patients failing conservative management.
Preliminary concepts for detecting national diversion of LWR spent fuel
International Nuclear Information System (INIS)
Sonnier, C.S.; Cravens, M.N.
1978-04-01
Preliminary concepts for detecting national diversion of LWR spent fuel during storage, handling and transportation are presented. Principal emphasis is placed on means to achieve timely detection by an international authority. This work was sponsored by the Department of Energy/Office of Safeguards and Security (DOE/OSS) as part of the overall Sandia Fixed Facility Physical Protection Program
Safety criteria related to microheterogeneities in LWR mixed oxide fuels
International Nuclear Information System (INIS)
Renard, A.; Mostin, N.
1978-01-01
The main safety aspets of PuO 2 microheterogeneities in the pellets of LWR mixed oxide fuels are reviewed. Points of interest are studied, especially the transient behaviour in accidental conditions and criteria are deduced for use in the specification and quality control of the fabricated product. (author)
Nondestructive evaluation of LWR spent fuel shipping casks
International Nuclear Information System (INIS)
Ballard, D.W.
1978-02-01
An analysis of nondestructive testing (NDT) methods currently being used to evaluate the integrity of Light Water Reactor (LWR) spent fuel shipping casks is presented. An assessment of anticipated NDT needs related to breeder reactor cask requirements is included. Specific R and D approaches to probable NDT problem areas such as the evaluation of austenitic stainless steel weldments are outlined
Materials choices for the advanced LWR steam generators
International Nuclear Information System (INIS)
Paine, J.P.N.; Shoemaker, C.E.; McIlree, A.R.
1987-01-01
Current light water reactor (LWR) steam generators have been affected by a variety of corrosion and mechanical damage degradation mechanisms. Included are wear caused by tube vibration, intergranular corrosion, pitting, and thinning or wastage of the steam generator tubing and accelerated corrosion of carbon steel supports (denting). The Electric Power Research Institute (EPRI) and the Steam Generator Owners Groups (I, II) have sponsored laboratory and field studies to provide ameliorative actions for the majority of the damage forms experienced to date. Some of the current corrosion mechanisms are aggravated or caused by unique materials choices or materials interactions. New materials have been proposed and at least partially qualified for use in replacement model steam generators, including an advanced LWR design. In so far as possible, the materials choices for the advanced LWR steam generator avoid the corrosion pitfalls seemingly inherent in the current designs. The EPRI Steam Generator Project staff has recommended materials and design choices for a new steam generator. Based on these recommendations we believe that the advanced LWR steam generators will be much less affected by corrosion and mechanical damage mechanisms than are now experienced
Contributions to LWR spent fuel storage and transport
International Nuclear Information System (INIS)
The papers included in this document describe the aspects of spent LWR fuel storage and transport-behaviour of spent fuel during storage; use of compact storage packs; safety of storage; design of storage facilities AR and AFR; description of transport casks and transport procedures
Notes on the Mesh Handler and Mesh Data Conversion
International Nuclear Information System (INIS)
Lee, Sang Yong; Park, Chan Eok
2009-01-01
At the outset of the development of the thermal-hydraulic code (THC), efforts have been made to utilize the recent technology of the computational fluid dynamics. Among many of them, the unstructured mesh approach was adopted to alleviate the restriction of the grid handling system. As a natural consequence, a mesh handler (MH) has been developed to manipulate the complex mesh data from the mesh generator. The mesh generator, Gambit, was chosen at the beginning of the development of the code. But a new mesh generator, Pointwise, was introduced to get more flexible mesh generation capability. An open source code, Paraview, was chosen as a post processor, which can handle unstructured as well as structured mesh data. Overall data processing system for THC is shown in Figure-1. There are various file formats to save the mesh data in the permanent storage media. A couple of dozen of file formats are found even in the above mentioned programs. A competent mesh handler should have the capability to import or export mesh data as many as possible formats. But, in reality, there are two aspects that make it difficult to achieve the competence. The first aspect to consider is the time and efforts to program the interface code. And the second aspect, which is even more difficult one, is the fact that many mesh data file formats are proprietary information. In this paper, some experience of the development of the format conversion programs will be presented. File formats involved are Gambit neutral format, Ansys-CFX grid file format, VTK legacy file format, Nastran format and CGNS
Directory of Open Access Journals (Sweden)
Juan J. Garcia-Cantero
2017-06-01
Full Text Available Gaining a better understanding of the human brain continues to be one of the greatest challenges for science, largely because of the overwhelming complexity of the brain and the difficulty of analyzing the features and behavior of dense neural networks. Regarding analysis, 3D visualization has proven to be a useful tool for the evaluation of complex systems. However, the large number of neurons in non-trivial circuits, together with their intricate geometry, makes the visualization of a neuronal scenario an extremely challenging computational problem. Previous work in this area dealt with the generation of 3D polygonal meshes that approximated the cells’ overall anatomy but did not attempt to deal with the extremely high storage and computational cost required to manage a complex scene. This paper presents NeuroTessMesh, a tool specifically designed to cope with many of the problems associated with the visualization of neural circuits that are comprised of large numbers of cells. In addition, this method facilitates the recovery and visualization of the 3D geometry of cells included in databases, such as NeuroMorpho, and provides the tools needed to approximate missing information such as the soma’s morphology. This method takes as its only input the available compact, yet incomplete, morphological tracings of the cells as acquired by neuroscientists. It uses a multiresolution approach that combines an initial, coarse mesh generation with subsequent on-the-fly adaptive mesh refinement stages using tessellation shaders. For the coarse mesh generation, a novel approach, based on the Finite Element Method, allows approximation of the 3D shape of the soma from its incomplete description. Subsequently, the adaptive refinement process performed in the graphic card generates meshes that provide good visual quality geometries at a reasonable computational cost, both in terms of memory and rendering time. All the described techniques have been
Streaming simplification of tetrahedral meshes.
Vo, Huy T; Callahan, Steven P; Lindstrom, Peter; Pascucci, Valerio; Silva, Cláudio T
2007-01-01
Unstructured tetrahedral meshes are commonly used in scientific computing to represent scalar, vector, and tensor fields in three dimensions. Visualization of these meshes can be difficult to perform interactively due to their size and complexity. By reducing the size of the data, we can accomplish real-time visualization necessary for scientific analysis. We propose a two-step approach for streaming simplification of large tetrahedral meshes. Our algorithm arranges the data on disk in a streaming, I/O-efficient format that allows coherent access to the tetrahedral cells. A quadric-based simplification is sequentially performed on small portions of the mesh in-core. Our output is a coherent streaming mesh which facilitates future processing. Our technique is fast, produces high quality approximations, and operates out-of-core to process meshes too large for main memory.
Performance and reliability of LWR fuel
International Nuclear Information System (INIS)
Bairiot, H.; Deramaix, P.; Vandenberg, C.
1977-01-01
The main requirements for fuel reloads are: good reliability, minimum fuel cycle costs and flexibility of operation. Fulfilling these goals requires a background of experience. The approach to the acquisition of this experience in the particular case of BN has included over the last 15 years a proper development and cross-checking of the design methods and criteria, a continuous updating of the drawings and specifications and the qualification of adequate fabrication plants. This approach can best be outlined on the basis of the gradual implementation of the modern features of the LWR fuel. The first fuel clad with stainless steel was loaded in the BR 3 (11 MWe) in 1969 and later on (since 1974) in the SENA plant (310 MWe). Similarly, Zircaloy 4 cladding was first introduced in a reactor reload in 1969 as autoclaved cladding and later on (in 1971) the autoclaving was suppressed for the further reloads. Zircaloy 2 was loaded in DODEWAARD (51.5 MWe) in 1970. The first demonstration assembly in a PWR was a Pu-island assembly loaded in the BR 3 in 1963. It was followed by an all-Pu assembly in the same reactor in 1965 and by the loading of Pu fuels in four prototype assemblies in GARIGLIANO (160 MWe) in 1968. A full reload incorporating Pu fuel has been experienced by the supply of fuel for GARIGLIANO (BOL: 1975) and for BR 3 (BOL: 1972 and 1976). While in the early sixties the brazed design was still being utilized, the first assembly incorporating grids with springs was introduced in BR 3 in 1963. The first Inconel grids were loaded in the same reactor in 1969 and the first Zircaloy grids in 1972 (the first Zr grid has been loaded in a BWR in 1973). The experience covered successively the shrouded design (BOL: 1963), the shroudless design (BOL: 1969), a BWR assembly (BOL: 1971), a typical RCC assembly first with large diameter fuel rods (1972) and later on with small diameter fuel rods (1974). The experience on the reactivity control covered successively diluted
Surface meshing with curvature convergence
Li, Huibin; Zeng, Wei; Morvan, Jean-Marie; Chen, Liming; Gu, Xianfengdavid
2014-01-01
Surface meshing plays a fundamental role in graphics and visualization. Many geometric processing tasks involve solving geometric PDEs on meshes. The numerical stability, convergence rates and approximation errors are largely determined by the mesh qualities. In practice, Delaunay refinement algorithms offer satisfactory solutions to high quality mesh generations. The theoretical proofs for volume based and surface based Delaunay refinement algorithms have been established, but those for conformal parameterization based ones remain wide open. This work focuses on the curvature measure convergence for the conformal parameterization based Delaunay refinement algorithms. Given a metric surface, the proposed approach triangulates its conformal uniformization domain by the planar Delaunay refinement algorithms, and produces a high quality mesh. We give explicit estimates for the Hausdorff distance, the normal deviation, and the differences in curvature measures between the surface and the mesh. In contrast to the conventional results based on volumetric Delaunay refinement, our stronger estimates are independent of the mesh structure and directly guarantee the convergence of curvature measures. Meanwhile, our result on Gaussian curvature measure is intrinsic to the Riemannian metric and independent of the embedding. In practice, our meshing algorithm is much easier to implement and much more efficient. The experimental results verified our theoretical results and demonstrated the efficiency of the meshing algorithm. © 2014 IEEE.
Surface meshing with curvature convergence
Li, Huibin
2014-06-01
Surface meshing plays a fundamental role in graphics and visualization. Many geometric processing tasks involve solving geometric PDEs on meshes. The numerical stability, convergence rates and approximation errors are largely determined by the mesh qualities. In practice, Delaunay refinement algorithms offer satisfactory solutions to high quality mesh generations. The theoretical proofs for volume based and surface based Delaunay refinement algorithms have been established, but those for conformal parameterization based ones remain wide open. This work focuses on the curvature measure convergence for the conformal parameterization based Delaunay refinement algorithms. Given a metric surface, the proposed approach triangulates its conformal uniformization domain by the planar Delaunay refinement algorithms, and produces a high quality mesh. We give explicit estimates for the Hausdorff distance, the normal deviation, and the differences in curvature measures between the surface and the mesh. In contrast to the conventional results based on volumetric Delaunay refinement, our stronger estimates are independent of the mesh structure and directly guarantee the convergence of curvature measures. Meanwhile, our result on Gaussian curvature measure is intrinsic to the Riemannian metric and independent of the embedding. In practice, our meshing algorithm is much easier to implement and much more efficient. The experimental results verified our theoretical results and demonstrated the efficiency of the meshing algorithm. © 2014 IEEE.
Coarse-graining complex dynamics
DEFF Research Database (Denmark)
Sibani, Paolo
2013-01-01
Continuous Time Random Walks (CTRW) are widely used to coarse-grain the evolution of systems jumping from a metastable sub-set of their configuration space, or trap, to another via rare intermittent events. The multi-scaled behavior typical of complex dynamics is provided by a fat...... macroscopic variables all produce identical long time relaxation behaviors. Hence, CTRW shed no light on the link between microscopic and macroscopic dynamics. We then highlight how a more recent approach, Record Dynamics (RD) provides a viable alternative, based on a very different set of physical ideas......: while CTRW make use of a renewal process involving identical traps of infinite size, RD embodies a dynamical entrenchment into a hierarchy of traps which are finite in size and possess different degrees of meta-stability. We show in particular how RD produces the stretched exponential, power...
Hierarchical coarse-graining transform.
Pancaldi, Vera; King, Peter R; Christensen, Kim
2009-03-01
We present a hierarchical transform that can be applied to Laplace-like differential equations such as Darcy's equation for single-phase flow in a porous medium. A finite-difference discretization scheme is used to set the equation in the form of an eigenvalue problem. Within the formalism suggested, the pressure field is decomposed into an average value and fluctuations of different kinds and at different scales. The application of the transform to the equation allows us to calculate the unknown pressure with a varying level of detail. A procedure is suggested to localize important features in the pressure field based only on the fine-scale permeability, and hence we develop a form of adaptive coarse graining. The formalism and method are described and demonstrated using two synthetic toy problems.
Impact of Variable-Resolution Meshes on Regional Climate Simulations
Fowler, L. D.; Skamarock, W. C.; Bruyere, C. L.
2014-12-01
The Model for Prediction Across Scales (MPAS) is currently being used for seasonal-scale simulations on globally-uniform and regionally-refined meshes. Our ongoing research aims at analyzing simulations of tropical convective activity and tropical cyclone development during one hurricane season over the North Atlantic Ocean, contrasting statistics obtained with a variable-resolution mesh against those obtained with a quasi-uniform mesh. Analyses focus on the spatial distribution, frequency, and intensity of convective and grid-scale precipitations, and their relative contributions to the total precipitation as a function of the horizontal scale. Multi-month simulations initialized on May 1st 2005 using ERA-Interim re-analyses indicate that MPAS performs satisfactorily as a regional climate model for different combinations of horizontal resolutions and transitions between the coarse and refined meshes. Results highlight seamless transitions for convection, cloud microphysics, radiation, and land-surface processes between the quasi-uniform and locally- refined meshes, despite the fact that the physics parameterizations were not developed for variable resolution meshes. Our goal of analyzing the performance of MPAS is twofold. First, we want to establish that MPAS can be successfully used as a regional climate model, bypassing the need for nesting and nudging techniques at the edges of the computational domain as done in traditional regional climate modeling. Second, we want to assess the performance of our convective and cloud microphysics parameterizations as the horizontal resolution varies between the lower-resolution quasi-uniform and higher-resolution locally-refined areas of the global domain.
Energy Technology Data Exchange (ETDEWEB)
Glass, H. [Cellnet, Alpharetta, GA (United States)
2006-07-01
Mesh network applications are used by utilities for metering, demand response, and mobile workforce management. This presentation provided an overview of a multi-dimensional mesh application designed to offer improved scalability and higher throughput in advanced metering infrastructure (AMI) systems. Mesh applications can be used in AMI for load balancing and forecasting, as well as for distribution and transmission planning. New revenue opportunities can be realized through the application's ability to improve notification and monitoring services, and customer service communications. Mesh network security features include data encryption, data fragmentation and the automatic re-routing of data. In order to use mesh network applications, networks must have sufficient bandwidth and provide flexibility at the endpoint layer to support multiple devices from multiple vendors, as well as support multiple protocols. It was concluded that smart meters will not enable energy response solutions without an underlying AMI that is reliable, scalable and self-healing. .refs., tabs., figs.
Safety aspects and operating experience of LWR plants in Japan
International Nuclear Information System (INIS)
Aoki, S.; Yoshioka, T.; Toyota, M.; Hinoki, M.
1977-01-01
To develop nuclear power generation for the future, it is necessary to put further emphasis on safety assurance and to endeavour to devise measures to improve plant availability, based on the careful analysis of causes that reduce plant availability. The paper discusses the results of studies on the following items from such viewpoints: (1) Safety and operating experience of LWR nuclear power plants in Japan: operating experience with LWRs; improvements in LWR design during the past ten years; analysis of the factors affecting plant availability; (2) Assurance of safety and measures to increase availability: measures for safety and environmental protection; measures to reduce radiation exposure of employees; appropriateness of maintenance and inspection work; measures to increase plant availability; measures to improve reliability of equipment and components; (3) Future technical problems. (author)
Improving the safety of LWR power plants. Final report
International Nuclear Information System (INIS)
1980-04-01
This report documents the results of the Study to identify current, potential research issues and efforts for improving the safety of Light Water Reactor (LWR) power plants. This final report describes the work accomplished, the results obtained, the problem areas, and the recommended solutions. Specifically, for each of the issues identified in this report for improving the safety of LWR power plants, a description is provided in detail of the safety significance, the current status (including information sources, status of technical knowledge, problem solution and current activities), and the suggestions for further research and development. Further, the issues are ranked for action into high, medium, and low priority with respect to primarily (a) improved safety (e.g. potential reduction in public risk and occupational exposure), and secondly (b) reduction in safety-related costs
Development of top nozzle for Korean standard LWR fuel
Energy Technology Data Exchange (ETDEWEB)
Lee, S. K.; Kim, I. K.; Choi, K. S.; Kim, Y. H.; Lee, J. N.; Kim, H. K. [KNFC, Taejon (Korea, Republic of)
2001-10-01
Performance evaluation was executed for each component and its assembly for the deduced Top Nozzles to develop the new Top Nozzle for LWR. This new Top Nozzle is composed of the optimum components among the derived Top Nozzles that have been evaluated in the viewpoint of structural integrity, simpleness of dismantle and assembly, manufacturability etc. In this study, the developed Top Nozzle satisfied all the related design criteria. In special, it makes fuel repair time reduced by assembling and disassembling itself as one body, and improves Fuel Assembly holddown ability by revising the design parameters of its spring and the structural integrity through the betterment of its geometrical shpae of Flange and Holddown Plate as compared with the existing LWR Top Nozzles.
LWR aerosol containment experiments (LACE) program and initial test results
International Nuclear Information System (INIS)
Muhlestein, L.D.; Hilliard, R.K.; Bloom, G.R.; McCormack, J.D.; Rahn, F.J.
1985-01-01
The LWR aerosol containment experiments (LACE) program is described. The LACE program is being performed at the Hanford Engineer Development Laboratory (operated by Westinghouse Hanford Company) and the initial tests are sponsored by EPRI. The objectives of the LACE program are: to demonstrate, at large-scale, inherent radioactive aerosol retention behavior for postulated high consequence LWR accident situations; and to provide a data base to be used for aerosol behavior . Test results from the first phase of the LACE program are presented and discussed. Three large-scale scoping tests, simulating a containment bypass accident sequence, demonstrated the extent of agglomeration and deposition of aerosols occurring in the pipe pathway and vented auxiliary building under realistic accident conditions. Parameters varied during the scoping tests were aerosol type and steam condensation
Development of LWR fuel performance code FEMAXI-6
International Nuclear Information System (INIS)
Suzuki, Motoe
2006-01-01
LWR fuel performance code: FEMAXI-6 (Finite Element Method in AXIs-symmetric system) is a representative fuel analysis code in Japan. Development history, background, design idea, features of model, and future are stated. Characteristic performance of LWR fuel and analysis code, what is model, development history of FEMAXI, use of FEMAXI code, fuel model, and a special feature of FEMAXI model is described. As examples of analysis, PCMI (Pellet-Clad Mechanical Interaction), fission gas release, gap bonding, and fission gas bubble swelling are reported. Thermal analysis and dynamic analysis system of FEMAXI-6, function block at one time step of FEMAXI-6, analytical example of PCMI in the output increase test by FEMAXI-III, analysis of fission gas release in Halden reactor by FEMAXI-V, comparison of the center temperature of fuel in Halden reactor, and analysis of change of diameter of fuel rod in high burn up BWR fuel are shown. (S.Y.)
Review and comparison of WWER and LWR Codes and Standards
International Nuclear Information System (INIS)
Buckthorpe, D.; Tashkinov, A.; Brynda, J.; Davies, L.M.; Cueto-Felgeueroso, C.; Detroux, P.; Bieniussa, K.; Guinovart, J.
2003-01-01
The results of work on a collaborative project on comparison of Codes and Standards used for safety related components of the WWER and LWR type reactors is presented. This work was performed on behalf of the European Commission, Working Group Codes and Standards and considers areas such as rules, criteria and provisions, failure mechanisms , derivation and understanding behind the fatigue curves, piping, materials and aging, manufacturing and ISI. WWERs are essentially designed and constructed using the Russian PNAE Code together with special provisions in a few countries (e.g. Czech Republic) from national standards. The LWR Codes have a strong dependence on the ASME Code. Also within Western Europe other codes are used including RCC-M, KTA and British Standards. A comparison of procedures used in all these codes and standards have been made to investigate the potential for equivalencies between the codes and any grounds for future cooperation between eastern and western experts in this field. (author)
Thermal conductivity of heterogeneous LWR MOX fuels
Staicu, D.; Barker, M.
2013-11-01
It is generally observed that the thermal conductivity of LWR MOX fuel is lower than that of pure UO2. For MOX, the degradation is usually only interpreted as an effect of the substitution of U atoms by Pu. This hypothesis is however in contradiction with the observations of Duriez and Philiponneau showing that the thermal conductivity of MOX is independent of the Pu content in the ranges 3-15 and 15-30 wt.% PuO2 respectively. Attributing this degradation to Pu only implies that stoichiometric heterogeneous MOX can be obtained, while we show that any heterogeneity in the plutonium distribution in the sample introduces a variation in the local stoichiometry which in turn has a strong impact on the thermal conductivity. A model quantifying this effect is obtained and a new set of experimental results for homogeneous and heterogeneous MOX fuels is presented and used to validate the proposed model. In irradiated fuels, this effect is predicted to disappear early during irradiation. The 3, 6 and 10 wt.% Pu samples have a similar thermal conductivity. Comparison of the results for this homogeneous microstructure with MIMAS (heterogeneous) fuel of the same composition showed no difference for the Pu contents of 3, 5.9, 6, 7.87 and 10 wt.%. A small increase of the thermal conductivity was obtained for 15 wt.% Pu. This increase is of about 6% when compared to the average of the values obtained for 3, 6 and 10 wt.% Pu. For comparison purposes, Duriez also measured the thermal conductivity of FBR MOX with 21.4 wt.% Pu with O/M = 1.982 and a density close to 95% TD and found a value in good agreement with the estimation obtained using the formula of Philipponneau [8] for FBR MOX, and significantly lower than his results corresponding to the range 3-15 wt.% Pu. This difference in thermal conductivity is of about 20%, i.e. higher than the measurement uncertainties.Thus, a significant difference was observed between FBR and PWR MOX fuels, but was not explained. This difference
Equipment designs for the spent LWR fuel dry storage demonstration
International Nuclear Information System (INIS)
Steffen, R.J.; Kurasch, D.H.; Hardin, R.T.; Schmitten, P.F.
1980-01-01
In conjunction with the Spent Fuel Handling and Packaging Program (SFHPP) equipment has been designed, fabricated and successfully utilized to demonstrate the packaging and interim dry storage of spent LWR fuel. Surface and near surface storage configurations containing PWR fuel assemblies are currently on test and generating baseline data. Specific areas of hardware design focused upon include storage cell components and the support related equipment associated with encapsulation, leak testing, lag storage, and emplacement operations
Safety-related LWR research. Annual report 1993
International Nuclear Information System (INIS)
Hueper, R.
1994-06-01
The reactor safety R and D work of the Karlsruhe Nuclear Research Centre (KfK) has been part of the Nuclear Safety Research Project (PSF) since 1990. The present annual report 1993 summarizes the results on LWR safety. The research tasks are coordinated in agreement with internal and external working groups. The contributions to this report correspond to the status at the end of 1993. (orig./HP) [de
Baseline descriptions for LWR spent fuel storage, handling, and transportation
International Nuclear Information System (INIS)
Moyer, J.W.; Sonnier, C.S.
1978-04-01
Baseline descriptions for the storage, handling, and transportation of reactor spent fuel are provided. The storage modes described include light water reactor (LWR) pools, away-from-reactor basins, dry surface storage, reprocessing-facility interim storage pools, and deep geologic storage. Land and water transportation are also discussed. This work was sponsored by the Department of Energy/Office of Safeguards and Security as part of the Sandia Laboratories Fixed Facility Physical Protection Program. 45 figs, 4 tables
Baseline descriptions for LWR spent fuel storage, handling, and transportation
Energy Technology Data Exchange (ETDEWEB)
Moyer, J.W.; Sonnier, C.S.
1978-04-01
Baseline descriptions for the storage, handling, and transportation of reactor spent fuel are provided. The storage modes described include light water reactor (LWR) pools, away-from-reactor basins, dry surface storage, reprocessing-facility interim storage pools, and deep geologic storage. Land and water transportation are also discussed. This work was sponsored by the Department of Energy/Office of Safeguards and Security as part of the Sandia Laboratories Fixed Facility Physical Protection Program. 45 figs, 4 tables.
Standard casks for the transport of LWR spent fuel
International Nuclear Information System (INIS)
Blum, P.
1986-01-01
During the past decade, TRANSNUCLEAIRE has developed, licensed and marketed a family of standard casks for the transport of spent fuel from LWR reactors to reprocessing plants and the ancillary equipments necessary for their operation and transport. A large number of these casks have been manufactured in different countries and are presently used for european and intercontinental transports. The main advantages of these casks are: large payload, moderate cost, reliability, standardisation facilitating fabrication, operation and spare part supply [fr
Investigation of valve failure problems in LWR power plants
International Nuclear Information System (INIS)
1980-04-01
An analysis of component failures from information in the computerized Nuclear Safety Information Center (NSIC) data bank shows that for both PWR and BWR plants the component category most responsible for approximately 19.3% of light water reactor (LWR) power plant shutdowns. This investigation by Burns and Roe, Inc. shows that the greatest cause of shutdowns in LWRs due to valve failures is leakage from valve stem packing. Both BWR plants and PWR plants have stem leakage problems
Modular approach to LWR in-core fuel management
International Nuclear Information System (INIS)
Urli, N.; Pevec, D.; Coffou, E.; Petrovic, B.
1980-01-01
The most important methods in the LWR in-core fuel management are reviewed. A modular approach and optimization by use of infinite multiplication factor and power form-factor are favoured. A computer program for rotation of fuel assemblies at reloads has been developed which improves further fuel economy and reliability of nuclear power plants. The program has been tested on the PWR core and showed to decrease the power form-factors and flatten the radial power distribution. (author)
Coarse graining for synchronization in directed networks
Zeng, An; Lü, Linyuan
2011-05-01
Coarse-graining model is a promising way to analyze and visualize large-scale networks. The coarse-grained networks are required to preserve statistical properties as well as the dynamic behaviors of the initial networks. Some methods have been proposed and found effective in undirected networks, while the study on coarse-graining directed networks lacks of consideration. In this paper we proposed a path-based coarse-graining (PCG) method to coarse grain the directed networks. Performing the linear stability analysis of synchronization and numerical simulation of the Kuramoto model on four kinds of directed networks, including tree networks and variants of Barabási-Albert networks, Watts-Strogatz networks, and Erdös-Rényi networks, we find our method can effectively preserve the network synchronizability.
Development of information management system on LWR spent fuel
International Nuclear Information System (INIS)
Lee, B. D.; Lee, S. H.; Song, D. Y.; Jeon, I.; Park, S. J.; Seo, D. S.
2002-01-01
LWRs in Korea should manage all the information of spent fuel to implement the obligations under Korea-IAEA safeguards agreement and to perform the nuclear material accountancy work at the facility level. The information management system on LWR spent fuel was developed to manage all movement records from receipt to shipment of LWR fuels, and to get the necessary information such as nuclear fuel inventory lists and status, maps of fresh fuel storage, reactor and spent fuel pool, receipt and shipment records and so on. This information management system has a function to setup the system environments to cover the various kinds of storage types for all LWRs ; reactor, spent fuel pool and fresh fuel storage. The movements of nuclear fuel between the storages can be easily done by double click of the mouse to the destination. It also has a several error checking routines for maintaining the correct accounting data. Using this information management system of LWR spent fuel, facility operators can perform efficiently and effectively the safeguards related works including nuclear material accountancy at each facility
Preliminary Study for Radioactivity Evaluation of MSR compared with LWR
Energy Technology Data Exchange (ETDEWEB)
Lee, Geun Hyeong; Kim, Hee Reyoung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)
2014-05-15
LWR uses fuel as {sup 235}U and fissile material as solid (enriched uranium). Those cannot control its component artificially and hard to change fuel frequently. Therefore this fuel remains as much as possible. That makes risk of high radiation leakage because of long neutron irradiation time. On the other hand, MSR (Molten Salt Reactor) uses fuel as thorium-uranium; fissile {sup 233}U when {sup 232}Th absorbs one neutron, and fissile material as liquid (molten salt). It has plenty of benefits respect to radioactive safety. It leads nuclear fuel dump when accident happens, diminishes basic fission substances' radiation and even the cost (Th exist 3∼4 times more on the earth compared with natural uranium). Source term is much lower than conventional LWR in order to processing time. Radiation exposure from volatile fission products in severe accidents is thought to be negligible due to the continuous removal mechanism. The generation of high level radioactive wastes from MSR is estimated to be much smaller than that of conventional LWR because of its less converting probability of thorium to minor actinides. It was thought the fundamental approach to MSR would make it possible to realize the safety of reactor when considering the severe accidents affecting on nuclear power plants due to natural disaster.
Preliminary Study for Radioactivity Evaluation of MSR compared with LWR
International Nuclear Information System (INIS)
Lee, Geun Hyeong; Kim, Hee Reyoung
2014-01-01
LWR uses fuel as 235 U and fissile material as solid (enriched uranium). Those cannot control its component artificially and hard to change fuel frequently. Therefore this fuel remains as much as possible. That makes risk of high radiation leakage because of long neutron irradiation time. On the other hand, MSR (Molten Salt Reactor) uses fuel as thorium-uranium; fissile 233 U when 232 Th absorbs one neutron, and fissile material as liquid (molten salt). It has plenty of benefits respect to radioactive safety. It leads nuclear fuel dump when accident happens, diminishes basic fission substances' radiation and even the cost (Th exist 3∼4 times more on the earth compared with natural uranium). Source term is much lower than conventional LWR in order to processing time. Radiation exposure from volatile fission products in severe accidents is thought to be negligible due to the continuous removal mechanism. The generation of high level radioactive wastes from MSR is estimated to be much smaller than that of conventional LWR because of its less converting probability of thorium to minor actinides. It was thought the fundamental approach to MSR would make it possible to realize the safety of reactor when considering the severe accidents affecting on nuclear power plants due to natural disaster
Energy profit ratio on LWR by uranium recycles
International Nuclear Information System (INIS)
Amano, Osamu; Uno, Takeki; Matsushima, Jun
2009-01-01
Energy profit ratio is defined as the ratio of output energy/input system total energy. In case of electric power generation, input energy is a total for fuel such as uranium mining and enrichment, fuel transportation, build nuclear power plant, M and O and for disposal waste and decommission of reactor vessel. Output energy is the total electricity on LWR during the plant life. EPR on both PWR and BWR is high value using gas centrifuge enrichment compared other type of electric power generation such as a thermal power, a hydraulic power, a wind power and a photovoltaic power. How is the EPR on LWR by MOX? We need understanding the energy of reprocessing spent fuel, MOX fuel fabrication, low level waste disposal and high level radioactive glass disposal. As we show the material balance for two cases, the first is the case of long term storage and reprocessing before FBR, the second is the MOX fuel cycle on LWR plant. The MOX fuel recycle is better EPR value rather than the case of long term storage and reprocessing before FBR (LTSRBF). At the gaseous diffusion enrichment case, MOX fuel recycle has 15 to 18% higher EPR value than LTSRBF. At the gas centrifuge enrichment case the MOX fuel recycle has 17 to 18 higher EPR value than LTSRBF. MOX fuel recycle decreases the uranium mining and refine mass, enrichment separative work and the spent fuel interim storage. It tells us the MOX fuel recycle is good way from view of EPR. (author)
Development of information management system on LWR spent fuel
Energy Technology Data Exchange (ETDEWEB)
Lee, B. D.; Lee, S. H.; Song, D. Y.; Jeon, I.; Park, S. J.; Seo, D. S. [KAERI, Taejon (Korea, Republic of)
2002-10-01
LWRs in Korea should manage all the information of spent fuel to implement the obligations under Korea-IAEA safeguards agreement and to perform the nuclear material accountancy work at the facility level. The information management system on LWR spent fuel was developed to manage all movement records from receipt to shipment of LWR fuels, and to get the necessary information such as nuclear fuel inventory lists and status, maps of fresh fuel storage, reactor and spent fuel pool, receipt and shipment records and so on. This information management system has a function to setup the system environments to cover the various kinds of storage types for all LWRs ; reactor, spent fuel pool and fresh fuel storage. The movements of nuclear fuel between the storages can be easily done by double click of the mouse to the destination. It also has a several error checking routines for maintaining the correct accounting data. Using this information management system of LWR spent fuel, facility operators can perform efficiently and effectively the safeguards related works including nuclear material accountancy at each facility.
Simulated Fission Gas Behavior in Silicide Fuel at LWR Conditions
Energy Technology Data Exchange (ETDEWEB)
Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States); Mo, Kun [Argonne National Lab. (ANL), Argonne, IL (United States); Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States); Harp, Jason [Argonne National Lab. (ANL), Argonne, IL (United States)
2016-09-15
As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U_{3}Si_{2}) at LWR conditions needs to be well-understood. However, existing experimental post-irradiation examination (PIE) data are limited to the research reactor conditions, which involve lower fuel temperature compared to LWR conditions. This lack of appropriate experimental data significantly affects the development of fuel performance codes that can precisely predict the microstructure evolution and property degradation at LWR conditions, and therefore evaluate the qualification of U_{3}Si_{2} as an AFT for LWRs. Considering the high cost, long timescale, and restrictive access of the in-pile irradiation experiments, this study aims to utilize ion irradiation to simulate the inpile behavior of the U_{3}Si_{2} fuel. Both in situ TEM ion irradiation and ex situ high-energy ATLAS ion irradiation experiments were employed to simulate different types of microstructure modifications in U_{3}Si_{2}. Multiple PIE techniques were used or will be used to quantitatively analyze the microstructure evolution induced by ion irradiation so as to provide valuable reference for the development of fuel performance code prior to the availability of the in-pile irradiation data.
Assessment of LWR piping design loading based on plant operating experience
International Nuclear Information System (INIS)
Svensson, P.O.
1980-08-01
The objective of this study has been to: (1) identify current Light Water Reactor (LWR) piping design load parameters, (2) identify significant actual LWR piping loads from plant operating experience, (3) perform a comparison of these two sets of data and determine the significance of any differences, and (4) make an evaluation of the load representation in current LWR piping design practice, in view of plant operating experience with respect to piping behavior and response to loading
Streaming Compression of Hexahedral Meshes
Energy Technology Data Exchange (ETDEWEB)
Isenburg, M; Courbet, C
2010-02-03
We describe a method for streaming compression of hexahedral meshes. Given an interleaved stream of vertices and hexahedral our coder incrementally compresses the mesh in the presented order. Our coder is extremely memory efficient when the input stream documents when vertices are referenced for the last time (i.e. when it contains topological finalization tags). Our coder then continuously releases and reuses data structures that no longer contribute to compressing the remainder of the stream. This means in practice that our coder has only a small fraction of the whole mesh in memory at any time. We can therefore compress very large meshes - even meshes that do not file in memory. Compared to traditional, non-streaming approaches that load the entire mesh and globally reorder it during compression, our algorithm trades a less compact compressed representation for significant gains in speed, memory, and I/O efficiency. For example, on the 456k hexahedra 'blade' mesh, our coder is twice as fast and uses 88 times less memory (only 3.1 MB) with the compressed file increasing about 3% in size. We also present the first scheme for predictive compression of properties associated with hexahedral cells.
Mesh Adaptation and Shape Optimization on Unstructured Meshes, Phase I
National Aeronautics and Space Administration — In this SBIR CRM proposes to implement the entropy adjoint method for solution adaptive mesh refinement into the Loci/CHEM unstructured flow solver. The scheme will...
Parareal in time 3D numerical solver for the LWR Benchmark neutron diffusion transient model
Energy Technology Data Exchange (ETDEWEB)
Baudron, Anne-Marie, E-mail: anne-marie.baudron@cea.fr [Laboratoire de Recherche Conventionné MANON, CEA/DEN/DANS/DM2S and UPMC-CNRS/LJLL (France); CEA-DRN/DMT/SERMA, CEN-Saclay, 91191 Gif sur Yvette Cedex (France); Lautard, Jean-Jacques, E-mail: jean-jacques.lautard@cea.fr [Laboratoire de Recherche Conventionné MANON, CEA/DEN/DANS/DM2S and UPMC-CNRS/LJLL (France); CEA-DRN/DMT/SERMA, CEN-Saclay, 91191 Gif sur Yvette Cedex (France); Maday, Yvon, E-mail: maday@ann.jussieu.fr [Sorbonne Universités, UPMC Univ Paris 06, UMR 7598, Laboratoire Jacques-Louis Lions and Institut Universitaire de France, F-75005, Paris (France); Laboratoire de Recherche Conventionné MANON, CEA/DEN/DANS/DM2S and UPMC-CNRS/LJLL (France); Brown Univ, Division of Applied Maths, Providence, RI (United States); Riahi, Mohamed Kamel, E-mail: riahi@cmap.polytechnique.fr [Laboratoire de Recherche Conventionné MANON, CEA/DEN/DANS/DM2S and UPMC-CNRS/LJLL (France); CMAP, Inria-Saclay and X-Ecole Polytechnique, Route de Saclay, 91128 Palaiseau Cedex (France); Salomon, Julien, E-mail: salomon@ceremade.dauphine.fr [CEREMADE, Univ Paris-Dauphine, Pl. du Mal. de Lattre de Tassigny, F-75016, Paris (France)
2014-12-15
In this paper we present a time-parallel algorithm for the 3D neutrons calculation of a transient model in a nuclear reactor core. The neutrons calculation consists in numerically solving the time dependent diffusion approximation equation, which is a simplified transport equation. The numerical resolution is done with finite elements method based on a tetrahedral meshing of the computational domain, representing the reactor core, and time discretization is achieved using a θ-scheme. The transient model presents moving control rods during the time of the reaction. Therefore, cross-sections (piecewise constants) are taken into account by interpolations with respect to the velocity of the control rods. The parallelism across the time is achieved by an adequate use of the parareal in time algorithm to the handled problem. This parallel method is a predictor corrector scheme that iteratively combines the use of two kinds of numerical propagators, one coarse and one fine. Our method is made efficient by means of a coarse solver defined with large time step and fixed position control rods model, while the fine propagator is assumed to be a high order numerical approximation of the full model. The parallel implementation of our method provides a good scalability of the algorithm. Numerical results show the efficiency of the parareal method on large light water reactor transient model corresponding to the Langenbuch–Maurer–Werner benchmark.
Learning to Play Efficient Coarse Correlated Equilibria
Borowski, Holly P.
2018-03-10
The majority of the distributed learning literature focuses on convergence to Nash equilibria. Coarse correlated equilibria, on the other hand, can often characterize more efficient collective behavior than even the best Nash equilibrium. However, there are no existing distributed learning algorithms that converge to specific coarse correlated equilibria. In this paper, we provide one such algorithm, which guarantees that the agents’ collective joint strategy will constitute an efficient coarse correlated equilibrium with high probability. The key to attaining efficient correlated behavior through distributed learning involves incorporating a common random signal into the learning environment.
Mersiline mesh in premaxillary augmentation.
Foda, Hossam M T
2005-01-01
Premaxillary retrusion may distort the aesthetic appearance of the columella, lip, and nasal tip. This defect is characteristically seen in, but not limited to, patients with cleft lip nasal deformity. This study investigated 60 patients presenting with premaxillary deficiencies in which Mersiline mesh was used to augment the premaxilla. All the cases had surgery using the external rhinoplasty technique. Two methods of augmentation with Mersiline mesh were used: the Mersiline roll technique, for the cases with central symmetric deficiencies, and the Mersiline packing technique, for the cases with asymmetric deficiencies. Premaxillary augmentation with Mersiline mesh proved to be simple technically, easy to perform, and not associated with any complications. Periodic follow-up evaluation for a mean period of 32 months (range, 12-98 months) showed that an adequate degree of premaxillary augmentation was maintained with no clinically detectable resorption of the mesh implant.
GENERATION OF IRREGULAR HEXAGONAL MESHES
Directory of Open Access Journals (Sweden)
Vlasov Aleksandr Nikolaevich
2012-07-01
Decomposition is performed in a constructive way and, as option, it involves meshless representation. Further, this mapping method is used to generate the calculation mesh. In this paper, the authors analyze different cases of mapping onto simply connected and bi-connected canonical domains. They represent forward and backward mapping techniques. Their potential application for generation of nonuniform meshes within the framework of the asymptotic homogenization theory is also performed to assess and project effective characteristics of heterogeneous materials (composites.
Cignoni, Paolo; Pietroni, Nico; Malomo, Luigi
2014-01-01
Mesh joinery is an innovative method to produce illustrative shape approximations suitable for fabrication. Mesh joinery is capable of producing complex fabricable structures in an efficient and visually pleasing manner. We represent an input geometry as a set of planar pieces arranged to compose a rigid structure, by exploiting an efficient slit mechanism. Since slices are planar, to fabricate them a standard 2D cutting system is enough. We automatically arrange slices according to a smooth ...
Method and system for mesh network embedded devices
Wang, Ray (Inventor)
2009-01-01
A method and system for managing mesh network devices. A mesh network device with integrated features creates an N-way mesh network with a full mesh network topology or a partial mesh network topology.
Mesh versus non-mesh repair of ventral abdominal hernias
International Nuclear Information System (INIS)
Jawaid, M.A.; Talpur, A.H.
2008-01-01
To investigate the relative effectiveness of mesh and suture repair of ventral abdominal hernias in terms of clinical outcome, quality of life and rate of recurrence in both the techniques. This is a retrospective descriptive analysis of 236 patients with mesh and non-mesh repair of primary ventral hernias performed between January 2000 to December 2004 at Surgery Department, Liaquat University of Medical and Health Sciences, Jamshoro. The record sheets of the patients were analyzed and data retrieved to compare the results of both techniques for short-term and long-term results. The data retrieved is statistically analyzed on SPSS version 11. There were 43 (18.22%) males and 193 (81.77%) females with a mean age of 51.79 years and a range of 59 (81-22). Para-umbilical hernia was the commonest of ventral hernia and accounted for 49.8% (n=118) of the total study population followed by incisional hernia comprising 24% (n=57) of the total number. There was a significant difference in the recurrent rate at 3 years interval with 23/101 (22.77%) recurrences in suture-repaired subjects compared to 10/135 (7.40%) in mesh repair group. Chronic pain lasting up to 1-2 years was noted in 14 patients with suture repair. Wound infection is comparatively more common (8.14%) in mesh group. The other variables such as operative and postoperative complications, total hospital stay and quality of life is also discussed. Mesh repair of ventral hernia is much superior to non-mesh suture repair in terms of recurrence and overall outcome. (author)
Learning to Play Efficient Coarse Correlated Equilibria
Borowski, Holly P.; Marden, Jason R.; Shamma, Jeff S.
2018-01-01
The majority of the distributed learning literature focuses on convergence to Nash equilibria. Coarse correlated equilibria, on the other hand, can often characterize more efficient collective behavior than even the best Nash equilibrium. However
NEW RSW & Wall Coarse Tet Only Grid
National Aeronautics and Space Administration — This is the RSW Coarse Tet Only grid with the root viscous tunnel wall. This grid is for a node-based unstructured solver. Quad Surface Faces= 0 Tria Surface Faces=...
NEW RSW & Wall Coarse Mixed Element Grid
National Aeronautics and Space Administration — This is the Coarse Mixed Element Grid for the RSW with a viscous wall at the root. This grid is for a node-based unstructured solver. Quad Surface Faces= 9728 Tria...
Coarse Thinking and Pricing a Financial Option
Siddiqi, Hammad
2009-01-01
Mullainathan et al [Quarterly Journal of Economics, May 2008] present a formalization of the concept of coarse thinking in the context of a model of persuasion. The essential idea behind coarse thinking is that people put situations into categories and the values assigned to attributes in a given situation are affected by the values of corresponding attributes in other co-categorized situations. We derive a new option pricing formula based on the assumption that the market consists of coars...
User Manual for the PROTEUS Mesh Tools
Energy Technology Data Exchange (ETDEWEB)
Smith, Micheal A. [Argonne National Lab. (ANL), Argonne, IL (United States); Shemon, Emily R. [Argonne National Lab. (ANL), Argonne, IL (United States)
2015-06-01
This report describes the various mesh tools that are provided with the PROTEUS code giving both descriptions of the input and output. In many cases the examples are provided with a regression test of the mesh tools. The most important mesh tools for any user to consider using are the MT_MeshToMesh.x and the MT_RadialLattice.x codes. The former allows the conversion between most mesh types handled by PROTEUS while the second allows the merging of multiple (assembly) meshes into a radial structured grid. Note that the mesh generation process is recursive in nature and that each input specific for a given mesh tool (such as .axial or .merge) can be used as “mesh” input for any of the mesh tools discussed in this manual.
User Manual for the PROTEUS Mesh Tools
International Nuclear Information System (INIS)
Smith, Micheal A.; Shemon, Emily R.
2015-01-01
This report describes the various mesh tools that are provided with the PROTEUS code giving both descriptions of the input and output. In many cases the examples are provided with a regression test of the mesh tools. The most important mesh tools for any user to consider using are the MT M eshToMesh.x and the MT R adialLattice.x codes. The former allows the conversion between most mesh types handled by PROTEUS while the second allows the merging of multiple (assembly) meshes into a radial structured grid. Note that the mesh generation process is recursive in nature and that each input specific for a given mesh tool (such as .axial or .merge) can be used as ''mesh'' input for any of the mesh tools discussed in this manual.
Measurement and characterization of fission products released from LWR fuel
International Nuclear Information System (INIS)
Osborne, M.F.; Collins, J.L.; Lorenz, R.A.; Norwood, K.S.; Strain, R.V.
1984-01-01
Samples of commercial LWR fuel have been heated under simulated accident conditions to determine the extent and the chemical forms of fission product release. This project was sponsored by the USNRC under a broad program of reactor safety studies. Of the five tests discussed, the fractional releases of Kr, I, and Cs varied from approx. 2% at 1400 0 C to >50% at 2000 0 C; much smaller fractions of Ru, Ag, Sb, and Te were measured in some tests. The major chemical forms in the effluent appeared to include CsI, CsOH, Sb, Te, and Ag
Conceptual design of a spent LWR fuel recycle complex
International Nuclear Information System (INIS)
Kirk, B.H.
1980-01-01
Purpose was to design a licensable facility, to make cost-benefit analyses of alternatives, and to aid in developing licensing criteria. The Savannah River Plant was taken to be the site for the recycle complex. The spent LWR fuel will be processed through the plant at the rate of 3000 metric tons of heavy metal per year. The following aspects of the complex are discussed: operation, maintenance, co-conversion (Coprecal), waste disposal, off-gas treatment, ventilation, safeguards, accounting, equipment and fuel fabrication. Differences between the co-processing case and the separated streams case are discussed. 44 figures
Issues in risk analysis of passive LWR designs
International Nuclear Information System (INIS)
Youngblood, R.W.; Pratt, W.T.; Amico, P.J.; Gallagher, D.
1992-01-01
This paper discusses issues which bear on the question of how safety is to be demonstrated for ''simplified passive'' light water reactor (LWR) designs. First, a very simplified comparison is made between certain systems in today's plants. comparable systems in evolutionary designs, and comparable systems in the simplified passives. in order to introduce the issues. This discussion is not intended to describe the designs comprehensively, but is offered only to show why certain issues seem to be important in these particular designs. Next, an important class of accident sequences is described; finally, based on this discussion, some priorities in risk analysis are presented and discussed
Hamor-2: a computer code for LWR inventory calculation
International Nuclear Information System (INIS)
Guimaraes, L.N.F.; Marzo, M.A.S.
1985-01-01
A method for calculating the accuracy inventory of LWR reactors is presented. This method uses the Hamor-2 computer code. Hamor-2 is obtained from the coupling of two other computer codes Hammer-Techion and Origen-2 for testing Hamor-2, its results were compared to concentration values measured from activides of two PWR reactors; Kernkraftwerk Obrighein (KWO) and H.B. Robinson (HBR). These actinides are U 235 , U 236 , U 238 , Pu 239 , Pu 241 and PU 242 . The computer code Hammor-2 shows better results than the computer code Origem-2, when both are compared with experimental results. (E.G.) [pt
Measurement and characterization of fission products released from LWR fuel
International Nuclear Information System (INIS)
Osborne, M.F.; Collins, J.L.; Lorenz, R.A.; Norwood, K.S.; Strain, R.V.
1984-01-01
Samples of commercial LWR fuel have been heated under simulated accident conditions to determine the extent and the chemical forms of fission product release. Of the five tests discussed, the fractional releases of Kr, I, and Cs varied from proportional 2% at 1400 0 C to >50% at 2000 0 C; much smaller fractions of Ru, Ag, Sb, and Te were measured in some tests. The major chemical forms in the effluent appeared to include CsI, CsOH, Sb, Te, and Ag. (orig./HP)
Nondestructive evaluation of LWR spent fuel shipping casks
International Nuclear Information System (INIS)
Ballard, D.W.
1978-02-01
An analysis of nondestructve testing (NDT) methods currently being used to evaluate the integrity of Light Water Reactor (LWR) spent fuel shipping casks is presented. An assessment of anticipated NDT needs related to breeder reactor cask requirements is included. Specific R and D approaches to probable NDT problem areas such as the evaluation of austenitic stainless steel weldments are outlined. A comprehensive bibliography of current NDT methods for cask evaluation in the USA, Great Britain, Japan and West Germany was compiled for this study
Standard casks for the transport of LWR spent fuel
International Nuclear Information System (INIS)
Blum, P.
1985-01-01
During the past decade, TRANSNUCLEAIRE has developed, licensed and marketed a family of standard casks for the transport of spent fuel from LWR reactors to reprocessing plants and the ancillary equipments necessary for their operation and transport. A large number of these casks have been manufacturer under TRANSNUCLEAIRE supervision in different countries and are presently used for European and intercontinental transports. The main advantages of these casks are: - large payload for considered modes of transport, - moderate cost, - reliability due to the large experience gained by TRANSNUCLEAIRE as concerns fabrication and operation problems, - standardisation facilitating fabrication, operation and spare part supply [fr
A study for small-medium LWR development of JAPC
International Nuclear Information System (INIS)
Okazaki, Toshihiko; Hida, Takahiko; Hoshi, Takashi; Kawahara, Hiroto; Tominaga, Kenji; Asano, Hiromitsu
2011-01-01
LWR (Light Water Reactor) power stations have accumulated many experiences of design, construction and operation. In addition, large-sized reactors have an advantage of economy of scale and 1,000 MWe LWR has therefore become the mainstream reactor in Japan. Meanwhile, introduction of the medium and small-sized LWRs (SMRs) has also been under review in Japan in order to respond to stagnant growth in electricity demand and electricity market liberalization or for investment risk mitigation; however, it has not been realized due to the economic disadvantage of scale. Therefore, JAPC has been developing the concept of SMR (300 MWe - 600 MWe) which is competitive to the large-sized LWR cooperating with Japanese plant makers (Hitachi, Toshiba Corporation and Mitsubishi Heavy Industries), assessing the possibility of realization of SMRs as one of the electric power sources in the future. As the result of the JAPC's study, we developed SMR concepts whose cost and safety are almost equal to large-sized LWR and confirmed technical feasibility of the concept in order to start developing basic design. In this paper, the outline of the SMR concepts and the current development status are presented. Concepts have been developed for two types of SMRs (i.e. BWR and PWR). As for the BWR type, reactor system is simplified by adopting natural circulation core method and CRD falling under gravity in order to downsize the reactor containments. As for the PWR type, the risk of LOCA occurrence is eliminated by unifying the primary system (e.g. incorporating steam generator into reactor). Furthermore, the primary system is simplified by adopting natural circulation core method in operation and containment vessel also become compact for the PWR. As for JAPC's further development of SMRs, key elements of SMR concepts are studied. In addition, the environment surrounding the SMRs has changed in recent years and the one with capacity exceeding 300-600 MWe class or small-sized reactor with
Hydrogen mixing study (HMS) in LWR type containments
International Nuclear Information System (INIS)
Travis, J.R.
1983-01-01
A numerical technique has been developed for calculating the full three-dimensional time-dependent Navier-Stokes equations with multiple speies transport. The method is a modified form of the Implicit Continuous-fluid Eulerian (ICE) technique to solve the governing equations for low Mach number flows where pressure waves and local variations in compression and expansion are not significant. Large density variations, due to thermal and species concentration gradients, are accounted for without the restrictions of the classical Boussinesq approximation. Calculations of the EPRI/HEDL standard problems verify the feasibility of using this finite-difference technique for analyzing hydrogen mixing within LWR containments
Investigation of valve failure problems in LWR power plants
Energy Technology Data Exchange (ETDEWEB)
None
1980-04-01
An analysis of component failures from information in the computerized Nuclear Safety Information Center (NSIC) data bank shows that for both PWR and BWR plants the component category most responsible for approximately 19.3% of light water reactor (LWR) power plant shutdowns. This investigation by Burns and Roe, Inc. shows that the greatest cause of shutdowns in LWRs due to valve failures is leakage from valve stem packing. Both BWR plants and PWR plants have stem leakage problems (BWRs, 21% and PWRs, 34%).
Transmutation of LWR waste actinides in thermal reactors
International Nuclear Information System (INIS)
Gorrell, T.C.
1979-01-01
Recycle of actinides to a reactor for transmutation to fission products is being considered as a possible means of waste disposal. Actinide transmutation calculations were made for two irradiation options in a thermal (LWR) reactor. The cases considered were: all actinides recycled in regular uranium fuel assemblies, and transuranic actinides recycled in separate mixed oxide (MOX) assemblies. When all actinides were recycled in a uranium lattice, a reduction of 62% in the transuranic inventory was achieved after 10 recycles, compared to the inventory accumulated without recycle. When the transuranics from 2 regular uranium assemblies were combined with those recycled from a MOX assembly, the transuranic inventory was reduced 50% after 5 recycles
Integrity of neutron-absorbing components of LWR fuel systems
International Nuclear Information System (INIS)
Bailey, W.J.; Berting, F.M.
1991-03-01
A study of the integrity and behavior of neutron-absorbing components of light-water (LWR) fuel systems was performed by Pacific Northwest Laboratory (PNL) and sponsored by the US Department of Energy (DOE). The components studies include control blades (cruciforms) for boiling-water reactors (BWRs) and rod cluster control assemblies for pressurized-water reactors (PWRs). The results of this study can be useful for understanding the degradation of neutron-absorbing components and for waste management planning and repository design. The report includes examples of the types of degradation, damage, or failures that have been encountered. Conclusions and recommendations are listed. 84 refs
Pie technique of LWR fuel cladding fracture toughness test
International Nuclear Information System (INIS)
Endo, Shinya; Usami, Koji; Nakata, Masahito; Fukuda, Takuji; Numata, Masami; Kizaki, Minoru; Nishino, Yasuharu
2006-01-01
Remote-handling techniques were developed by cooperative research between the Department of Hot Laboratories in the Japan Atomic Energy Research Institute (JAERI) and the Nuclear Fuel Industries Ltd. (NFI) for evaluating the fracture toughness on irradiated LWR fuel cladding. The developed techniques, sample machining by using the electrical discharge machine (EDM), pre-cracking by fatigue tester, sample assembling to the compact tension (CT) shaped test fixture gave a satisfied result for a fracture toughness test developed by NFL. And post-irradiation examination (PIE) using the remote-handling techniques were carried out to evaluate the fracture toughness on BWR spent fuel cladding in the Waste Safety Testing Facility (WASTEF). (author)
International Nuclear Information System (INIS)
Nakai, Satoru; Nishio, Ryuichi; Uchihashi, Masaya; Kaneko, Yoshihisa; Yamashita, Hironobu; Yamaguchi, Atsunori; Aoki, Takayuki
2014-01-01
A sodium cooled fast breeder reactor (FBR) has unique systems and components and different degradation mechanism from light water reactor (LWR) so that need to establish maintenance technology in accordance with its features. The examination of the FBR maintenance technology is carried out in the special committee for considering the maintenance for Monju established in the Japan Society of Maintenology (JSM). As a result of the study such as extraction of Monju maintenance feature, maintenance technology benchmark between Monju and LWR components and survey of LWR maintenance experience, it is clear that principles of maintenance are same as LWR, necessity of LWR maintenance experience reflection and points to be considered in Monju maintenance. The road map to establish a FBR maintenance technology in the technical aspect became clear and it is vital to acquire operation and maintenance experience of the plant to implement this road map, and to establish a fast reactor maintenance. (author)
International Nuclear Information System (INIS)
Yoon, S; Lindstrom, P; Pascucci, V; Manocha, D
2005-01-01
We present a novel method for computing cache-oblivious layouts of large meshes that improve the performance of interactive visualization and geometric processing algorithms. Given that the mesh is accessed in a reasonably coherent manner, we assume no particular data access patterns or cache parameters of the memory hierarchy involved in the computation. Furthermore, our formulation extends directly to computing layouts of multi-resolution and bounding volume hierarchies of large meshes. We develop a simple and practical cache-oblivious metric for estimating cache misses. Computing a coherent mesh layout is reduced to a combinatorial optimization problem. We designed and implemented an out-of-core multilevel minimization algorithm and tested its performance on unstructured meshes composed of tens to hundreds of millions of triangles. Our layouts can significantly reduce the number of cache misses. We have observed 2-20 times speedups in view-dependent rendering, collision detection, and isocontour extraction without any modification of the algorithms or runtime applications
Connectivity editing for quadrilateral meshes
Peng, Chihan; Zhang, Eugene; Kobayashi, Yoshihiro; Wonka, Peter
2011-01-01
We propose new connectivity editing operations for quadrilateral meshes with the unique ability to explicitly control the location, orientation, type, and number of the irregular vertices (valence not equal to four) in the mesh while preserving sharp edges. We provide theoretical analysis on what editing operations are possible and impossible and introduce three fundamental operations to move and re-orient a pair of irregular vertices. We argue that our editing operations are fundamental, because they only change the quad mesh in the smallest possible region and involve the fewest irregular vertices (i.e., two). The irregular vertex movement operations are supplemented by operations for the splitting, merging, canceling, and aligning of irregular vertices. We explain how the proposed highlevel operations are realized through graph-level editing operations such as quad collapses, edge flips, and edge splits. The utility of these mesh editing operations are demonstrated by improving the connectivity of quad meshes generated from state-of-art quadrangulation techniques. © 2011 ACM.
Connectivity editing for quadrilateral meshes
Peng, Chihan
2011-12-12
We propose new connectivity editing operations for quadrilateral meshes with the unique ability to explicitly control the location, orientation, type, and number of the irregular vertices (valence not equal to four) in the mesh while preserving sharp edges. We provide theoretical analysis on what editing operations are possible and impossible and introduce three fundamental operations to move and re-orient a pair of irregular vertices. We argue that our editing operations are fundamental, because they only change the quad mesh in the smallest possible region and involve the fewest irregular vertices (i.e., two). The irregular vertex movement operations are supplemented by operations for the splitting, merging, canceling, and aligning of irregular vertices. We explain how the proposed highlevel operations are realized through graph-level editing operations such as quad collapses, edge flips, and edge splits. The utility of these mesh editing operations are demonstrated by improving the connectivity of quad meshes generated from state-of-art quadrangulation techniques. © 2011 ACM.
Vlyssides, Apostolos G; Mai, Sofia T H; Barampouti, Elli Maria P; Loukakis, Haralampos N
2009-07-01
To estimate the influence of gravel mesh (fine and coarse) and vegetation (Phragmites and Arundo) on the efficiency of a reed bed, a pilot plant was included after the wastewater treatment plant of a cosmetic industry treatment system according to a 22 factorial experimental design. The maximum biochemical oxygen demand (BOD5), chemical oxygen demand (COD) and total phosphorous (TP) reduction was observed in the reactor, where Phragmites and fine gravel were used. In the reactor with Phragmites and coarse gravel, the maximum total Kjeldahl nitrogen (TKN) and total suspended solids (TSS) reduction was observed. The maximum total solids reduction was measured in the reed bed, which was filled with Arundo and coarse gravel. Conclusively, the treatment of a cosmetic industry's wastewater by reed beds as a tertiary treatment method is quite effective.
The dupic fuel cycle synergism between LWR and HWR
International Nuclear Information System (INIS)
Lee, J.S.; Yang, M.S.; Park, H.S.; Lee, H.H.; Kim, K.P.; Sullivan, J.D.; Boczar, P.G.; Gadsby, R.D.
1999-01-01
The DUPIC fuel cycle can be developed as an alternative to the conventional spent fuel management options of direct disposal or plutonium recycle. Spent LWR fuel can be burned again in a HWR by direct refabrication into CANDU-compatible DUPIC fuel bundles. Such a linkage between LWR and HWR can result in a multitude of synergistic effects, ranging from savings of natural uranium to reductions in the amount of spent fuel to be buried in the earth, for a given amount of nuclear electricity generated. A special feature of the DUPIC fuel cycle is its compliance with the 'Spent Fuel Standard' criteria for diversion resistance, throughout the entire fuel cycle. The DUPIC cycle thus has a very high degree of proliferation resistance. The cost penalty due to this technical factor needs to be considered in balance with the overall benefits of the DUPIC fuel cycle. The DUPIC alternative may be able to make a significant contribution to reducing spent nuclear fuel burial in the geosphere, in a manner similar to the contribution of the nuclear energy alternative in reducing atmospheric pollution from fossil fuel combustion. (author)
Effects of cooling time on a closed LWR fuel cycle
International Nuclear Information System (INIS)
Arnold, R. P.; Forsberg, C. W.; Shwageraus, E.
2012-01-01
In this study, the effects of cooling time prior to reprocessing spent LWR fuel has on the reactor physics characteristics of a PWR fully loaded with homogeneously mixed U-Pu or U-TRU oxide (MOX) fuel is examined. A reactor physics analysis was completed using the CASM04e code. A void reactivity feedback coefficient analysis was also completed for an infinite lattice of fresh fuel assemblies. Some useful conclusions can be made regarding the effect that cooling time prior to reprocessing spent LWR fuel has on a closed homogeneous MOX fuel cycle. The computational analysis shows that it is more neutronically efficient to reprocess cooled spent fuel into homogeneous MOX fuel rods earlier rather than later as the fissile fuel content decreases with time. Also, the number of spent fuel rods needed to fabricate one MOX fuel rod increases as cooling time increases. In the case of TRU MOX fuel, with time, there is an economic tradeoff between fuel handling difficulty and higher throughput of fuel to be reprocessed. The void coefficient analysis shows that the void coefficient becomes progressively more restrictive on fuel Pu content with increasing spent fuel cooling time before reprocessing. (authors)
FMDP reactor alternative summary report: Volume 4, Evolutionary LWR alternative
International Nuclear Information System (INIS)
1996-09-01
Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] have become surplus to national defense needs both in the United States and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety, and health (ES ampersand H) consequences if surplus fissile materials are not properly managed. The purpose of this report is to provide schedule, cost, and technical information that will be used to support the Record of Process (ROD). Following the screening process, DOE/MD via its national laboratories initiated a more detailed analysis activity to further evaluate each of the ten plutonium disposition alternatives that survived the screening process. Three ''Alternative Teams,'' chartered by DOE and comprised of technical experts from across the DOE national laboratory complex, conducted these analyses. One team was chartered for each of the major disposition classes (borehole, immobilization, and reactors). During the last year and a half, the Fissile Materials Disposition Program (FMDP) Reactor Alternative Team (RxAT) has conducted extensive analyses of the cost, schedule, technical maturity, S ampersand S, and other characteristics of reactor-based plutonium disposition. The results of the RxAT's analyses of the existing LWR, CANDU, and partially complete LWR alternatives are documented in Volumes 1-3 of this report. This document (Volume 4) summarizes the results of these analyses for the ELWR-based plutonium disposition option
Qualification of ARROTTA code for LWR accident analysis
International Nuclear Information System (INIS)
Huang, P.-H.; Peng, K.Y.; Lin, W.-C.; Wu, J.-Y.
2004-01-01
This paper presents the qualification efforts performed by TPC and INER for the 3-D spatial kinetics code ARROTTA for LWR core transient analysis. TPC and INER started a joint 5 year project in 1989 to establish independent capabilities to perform reload design and transient analysis utilizing state-of-the-art computer programs. As part of the effort, the ARROTTA code was chosen to perform multi-dimensional kinetics calculations such as rod ejection for PWR and rod drop for BWR. To qualify ARROTTA for analysis of FSAR licensing basis core transients, ARROTTA has been benchmarked for the static core analysis against plant measured data and SIMULATE-3 predictions, and for the kinetic analysis against available benchmark problems. The static calculations compared include critical boron concentration, core power distribution, and control rod worth. The results indicated that ARROTTA predictions match very well with plant measured data and SIMULATE-3 predictions. The kinetic benchmark problems validated include NEACRP rod ejection problem, 3-D LMW LWR rod withdrawal/insertion problem, and 3-D LRA BWR transient benchmark problem. The results indicate that ARROTTA's accuracy and stability are excellent as compared to other space-time kinetics codes. It is therefore concluded that ARROTTA provides accurate predictions for multi-dimensional core transient for LWRs. (author)
FMDP reactor alternative summary report: Volume 4, Evolutionary LWR alternative
Energy Technology Data Exchange (ETDEWEB)
NONE
1996-09-01
Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] have become surplus to national defense needs both in the United States and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety, and health (ES&H) consequences if surplus fissile materials are not properly managed. The purpose of this report is to provide schedule, cost, and technical information that will be used to support the Record of Process (ROD). Following the screening process, DOE/MD via its national laboratories initiated a more detailed analysis activity to further evaluate each of the ten plutonium disposition alternatives that survived the screening process. Three ``Alternative Teams,`` chartered by DOE and comprised of technical experts from across the DOE national laboratory complex, conducted these analyses. One team was chartered for each of the major disposition classes (borehole, immobilization, and reactors). During the last year and a half, the Fissile Materials Disposition Program (FMDP) Reactor Alternative Team (RxAT) has conducted extensive analyses of the cost, schedule, technical maturity, S&S, and other characteristics of reactor-based plutonium disposition. The results of the RxAT`s analyses of the existing LWR, CANDU, and partially complete LWR alternatives are documented in Volumes 1-3 of this report. This document (Volume 4) summarizes the results of these analyses for the ELWR-based plutonium disposition option.
Evaluation of LWR fuel rod behavior under operational transient conditions
International Nuclear Information System (INIS)
Nakamura, M.; Hiramoto, K.; Maru, A.
1984-01-01
To evaluate the effects of fission gas flow and diffusion in the fuel-cladding gap on fuel rod thermal and mechanical behaviors in light water reactor (LWR) fuel rods under operational transient conditions, computer sub-programs which can calculate the gas flow and diffusion have been developed and integrated into the LWR fuel rod performance code BEAF. This integrated code also calculates transient temperature distribution in the fuel-pellet and cladding. The integrated code was applied to an analysis of Inter Ramp Project data, which showed that by taking into account the gas flow and diffusion effects, the calculated cladding damage indices predicted for the failed rods in the ramp test were consistent with iodine-SCC (Stress Corrosion Cracking) failure conditions which were obtained from out-of-reactor pressurized tube experiments with irradiated Zircaloy claddings. This consistency was not seen if the gas flow and diffusion effects were neglected. Evaluation were also made for the BWR 8x8 RJ fuel rod temperatures under power ramp conditions. (orig.)
Modelling airborne dispersion of coarse particulate material
International Nuclear Information System (INIS)
Apsley, D.D.
1989-03-01
Methods of modelling the airborne dispersion and deposition of coarse particulates are presented, with the emphasis on the heavy particles identified as possible constituents of releases from damaged AGR fuel. The first part of this report establishes the physical characteristics of the irradiated particulate in airborne emissions from AGR stations. The second part is less specific and describes procedures for extending current dispersion/deposition models to incorporate a coarse particulate component: the adjustment to plume spread parameters, dispersion from elevated sources and dispersion in conjunction with building effects and plume rise. (author)
Flexible fuel cycle system for the transition from LWR to FBR
International Nuclear Information System (INIS)
Fukasawa, Tetsuo; Yamashita, Junichi; Hoshino, Kuniyoshi; Sasahira, Akira; Inoue, Tadashi; Minato, Kazuo; Sato, Seichi
2009-01-01
Japan will deploy commercial fast breeder reactor (FBR) from around 2050 under the suitable conditions for the replacement of light water reactor (LWR) with FBR. The transition scenario from LWR to FBR is investigated in detail and the flexible fuel cycle initiative (FFCI) system has been proposed as a optimum transition system. The FFCI removes ∼95% uranium from LWR spent fuel (SF) in LWR reprocessing and residual material named Recycle Material (RM), which is ∼1/10 volume of original SF and contains ∼50% U, ∼10% Pu and ∼40% other nuclides, is treated in FBR reprocessing to recover Pu and U. If the FBR deployment speed becomes lower, the RM will be stored until the higher speed again. The FFCI has some merits compared with ordinary system that consists of full reprocessing facilities for both LWR and FBR SF during the transition period. The economy is better for FFCI due to the smaller LWR reprocessing facility (no Pu/U recovery and fabrication). The FFCI can supply high Pu concentration RM, which has high proliferation resistance and flexibly respond to FBR introduction rate changes. Volume minimization of LWR SF is possible for FFCI by its conversion to RM. Several features of FFCI were quantitatively evaluated such as Pu mass balance, reprocessing capacities, LWR SF amounts, RM amounts, and proliferation resistance to compare the effectiveness of the FFCI system with other systems. The calculated Pu balance revealed that the FFCI could supply enough but no excess Pu to FBR. These evaluations demonstrated the applicability of FFCI system to the transition period from LWR to FBR cycles. (author)
Multigrid for refined triangle meshes
Energy Technology Data Exchange (ETDEWEB)
Shapira, Yair
1997-02-01
A two-level preconditioning method for the solution of (locally) refined finite element schemes using triangle meshes is introduced. In the isotropic SPD case, it is shown that the condition number of the preconditioned stiffness matrix is bounded uniformly for all sufficiently regular triangulations. This is also verified numerically for an isotropic diffusion problem with highly discontinuous coefficients.
Validating the BISON fuel performance code to integral LWR experiments
Energy Technology Data Exchange (ETDEWEB)
Williamson, R.L., E-mail: Richard.Williamson@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Gamble, K.A., E-mail: Kyle.Gamble@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Perez, D.M., E-mail: Danielle.Perez@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Novascone, S.R., E-mail: Stephen.Novascone@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Pastore, G., E-mail: Giovanni.Pastore@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Gardner, R.J., E-mail: Russell.Gardner@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Hales, J.D., E-mail: Jason.Hales@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Liu, W., E-mail: Wenfeng.Liu@anatech.com [ANATECH Corporation, 5435 Oberlin Dr., San Diego, CA 92121 (United States); Mai, A., E-mail: Anh.Mai@anatech.com [ANATECH Corporation, 5435 Oberlin Dr., San Diego, CA 92121 (United States)
2016-05-15
Highlights: • The BISON multidimensional fuel performance code is being validated to integral LWR experiments. • Code and solution verification are necessary prerequisites to validation. • Fuel centerline temperature comparisons through all phases of fuel life are very reasonable. • Accuracy in predicting fission gas release is consistent with state-of-the-art modeling and the involved uncertainties. • Rod diameter comparisons are not satisfactory and further investigation is underway. - Abstract: BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON's computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to date for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Results demonstrate that (1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, (2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and (3) comparison
Resterilized Polypropylene Mesh for Inguinal Hernia Repair
African Journals Online (AJOL)
2018-04-19
Apr 19, 2018 ... Conclusion: The use of sterilized polypropylene mesh for the repair of inguinal ... and nonabsorbable materials to reduce the tissue–mesh. INTRODUCTION ... which we have been practicing in our center since we introduced ...
Management of complications of mesh surgery.
Lee, Dominic; Zimmern, Philippe E
2015-07-01
Transvaginal placements of synthetic mid-urethral slings and vaginal meshes have largely superseded traditional tissue repairs in the current era because of presumed efficacy and ease of implant with device 'kits'. The use of synthetic material has generated novel complications including mesh extrusion, pelvic and vaginal pain and mesh contraction. In this review, our aim is to discuss the management, surgical techniques and outcomes associated with mesh removal. Recent publications have seen an increase in presentation of these mesh-related complications, and reports from multiple tertiary centers have suggested that not all patients benefit from surgical intervention. Although the true incidence of mesh complications is unknown, recent publications can serve to guide physicians and inform patients of the surgical outcomes from mesh-related complications. In addition, the literature highlights the growing need for a registry to account for a more accurate reporting of these events and to counsel patients on the risk and benefits before proceeding with mesh surgeries.
Characteristics Data Base: Programmer's guide to the LWR Quantities Data Base
International Nuclear Information System (INIS)
Jones, K.E.; Moore, R.S.
1990-08-01
The LWR Quantities Data Base is a menu-driven PC data base developed as part of OCRWM's waste, technical data base on the characteristics of potential repository wastes, which also includes non-LWR spent fuel, high-level and other materials. This programmer's guide completes the documentation for the LWR Quantities Data Base, the user's guide having been published previously. The PC data base itself may be requested from the Oak Ridge National Laboratory, using the order form provided in Volume 1 of publication DOE/RW-0184
6th International Meshing Roundtable '97
Energy Technology Data Exchange (ETDEWEB)
White, D.
1997-09-01
The goal of the 6th International Meshing Roundtable is to bring together researchers and developers from industry, academia, and government labs in a stimulating, open environment for the exchange of technical information related to the meshing process. In the pas~ the Roundtable has enjoyed significant participation born each of these groups from a wide variety of countries. The Roundtable will consist of technical presentations from contributed papers and abstracts, two invited speakers, and two invited panels of experts discussing topics related to the development and use of automatic mesh generation tools. In addition, this year we will feature a "Bring Your Best Mesh" competition and poster session to encourage discussion and participation from a wide variety of mesh generation tool users. The schedule and evening social events are designed to provide numerous opportunities for informal dialog. A proceedings will be published by Sandia National Laboratories and distributed at the Roundtable. In addition, papers of exceptionally high quaIity will be submitted to a special issue of the International Journal of Computational Geometry and Applications. Papers and one page abstracts were sought that present original results on the meshing process. Potential topics include but are got limited to: Unstructured triangular and tetrahedral mesh generation Unstructured quadrilateral and hexahedral mesh generation Automated blocking and structured mesh generation Mixed element meshing Surface mesh generation Geometry decomposition and clean-up techniques Geometry modification techniques related to meshing Adaptive mesh refinement and mesh quality control Mesh visualization Special purpose meshing algorithms for particular applications Theoretical or novel ideas with practical potential Technical presentations from industrial researchers.
Development of a data bank system for LWR integral experiment
International Nuclear Information System (INIS)
Naito, Yoshitaka; Aoyagi, Hideo
1983-01-01
A data bank system for LWR integral experiment has been developed for the purpose of alleviating various efforts associated with the verification of computer codes. The final aim of this system is such that the imput data for the code to be verified can be easily obtained, and the results of calculation can be obtained in the form of the comparison with measurement. Geometry and material composition as well as measured data are stored in the data bank. This data bank system is composed of four sub-programs; (1) registration program, (2) information retrieval program, (3) maintenance program, and (4) figure representation program. In this report, the structure of this data bank system and how to use the system are explained. An example of the use of this system is also included. (Aoki, K.)
Alternatives for managing post LWR reactor nuclear wastes
International Nuclear Information System (INIS)
Platt, A.M.
1976-01-01
The two extremes in the LWR fuel cycle are discarding the spent fuel and recycling the U and Pu to the maximum extent possible. The waste volumes from the two alternatives are compared. A preliminary evaluation is made of the technology available for handling wastes from each step of the fuel cycle. The wastes considered are fuel materials, high--level wastes, other liquids, combustible and non-combustible solids, and non--high--level wastes. Evaluation of processing gaseous wastes indicates that technology is available for capture of Kr and I 2 , but further development is needed for T 2 . Technology for interim storage and geological isolation is considered adequate. An outline is given of the steps in the selection of a final storage site
LWR-PV Surveillance Dosimetry Improvement Program review graphics
International Nuclear Information System (INIS)
McElroy, W.N.; Gold, R.; Gutherie, G.L.
1979-10-01
A primary objective of the multilaboratory program is to prepare an updated and improved set of dosimetry, damage correlation, and the associated reactor analysis ASTM standards for LWR-PV irradiation surveillance programs. Supporting this objective are a series of analytical and experimental validation and calibration studies in Benchmark Neutron Fields, reactor Test Regions, and operating power reactor Surveillance Positions. These studies will establish and certify the precision and accuracy of the measurement and predictive methods which are recommended for use in these standards. Consistent and accurate measurement and data analysis techniques and methods, therefore, will have been developed and validated along with guidelines for required neutron field calculations that are used to (1) correlate changes in material properties with the characteristics of the neutron radiation field and (2) predict pressure vessel steel toughness and embrittlement from power reactor surveillance data
LWR risk management by safety R and D
International Nuclear Information System (INIS)
El-Sheikh, K.A.; Damon, D.R.; Temme, M.I.
1982-01-01
This paper presents a methodology which has been developed for selecting LWR safety RandD projects. The methodology provides ranking of the RandD projects and the RandD budget allocation which minimizes public risk. The methodology contains procedures to identify institutional, organizational, legal, and contractual factors which affect the probabilities of success and use of RandD projects so that these factors can be evaluated and possibly managed.The methodology also contains a nonlinear optimization code to provide the optimum selection of RandD projects and evaluate the sensitivity of this selection to uncertainity in the input data. Application of the methodology to a test case has shown that: 1) commonly used schemes for ranking RandD projects do not necessarily lead to the optimum selection, and 2) the optimum selection is not necessarily strongly sensitive to uncertainty in the input data
Spent LWR fuel leach tests: Waste Isolation Safety Assessment program
International Nuclear Information System (INIS)
Katayama, Y.B.
1979-04-01
Spent light-water-reactor (LWR) fuels with burnups of 54.5, 28 and 9 MWd/kgU were leach-tested in deionized water at 25 0 C. Fuel burnup has no apparent effect on the calculated leach rates based upon the behavior of 137 Cs and 239+240 Pu. A leach test of 54.5 MWd/kgU spent fuel in synthetic sea brine showed that the cesium-based leach rate is lower in sea brine than in deionized water. A rise in the leach rate was observed after approximately 600 d of cumulative leaching. During the rise, the leach rate for all the measured radionuclides become nearly equal. Evidence suggests that exposure of new surfaces to the leachant may cause the increase. As a result, experimental work to study leaching mechanisms of spent fuel has been initiated. 22 figures
Evaluation of management alternatives for LWR hulls and caps
International Nuclear Information System (INIS)
Chaudon, L.; Mehling, O.; Cecille, L.; Thiels, G.; Kowa, S.
1993-01-01
Hulls and caps resulting from the reprocessing of LWR spent fuels represent one of the major sources of alpha-bearing solid waste generated during the nuclear fuel cycle. The Commission of the European Communities has undertaken considerable R and D efforts on the development of advanced treatment and conditioning methods for this type of waste. In view of the encouraging results achieved, the Commission launched a theoretical assessment study on cladding waste management. Six practical or potential schemes were identified and elaborated: direct cementation, decontamination prior to cementation, rolling before cementation, rolling followed by embedding in graphite, compaction, and melting in a cold crucible. The economic aspects of each management option were also investigated. This included the assessment of the plant (treatment, conditioning and interim storage), transport and disposal costs. Further consideration will be required to define the best management option for 'cap' wastes. Transport and disposal costs will also require further analysis from an industrial standpoint
Cost optimization of long-cycle LWR operation
International Nuclear Information System (INIS)
Handwerk, C.S.; Driscoll, M.J.; McMahon, M.V.; Todreas, N.E.
1997-01-01
The continuing emphasis on improvement of plant capacity factor, as a major means to make nuclear energy more cost competitive in the current deregulatory environment, motivates heightened interest in long intra-refueling intervals and high burnup in LWR units. This study examines the economic implications of these trends, to determine the envelope of profitable fuel management tactics. One batch management is found to be significantly more expensive than two-batch management. Parametric studies were carried out varying the most important input parameters. If ultra-high burnup can be achieved, then n = 3 or even n = 4 management may be preferable. For n = 1 or 2, economic performance declines at higher burnups, hence providing no great incentive for moving further in that direction. Values for n > 2 are also attractive because, for a given burnup target, required enrichment decreases as n increases. This study was limited to average batch burnups below 60,000 MWd/MT
Irradiation effects on thermal properties of LWR hydride fuel
Energy Technology Data Exchange (ETDEWEB)
Terrani, Kurt, E-mail: terrani@berkeley.edu [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States); Balooch, Mehdi [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States); Carpenter, David; Kohse, Gordon [Massachusetts Institute of Technology, 138 Albany St., Cambridge, MA 02139 (United States); Keiser, Dennis; Meyer, Mitchell [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Olander, Donald [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States)
2017-04-01
Three hydride mini-fuel rods were fabricated and irradiated at the MIT nuclear reactor with a maximum burnup of 0.31% FIMA or ∼5 MWd/kgU equivalent oxide fuel burnup. Fuel rods consisted of uranium-zirconium hydride (U (30 wt%)ZrH{sub 1.6}) pellets clad inside a LWR Zircaloy-2 tubing. The gap between the fuel and the cladding was filled with lead-bismuth eutectic alloy to eliminate the gas gap and the large temperature drop across it. Each mini-fuel rod was instrumented with two thermocouples with tips that are axially located halfway through the fuel centerline and cladding surface. In-pile temperature measurements enabled calculation of thermal conductivity in this fuel as a function of temperature and burnup. In-pile thermal conductivity at the beginning of test agreed well with out-of-pile measurements on unirradiated fuel and decreased rapidly with burnup.
LIFE vs. LWR: End of the Fuel Cycle
International Nuclear Information System (INIS)
Farmer, J.C.; Blink, J.A.; Shaw, H.F.
2008-01-01
The worldwide energy consumption in 2003 was 421 quadrillion Btu (Quads), and included 162 quads for oil, 99 quads for natural gas, 100 quads for coal, 27 quads for nuclear energy, and 33 quads for renewable sources. The projected worldwide energy consumption for 2030 is 722 quads, corresponding to an increase of 71% over the consumption in 2003. The projected consumption for 2030 includes 239 quads for oil, 190 quads for natural gas, 196 quads for coal, 35 quads for nuclear energy, and 62 quads for renewable sources (International Energy Outlook, DOE/EIA-0484, Table D1 (2006) p. 133]. The current fleet of light water reactors (LRWs) provides about 20% of current U.S. electricity, and about 16% of current world electricity. The demand for electricity is expected to grow steeply in this century, as the developing world increases its standard of living. With the increasing price for oil and gasoline within the United States, as well as fear that our CO2 production may be driving intolerable global warming, there is growing pressure to move away from oil, natural gas, and coal towards nuclear energy. Although there is a clear need for nuclear energy, issues facing waste disposal have not been adequately dealt with, either domestically or internationally. Better technological approaches, with better public acceptance, are needed. Nuclear power has been criticized on both safety and waste disposal bases. The safety issues are based on the potential for plant damage and environmental effects due to either nuclear criticality excursions or loss of cooling. Redundant safety systems are used to reduce the probability and consequences of these risks for LWRs. LIFE engines are inherently subcritical, reducing the need for systems to control the fission reactivity. LIFE engines also have a fuel type that tolerates much higher temperatures than LWR fuel, and has two safety systems to remove decay heat in the event of loss of coolant or loss of coolant flow. These features of
Fracture probability evaluation of a LWR pressure vessel
International Nuclear Information System (INIS)
Grandemange, J.; Pellissier-Tanon, A.; Quero, J.; Carnino, A.; Dufresne, J.
1978-01-01
Fracture probability evaluation, of a LWR pressure vessel have been performed in the past, using statistical data from conventional plant. A more accurate evaluation has been requested in 1976 from the SCSIN to the CEA. With this object, a joint collaboration agreement has been signed between CEA, EURATOM/ISPRA and FRAMATOME. The whole program proceeding from this agreement is managed by a joint board including the three partners. The basic objective of this program is to develop a method which integrates, or makes it possible to integrate at a later stage, the greatest number of significant parameters. Also, in order to prepare the practical applications, a special effort is being made to collect the data corresponding to these parameters. Parallel basic research program have been launched in order to clarify our knowledge on some important parts of the main factors contributing to the evaluation. The results of this research will be progressively introduced into the method or will help checking its validity
Dual-purpose LWR supplying heat for desalination
International Nuclear Information System (INIS)
Waplington, G.; Fitcher, H.
1977-01-01
A number of desalination processes are at present in various stages of development but distillation is the only serious choice for a large-scale project. The distillation process temperature requirement is low compared with the temperature of steam normally delivered to the turbine in a power generation plant. This gives the possibility for combining the functions of electricity generation with water distillation. The brine heater of the multi-stage flash distillation plant can be supplied with steam after partial expansion through a turbine. Such an arrangement allows the use of a standard nuclear steam supply system and makes fuller use of the energy output than would either single purpose role. The LWR represents a safe, reliable and economic system, and is easily able to provide heat of a quality adequate for the desalination process. (M.S.)
Analysis of alternative light water reactor (LWR) fuel cycles
International Nuclear Information System (INIS)
Heeb, C.M.; Aaberg, R.L.; Boegel, A.J.; Jenquin, U.P.; Kottwitz, D.A.; Lewallen, M.A.; Merrill, E.T.; Nolan, A.M.
1979-12-01
Nine alternative LWR fuel cycles are analyzed in terms of the isotopic content of the fuel material, the relative amounts of primary and recycled material, the uranium and thorium requirements, the fuel cycle costs and the fraction of energy which must be generated at secured sites. The fuel materials include low-enriched uranium (LEU), plutonium-uranium (MOX), highly-enriched uranium-thorium (HEU-Th), denatured uranium-thorium (DU-Th) and plutonium-thorium (Pu-Th). The analysis is based on tracing the material requirements of a generic pressurized water reactor (PWR) for a 30-year period at constant annual energy output. During this time period all the created fissile material is recycled unless its reactivity worth is less than 0.2% uranium enrichment plant tails
A study on the behavior of defected LWR spent fuel
International Nuclear Information System (INIS)
You, Gil Sung; Kim, Eun Ka; Kim, Keon Sik; Suh, Hang Suck; Kim, Seung Jung; Ro, Seung Gy; Park, Chong Mook; Ji, Pyung Gook
1992-03-01
To investigate the storage behavior of the defective LWR spent fuel rods, the characteristic changes of fuel and cladding are to be measured and analyzed. In addition, the oxidation study in air on non-irradiated and irradiated U0 2 was performed. No changes were observed in the tested fuel rods after 30 month storage. The Cs-134, 137 released rapidly during the initial 3 months of storage, but remained in constant value after 3 month storage and the release was almost ceased after 30 month storage. The weight gain of non-irradiated U0 2 samples showed a trend of S type curves and the activation energies were 11OKJ/mol above 350 deg C. and 143KJ/mol below 350 deg C. But irradiated U0 2 showed a rapid increase at initial stage of oxidation and a decrease at later stage when compared with the results of non-irradiated U0 2 . (Author)
Development of PIE techniques for irradiated LWR pressure vessel steels
International Nuclear Information System (INIS)
Nishi, Masahiro; Kizaki, Minoru; Sukegawa, Tomohide
1999-01-01
For the evaluation of safety and integrity of light water reactors (LWRs), various post irradiation examinations (PIEs) of reactor pressure vessel (RPV) steels and fuel claddings have been carried out in the Research Hot Laboratory (RHL). In recent years, the instrumented Charpy impact testing machine was remodeled aiming at the improvement of accuracy and reliability. By this remodeling, absorbed energy and other useful information on impact properties can be delivered from the force-displacement curve for the evaluation of neutron irradiation embrittlement behavior of LWR-RPV steels at one-time striking. In addition, two advanced PIE technologies are now under development. One is the remote machining of mechanical test pieces from actual irradiated pressure vessel steels. The other is development of low-cycle and high-cycle fatigue test technology in order to clarify the post-irradiation fatigue characteristics of structural and fuel cladding materials. (author)
Progress in Development of I2S-LWR Concept
International Nuclear Information System (INIS)
Petrovic, Bojan
2014-01-01
The paper will present the progress in developing the Integral Inherently Safe Light Water Reactor (12S-LWR) concept. This new concept aims to combine the competitive economics of a large nuclear power plant, with enhanced safety achieved by the integral primary circuit configuration (previously considered only for PWRs with power levels not exceeding several hundred MWc), and with enhanced accident tolerance (to address concerns after the Fukushima Dai-lchi accidents). Several new technologies are being developed to enable this concept, including novel silicide fuel and micro-channel primary heat exchangers. This project is performed by a multi-disciplinary multi-organization team led by Georgia Tech, including academia, a national laboratory, nuclear industry, and a power utility, wit expected participation of the University of Zagreb. (author)
Quality assurance in the course of fabrication of LWR fuel
International Nuclear Information System (INIS)
Dressler, G.; Perry, J.A.
1982-01-01
A high quality level of LWR fuel elements can only be assured by a system of Quality Assurance measures purposefully designed, balanced, and appropriately applied. This includes application of and the appropriate balance between both system and product oriented measures. A prerequisite to the establishment of these measures is a precise analysis of the various influences of the individual process steps on the quality characteristics of the starting materials, semi-finished and finished products. In addition, these characteristics require classification criteria relative to their significance. The described classification is used to establish sampling plans and to disposition non-conformances. The EXXON Nuclear Quality Assurance system which is based on these principles is described and illustrated with some examples. (orig.)
Automatic test equipment for C and I of compact LWR
International Nuclear Information System (INIS)
Mayya, Anuradha; Marathe, P.P.; Madala, Kalyan C.
2014-01-01
The C and I of compact LWR consist of a wide variety of electronic modules. Testing of these modules manually was found to be very cumbersome. To ease the testing of these modules, Automatic Test Equipments (ATE) were developed jointly by BARC and ECIL. This paper describes the design of two ATEs for testing 69 types of modules. A power supply ATE was developed for 43 types of power supply modules of type AC-AC, AC-DC, DC-DC and signal conditioning modules. A VME ATE was developed to test 26 types of VME bus based and other microcontroller based non-bussed modules. These ATEs are used for the automated black box testing of modules by feeding power and control inputs and checking the outputs without operator intervention. This paper describes the important considerations in design and the major design challenges. (author)
Coarse grained model for semiquantitative lipid simulations
Marrink, SJ; de Vries, AH; Mark, AE
2004-01-01
This paper describes the parametrization of a new coarse grained (CG) model for lipid and surfactant systems. Reduction of the number of degrees of freedom together with the use of short range potentials makes it computationally very efficient. Compared to atomistic models a gain of 3-4 orders of
The Martini Coarse-Grained Force Field
Periole, X.; Marrink, S.J.; Monticelli, Luca; Salonen, Emppu
2013-01-01
The Martini force field is a coarse-grained force field suited for molecular dynamics simulations of biomolecular systems. The force field has been parameterized in a systematic way, based on the reproduction of partitioning free energies between polar and apolar phases of a large number of chemical
Core design of super LWR with double tube water rods
International Nuclear Information System (INIS)
Wu, Jianhui; Oka, Yoshiaki
2014-01-01
Highlights: • Supercritical light water cooled and moderated reactor with double tube water rods is developed. • Double-row fuel rod assembly and out-in fuel loading pattern are applied. • Separation plates in peripheral assemblies increase average outlet temperature. • Neutronic and thermal design criteria are satisfied during the cycle. - Abstract: Double tube water rods are employed in core design of super LWR to simplify the upper core structure and refueling procedure. The light water moderator flows up in the inner tube from the bottom of the core, then, changes the flow direction at the top of the core into the outer tube and flows out at the bottom of the core. It eliminates the moderator guide/distribution tubes into the single tube water rods from the top dome of the reactor pressure vessel of the previous super LWR design. Two rows of fuel rods are filled between the water rods in the fuel assembly. Out-in refueling pattern is adopted to flatten radial power distribution. The peripheral fuel assemblies of the core are divided into four flow zones by separation plates for increasing the average core outlet temperature. Three enrichment zones are used for axial power flattening. The equilibrium core is analyzed based on neutronic/thermal-hydraulic coupled model. The results show that, by applying the separation plates in peripheral fuel assemblies and low gadolinia enrichment, the maximum cladding surface temperature (MCST) is limited to 653 °C with the average outlet temperature of 500 °C. The inherent safety is satisfied by the negative void reactivity effects and sufficient shutdown margin
User Manual for the PROTEUS Mesh Tools
Energy Technology Data Exchange (ETDEWEB)
Smith, Micheal A. [Argonne National Lab. (ANL), Argonne, IL (United States); Shemon, Emily R [Argonne National Lab. (ANL), Argonne, IL (United States)
2016-09-19
PROTEUS is built around a finite element representation of the geometry for visualization. In addition, the PROTEUS-SN solver was built to solve the even-parity transport equation on a finite element mesh provided as input. Similarly, PROTEUS-MOC and PROTEUS-NEMO were built to apply the method of characteristics on unstructured finite element meshes. Given the complexity of real world problems, experience has shown that using commercial mesh generator to create rather simple input geometries is overly complex and slow. As a consequence, significant effort has been put into place to create multiple codes that help assist in the mesh generation and manipulation. There are three input means to create a mesh in PROTEUS: UFMESH, GRID, and NEMESH. At present, the UFMESH is a simple way to generate two-dimensional Cartesian and hexagonal fuel assembly geometries. The UFmesh input allows for simple assembly mesh generation while the GRID input allows the generation of Cartesian, hexagonal, and regular triangular structured grid geometry options. The NEMESH is a way for the user to create their own mesh or convert another mesh file format into a PROTEUS input format. Given that one has an input mesh format acceptable for PROTEUS, we have constructed several tools which allow further mesh and geometry construction (i.e. mesh extrusion and merging). This report describes the various mesh tools that are provided with the PROTEUS code giving both descriptions of the input and output. In many cases the examples are provided with a regression test of the mesh tools. The most important mesh tools for any user to consider using are the MT_MeshToMesh.x and the MT_RadialLattice.x codes. The former allows the conversion between most mesh types handled by PROTEUS while the second allows the merging of multiple (assembly) meshes into a radial structured grid. Note that the mesh generation process is recursive in nature and that each input specific for a given mesh tool (such as .axial
Quantum theory of multiscale coarse-graining.
Han, Yining; Jin, Jaehyeok; Wagner, Jacob W; Voth, Gregory A
2018-03-14
Coarse-grained (CG) models serve as a powerful tool to simulate molecular systems at much longer temporal and spatial scales. Previously, CG models and methods have been built upon classical statistical mechanics. The present paper develops a theory and numerical methodology for coarse-graining in quantum statistical mechanics, by generalizing the multiscale coarse-graining (MS-CG) method to quantum Boltzmann statistics. A rigorous derivation of the sufficient thermodynamic consistency condition is first presented via imaginary time Feynman path integrals. It identifies the optimal choice of CG action functional and effective quantum CG (qCG) force field to generate a quantum MS-CG (qMS-CG) description of the equilibrium system that is consistent with the quantum fine-grained model projected onto the CG variables. A variational principle then provides a class of algorithms for optimally approximating the qMS-CG force fields. Specifically, a variational method based on force matching, which was also adopted in the classical MS-CG theory, is generalized to quantum Boltzmann statistics. The qMS-CG numerical algorithms and practical issues in implementing this variational minimization procedure are also discussed. Then, two numerical examples are presented to demonstrate the method. Finally, as an alternative strategy, a quasi-classical approximation for the thermal density matrix expressed in the CG variables is derived. This approach provides an interesting physical picture for coarse-graining in quantum Boltzmann statistical mechanics in which the consistency with the quantum particle delocalization is obviously manifest, and it opens up an avenue for using path integral centroid-based effective classical force fields in a coarse-graining methodology.
Quantum theory of multiscale coarse-graining
Han, Yining; Jin, Jaehyeok; Wagner, Jacob W.; Voth, Gregory A.
2018-03-01
Coarse-grained (CG) models serve as a powerful tool to simulate molecular systems at much longer temporal and spatial scales. Previously, CG models and methods have been built upon classical statistical mechanics. The present paper develops a theory and numerical methodology for coarse-graining in quantum statistical mechanics, by generalizing the multiscale coarse-graining (MS-CG) method to quantum Boltzmann statistics. A rigorous derivation of the sufficient thermodynamic consistency condition is first presented via imaginary time Feynman path integrals. It identifies the optimal choice of CG action functional and effective quantum CG (qCG) force field to generate a quantum MS-CG (qMS-CG) description of the equilibrium system that is consistent with the quantum fine-grained model projected onto the CG variables. A variational principle then provides a class of algorithms for optimally approximating the qMS-CG force fields. Specifically, a variational method based on force matching, which was also adopted in the classical MS-CG theory, is generalized to quantum Boltzmann statistics. The qMS-CG numerical algorithms and practical issues in implementing this variational minimization procedure are also discussed. Then, two numerical examples are presented to demonstrate the method. Finally, as an alternative strategy, a quasi-classical approximation for the thermal density matrix expressed in the CG variables is derived. This approach provides an interesting physical picture for coarse-graining in quantum Boltzmann statistical mechanics in which the consistency with the quantum particle delocalization is obviously manifest, and it opens up an avenue for using path integral centroid-based effective classical force fields in a coarse-graining methodology.
A Brief Assessment of North Korea's Capacities for Building an Experimental LWR
International Nuclear Information System (INIS)
Lee, Jung Hyu; An, Jin Soo
2011-01-01
On November 2010, North Korea revealed the construction site of 100 MWt (thermal) experimental LWR in the early stage with a target operation date of 2012. And they claimed that their first LWR construction project is proceeding with strictly domestic talent and resources. Introduction of LWR imposes various technical challenges, even though North Korea has experiences in the construction and management of graphite-moderated and gas-cooled reactor. So, there are doubts about whether they can successfully complete the project in time without any external support. In this paper, to estimate the fate of the LWR construction, we focused on the North Korea's capability to deal with the technical challenges which differ from those of gas-graphite reactor
A Brief Assessment of North Korea's Capacities for Building an Experimental LWR
Energy Technology Data Exchange (ETDEWEB)
Lee, Jung Hyu; An, Jin Soo [Korea Institute of Nuclear Nonproliferation and Control, Daejeon (Korea, Republic of)
2011-10-15
On November 2010, North Korea revealed the construction site of 100 MWt (thermal) experimental LWR in the early stage with a target operation date of 2012. And they claimed that their first LWR construction project is proceeding with strictly domestic talent and resources. Introduction of LWR imposes various technical challenges, even though North Korea has experiences in the construction and management of graphite-moderated and gas-cooled reactor. So, there are doubts about whether they can successfully complete the project in time without any external support. In this paper, to estimate the fate of the LWR construction, we focused on the North Korea's capability to deal with the technical challenges which differ from those of gas-graphite reactor
Energy Technology Data Exchange (ETDEWEB)
Guthrie, G L; McElroy, W N; Lippincott, E P; Gold, R
1978-12-01
Program objectives and progress to date by the national laboratories in LWR pressure vessel irradiation surveillance dosimetry are summarized. Participants in the program include: Rockwell International, Hanford Engineering Development Laboratory, National Bureau of Standards, and Oak Ridge National Laboratory.
International Nuclear Information System (INIS)
Hoeft, L.O.; Hofstra, J.S.; Karaskiewicz, R.J.; Wiser, G.
1984-01-01
The surface magnetic field attenuation of five types of shielded transparency (window) material was measured over the frequency range 10 kHz to 100 MHz by installing them on an .61 m x .61 m x .2 m enclosure, placing the enclosure on the wall of a TEM cell and measuring the surface and interior magnetic fields using a computer-controlled network analyzer system. The samples included two thicknesses of conductive grids on acrylic, hardware, cloth with 1/8 and 1/4-inch mesh, and a fine mesh laminated optical display window. These measurements are indicative of an enclosure with aperture coupling; namely, they become frequency-independent at high frequencies. Coarse mesh samples (1/8-1/4-inch mesh) were able to provide 50 to 60 dB of magnetic field reduction at tens of MHz, whereas the finer mesh did slightly better. This behavior is consistent with magnetic polarizability theory. Material thickness did not have an appreciable effect for frequencies above a MHz
Voltammetry at micro-mesh electrodes
Directory of Open Access Journals (Sweden)
Wadhawan Jay D.
2003-01-01
Full Text Available The voltammetry at three micro-mesh electrodes is explored. It is found that at sufficiently short experimental durations, the micro-mesh working electrode first behaves as an ensemble of microband electrodes, then follows the behaviour anticipated for an array of diffusion-independent micro-ring electrodes of the same perimeter as individual grid-squares within the mesh. During prolonged electrolysis, the micro-mesh electrode follows that behaviour anticipated theoretically for a cubically-packed partially-blocked electrode. Application of the micro-mesh electrode for the electrochemical determination of carbon dioxide in DMSO electrolyte solutions is further illustrated.
22nd International Meshing Roundtable
Staten, Matthew
2014-01-01
This volume contains the articles presented at the 22nd International Meshing Roundtable (IMR) organized, in part, by Sandia National Laboratories and was held on Oct 13-16, 2013 in Orlando, Florida, USA. The first IMR was held in 1992, and the conference series has been held annually since. Each year the IMR brings together researchers, developers, and application experts in a variety of disciplines, from all over the world, to present and discuss ideas on mesh generation and related topics. The technical papers in this volume present theoretical and novel ideas and algorithms with practical potential, as well as technical applications in science and engineering, geometric modeling, computer graphics and visualization.
21st International Meshing Roundtable
Weill, Jean-Christophe
2013-01-01
This volume contains the articles presented at the 21st International Meshing Roundtable (IMR) organized, in part, by Sandia National Laboratories and was held on October 7–10, 2012 in San Jose, CA, USA. The first IMR was held in 1992, and the conference series has been held annually since. Each year the IMR brings together researchers, developers, and application experts in a variety of disciplines, from all over the world, to present and discuss ideas on mesh generation and related topics. The technical papers in this volume present theoretical and novel ideas and algorithms with practical potential, as well as technical applications in science and engineering, geometric modeling, computer graphics, and visualization.
Multilevel Methods for Elliptic Problems with Highly Varying Coefficients on Nonaligned Coarse Grids
Energy Technology Data Exchange (ETDEWEB)
Scheichl, Robert [Univ. of Bath (United Kingdom). Dept. of Mathematical Sciences; Vassilevski, Panayot S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Zikatanov, Ludmil T. [Pennsylvania State Univ., University Park, PA (United States). Dept. of Mathematics
2012-06-21
We generalize the analysis of classical multigrid and two-level overlapping Schwarz methods for 2nd order elliptic boundary value problems to problems with large discontinuities in the coefficients that are not resolved by the coarse grids or the subdomain partition. The theoretical results provide a recipe for designing hierarchies of standard piecewise linear coarse spaces such that the multigrid convergence rate and the condition number of the Schwarz preconditioned system do not depend on the coefficient variation or on any mesh parameters. One assumption we have to make is that the coarse grids are sufficiently fine in the vicinity of cross points or where regions with large diffusion coefficients are separated by a narrow region where the coefficient is small. We do not need to align them with possible discontinuities in the coefficients. The proofs make use of novel stable splittings based on weighted quasi-interpolants and weighted Poincaré-type inequalities. Finally, numerical experiments are included that illustrate the sharpness of the theoretical bounds and the necessity of the technical assumptions.
Metholology for the selection of LWR safety R and D projects. Phase I, status report
International Nuclear Information System (INIS)
El-Sheikh, K.A.
1980-03-01
The objective of the LWR R and D Selection Methodology Program is to develop and demonstrate an R and D selection methodology appropriate for LWR safety technology. This report documents the development work from the program beginning in April, 1979 to the end of Fiscal Year 1979. The scope of work for this period included three tasks; methodology review (Task 1), measures development (Task 2), and methodology development for the first phase of application (Task 3). The accomplishments of these tasks are presented
Adaptive Mesh Refinement in CTH
International Nuclear Information System (INIS)
Crawford, David
1999-01-01
This paper reports progress on implementing a new capability of adaptive mesh refinement into the Eulerian multimaterial shock- physics code CTH. The adaptivity is block-based with refinement and unrefinement occurring in an isotropic 2:1 manner. The code is designed to run on serial, multiprocessor and massive parallel platforms. An approximate factor of three in memory and performance improvements over comparable resolution non-adaptive calculations has-been demonstrated for a number of problems
International Nuclear Information System (INIS)
Newman, D.F.; Fleischman, R.M.; White, M.K.
1977-02-01
The plutonium interface between the LWR and LMFBR fuel cycles is examined for typical nuclear growth projections both with and without plutonium recycle in LWRs. In order to guarantee a fuel supply for projected LMFBR growth rates, significant multiple Pu recycle in LWRs will not be possible. However, about 78% of the benefit of multiple plutonium recycle between now and the turn of the century is realized by one recycle and then stockpiling spent MOX for the LMFBR. LMFBR reprocessing schecules are estimated based on accumulation of reprocessing load. These schedules are used to estimate the amount of plutonium recovered from LMFBR fuels and determine the residual LWR plutonium required to meet LMFBR demand. The stockpile of LWR produced plutonium in spent MOX is sufficient to fuel the LMFBR until commercial LMFBR reprocessing can be justified. After that time, recycle of plutonium in LWRs will be significantly limited by a continuing LMFBR demand for LWR plutonium due to the projected high LMFBR growth rate. LWR reprocessing requirements are estimated for the assumed condition that LWR plutonium recycle is not approved, but the LMFBR is still pursued as an energy option. The uncertainties presented by this condition are addressed qualitatively. However, in our judgment these uncertainties in the plutonium market would likely delay LMFBR growth to levels significantly below current projections
Fuel-steel mixing and radial mesh effects in power excursion simulations
International Nuclear Information System (INIS)
Chen, X.-N.; Rineiski, A.; Gabrielli, F.; Andriolo, L.; Vezzoni, B.; Li, R.; Maschek, W.; Kiefhaber, E.
2016-01-01
Highlights: • Fuel-steel mixing and radial mesh effects are significant on power excursion. • The earliest power peak is reduced and retarded by these two effects. • Unprotected loss of coolant transients in ESFR core are calculated. - Abstract: This paper deals with SIMMER-III once-through simulations of the earliest power excursion initiated by an unprotected loss of flow (ULOF) in the Working Horse design of the European Sodium Cooled Fast Reactor (ESFR). Since the sodium void effect is strictly positive in this core and dominant in the transient, a power excursion is initiated by sodium boiling in the ULOF case. Two major effects, namely (1) reactivity effects due to fuel-steel mixing after melting and (2) the radial mesh size, which were not considered originally in SIMMER simulations for ESFR, are studied. The first effect concerns the reactivity difference between the heterogeneous fuel/clad/wrapper configuration and the homogeneous mixture of steel and fuel. The full core homogenization (due to melting) effect is −2 $, though a smaller effect takes place in case of partial core melting. The second effect is due to the SIMMER sub-assembly (SA) coarse mesh treatment, where a simultaneous sodium boiling onset in all SAs belonging to one ring leads to an overestimated reactivity ramp. For investigating the influence of fuel/steel mixing effects, a lumped “homogenization” reactivity feedback has been introduced, being proportional to the molten steel mass. For improving the coarse mesh treatment, we employ finer radial meshes to take the subchannel effects into account, where the side and interior channels have different coolant velocities and temperatures. The simulation results show that these two effects have significant impacts on the earliest power excursion after the sodium boiling.
International Nuclear Information System (INIS)
Thiels, G.M.; Kowa, S.
1993-01-01
This report deals with the cost determination of a number of schemes for the treatment, conditioning, packaging, interim storage and transport operations of LWR wastes drawn up on the basis of Belgian, French and German practices in this particular area. In addition to the general procedure elaborated for determining, actualizing and scaling of plant and transport costs associated with the various schemes, in-depth calculations of each intermediate management stage are included in this report. This study is part of an overall theoretical exercise aimed at evaluating a selection of management routes for LWR waste based on economical and radiological criteria
Spacetime coarse grainings in nonrelativistic quantum mechanics
International Nuclear Information System (INIS)
Hartle, J.B.
1991-01-01
Sum-over-histories generalizations of nonrelativistic quantum mechanics are explored in which probabilities are predicted, not just for alternatives defined on spacelike surfaces, but for alternatives defined by the behavior of spacetime histories with respect to spacetime regions. Closed, nonrelativistic systems are discussed whose histories are paths in a given configuration space. The action and the initial quantum state are assumed fixed and given. A formulation of quantum mechanics is used which assigns probabilities to members of sets of alternative coarse-grained histories of the system, that is, to the individual classes of a partition of its paths into exhaustive and exclusive classes. Probabilities are assigned to those sets which decohere, that is, whose probabilities are consistent with the sum rules of probability theory. Coarse graining by the behavior of paths with respect to regions of spacetime is described. For example, given a single region, the set of all paths may be partitioned into those which never pass through the region and those which pass through the region at least once. A sum-over-histories decoherence functional is defined for sets of alternative histories coarse-grained by spacetime regions. Techniques for the definition and effective computation of the relevant sums over histories by operator-product formulas are described and illustrated by examples. Methods based on Euclidean stochastic processes are also discussed and illustrated. Models of decoherence and measurement for spacetime coarse grainings are described. Issues of causality are investigated. Such spacetime generalizations of nonrelativistic quantum mechanics may be useful models for a generalized quantum mechanics of spacetime geometry
Coarse graining flow of spin foam intertwiners
Dittrich, Bianca; Schnetter, Erik; Seth, Cameron J.; Steinhaus, Sebastian
2016-12-01
Simplicity constraints play a crucial role in the construction of spin foam models, yet their effective behavior on larger scales is scarcely explored. In this article we introduce intertwiner and spin net models for the quantum group SU (2 )k×SU (2 )k, which implement the simplicity constraints analogous to four-dimensional Euclidean spin foam models, namely the Barrett-Crane (BC) and the Engle-Pereira-Rovelli-Livine/Freidel-Krasnov (EPRL/FK) model. These models are numerically coarse grained via tensor network renormalization, allowing us to trace the flow of simplicity constraints to larger scales. In order to perform these simulations we have substantially adapted tensor network algorithms, which we discuss in detail as they can be of use in other contexts. The BC and the EPRL/FK model behave very differently under coarse graining: While the unique BC intertwiner model is a fixed point and therefore constitutes a two-dimensional topological phase, BC spin net models flow away from the initial simplicity constraints and converge to several different topological phases. Most of these phases correspond to decoupling spin foam vertices; however we find also a new phase in which this is not the case, and in which a nontrivial version of the simplicity constraints holds. The coarse graining flow of the BC spin net models indicates furthermore that the transitions between these phases are not of second order. The EPRL/FK model by contrast reveals a far more intricate and complex dynamics. We observe an immediate flow away from the original simplicity constraints; however, with the truncation employed here, the models generically do not converge to a fixed point. The results show that the imposition of simplicity constraints can indeed lead to interesting and also very complex dynamics. Thus we need to further develop coarse graining tools to efficiently study the large scale behavior of spin foam models, in particular for the EPRL/FK model.
Measuring Crack Length in Coarse Grain Ceramics
Salem, Jonathan A.; Ghosn, Louis J.
2010-01-01
Due to a coarse grain structure, crack lengths in precracked spinel specimens could not be measured optically, so the crack lengths and fracture toughness were estimated by strain gage measurements. An expression was developed via finite element analysis to correlate the measured strain with crack length in four-point flexure. The fracture toughness estimated by the strain gaged samples and another standardized method were in agreement.
Behavior of LWR fuel elements under accident conditions
International Nuclear Information System (INIS)
Albrecht, H.; Bocek, M.; Erbacher, F.; Fiege, A.; Fischer, M.; Hagen, S.; Hofmann, P.; Holleck, H.; Karb, E.; Leistikow, S.; Melang, S.; Ondracek, G.; Thuemmler, F.; Wiehr, K.
1977-01-01
In the frame of the German reactor safety research program, the Kernforschungszentrum Karlsruhe is carrying out a comprehensive program on the behavior of LWR fuel elements under a variety of power cooling mismatch conditions in particular during loss-of-coolant accidents. The major objectives are to establish a detailed quantitative understanding of fuel rod failures mechanisms and their thresholds, to evaluate the safety margins of power reactor cores under accident conditions and to investigate the feedback of fuel rod failures on the efficiency of emergency core cooling systems. This detailed quantitative understanding is achieved through extensive basic and integral experiments and is incorporated in a fuel behavior code. On the basis of these results the design of power reactor fuel elements and of safety devices can be further improved. The results of investigations on the inelastic deformation (ballooning) behavior of Zircaloy 4 cladding at LOCA temperatures in oxidizing atmosphere are presented. Depending upon strain rate and temperature superplastic deformation behavior was observed. In the equation of state of Zry 4 the strain rate sensitivity index depends strongly upon strain and in the superplastic region upon sample anisotropy. Oxidation kinetics experiments with Zry-tubes at 900-1300 0 C showed that the Baker-Just correlation describes the reality quite conservative. Therefore a reduction of the amount of Zry oxidation can be assumed in the course of a LOCA. The external oxidation of Zry-cladding by steam as well as internal oxidation by the oxygen in oxide fuel and fission products (Cs, I, Te) have an influence on the strain and rupture behavior of Zry-cladding at LOCA temperatures. In out-of-pile and inpile experiments the mechanical and thermal behavior of fuel rods during the blowdown, the heatup and the reflood phases of a LOCA are investigated under representative and controlled thermohydraulic conditions. The task of the inpile experiments is
Directory of Open Access Journals (Sweden)
Jennings Jason
2010-01-01
Full Text Available Laparoscopic inguinal herniorraphy via a transabdominal preperitoneal (TAPP approach using Polypropylene Mesh (Mesh and staples is an accepted technique. Mesh induces a localised inflammatory response that may extend to, and involve, adjacent abdominal and pelvic viscera such as the appendix. We present an interesting case of suspected Mesh-induced appendicitis treated successfully with laparoscopic appendicectomy, without Mesh removal, in an elderly gentleman who presented with symptoms and signs of acute appendicitis 18 months after laparoscopic inguinal hernia repair. Possible mechanisms for Mesh-induced appendicitis are briefly discussed.
Adapting LWR to future needs: SECURE-P (PIUS)
International Nuclear Information System (INIS)
Hannerz, K.
1984-01-01
Advanced nuclear technology based on breeder reactors and fuel reprocessing may eventually be applied on a large scale, although the timing for this appears uncertain. However, in many parts of the world societal conditions and technological infrastructure mandate the use of a less complicated technology if the benefits of clean, safe nuclear power are to be available. Such a technology must be based on thermal reactors. Lack of fuel resources for their operation through most of the next century is unlikely to be a serious limitation. A natural contender would be the light water reactor, but today's designs lack many of the desired characteristics. However, introduction of certain new design features can eliminate the shortcomings and make the LWR the prime longterm candidate for a simple, technologically unsophisticated generation of nuclear power. Availability of such an option will also be a major asset for utilities in the large industrial countries before the advent of the era of advanced 'second generation' nuclear power. The costs of demonstrating the new design features are miniscule in relation to the benefits that should accrue. (author)
Democratic People's Republic of Korea LWR project status
International Nuclear Information System (INIS)
Mulligan, J.B.
1996-01-01
In October 1994, at Geneva, the United States and the Democratic People's Republic of Korea (DPRK) signed an Agreed Framework as a first step toward resolving international concerns about nuclear activities in the DPRK. This Agreement, when implemented, will ultimately lead to the complete dismantlement of those aspects of the DPRK's nuclear program, including reprocessing-related facilities, that have undermined the viability of the international nuclear non-proliferation regime and the stability of the Asia-Pacific region. The essence of the Agreement is that the DPRK will take near-term action to cease the activities of concern and permit some International Atomic Energy Agency (IAEA) verification inspection. In the future, it will dismantle its production reactors and accept full-scope IAWA safeguards. In return, the United Stated agreed to lead an international effort to supply the DPRK with light-water reactors which are less of proliferation concern than are graphite-moderated production reactors. Until the first LWR is in operation the DPRK will receive shipments of heavy oil to replace the energy lost by shutting down the production reactors
Evaluation of inorganic sorbent treatment for LWR coolant process streams
International Nuclear Information System (INIS)
Roddy, J.W.
1984-03-01
This report presents results of a survey of the literature and of experience at selected nuclear installations to provide information on the feasibility of replacing organic ion exchangers with inorganic sorbents at light-water-cooled nuclear power plants. Radioactive contents of the various streams in boiling water reactors and pressurized water reactors were examined. In addition, the methods and performances of current methods used for controlling water quality at these plants were evaluated. The study also includes a brief review of the physical and chemical properties of selected inorganic sorbents. Some attributes of inorganic sorbents would be useful in processing light water reactor (LWR) streams. The inorganic resins are highly resistant to damage from ionizing radiation, and their exchange capacities are generally equivalent to those of organic ion exchangers. However, they are more limited in application, and there are problems with physical integrity, especially in acidic solutions. Research is also needed in the areas of selectivity and anion removal before inorganic sorbents can be considered as replacements for the synthetic organic resins presently used in LWRs. 11 figures, 14 tables
LWR reactivity/isotopics code for pedagogical and scoping applications
International Nuclear Information System (INIS)
AbuZaied, G.; Driscoll, M.J.
1986-01-01
A program designated BRICC (Burnup Reactivity and Isotopic Composition Computation), has been programmed for use on microcomputers to permit rapid parametric studies of the neutronics of light water reactor (LWR) assemblies. It is currently employed as a teaching tool in a graduate-level subject on nuclear fuel management, and has proven to be of sufficient accuracy to permit its use as a submodule in a more comprehensive program used to evaluate various mechanical spectral shift concepts for pressurized water reactor control. It should also prove useful in teaching reactor physics as it will fill an important gap between hand calculations of inadequate accuracy and state-of-the-art multigroup programs of daunting complexity. The BRICC program combines a minimum adequate set of old-fashioned phenomenological submodels that describe key physics attributed in an integral fashion, thereby providing the student or researcher with convenient mental pictures to serve as the basis for deductive reasoning. The program is short, written in a simplistic version of the Basic language, with many interspersed Remark statements, and is therefore easy to tinker with for various constructive purposes
CCGT + LWR = the power plant of the future?
International Nuclear Information System (INIS)
Tsiklauri, G.
1997-01-01
The thermal efficiency of LWR type reactors can be increased making use of the Tsikl-Durst cycle, where the gas turbine is combined with the nuclear reactor using a steam mixer. The principle of this combined cycle is outlined. It is envisaged that the overall thermal efficiency of the power plant can be increased to 41 - 44%. The total output would be two to three times higher. With advanced light-water reactors (ABWR, AP-600) and advanced gas turbines in combination with the one-way steam generator as developed by Solar Turbines Inc., producing steam at 650 degC to 750 degC, it is feasible to attain a total thermal efficiency of 55%. The combination of two kinds of fuel (nuclear fuel and natural gas) improves operating flexibility of the cycle in various regimes so as to respond to natural gas prices and electricity demands. The gas turbine adds to the nuclear power plant an independent source of power, so that standby dieselgenerators are no more necessary. (P.A.). 1 tab., 2 figs
Feasibility study on the development of advanced LWR fuel technology
International Nuclear Information System (INIS)
Jung, Youn Ho; Sohn, D. S.; Jeong, Y. H.; Song, K. W.; Song, K. N.; Chun, T. H.; Bang, J. G.; Bae, K. K.; Kim, D. H. and others.
1997-07-01
Worldwide R and D trends related to core technology of LWR fuels and status of patents have been surveyed for the feasibility study. In addition, various fuel cycle schemes have been studied to establish the target performance parameters. For the development of cladding material, establishment of long-term research plan for alloy development and optimization of melting process and manufacturing technology were conducted. A work which could characterize the effect of sintering additives on the microstructure of UO 2 pellet has been experimentally undertaken, and major sintering variables and their ranges have been found in the sintering process of UO 2 -Gd 2 O 3 burnable absorber pellet. The analysis of state of the art technology related to flow mixing device for spacer grid and debris filtering device for bottom nozzle and the investigation of the physical phenomena related to CHF enhancement and the establishment of the data base for thermal-hydraulic performance tests has been done in this study. In addition, survey on the documents of the up-to-date PWR fuel assemblies developed by foreign vendors have been carried out to understand their R and D trends and establish the direction of R and D for these structural components. And, to set the performance target of the new fuel, to be developed, fuel burnup and economy under the extended fuel cycle length scheme were estimated. A preliminary study on the failure mechanism of CANDU fuel, key technology and advanced coating has been performed. (author). 190 refs., 31 tabs., 129 figs
Qualification of the neutronic evolution of LWR fuels in MELUSINE
International Nuclear Information System (INIS)
Beretz, D.; Garcin, J.; Ducros, G.; Vanhumbeeck, D.; Chaucheprat, P.
1984-09-01
MELUSINE, a swimming pool type reactor, in Grenoble, for research and technological irradiations is well fitted to the neutronic evolution qualification of the LWR fuel. Thus, with an adjustment of the lattice pitch, representative neutron spectrum locations are available. The re-leading management and the regulation mode flexibility of MELUSINE lead to reproductible neutronic parameters configurations without restricting the reactor to this purpose only. Under these conditions, simple calculations can be carried out for interpretation, without taking into account the whole core. An instrumentation by Self Power Neutron Detectors (collectrons) gives on-line information on the fluxes at the periphery of the device. When required by the neutronicians, experimental pins can be unloaded during the irradiation process and scanned on a gammametry bench immersed in the reactor-pool itself, before their isotopic composition analysis. Thus, within the framework of neutronic evolution qualification, are studied fuel pins for advanced assemblies for the light water reactors or their derivatives, with large advantages over irradiations in power reactors [fr
Convergence studies of deterministic methods for LWR explicit reflector methodology
International Nuclear Information System (INIS)
Canepa, S.; Hursin, M.; Ferroukhi, H.; Pautz, A.
2013-01-01
The standard approach in modem 3-D core simulators, employed either for steady-state or transient simulations, is to use Albedo coefficients or explicit reflectors at the core axial and radial boundaries. In the latter approach, few-group homogenized nuclear data are a priori produced with lattice transport codes using 2-D reflector models. Recently, the explicit reflector methodology of the deterministic CASMO-4/SIMULATE-3 code system was identified to potentially constitute one of the main sources of errors for core analyses of the Swiss operating LWRs, which are all belonging to GII design. Considering that some of the new GIII designs will rely on very different reflector concepts, a review and assessment of the reflector methodology for various LWR designs appeared as relevant. Therefore, the purpose of this paper is to first recall the concepts of the explicit reflector modelling approach as employed by CASMO/SIMULATE. Then, for selected reflector configurations representative of both GII and GUI designs, a benchmarking of the few-group nuclear data produced with the deterministic lattice code CASMO-4 and its successor CASMO-5, is conducted. On this basis, a convergence study with regards to geometrical requirements when using deterministic methods with 2-D homogenous models is conducted and the effect on the downstream 3-D core analysis accuracy is evaluated for a typical GII deflector design in order to assess the results against available plant measurements. (authors)
Research on ultrasonic flow detection techniques for LWR facilities
International Nuclear Information System (INIS)
Kimura, Katsumi; Fukuhara, Hiroaki; Hoshimoto, Kenichi; Matsumoto, Shojiro; Yamawaki, Hisashi; Ito, Hideyuki; Uetake, Ichizo
1986-01-01
Aiming at establishing the techniques for inspecting the inside of LWR pressure vessels by ultrasonic flaw detection from the outside of the vessels, the development of a probe suitable to the flaw detection in the thick steel plates with stainless steel overlay and the method of its driving, the examination of the ultrasonic characteristics of austenitic stainless steel welded metal used for overlay, and the improvement of the detectability of defects and the accuracy of measuring dimensions by the application of signal processing techniques to ultrasonic flaw detection were attempted. In order to cope with the impedance lowering accompanying the increase of oscillator size, the oscillator was divided into the rings with equal area, and the driving and signal receiving were carried out individually, in this way, the good results were obtained by summing the signals. It was theoretically proved that it is rational to use longitudinal waves for the flaw detection in overlay. It was found that by displaying the results of flaw detection as pictures using a microcomputer, the capability of defect detection was increased. Also by the signal processing combining Fourier transformation and filtering, noise removal and the heightening of the accuracy of measuring dimensions were able to be attained. (Kako, I.)
Modelling of a LWR open fuel cycle using the message
Energy Technology Data Exchange (ETDEWEB)
Estanislau, Fidéllis B.G.L. e; Jonusan, Raoni A.S.; Costa, Antonella L.; Pereira, Claubia, E-mail: fidellis01@hotmail.com, E-mail: rjonusan@gmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear
2017-07-01
The main goal of the national energy planning is the development of a short and long-term strategies based on a holistic evaluation of all available energy sources guiding trends and delimiting expansion alternatives in the energetic sector. For a better understanding of the future possibilities, energy systems analyses are indispensable and support in the decision making related to the long term strategy and energy planning. Due to the projections for increased energy consumption according to the Energy Decennial Plan (year 2015) and the need to reduce greenhouse gas emissions presented by Brazil in the UNFCCC (United Nations Framework Convention on Climate Change), alternative energy sources such as solar, wind, nuclear and biomass sources have played an important role in the world energy matrix. In this way, since the nuclear energy is an option for the national energy mix, the present work aims to use the modelling tool MESSAGE (Model for Energy Supply System Alternatives and Their General Environmental Impact) to analyze and evaluate a nuclear power plant in an energy system. This tool is an optimization model for medium and long-term energy planning taking into account conversion and distribution technologies, energy policies and scenarios to satisfy a determined demand and systems constraints. In this work, a reproduction of results considering an LWR (Light Water Reactor) open-cycle are presented using a model in the MESSAGE code. (author)
MCNP analysis of the nine-cell LWR gadolinium benchmark
International Nuclear Information System (INIS)
Arkuszewski, J.J.
1988-01-01
The Monte Carlo results for a 9-cell fragment of the light water reactor square lattice with a central gadolinium-loaded pin are presented. The calculations are performed with the code MCNP-3A and the ENDF-B/5 library and compared with the results obtained from the BOXER code system and the JEF-1 library. The objective of this exercise is to study the feasibility of BOXER for the analysis of a Gd-loaded LWR lattice in the broader framework of GAP International Benchmark Analysis. A comparison of results indicates that, apart from unavoidable discrepancies originating from different data evaluations, the BOXER code overestimates the multiplication factor by 1.4 % and underestimates the power release in a Gd cell by 4.66 %. It is hoped that further similar studies with use of the JEF-1 library for both BOXER and MCNP will help to isolate and explain these discrepancies in a cleaner way. (author) 4 refs., 9 figs., 10 tabs
The scale analysis sequence for LWR fuel depletion
International Nuclear Information System (INIS)
Hermann, O.W.; Parks, C.V.
1991-01-01
The SCALE (Standardized Computer Analyses for Licensing Evaluation) code system is used extensively to perform away-from-reactor safety analysis (particularly criticality safety, shielding, heat transfer analyses) for spent light water reactor (LWR) fuel. Spent fuel characteristics such as radiation sources, heat generation sources, and isotopic concentrations can be computed within SCALE using the SAS2 control module. A significantly enhanced version of the SAS2 control module, which is denoted as SAS2H, has been made available with the release of SCALE-4. For each time-dependent fuel composition, SAS2H performs one-dimensional (1-D) neutron transport analyses (via XSDRNPM-S) of the reactor fuel assembly using a two-part procedure with two separate unit-cell-lattice models. The cross sections derived from a transport analysis at each time step are used in a point-depletion computation (via ORIGEN-S) that produces the burnup-dependent fuel composition to be used in the next spectral calculation. A final ORIGEN-S case is used to perform the complete depletion/decay analysis using the burnup-dependent cross sections. The techniques used by SAS2H and two recent applications of the code are reviewed in this paper. 17 refs., 5 figs., 5 tabs
Validation of KENOREST with LWR-PROTEUS phase II samples
Energy Technology Data Exchange (ETDEWEB)
Wagner, M.; Kilger, R.; Pautz, A.; Zwermann, W. [GRS, Garching (Germany); Grimm, P.; Vasiliev, A.; Ferroukhi, H. [Paul Scherrer Institut, Villigen (Switzerland)
2012-11-01
In order to broaden the validation basis of the reactivity and nuclide inventory code KENOREST two samples of the LWR-PROTEUS phase II program have been calculated and compared to the experimental results. In general most nuclides are reproduced very well and agree within about ten percent with the experiment. Some already known problems, the overprediction of metallic fission products and the underprediction of the higher curium isotopes, have been confirmed. One of the largest uncertainties in the calculation was the burnup of the samples due to differences between a core simulation of the fuel vendor and the burnup determined from the measured values of the burnup indicator Nd-148. Two different models taking into account the environment for a peripheral fuel rod have been studied. The more detailed model included the three direct neighbor fuel assemblies depleted along with the fuel rod of interest. The influence on the results has been found to be very small. Compared to the uncertainties from the burnup, this effect can be considered negligible. The reason for the low influence was basically that the spectrum did not get considerably harder with increasing burnup beyond about 20GWd/tHM. Since the sample reached burnups far beyond that value, an effect could not be seen. In the near future an update of the used libraries is planned and it will be very interesting to study the effect on the results, especially for Curium. (orig.)
Preliminary concepts for detecting diversion of LWR spent fuel
International Nuclear Information System (INIS)
Sellers, T.A.
Sandia Laboratories, under the sponsorship of the Department of Energy, Office of Safeguards and Security, has been developing conceptual designs of advanced systems to rapidly detect diversion of LWR spent fuel. Three detection options have been identified and compared on the basis of timeliness of detection and cost. Option 1 is based upon inspectors visiting each facility on a periodic basis to obtain and review data acquired by surveillance instruments and to verify the inventory. Option 2 is based upon continuous inspector presence, aided by surveillance instruments. Option 3 is based upon the collection of data from surveillance instruments with periodic readout either at the facility or at a remote central monitoring and display module and occasional inspection. Surveillance instruments are included in each option to assure a sufficiently high probability of detection. An analysis technique with an example logic tree that was used to identify performance requirements is described. A conceptual design has been developed for Option 3 and the essential hardware elements are not being developed. These elements include radiation, crane and pool acoustic sensors, a Data Collection Module, a Local Collection Module, a Local Display Module and a Central Monitoring and Display Module. A demonstration, in operating facilities, of the overall system concept is planned for the March to June 1979 time frame
Discussion of OECD LWR Uncertainty Analysis in Modelling Benchmark
International Nuclear Information System (INIS)
Ivanov, K.; Avramova, M.; Royer, E.; Gillford, J.
2013-01-01
The demand for best estimate calculations in nuclear reactor design and safety evaluations has increased in recent years. Uncertainty quantification has been highlighted as part of the best estimate calculations. The modelling aspects of uncertainty and sensitivity analysis are to be further developed and validated on scientific grounds in support of their performance and application to multi-physics reactor simulations. The Organization for Economic Co-operation and Development (OECD) / Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC) has endorsed the creation of an Expert Group on Uncertainty Analysis in Modelling (EGUAM). Within the framework of activities of EGUAM/NSC the OECD/NEA initiated the Benchmark for Uncertainty Analysis in Modelling for Design, Operation, and Safety Analysis of Light Water Reactor (OECD LWR UAM benchmark). The general objective of the benchmark is to propagate the predictive uncertainties of code results through complex coupled multi-physics and multi-scale simulations. The benchmark is divided into three phases with Phase I highlighting the uncertainty propagation in stand-alone neutronics calculations, while Phase II and III are focused on uncertainty analysis of reactor core and system respectively. This paper discusses the progress made in Phase I calculations, the Specifications for Phase II and the incoming challenges in defining Phase 3 exercises. The challenges of applying uncertainty quantification to complex code systems, in particular the time-dependent coupled physics models are the large computational burden and the utilization of non-linear models (expected due to the physics coupling). (authors)
Sierra toolkit computational mesh conceptual model
International Nuclear Information System (INIS)
Baur, David G.; Edwards, Harold Carter; Cochran, William K.; Williams, Alan B.; Sjaardema, Gregory D.
2010-01-01
The Sierra Toolkit computational mesh is a software library intended to support massively parallel multi-physics computations on dynamically changing unstructured meshes. This domain of intended use is inherently complex due to distributed memory parallelism, parallel scalability, heterogeneity of physics, heterogeneous discretization of an unstructured mesh, and runtime adaptation of the mesh. Management of this inherent complexity begins with a conceptual analysis and modeling of this domain of intended use; i.e., development of a domain model. The Sierra Toolkit computational mesh software library is designed and implemented based upon this domain model. Software developers using, maintaining, or extending the Sierra Toolkit computational mesh library must be familiar with the concepts/domain model presented in this report.
Anisotropic evaluation of synthetic surgical meshes.
Saberski, E R; Orenstein, S B; Novitsky, Y W
2011-02-01
The material properties of meshes used in hernia repair contribute to the overall mechanical behavior of the repair. The anisotropic potential of synthetic meshes, representing a difference in material properties (e.g., elasticity) in different material axes, is not well defined to date. Haphazard orientation of anisotropic mesh material can contribute to inconsistent surgical outcomes. We aimed to characterize and compare anisotropic properties of commonly used synthetic meshes. Six different polypropylene (Trelex(®), ProLite™, Ultrapro™), polyester (Parietex™), and PTFE-based (Dualmesh(®), Infinit) synthetic meshes were selected. Longitudinal and transverse axes were defined for each mesh, and samples were cut in each axis orientation. Samples underwent uniaxial tensile testing, from which the elastic modulus (E) in each axis was determined. The degree of anisotropy (λ) was calculated as a logarithmic expression of the ratio between the elastic modulus in each axis. Five of six meshes displayed significant anisotropic behavior. Ultrapro™ and Infinit exhibited approximately 12- and 20-fold differences between perpendicular axes, respectively. Trelex(®), ProLite™, and Parietex™ were 2.3-2.4 times. Dualmesh(®) was the least anisotropic mesh, without marked difference between the axes. Anisotropy of synthetic meshes has been underappreciated. In this study, we found striking differences between elastic properties of perpendicular axes for most commonly used synthetic meshes. Indiscriminate orientation of anisotropic mesh may adversely affect hernia repairs. Proper labeling of all implants by manufacturers should be mandatory. Understanding the specific anisotropic behavior of synthetic meshes should allow surgeons to employ rational implant orientation to maximize outcomes of hernia repair.
Huang, Zhihua; Li, Zhihong; Su, Yongjian; Zhu, Yongfeng; Zeng, Wei; Chen, Guiguang; Liang, Zhiqun
2018-02-13
The coarse perlite 40-80 mesh was selected as an immobilizing material and put into a packed bed reactor (PBR) to continuously convert maltose to isomalto-oligosaccharides (IMOs). The PBR was prepared by mixing the thermo-inactivated cells (TIC) from Aspergillus niger J2 strain with the coarse perlite, then the mixture was put into an overpressure-resistant column. Compared with diatomite 40-80 mesh and thin perlite 80-120 mesh in PBR, coarse perlite was chosen as the best filtration aid, when the ratio of coarse perlite versus TIC was 1:1. The thermal and pH stability of the free and immobilized TIC and the optimum conditions for the transglycosylation reactions were determined. The results show that approximately 75 and 82% and 87 and 91% of α-glucosidase activity were reserved for free and immobilized TIC at temperatures from 30 to 60 °C and pH from 3.00 to 7.00 for 12 h, respectively. With 30% malt syrup under the conditions of 50 °C and pH 4.00, a mini-scale packed bed reactor (Mi-PBR) and medium-scale packed bed reactor (Me-PBR) could continuously produce IMO over 25 and 34 days with the yield of effective IMO (eIMO) ≥ 35% and total IMO (tIMO) ≥ 50%, respectively. The strategy of mixing the coarse perlite with TIC in PBR is a novel approach to continuously produce IMO and has great application potential in industry.
Unstructured mesh adaptivity for urban flooding modelling
Hu, R.; Fang, F.; Salinas, P.; Pain, C. C.
2018-05-01
Over the past few decades, urban floods have been gaining more attention due to their increase in frequency. To provide reliable flooding predictions in urban areas, various numerical models have been developed to perform high-resolution flood simulations. However, the use of high-resolution meshes across the whole computational domain causes a high computational burden. In this paper, a 2D control-volume and finite-element flood model using adaptive unstructured mesh technology has been developed. This adaptive unstructured mesh technique enables meshes to be adapted optimally in time and space in response to the evolving flow features, thus providing sufficient mesh resolution where and when it is required. It has the advantage of capturing the details of local flows and wetting and drying front while reducing the computational cost. Complex topographic features are represented accurately during the flooding process. For example, the high-resolution meshes around the buildings and steep regions are placed when the flooding water reaches these regions. In this work a flooding event that happened in 2002 in Glasgow, Scotland, United Kingdom has been simulated to demonstrate the capability of the adaptive unstructured mesh flooding model. The simulations have been performed using both fixed and adaptive unstructured meshes, and then results have been compared with those published 2D and 3D results. The presented method shows that the 2D adaptive mesh model provides accurate results while having a low computational cost.
Adaptive hybrid mesh refinement for multiphysics applications
International Nuclear Information System (INIS)
Khamayseh, Ahmed; Almeida, Valmor de
2007-01-01
The accuracy and convergence of computational solutions of mesh-based methods is strongly dependent on the quality of the mesh used. We have developed methods for optimizing meshes that are comprised of elements of arbitrary polygonal and polyhedral type. We present in this research the development of r-h hybrid adaptive meshing technology tailored to application areas relevant to multi-physics modeling and simulation. Solution-based adaptation methods are used to reposition mesh nodes (r-adaptation) or to refine the mesh cells (h-adaptation) to minimize solution error. The numerical methods perform either the r-adaptive mesh optimization or the h-adaptive mesh refinement method on the initial isotropic or anisotropic meshes to equidistribute weighted geometric and/or solution error function. We have successfully introduced r-h adaptivity to a least-squares method with spherical harmonics basis functions for the solution of the spherical shallow atmosphere model used in climate modeling. In addition, application of this technology also covers a wide range of disciplines in computational sciences, most notably, time-dependent multi-physics, multi-scale modeling and simulation
Meshes optimized for discrete exterior calculus (DEC).
Energy Technology Data Exchange (ETDEWEB)
Mousley, Sarah C. [Univ. of Illinois, Urbana-Champaign, IL (United States); Deakin, Michael [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Knupp, Patrick [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Mitchell, Scott A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
2017-12-01
We study the optimization of an energy function used by the meshing community to measure and improve mesh quality. This energy is non-traditional because it is dependent on both the primal triangulation and its dual Voronoi (power) diagram. The energy is a measure of the mesh's quality for usage in Discrete Exterior Calculus (DEC), a method for numerically solving PDEs. In DEC, the PDE domain is triangulated and this mesh is used to obtain discrete approximations of the continuous operators in the PDE. The energy of a mesh gives an upper bound on the error of the discrete diagonal approximation of the Hodge star operator. In practice, one begins with an initial mesh and then makes adjustments to produce a mesh of lower energy. However, we have discovered several shortcomings in directly optimizing this energy, e.g. its non-convexity, and we show that the search for an optimized mesh may lead to mesh inversion (malformed triangles). We propose a new energy function to address some of these issues.
Transrectal Mesh Erosion Requiring Bowel Resection.
Kemp, Marta Maria; Slim, Karem; Rabischong, Benoît; Bourdel, Nicolas; Canis, Michel; Botchorishvili, Revaz
To report a case of a transrectal mesh erosion as complication of laparoscopic promontofixation with mesh repair, necessitating bowel resection and subsequent surgical interventions. Sacrocolpopexy has become a standard procedure for vaginal vault prolapse [1], and the laparoscopic approach has gained popularity owing to more rapid recovery and less morbidity [2,3]. Mesh erosion is a well-known complication of surgical treatment for prolapse as reported in several negative evaluations, including a report from the US Food and Drug Administration in 2011 [4]. Mesh complications are more common after surgeries via the vaginal approach [5]; nonetheless, the incidence of vaginal mesh erosion after laparoscopic procedures is as high as 9% [6]. The incidence of transrectal mesh exposure after laparoscopic ventral rectopexy is roughly 1% [7]. The diagnosis may be delayed because of its rarity and variable presentation. In addition, polyester meshes, such as the mesh used in this case, carry a higher risk of exposure [8]. A 57-year-old woman experiencing genital prolapse, with the cervix classified as +3 according to the Pelvic Organ Prolapse Quantification system, underwent laparoscopic standard sacrocolpopexy using polyester mesh. Subtotal hysterectomy and bilateral adnexectomy were performed concomitantly. A 3-year follow-up consultation demonstrated no signs or symptoms of erosion of any type. At 7 years after the surgery, however, the patient presented with rectal discharge, diagnosed as infectious rectocolitis with the isolation of Clostridium difficile. She underwent a total of 5 repair surgeries in a period of 4 months, including transrectal resection of exposed mesh, laparoscopic ablation of mesh with digestive resection, exploratory laparoscopy with abscess drainage, and exploratory laparoscopy with ablation of residual mesh and transverse colostomy. She recovered well after the last intervention, exhibiting no signs of vaginal or rectal fistula and no recurrence
RGG: Reactor geometry (and mesh) generator
International Nuclear Information System (INIS)
Jain, R.; Tautges, T.
2012-01-01
The reactor geometry (and mesh) generator RGG takes advantage of information about repeated structures in both assembly and core lattices to simplify the creation of geometry and mesh. It is released as open source software as a part of the MeshKit mesh generation library. The methodology operates in three stages. First, assembly geometry models of various types are generated by a tool called AssyGen. Next, the assembly model or models are meshed by using MeshKit tools or the CUBIT mesh generation tool-kit, optionally based on a journal file output by AssyGen. After one or more assembly model meshes have been constructed, a tool called CoreGen uses a copy/move/merge process to arrange the model meshes into a core model. In this paper, we present the current state of tools and new features in RGG. We also discuss the parallel-enabled CoreGen, which in several cases achieves super-linear speedups since the problems fit in available RAM at higher processor counts. Several RGG applications - 1/6 VHTR model, 1/4 PWR reactor core, and a full-core model for Monju - are reported. (authors)
Parallel adaptive simulations on unstructured meshes
International Nuclear Information System (INIS)
Shephard, M S; Jansen, K E; Sahni, O; Diachin, L A
2007-01-01
This paper discusses methods being developed by the ITAPS center to support the execution of parallel adaptive simulations on unstructured meshes. The paper first outlines the ITAPS approach to the development of interoperable mesh, geometry and field services to support the needs of SciDAC application in these areas. The paper then demonstrates the ability of unstructured adaptive meshing methods built on such interoperable services to effectively solve important physics problems. Attention is then focused on ITAPs' developing ability to solve adaptive unstructured mesh problems on massively parallel computers
PASCAL, Probabilistic Fracture Mechanics Analysis of Structural Components in Aging LWR
International Nuclear Information System (INIS)
Shibata, Katsuyuki; Onizawa, Kunio; Li, Yinsheng; Kato, Daisuke
2005-01-01
A - Description of program or function: PASCAL (PFM analysis of Structural Components in Aging LWR) is a PFM (Probabilistic Fracture Mechanics) code for evaluating the failure probability of aged pressure components. PASCAL has been developed as a part of the JAERI's research program on aging and structural integrity of LWR components, in order to respond to the increasing need of the probabilistic methodology in the regulation and inspection of nuclear components with the objective to provide a rational tool for the evaluation of the reliability and integrity of structural components. In order to improve the accuracy and reliability of the analysis code, some new fracture mechanics models or computational techniques are introduced considering the recent progress in the state of the art and performance of PC. Thus some new analysis models and original methodologies were introduced in PASCAL such as the elastic-plastic fracture criterion based on R6 method, a new crack extension model of semi-elliptical crack evaluation and so on. Moreover a function to evaluate the effect of embrittlement recovery by annealing of irradiated RPV is also introduced in the code based on the USNRC R.G. 1.162(1996). The code has been verified through various failure analysis results and international PTS round robin analysis ICAS which had been organized by the Principal Working Group 3 of OECD/NEA/CSNI. In order to attain a high usability, PASCAL Ver.1 with GUI provides an exclusive FEM pre-processor Pre-PASCAL for generating the input load transient data, a GUI system for generating the input data for PASCAL main processor of main solver and post-processor for output data. - Pre-PASCAL: Pre-PASCAL is an exclusive 3-D FEM pre-processor for generating the input transient data provided with 3 RPV mesh models and two simple specimen mesh models, i.e. CT and CCP. Almost the same input data format with that of PASCAL main processor is used. Output data of temperature and stress distribution
Coarse mode aerosols in the High Arctic
Baibakov, K.; O'Neill, N. T.; Chaubey, J. P.; Saha, A.; Duck, T. J.; Eloranta, E. W.
2014-12-01
Fine mode (submicron) aerosols in the Arctic have received a fair amount of scientific attention in terms of smoke intrusions during the polar summer and Arctic haze pollution during the polar winter. Relatively little is known about coarse mode (supermicron) aerosols, notably dust, volcanic ash and sea salt. Asian dust is a regular springtime event whose optical and radiative forcing effects have been fairly well documented at the lower latitudes over North America but rarely reported for the Arctic. Volcanic ash, whose socio-economic importance has grown dramatically since the fear of its effects on aircraft engines resulted in the virtual shutdown of European civil aviation in the spring of 2010 has rarely been reported in the Arctic in spite of the likely probability that ash from Iceland and the Aleutian Islands makes its way into the Arctic and possibly the high Arctic. Little is known about Arctic sea salt aerosols and we are not aware of any literature on the optical measurement of these aerosols. In this work we present preliminary results of the combined sunphotometry-lidar analysis at two High Arctic stations in North America: PEARL (80°N, 86°W) for 2007-2011 and Barrow (71°N,156°W) for 2011-2014. The multi-years datasets were analyzed to single out potential coarse mode incursions and study their optical characteristics. In particular, CIMEL sunphotometers provided coarse mode optical depths as well as information on particle size and refractive index. Lidar measurements from High Spectral Resolution lidars (AHSRL at PEARL and NSHSRL at Barrow) yielded vertically resolved aerosol profiles and gave an indication of particle shape and size from the depolarization ratio and color ratio profiles. Additionally, we employed supplementary analyses of HYSPLIT backtrajectories, OMI aerosol index, and NAAPS (Navy Aerosol Analysis and Prediction System) outputs to study the spatial context of given events.
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
Tensile Behaviour of Welded Wire Mesh and Hexagonal Metal Mesh for Ferrocement Application
Tanawade, A. G.; Modhera, C. D.
2017-08-01
Tension tests were conducted on welded mesh and hexagonal Metal mesh. Welded Mesh is available in the market in different sizes. The two types are analysed viz. Ø 2.3 mm and Ø 2.7 mm welded mesh, having opening size 31.75 mm × 31.75 mm and 25.4 mm × 25.4 mm respectively. Tensile strength test was performed on samples of welded mesh in three different orientations namely 0°, 30° and 45° degrees with the loading axis and hexagonal Metal mesh of Ø 0.7 mm, having opening 19.05 × 19.05 mm. Experimental tests were conducted on samples of these meshes. The objective of this study was to investigate the behaviour of the welded mesh and hexagonal Metal mesh. The result shows that the tension load carrying capacity of welded mesh of Ø 2.7 mm of 0° orientation is good as compared to Ø2.3 mm mesh and ductility of hexagonal Metal mesh is good in behaviour.
Zhang, Fang; Merrill, Matthew D.; Tokash, Justin C.; Saito, Tomonori; Cheng, Shaoan; Hickner, Michael A.; Logan, Bruce E.
2011-01-01
that the mesh properties of these cathodes can significantly affect performance. Cathodes made from the coarsest mesh (30-mesh) achieved the highest maximum power of 1616 ± 25 mW m-2 (normalized to cathode projected surface area; 47.1 ± 0.7 W m-3 based on liquid
Intravesical midurethral sling mesh erosion secondary to transvaginal mesh reconstructive surgery
Directory of Open Access Journals (Sweden)
Sukanda Bin Jaili
2015-05-01
Conclusion: Repeated vaginal reconstructive surgery may jeopardize a primary mesh or sling, and pose a high risk of mesh erosion, which may be delayed for several years. Removal of the mesh erosion and bladder repair are feasible pervaginally with good outcome.
Yang, Dikun; Oldenburg, Douglas W.; Haber, Eldad
2014-03-01
Airborne electromagnetic (AEM) methods are highly efficient tools for assessing the Earth's conductivity structures in a large area at low cost. However, the configuration of AEM measurements, which typically have widely distributed transmitter-receiver pairs, makes the rigorous modelling and interpretation extremely time-consuming in 3-D. Excessive overcomputing can occur when working on a large mesh covering the entire survey area and inverting all soundings in the data set. We propose two improvements. The first is to use a locally optimized mesh for each AEM sounding for the forward modelling and calculation of sensitivity. This dedicated local mesh is small with fine cells near the sounding location and coarse cells far away in accordance with EM diffusion and the geometric decay of the signals. Once the forward problem is solved on the local meshes, the sensitivity for the inversion on the global mesh is available through quick interpolation. Using local meshes for AEM forward modelling avoids unnecessary computing on fine cells on a global mesh that are far away from the sounding location. Since local meshes are highly independent, the forward modelling can be efficiently parallelized over an array of processors. The second improvement is random and dynamic down-sampling of the soundings. Each inversion iteration only uses a random subset of the soundings, and the subset is reselected for every iteration. The number of soundings in the random subset, determined by an adaptive algorithm, is tied to the degree of model regularization. This minimizes the overcomputing caused by working with redundant soundings. Our methods are compared against conventional methods and tested with a synthetic example. We also invert a field data set that was previously considered to be too large to be practically inverted in 3-D. These examples show that our methodology can dramatically reduce the processing time of 3-D inversion to a practical level without losing resolution
Tetrahedral-Mesh Simulation of Turbulent Flows with the Space-Time Conservative Schemes
Chang, Chau-Lyan; Venkatachari, Balaji; Cheng, Gary C.
2015-01-01
Direct numerical simulations of turbulent flows are predominantly carried out using structured, hexahedral meshes despite decades of development in unstructured mesh methods. Tetrahedral meshes offer ease of mesh generation around complex geometries and the potential of an orientation free grid that would provide un-biased small-scale dissipation and more accurate intermediate scale solutions. However, due to the lack of consistent multi-dimensional numerical formulations in conventional schemes for triangular and tetrahedral meshes at the cell interfaces, numerical issues exist when flow discontinuities or stagnation regions are present. The space-time conservative conservation element solution element (CESE) method - due to its Riemann-solver-free shock capturing capabilities, non-dissipative baseline schemes, and flux conservation in time as well as space - has the potential to more accurately simulate turbulent flows using unstructured tetrahedral meshes. To pave the way towards accurate simulation of shock/turbulent boundary-layer interaction, a series of wave and shock interaction benchmark problems that increase in complexity, are computed in this paper with triangular/tetrahedral meshes. Preliminary computations for the normal shock/turbulence interactions are carried out with a relatively coarse mesh, by direct numerical simulations standards, in order to assess other effects such as boundary conditions and the necessity of a buffer domain. The results indicate that qualitative agreement with previous studies can be obtained for flows where, strong shocks co-exist along with unsteady waves that display a broad range of scales, with a relatively compact computational domain and less stringent requirements for grid clustering near the shock. With the space-time conservation properties, stable solutions without any spurious wave reflections can be obtained without a need for buffer domains near the outflow/farfield boundaries. Computational results for the
Enhanced Accident Tolerant LWR Fuels National Metrics Workshop Report
Energy Technology Data Exchange (ETDEWEB)
Lori Braase
2013-01-01
The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), in collaboration with the nuclear industry, has been conducting research and development (R&D) activities on advanced Light Water Reactor (LWR) fuels for the last few years. The emphasis for these activities was on improving the fuel performance in terms of increased burnup for waste minimization and increased power density for power upgrades, as well as collaborating with industry on fuel reliability. After the events at the Fukushima Nuclear Power Plant in Japan in March 2011, enhancing the accident tolerance of LWRs became a topic of serious discussion. In the Consolidated Appropriations Act, 2012, Conference Report 112-75, the U.S. Congress directed DOE-NE to: • Give “priority to developing enhanced fuels and cladding for light water reactors to improve safety in the event of accidents in the reactor or spent fuel pools.” • Give “special technical emphasis and funding priority…to activities aimed at the development and near-term qualification of meltdown-resistant, accident-tolerant nuclear fuels that would enhance the safety of present and future generations of light water reactors.” • Report “to the Committee, within 90 days of enactment of this act, on its plan for development of meltdown-resistant fuels leading to reactor testing and utilization by 2020.” Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, and operational transients, as well as design-basis and beyond design-basis events. The overall draft strategy for development and demonstration is comprised of three phases: Feasibility Assessment and Down-selection; Development and Qualification; and
Early deterioration of coarse woody debris.
Energy Technology Data Exchange (ETDEWEB)
Tainter, Frank, H.; McMinn, James, W.
1999-02-16
Tainter, F.H., and J.W. McMinn. 1999. Early deterioration of coarse woody debris. In: Proc. Tenth Bien. South. Silv. Res. Conf. Shreveport, LA, February 16-18, 1999. Pp. 232-237 Abstract - Coarse woody debris (CWD) is an important structural component of southern forest ecosystems. CWD loading may be affected by different decomposition rates on sites of varying quality. Bolts of red oak and loblolly pine were placed on plots at each of three (hydric, mesic. and xerlc) sites at the Savannah River Site and sampled over a I6-week period. Major changes were in moisture content and nonstructural carbohydrate content (total carbohydrates, reducing sugars, and starch) of sapwood. Early changes in nonstructural carbohydrate levels following placement of the bolts were likely due to reallocation of these materials by sapwood parenchyma cells. These carbohydrates later formed pools increasingly metabolized by bacteria and invading fungi. Most prevalent fungi in sapwood were Ceratocysfis spp. in pine and Hypoxy/on spp. in oak. Although pine sapwood became blue stained and oak sapwood exhibited yellow soft decay with black zone lines, estimators of decay (specific gravity, sodium hydroxide solubility, and holocellulose content) were unchanged during the 16-week study period. A small effect of site was detected for starch content of sapwood of both species. Fungal biomass in sapwood of both species, as measured by ergosterol content, was detectable at week zero, increased somewhat by week three and increased significantly by week 16.
Flooding of a large, passive, pressure-tube LWR
Energy Technology Data Exchange (ETDEWEB)
Hejzlar, P.; Todreas, N.E.; Driscoll, M.J. [Massachusetts Institute of Technology, Cambridge, MA (United States)
1995-09-01
A reactor concept has been developed which can survive LOCA without scram and without replenishing primary coolant inventory. The proposed concept is a pressure tube type reactor similar to CANDU reactors, but differing in three key aspects: (1) a solid SiC-coated graphite fuel matrix is used in place of fuel pin bundles, (2) the heavy water coolant in the pressure tubes is replaced by light water, and (3) the calandria tank contains a low pressure gas instead of heavy water moderator. The gas displaces the light water from the calandria during normal operation, while during loss of coolant or loss of heat sink accidents, it allows passive calandria flooding. This paper describes the thermal hydraulic characteristics of the gravity driven calandria flooding process. Flooding the calandria space with light water is a unique and very important feature of the proposed pressure-tube LWR concept. The flooding of the top row of fuel channels must be accomplished fast enough so that none of the critical components of the fuel channel exceed their design limits. The flooding process has been modeled and shown to be rapid enough to maintain all components within their design limits. Two other considerations are important. The thermal shock experienced by the calandria and pressure tubes has been evaluated and shown to be within acceptable bounds. Finally, although complete flooding renders the reactor deeply subcritical, various steam/water densities can be hypothesized to be present during the flooding process which could cause reactivity to increase from the initially voided calandria case. One such hypothesis which leads to the maximum possible density of the steam/water mixture in the still unflooded calandria space is entrainment from the free surface. It is shown that the steam/water mixture density yielding the maximum reactivity peak cannot be achieved by entrainment because it exceeds thermohydraulically attainable densities of steam/water by an order of magnitude.
Survey of LWR environmental control technology performance and cost
International Nuclear Information System (INIS)
Heeb, C.M.; Aaberg, R.L.; Cole, B.M.; Engel, R.L.; Kennedy, W.E. Jr.; Lewallen, M.A.
1980-03-01
This study attempts to establish a ranking for species that are routinely released to the environment for a projected nuclear power growth scenario. Unlike comparisons made to existing standards, which are subject to frequent revision, the ranking of releases can be used to form a more logical basis for identifying the areas where further development of control technology could be required. This report describes projections of releases for several fuel cycle scenarios, identifies areas where alternative control technologies may be implemented, and discusses the available alternative control technologies. The release factors were used in a computer code system called ENFORM, which calculates the annual release of any species from any part of the LWR nuclear fuel cycle given a projection of installed nuclear generation capacity. This survey of fuel cycle releases was performed for three reprocessing scenarios (stowaway, reprocessing without recycle of Pu and reprocessing with full recycle of U and Pu) for a 100-year period beginning in 1977. The radioactivity releases were ranked on the basis of a relative ranking factor. The relative ranking factor is based on the 100-year summation of the 50-year population dose commitment from an annual release of radioactive effluents. The nonradioactive releases were ranked on the basis of dilution factor. The twenty highest ranking radioactive releases were identified and each of these was analyzed in terms of the basis for calculating the release and a description of the currently employed control method. Alternative control technology is then discussed, along with the available capital and operating cost figures for alternative control methods
Feasibility study on the development of advanced LWR fuel technology
Energy Technology Data Exchange (ETDEWEB)
Jung, Youn Ho; Sohn, D. S.; Jeong, Y. H.; Song, K. W.; Song, K. N.; Chun, T. H.; Bang, J. G.; Bae, K. K.; Kim, D. H. and others
1997-07-01
Worldwide R and D trends related to core technology of LWR fuels and status of patents have been surveyed for the feasibility study. In addition, various fuel cycle schemes have been studied to establish the target performance parameters. For the development of cladding material, establishment of long-term research plan for alloy development and optimization of melting process and manufacturing technology were conducted. A work which could characterize the effect of sintering additives on the microstructure of UO{sub 2} pellet has been experimentally undertaken, and major sintering variables and their ranges have been found in the sintering process of UO{sub 2}-Gd{sub 2}O{sub 3} burnable absorber pellet. The analysis of state of the art technology related to flow mixing device for spacer grid and debris filtering device for bottom nozzle and the investigation of the physical phenomena related to CHF enhancement and the establishment of the data base for thermal-hydraulic performance tests has been done in this study. In addition, survey on the documents of the up-to-date PWR fuel assemblies developed by foreign vendors have been carried out to understand their R and D trends and establish the direction of R and D for these structural components. And, to set the performance target of the new fuel, to be developed, fuel burnup and economy under the extended fuel cycle length scheme were estimated. A preliminary study on the failure mechanism of CANDU fuel, key technology and advanced coating has been performed. (author). 190 refs., 31 tabs., 129 figs.
Safety aspects and operating experience of LWR plants in Japan
International Nuclear Information System (INIS)
Aoki, S.; Hinoki, M.
1977-01-01
From the outset of nuclear power development in Japan, major emphasis has been placed on the safety of the nuclear power plants. There are now twelve nuclear power plants in operation with a total output of 6600 MWe. Their operating records were generally satisfactory, but in the 1974 to 1975 period, they experienced somewhat declined availability due to the repair work under the specific circumstances. After investigation of causes of troubles and the countermeasures thereof were made to ensure safety, they are now keeping good performance. In Japan, nuclear power plants are strictly subject to sufficient and careful inspection in compliance with the safety regulation, and are placed under stringent radiation control of employees. Under the various circumstances, however, the period of annual inspection tends to be prolonged more than originally planned, and this consequently is considered to be one of the causes of reduced availability. In order to develop nuclear power generation for the future, it is necessary to put further emphasis on the assurance of safety and to endeavor to devise measures to improve availability of the plants, based on the careful analysis of causes which reduce plant availability. This paper discusses the results of studies made for the following items from such viewpoints: (1) Safety and Operating Experience of LWR Nuclear Power Plants in Japan; a) Operating experience with light water reactors b) Improvements in design of light water reactors during the past ten years c) Analysis of the factors which affect plant availability; 2) Assurance of Safety and Measures to Increase Availability a) Measures for safety and environmental protection b) Measures to reduce radiation exposure of employees c) Appropriateness of maintenance and inspection work d) Measures to increase plant availability e) Measures to improve reliability of equipments and components; and 3) Future Technical Problems
A review on future trends of LWR fuel cycle costs
International Nuclear Information System (INIS)
Tamiya, S.; Otomo, T.; Meguro, T.
1977-01-01
In the cost estimations in the past, the main components of fuel cycle were mining and milling, uranium enrichment and fuel fabrication, and reprocessing charge deemed to be recovered by plutonium credit. Since the oil crisis, every component of fuel cycle cost has gone up in recent years as well as the construction cost of a power station. Recent analysis shows that the costs in the back end of fuel cycle are much higher than those anticipated several years ago, although their contribution to the electricity generating cost by nuclear would be small. The situation of the back end of the fuel cycle has been quite changed in recent years, and there are still many uncertainties in this field, that is, regulatory requirements for reprocessing plant such as safety, safeguards, environmental protection and high level waste management. So, it makes it more difficult to estimate the investment in this sector of fuel cycle, therefore, to estimate the cost of this sector. The institutional problems must be cleared in relation to the ultimate disposal of high level waste, too. Co-location of some parts of fuel cycle facilities may also affect on the fuel cycle costs. In this paper a review is made of the future trend of nuclear fuel cycle cost of LWR based on the recent analysis. Those factors which affect the fuel cycle costs are also discussed. In order to reduce the uncertainties of the cost estimations as soon as possible, the necessity is emphasized to discuss internationally such items as the treatment and disposal of high level radioactive wastes, siting issues of a reprocessing plant, physical protection of plutonium and the effects of plutonium on the environment
Laparoscopic Pelvic Floor Repair Using Polypropylene Mesh
Directory of Open Access Journals (Sweden)
Shih-Shien Weng
2008-09-01
Conclusion: Laparoscopic pelvic floor repair using a single piece of polypropylene mesh combined with uterosacral ligament suspension appears to be a feasible procedure for the treatment of advanced vaginal vault prolapse and enterocele. Fewer mesh erosions and postoperative pain syndromes were seen in patients who had no previous pelvic floor reconstructive surgery.
Robust diamond meshes with unique wettability properties.
Yang, Yizhou; Li, Hongdong; Cheng, Shaoheng; Zou, Guangtian; Wang, Chuanxi; Lin, Quan
2014-03-18
Robust diamond meshes with excellent superhydrophobic and superoleophilic properties have been fabricated. Superhydrophobicity is observed for water with varying pH from 1 to 14 with good recyclability. Reversible superhydrophobicity and hydrophilicity can be easily controlled. The diamond meshes show highly efficient water-oil separation and water pH droplet transference.
Mesh-graft urethroplasty: a case report
田中, 敏博; 滝川, 浩; 香川, 征; 長江, 浩朗
1987-01-01
We used a meshed free-foreskin transplant in a two-stage procedure for reconstruction of the extended stricture of urethra after direct vision urethrotomy. The results were excellent. Mesh-graft urethroplasty is a useful method for patients with extended strictures of the urethra or recurrent strictures after several operations.
7th International Meshing Roundtable '98
Energy Technology Data Exchange (ETDEWEB)
Eldred, T.J.
1998-10-01
The goal of the 7th International Meshing Roundtable is to bring together researchers and developers from industry, academia, and government labs in a stimulating, open environment for the exchange of technical information related to the meshing process. In the past, the Roundtable has enjoyed significant participation from each of these groups from a wide variety of countries.
Postoperative pain outcomes after transvaginal mesh revision.
Danford, Jill M; Osborn, David J; Reynolds, W Stuart; Biller, Daniel H; Dmochowski, Roger R
2015-01-01
Although the current literature discusses mesh complications including pain, as well as suggesting different techniques for removing mesh, there is little literature regarding pain outcomes after surgical removal or revision. The purpose of this study is to determine if surgical removal or revision of vaginal mesh improves patient's subjective complaints of pelvic pain associated with original placement of mesh. After obtaining approval from the Vanderbilt University Medical Center Institutional Review Board, a retrospective review of female patients with pain secondary to previous mesh placement who underwent excision or revision of vaginal mesh from January 2000 to August 2012 was performed. Patient age, relevant medical history including menopause status, previous hysterectomy, smoking status, and presence of diabetes, fibromyalgia, interstitial cystitis, and chronic pelvic pain, was obtained. Patients' postoperative pain complaints were assessed. Of the 481 patients who underwent surgery for mesh revision, removal or urethrolysis, 233 patients met our inclusion criteria. One hundred and sixty-nine patients (73 %) reported that their pain improved, 19 (8 %) reported that their pain worsened, and 45 (19 %) reported that their pain remained unchanged after surgery. Prior history of chronic pelvic pain was associated with increased risk of failure of the procedure to relieve pain (OR 0.28, 95 % CI 0.12-0.64, p = 0.003). Excision or revision of vaginal mesh appears to be effective in improving patients' pain symptoms most of the time. Patients with a history of chronic pelvic pain are at an increased risk of no improvement or of worsening pain.
Converting skeletal structures to quad dominant meshes
DEFF Research Database (Denmark)
Bærentzen, Jakob Andreas; Misztal, Marek Krzysztof; Welnicka, Katarzyna
2012-01-01
We propose the Skeleton to Quad-dominant polygonal Mesh algorithm (SQM), which converts skeletal structures to meshes composed entirely of polar and annular regions. Both types of regions have a regular structure where all faces are quads except for a single ring of triangles at the center of each...
Adaptive mesh refinement in titanium
Energy Technology Data Exchange (ETDEWEB)
Colella, Phillip; Wen, Tong
2005-01-21
In this paper, we evaluate Titanium's usability as a high-level parallel programming language through a case study, where we implement a subset of Chombo's functionality in Titanium. Chombo is a software package applying the Adaptive Mesh Refinement methodology to numerical Partial Differential Equations at the production level. In Chombo, the library approach is used to parallel programming (C++ and Fortran, with MPI), whereas Titanium is a Java dialect designed for high-performance scientific computing. The performance of our implementation is studied and compared with that of Chombo in solving Poisson's equation based on two grid configurations from a real application. Also provided are the counts of lines of code from both sides.
Technical development on burn-up credit for spent LWR fuels
International Nuclear Information System (INIS)
Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori
2000-10-01
Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled 'Technical Development on Criticality Safety Management for Spent LWR Fuels'. Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burn-up and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report. (author)
Technical Development on Burn-up Credit for Spent LWR Fuel
Energy Technology Data Exchange (ETDEWEB)
Gauld, I.C.
2001-12-26
Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled ''Technical Development on Criticality Safety Management for Spent LWR Fuels.'' Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burnup and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report.
Effects of Listening While Reading (LWR on Swahili Reading Fluency and Comprehension
Directory of Open Access Journals (Sweden)
Filipo Lubua
2016-10-01
Full Text Available A number of studies have examined the contribution of technology in teaching such languages as English, French, and Spanish, among many others. Contrarily, most LCTL’s, have received very little attention. This study investigates if listening while reading (LWR may expedite Swahili reading fluency and comprehension. The study employed the iBook Author tool to create weekly mediated and interactive reading texts, with comprehension exercises, which were eventually used to collect descriptive and qualitative data from four Elementary Swahili students. Participants participated in a seven week reading program, which provided them with some kind of directed self-learning, and met with the instructor for at least 30 minutes every week for observation and more reading activities. The teacher recorded their reading scores, and a number of themes on how LWR influenced reading fluency and comprehension are discussed here. It shows that participants have a positive attitude towards LWR and they suggest it for all the reading classes.
Technical development on burn-up credit for spent LWR fuels
Energy Technology Data Exchange (ETDEWEB)
Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2000-10-01
Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled 'Technical Development on Criticality Safety Management for Spent LWR Fuels'. Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burn-up and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report. (author)
EUV lithography for 22nm half pitch and beyond: exploring resolution, LWR, and sensitivity tradeoffs
Putna, E. Steve; Younkin, Todd R.; Leeson, Michael; Caudillo, Roman; Bacuita, Terence; Shah, Uday; Chandhok, Manish
2011-04-01
The International Technology Roadmap for Semiconductors (ITRS) denotes Extreme Ultraviolet (EUV) lithography as a leading technology option for realizing the 22nm half pitch node and beyond. According to recent assessments made at the 2010 EUVL Symposium, the readiness of EUV materials remains one of the top risk items for EUV adoption. The main development issue regarding EUV resists has been how to simultaneously achieve high resolution, high sensitivity, and low line width roughness (LWR). This paper describes our strategy, the current status of EUV materials, and the integrated post-development LWR reduction efforts made at Intel Corporation. Data collected utilizing Intel's Micro- Exposure Tool (MET) is presented in order to examine the feasibility of establishing a resist process that simultaneously exhibits <=22nm half-pitch (HP) L/S resolution at <=11.3mJ/cm2 with <=3nm LWR.
International Nuclear Information System (INIS)
Prince, B.E.; Hadley, S.W.
1983-01-01
This is the second of a two-part report intended as a critical review of certain issues involved with closing the Light Water Reactor (LWR) fuel cycle and establishing the basis for future transition to commercial breeder applications. The report is divided into four main sections consisting of (1) a review of the status of the LWR spent fuel management and storage problem; (2) an analysis of the economic incentives for instituting reprocessing and recycle in LWRs; (3) an analysis of the time-dependent aspects of plutonium economic value particularly as related to the LWR-breeder transition; and (4) an analysis of the time-dependent aspects of plutonium requirements and supply relative to this transition
Pretest aerosol code comparisons for LWR aerosol containment tests LA1 and LA2
International Nuclear Information System (INIS)
Wright, A.L.; Wilson, J.H.; Arwood, P.C.
1986-01-01
The Light-Water-Reactor (LWR) Aerosol Containment Experiments (LACE) are being performed in Richland, Washington, at the Hanford Engineering Development Laboratory (HEDL) under the leadership of an international project board and the Electric Power Research Institute. These tests have two objectives: (1) to investigate, at large scale, the inherent aerosol retention behavior in LWR containments under simulated severe accident conditions, and (2) to provide an experimental data base for validating aerosol behavior and thermal-hydraulic computer codes. Aerosol computer-code comparison activities are being coordinated at the Oak Ridge National Laboratory. For each of the six LACE tests, ''pretest'' calculations (for code-to-code comparisons) and ''posttest'' calculations (for code-to-test data comparisons) are being performed. The overall goals of the comparison effort are (1) to provide code users with experience in applying their codes to LWR accident-sequence conditions and (2) to evaluate and improve the code models
Automatic mesh generation with QMESH program
International Nuclear Information System (INIS)
Ise, Takeharu; Tsutsui, Tsuneo
1977-05-01
Usage of the two-dimensional self-organizing mesh generation program, QMESH, is presented together with the descriptions and the experience, as it has recently been converted and reconstructed from the NEACPL version to the FACOM. The program package consists of the QMESH code to generate quadrilaterial meshes with smoothing techniques, the QPLOT code to plot the data obtained from the QMESH on the graphic COM, and the RENUM code to renumber the meshes by using a bandwidth minimization procedure. The technique of mesh reconstructuring coupled with smoothing techniques is especially useful when one generates the meshes for computer codes based on the finite element method. Several typical examples are given for easy access to the QMESH program, which is registered in the R.B-disks of JAERI for users. (auth.)
Feasibility assessment of the once-through thorium fuel cycle for the PTVM LWR concept
International Nuclear Information System (INIS)
Rachamin, R.; Fridman, E.; Galperin, A.
2015-01-01
Highlights: • The PTVM LWR is an innovation reactor concept operating in a “breed & burn” mode. • An advanced once-through thorium fuel cycle for the PTVM LWR concept is proposed. • The PTVM LWR concept makes use of a seed-blanket geometry. • A novel fuel management scheme based on two separate fuel flow routes is analyzed. • The analysis indicates a potential for utilizing the fuel in an efficient manner. - Abstract: This paper investigates the feasibility of a once-through thorium fuel cycle for the novel reactor-design concept named the pressure tube light water reactor with variable moderator control (PTVM LWR). The PTVM LWR operates in a “breed & burn” mode, which makes it an attractive system for utilizing thorium fuel in a once-through mode. The “breed & burn” mode can emphasize the in situ generation as well as incineration of 233 U, which are the basic foundations of the once-through thorium fuel cycle. The PTVM LWR concept makes use of a seed–blanket geometry, whereby the core is divided into separated regions of thorium-based fuel channel assemblies (blanket) and low-enriched uranium (LEU) based fuel channel assemblies (seed). A novel fuel in-core management scheme based on two separate fuel flow routes (i.e., seed route and blanket route) is proposed and analyzed. Neutronic performance analysis indicates that the proposed novel fuel in-core management scheme has the potential to utilize both LEU- and thorium-based fuel in an efficient manner. The once-through thorium cycle, presented and discussed in this paper, provide interesting research leads and can serve as a bridge between current LEU-based fuel cycles and a thorium fuel cycle based on recycling of 233 U
Fog water collection effectiveness: Mesh intercomparisons
Fernandez, Daniel; Torregrosa, Alicia; Weiss-Penzias, Peter; Zhang, Bong June; Sorensen, Deckard; Cohen, Robert; McKinley, Gareth; Kleingartner, Justin; Oliphant, Andrew; Bowman, Matthew
2018-01-01
To explore fog water harvesting potential in California, we conducted long-term measurements involving three types of mesh using standard fog collectors (SFC). Volumetric fog water measurements from SFCs and wind data were collected and recorded in 15-minute intervals over three summertime fog seasons (2014–2016) at four California sites. SFCs were deployed with: standard 1.00 m2 double-layer 35% shade coefficient Raschel; stainless steel mesh coated with the MIT-14 hydrophobic formulation; and FogHa-Tin, a German manufactured, 3-dimensional spacer fabric deployed in two orientations. Analysis of 3419 volumetric samples from all sites showed strong relationships between mesh efficiency and wind speed. Raschel mesh collected 160% more fog water than FogHa-Tin at wind speeds less than 1 m s–1 and 45% less for wind speeds greater than 5 m s–1. MIT-14 coated stainless-steel mesh collected more fog water than Raschel mesh at all wind speeds. At low wind speeds of steel mesh collected 3% more and at wind speeds of 4–5 m s–1, it collected 41% more. FogHa-Tin collected 5% more fog water when the warp of the weave was oriented vertically, per manufacturer specification, than when the warp of the weave was oriented horizontally. Time series measurements of three distinct mesh across similar wind regimes revealed inconsistent lags in fog water collection and inconsistent performance. Since such differences occurred under similar wind-speed regimes, we conclude that other factors play important roles in mesh performance, including in-situ fog event and aerosol dynamics that affect droplet-size spectra and droplet-to-mesh surface interactions.
International Nuclear Information System (INIS)
1976-05-01
Volume I of the five-volume report contains executive and technical summaries of the entire report, background information of the LWR fuel cycle alternatives, descriptions of waste types, and projections of waste quantities. Overview characterizations of alternative LWR fuel cycle modes are also included
Review of literature on the TMI accident and correlation to the LWR Safety Technology Program
Energy Technology Data Exchange (ETDEWEB)
Miller, W.J.
1980-05-01
This report is the result of approximately two man-months of effort devoted to assimilating and comprehending significant publicly available material related to Three Mile Island Unit 2 and events during and subsequent to the accident experienced on March 28, 1979. Those events were then correlated with the Preliminary LWR Safety Technology Program Plan (Preliminary Program Plan) prepared for the US Department of Energy by Sandia National Lab. This report is being submitted simultaneously with the SAI report entitled Preliminary Prioritization of Tasks in the Draft LWR Safety Technology Program Plan.
ORIGEN2 libraries based on JENDL-3.2 for LWR-MOX fuels
Energy Technology Data Exchange (ETDEWEB)
Suyama, Kenya; Katakura, Jun-ichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Onoue, Masaaki; Matsumoto, Hideki [Mitsubishi Heavy Industries Ltd., Tokyo (Japan); Sasahara, Akihiro [Central Research Inst. of Electric Power Industry, Tokyo (Japan)
2000-11-01
A set of ORIGEN2 libraries for LWR MOX fuels was developed based on JENDL-3.2. The libraries were compiled with SWAT using the specification of MOX fuels that will be used in nuclear power reactors in Japan. The verification of the libraries were performed by the analyses of post irradiation examinations for the fuels from European PWR. By the analysis of PIE data from PWR in United States, the comparison was made between calculation and experimental results in the case of that parameters for making the libraries are different from irradiation conditions. These new libraries for LWR MOX fuels are packaged in ORLIBJ32, the libraries released in 1999. (author)
Review of literature on the TMI accident and correlation to the LWR Safety Technology Program
International Nuclear Information System (INIS)
Miller, W.J.
1980-05-01
This report is the result of approximately two man-months of effort devoted to assimilating and comprehending significant publicly available material related to Three Mile Island Unit 2 and events during and subsequent to the accident experienced on March 28, 1979. Those events were then correlated with the Preliminary LWR Safety Technology Program Plan (Preliminary Program Plan) prepared for the US Department of Energy by Sandia National Lab. This report is being submitted simultaneously with the SAI report entitled Preliminary Prioritization of Tasks in the Draft LWR Safety Technology Program Plan
Comparison of scale/triton and helios burnup calculations for high burnup LWR fuel
Energy Technology Data Exchange (ETDEWEB)
Tittelbach, S.; Mispagel, T.; Phlippen, P.W. [WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Juelich (Germany)
2009-07-01
The presented analyses provide information about the suitability of the lattice burnup code HELIOS and the recently developed code SCALE/TRITON for the prediction of isotopic compositions of high burnup LWR fuel. The accurate prediction of the isotopic inventory of high burnt spent fuel is a prerequisite for safety analyses in and outside of the reactor core, safe loading of spent fuel into storage casks, design of next generation spent fuel casks and for any consideration of burnup credit. Depletion analyses are performed with both burnup codes for PWR and BWR fuel samples which were irradiated far beyond 50 GWd/t within the LWR-PROTEUS Phase II project. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Rao, Appajosula S., E-mail: Appajosula.Rao@nrc.gov
2014-04-01
Austenitic stainless steels (SSs) are extensively being used in the fabrication of light water reactor (LWR) core internal components. It is because these steels have relatively high ductility, fracture toughness and moderate strength. However, the LWR internal components exposure to neutron irradiation over an extended period of plant operation degrades the materials mechanical properties such as the fracture toughness. This paper summarizes some of the results of the existing open literature data on irradiation assisted stress corrosion cracking (IASCC) of 316 CW steels that have been published by the United States Nuclear Regulatory Commission (USNRC), industry, academia, and other research agencies.
Generalized perturbation theory for LWR depletion analysis and core design applications
International Nuclear Information System (INIS)
White, J.R.; Frank, B.R.
1986-01-01
A comprehensive time-dependent perturbation theory formulation that includes macroscopic depletion, thermal-hydraulic and poison feedback effects, and a criticality reset mechanism is developed. The methodology is compatible with most current LWR design codes. This new development allows GTP/DTP methods to be used quantitatively in a variety of realistic LWR physics applications that were not possible prior to this work. A GTP-based optimization technique for incore fuel management analyses is addressed as a promising application of the new formulation
Testing for entanglement with periodic coarse graining
Tasca, D. S.; Rudnicki, Łukasz; Aspden, R. S.; Padgett, M. J.; Souto Ribeiro, P. H.; Walborn, S. P.
2018-04-01
Continuous-variable systems find valuable applications in quantum information processing. To deal with an infinite-dimensional Hilbert space, one in general has to handle large numbers of discretized measurements in tasks such as entanglement detection. Here we employ the continuous transverse spatial variables of photon pairs to experimentally demonstrate entanglement criteria based on a periodic structure of coarse-grained measurements. The periodization of the measurements allows an efficient evaluation of entanglement using spatial masks acting as mode analyzers over the entire transverse field distribution of the photons and without the need to reconstruct the probability densities of the conjugate continuous variables. Our experimental results demonstrate the utility of the derived criteria with a success rate in entanglement detection of ˜60 % relative to 7344 studied cases.
Field description of coarse bioclastic fabrics
Energy Technology Data Exchange (ETDEWEB)
Kidwell, S.M.; Holland, S.M. (Univ. of Chicago, IL (United States))
1991-08-01
Shell- and bone-bearing rocks can be readily categorized into 9 macroscopic fabric types using semi-quantitative scales for close-packing and size-sorting of bioclasts greater than 2 mm in diameter. Although designed to describe fossiliferous siliciclastics and volcaniclastics, this system of field description can also be used to enlarge upon standard petrographic descriptions of fossiliferous carbonates. In cross-sectional bed views, coarse bioclasts may be densely packed. These coarse bioclasts (>2 mm) may be well sorted (central 80% of bioclasts lie within 1 or 2 adjacent phi size-classes), bimodal (well sorted but with a distinct second mode), or poorly sorted (central 80% of bioclasts distributed over 3 or more adjacent size-classes). Despite the complicating effects of bioclast shape, novices show 90% accuracy in estimating close-packing from photographs. They have only 60% accuracy in estimating size-sorting (the most common error is underestimating goodness of sorting), underscoring the importance of size-tallies to cross-check visual estimates when first using this scheme. This packing/sorting approach provides a good visual image of the fabric, and narrows the range of possible modes of origin more than alternative criteria such as volumetric percent-abundance (which shows no one-to-one equivalence with close-packing), orientation, and fragmentation. However, detailed interpretations of fabrics usually require more information on these and other features of the deposit, including bioclast condition, associated sedimentary structures, life-habits of bioclast-producers, and stratigraphic context.
International Nuclear Information System (INIS)
Barrera, J.; Corisco, M.; Riverola, J.
2010-01-01
Since the construction of the first light water reactors (LWR) safety analysis has played a very important role in the operation and its evolution to come up with designs that are currently operating. With new tools available, this role will see increased allowing more efficient operation with security assessments in real time, and a more efficient designs both in terms of fuel efficiency and from the security of the plant during operation.
LWR safety research in the Federal Republic of Germany
International Nuclear Information System (INIS)
Seipel, H.G.
1977-01-01
The paper gives a review of the German LWR safety research programme. It describes how the programme was initiated and informs on its goals, development andpractical realization, and indicates how it is bound up with international collaboration. The contribution so far made by the programme to an enhancement of the understanding of major safety problems and to the improvement of safety technology is demonstrated by means of a few selected examples. Experiments relating to loss-of--coolant accidents have deepened our understanding of the heat transfer in the reactor core during blowdown as well as during the flooding phase. Investigations of the dynamic effects going on in dry full pressure containments and pressure suppression systems, following a loss-of--coolant accident, have indicated that existing computer models cannot satisfactorily predict all relevant physical phenomena. Yet, the experimental results obtained constitute a sufficient basis for safe containment design. Research work on core meltdown accidents has identified the particular importance of the type of concrete used for the containment structures and its foundation. If basaltic concrete is used, a substantial fission product release to the environment is extremely unlikely even in the case of a core meltdown accident. At least, it would take place much later than was previously assumed. Resrach on the safety of pressurized components has been concentrated on the problem of cracks in the heat-affected zone of welds. New methods were developed for the detection and analysis of the acceptability of microcrack fields. Additional investigations of specimens and components to increase the understanding of the long-term behaviour of components with microcracks are envisaged in the frame of a new major project on ''component safety''. Considerable progress has been made in the development of methods for automatic remote-control volumetric testing of reactor pressure vessels using ultrasonic techniques
Phoenix type concepts for transmutation of LWR waste minor actinides
International Nuclear Information System (INIS)
Segev, M.
1994-01-01
A number of variations on the original Phoenix theme were studied. The basic rationale of the Phoenix incinerator is making oxide fuel of the LWR waste minor actinides, loading it in an FFTF-like subcritical core, then bombarding the core with the high current beam accelerated protons to generate considerable energy through spallation and fission reactions. As originally assessed, if the machine is fed with 1600 MeV protons in a 102 mA current, then 8 core modules are driven to transmute the yearly minor actinides waste of 75 1000 MW LWRs into Pu 238 and fission products; in a 2 years cycle the energy extracted is 100000 MW d/T. This performance cannot be substantiated in a rigorous analysis. A calculational consistent methodology, based on a combined execution of the Hermes, NCNP, and Korigen codes, shows, nonetheless that changes in the original Phoenix parameters can upgrade its performance.The original Phoenix contains 26 tons minor actinides in 8 core modules; 1.15 m 3 module is shaped for 40% neutron leakage; with a beam of 102 mA the 8 modules are driven to 100000 MW/T in 10.5 years, burning out the yearly minor actinide waste of 15 LWRs; the operation must be assisted by grid electricity. If the 1.15 m 3 module is shaped to allow only 28% leakage, then a beam of 102 mA will drive the 8 modules to 100000 MW/T in 3.5 years, burning out the yearly minor actinides waste of 45 LWRs. Some net grid electricity will be generated. If 25 tons minor actinides are loaded into 5 modules, each 1.72 m 3 in volume and of 24% leakage, then a 97 mA beam will drive the module to 100000 MW/T in 2.5 years, burning out the yearly minor actinides waste of 70 LWRs. A considerable amount of net grid electricity will be generated. If the lattice is made of metal fuel, and 26 tons minor actinides are loaded into 32 small modules, 0.17 m 3 each, then a 102 mA beam will drive the modules to 100000 MW/T in 2 years, burning out the yearly minor actinides waste of 72 LWRs. A considerable
[CLINICAL EVALUATION OF THE NEW ANTISEPTIC MESHES].
Gogoladze, M; Kiladze, M; Chkhikvadze, T; Jiqia, D
2016-12-01
Improving the results of hernia treatment and prevention of complications became a goal of our research which included two parts - experimental and clinical. Histomorphological and bacteriological researches showed that the best result out of the 3 control groups was received in case of covering implant "Coladerm"+ with chlorhexidine. Based on the experiment results working process continued in clinics in order to test and introduce new "coladerm"+ chlorhexidine covered poliprophilene meshes into practice. For clinical illustration there were 60 patients introduced to the research who had hernioplasty procedures by different nets: I group - standard meshes+"coladerm"+chlorhexidine, 35 patients; II group - standard meshes +"coladerm", 15 patients; III group - standard meshes, 10 patients. Assessment of the wound and echo-control was done post-surgery on the 8th, 30th and 90th days. This clinical research based on the experimental results once again showed the best anti-microbe features of new antiseptic polymeric biocomposite meshes (standard meshes+"coladerm"+chlorhexidine); timely termination of regeneration and reparation processes without any post-surgery suppurative complications. We hope that new antiseptic polymeric biocomposite meshes presented by us will be successfully used in surgical practice of hernia treatment based on and supported by expermental-clinical research.
Fog water collection effectiveness: Mesh intercomparisons
Fernandez, Daniel; Torregrosa, Alicia; Weiss-Penzias, Peter; Zhang, Bong June; Sorensen, Deckard; Cohen, Robert; McKinley, Gareth; Kleingartner, Justin; Oliphant, Andrew; Bowman, Matthew
2018-01-01
To explore fog water harvesting potential in California, we conducted long-term measurements involving three types of mesh using standard fog collectors (SFC). Volumetric fog water measurements from SFCs and wind data were collected and recorded in 15-minute intervals over three summertime fog seasons (2014–2016) at four California sites. SFCs were deployed with: standard 1.00 m2 double-layer 35% shade coefficient Raschel; stainless steel mesh coated with the MIT-14 hydrophobic formulation; and FogHa-Tin, a German manufactured, 3-dimensional spacer fabric deployed in two orientations. Analysis of 3419 volumetric samples from all sites showed strong relationships between mesh efficiency and wind speed. Raschel mesh collected 160% more fog water than FogHa-Tin at wind speeds less than 1 m s–1 and 45% less for wind speeds greater than 5 m s–1. MIT-14 coated stainless-steel mesh collected more fog water than Raschel mesh at all wind speeds. At low wind speeds of wind speeds of 4–5 m s–1, it collected 41% more. FogHa-Tin collected 5% more fog water when the warp of the weave was oriented vertically, per manufacturer specification, than when the warp of the weave was oriented horizontally. Time series measurements of three distinct mesh across similar wind regimes revealed inconsistent lags in fog water collection and inconsistent performance. Since such differences occurred under similar wind-speed regimes, we conclude that other factors play important roles in mesh performance, including in-situ fog event and aerosol dynamics that affect droplet-size spectra and droplet-to-mesh surface interactions.
Transvaginal mesh procedures for pelvic organ prolapse.
Walter, Jens-Erik
2011-02-01
To provide an update on transvaginal mesh procedures, newly available minimally invasive surgical techniques for pelvic floor repair. The discussion is limited to minimally invasive transvaginal mesh procedures. PubMed and Medline were searched for articles published in English, using the key words "pelvic organ prolapse," transvaginal mesh," and "minimally invasive surgery." Results were restricted to systematic reviews, randomized control trials/controlled clinical trials, and observational studies. Searches were updated on a regular basis, and articles were incorporated in the guideline to May 2010. Grey (unpublished) literature was identified through searching the websites of health technology assessment and health technology assessment-related agencies, clinical practice guideline collections, clinical trial registries, and national and international medical specialty societies. The quality of evidence was rated using the criteria described in the Report of the Canadian Task Force on the Preventive Health Care. Recommendations for practice were ranked according to the method described in that report (Table 1). Counselling for the surgical treatment of pelvic organ prolapse should consider all benefits, harms, and costs of the surgical procedure, with particular emphasis on the use of mesh. 1. Patients should be counselled that transvaginal mesh procedures are considered novel techniques for pelvic floor repair that demonstrate high rates of anatomical cure in uncontrolled short-term case series. (II-2B) 2. Patients should be informed of the range of success rates until stronger evidence of superiority is published. (II-2B) 3. Training specific to transvaginal mesh procedures should be undertaken before procedures are performed. (III-C) 4. Patients should undergo thorough preoperative counselling regarding (a) the potential serious adverse sequelae of transvaginal mesh repairs, including mesh exposure, pain, and dyspareunia; and (b) the limited data available
Zhang, Fang
2011-02-01
Mesh current collectors made of stainless steel (SS) can be integrated into microbial fuel cell (MFC) cathodes constructed of a reactive carbon black and Pt catalyst mixture and a poly(dimethylsiloxane) (PDMS) diffusion layer. It is shown here that the mesh properties of these cathodes can significantly affect performance. Cathodes made from the coarsest mesh (30-mesh) achieved the highest maximum power of 1616 ± 25 mW m-2 (normalized to cathode projected surface area; 47.1 ± 0.7 W m-3 based on liquid volume), while the finest mesh (120-mesh) had the lowest power density (599 ± 57 mW m-2). Electrochemical impedance spectroscopy showed that charge transfer and diffusion resistances decreased with increasing mesh opening size. In MFC tests, the cathode performance was primarily limited by reaction kinetics, and not mass transfer. Oxygen permeability increased with mesh opening size, accounting for the decreased diffusion resistance. At higher current densities, diffusion became a limiting factor, especially for fine mesh with low oxygen transfer coefficients. These results demonstrate the critical nature of the mesh size used for constructing MFC cathodes. © 2010 Elsevier B.V. All rights reserved.
Polygonal Prism Mesh in the Viscous Layers for the Polyhedral Mesh Generator, PolyGen
International Nuclear Information System (INIS)
Lee, Sang Yong; Park, Chan Eok; Kim, Shin Whan
2015-01-01
Polyhedral mesh has been known to have some benefits over the tetrahedral mesh. Efforts have been made to set up a polyhedral mesh generation system with open source programs SALOME and TetGen. The evaluation has shown that the polyhedral mesh generation system is promising. But it is necessary to extend the capability of the system to handle the viscous layers to be a generalized mesh generator. A brief review to the previous works on the mesh generation for the viscous layers will be made in section 2. Several challenging issues for the polygonal prism mesh generation will be discussed as well. The procedure to generate a polygonal prism mesh will be discussed in detail in section 3. Conclusion will be followed in section 4. A procedure to generate meshes in the viscous layers with PolyGen has been successfully designed. But more efforts have to be exercised to find the best way for the generating meshes for viscous layers. Using the extrusion direction of the STL data will the first of the trials in the near future
International Nuclear Information System (INIS)
Mazumdar, Tanay; Degweker, S.B.
2017-01-01
Highlights: • In Method of Characteristics, the neutron source within a mesh is expanded up to linear term. • This expansion reduces the number of meshes as compared to flat source assumption. • Poor representation of circular geometry with coarser meshes is corrected. • Few benchmark problems are solved to show the advantages of linear expansion of source. • The advantage of the present formalism is quite visible in problems with large flux gradient. - Abstract: A common assumption in the solution of the neutron transport equation by the Method of Characteristics (MOC) is that the source (or flux) is constant within a mesh. This assumption is adequate provided the meshes are small enough so that the spatial variation of flux within a mesh may be ignored. Whether a mesh is small enough or not depends upon the flux gradient across a mesh, which in turn depends on factors like the presence of strong absorbers, localized sources or vacuum boundaries. The flat flux assumption often requires a very large number of meshes for solving the neutron transport equation with acceptable accuracy as was observed in our earlier work on the subject. A significant reduction in the required number of meshes is attainable by using a higher order representation of the flux within a mesh. In this paper, we expand the source within a mesh up to first order (linear) terms, which permits the use of larger sized (and therefore fewer) meshes and thereby reduces the computation time without compromising the accuracy of calculation. Since the division of the geometry into meshes is through an automatic triangulation procedure using the Bowyer-Watson algorithm, representation of circular objects (cylindrical fuel rods) with coarse meshes is poorer and causes geometry related errors. A numerical recipe is presented to make a correction to the automatic triangulation process and thereby eliminate this source of error. A number of benchmark problems are analyzed to emphasize the
Mesh Optimization for Ground Vehicle Aerodynamics
Adrian Gaylard; Essam F Abo-Serie; Nor Elyana Ahmad
2010-01-01
Mesh optimization strategy for estimating accurate drag of a ground vehicle is proposed based on examining the effect of different mesh parameters. The optimized mesh parameters were selected using design of experiment (DOE) method to be able to work in a...
Engagement of Metal Debris into Gear Mesh
handschuh, Robert F.; Krantz, Timothy L.
2010-01-01
A series of bench-top experiments was conducted to determine the effects of metallic debris being dragged through meshing gear teeth. A test rig that is typically used to conduct contact fatigue experiments was used for these tests. Several sizes of drill material, shim stock and pieces of gear teeth were introduced and then driven through the meshing region. The level of torque required to drive the "chip" through the gear mesh was measured. From the data gathered, chip size sufficient to jam the mechanism can be determined.
Bui, Huu Phuoc; Tomar, Satyendra; Courtecuisse, Hadrien; Audette, Michel; Cotin, Stéphane; Bordas, Stéphane P A
2018-05-01
An error-controlled mesh refinement procedure for needle insertion simulations is presented. As an example, the procedure is applied for simulations of electrode implantation for deep brain stimulation. We take into account the brain shift phenomena occurring when a craniotomy is performed. We observe that the error in the computation of the displacement and stress fields is localised around the needle tip and the needle shaft during needle insertion simulation. By suitably and adaptively refining the mesh in this region, our approach enables to control, and thus to reduce, the error whilst maintaining a coarser mesh in other parts of the domain. Through academic and practical examples we demonstrate that our adaptive approach, as compared with a uniform coarse mesh, increases the accuracy of the displacement and stress fields around the needle shaft and, while for a given accuracy, saves computational time with respect to a uniform finer mesh. This facilitates real-time simulations. The proposed methodology has direct implications in increasing the accuracy, and controlling the computational expense of the simulation of percutaneous procedures such as biopsy, brachytherapy, regional anaesthesia, or cryotherapy. Moreover, the proposed approach can be helpful in the development of robotic surgeries because the simulation taking place in the control loop of a robot needs to be accurate, and to occur in real time. Copyright © 2018 John Wiley & Sons, Ltd.
Determination of optimal LWR containment design, excluding accidents more severe than Class 8
International Nuclear Information System (INIS)
Cave, L.; Min, T.K.
1980-04-01
Information is presented concerning the restrictive effect of existing NRC requirements; definition of possible targets for containment; possible containment systems for LWR; optimization of containment design for class 3 through class 8 accidents (PWR); estimated costs of some possible containment arrangements for PWR relative to the standard dry containment system; estimated costs of BWR containment
FMDP Reactor Alternative Summary Report: Volume 3 - partially complete LWR alternative
International Nuclear Information System (INIS)
Greene, S.R.; Fisher, S.E.; Bevard, B.B.
1996-09-01
The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 3 of a four volume report summarizes the results of these analyses for the partially complete LWR (PCLWR) reactor based plutonium disposition alternative
Reduction of nuclear waste burden from LWR by deployment of the SCNES
International Nuclear Information System (INIS)
Arie, Kazuo; Watanabe, Junko; Mori, Kenji; Kubota, Kenichi; Kawashima, Masatoshi; Nakayama, Yoshiyuki; Nakazono, Ryuichi; Kuroda, Yuji; Fujiie, Yoichi
2009-01-01
Current efforts for enhancing capabilities for energy generation by LWR systems are efficient against the global warming crisis. In parallel to those movements, early realization of the SCNES concept can be the most viable solution to reduce nuclear waste burden produced by the current energy production system. (author)
Air quality impacts due to construction of LWR waste management facilities
International Nuclear Information System (INIS)
1977-06-01
Air quality impacts of construction activities and induced housing growth as a result of construction activities were evaluated for four possible facilities in the LWR fuel cycle: a fuel reprocessing facility, fuel storage facility, fuel fabrication plant, and a nuclear power plant. Since the fuel reprocessing facility would require the largest labor force, the impacts of construction of that facility were evaluated in detail
Computer program of iodine removal in the LWR containment vessel under LOCA conditions, MIRA-PB
International Nuclear Information System (INIS)
Nishio, Gunji; Tanaka, Mitsugu; Tamura, Tomohiko.
1978-03-01
LWR plants have a containment system for reactor safety consisting of spray and air cleaning filter. R.L.Ritzman of Battele Columbus Lab. developed computer code MIRAP/MIRAB for predicting iodine removal by containment system for PWR and BWR; which has some problem, however. The computer code MIRA-PB prepared by the authors is a modification of MIRAP/MIRAB. (auth.)
FMDP Reactor Alternative Summary Report: Volume 3 - partially complete LWR alternative
Energy Technology Data Exchange (ETDEWEB)
Greene, S.R.; Fisher, S.E.; Bevard, B.B. [and others
1996-09-01
The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 3 of a four volume report summarizes the results of these analyses for the partially complete LWR (PCLWR) reactor based plutonium disposition alternative.
Obtuse triangle suppression in anisotropic meshes
Sun, Feng; Choi, Yi King; Wang, Wen Ping; Yan, Dongming; Liu, Yang; Lé vy, Bruno L.
2011-01-01
Anisotropic triangle meshes are used for efficient approximation of surfaces and flow data in finite element analysis, and in these applications it is desirable to have as few obtuse triangles as possible to reduce the discretization error. We present a variational approach to suppressing obtuse triangles in anisotropic meshes. Specifically, we introduce a hexagonal Minkowski metric, which is sensitive to triangle orientation, to give a new formulation of the centroidal Voronoi tessellation (CVT) method. Furthermore, we prove several relevant properties of the CVT method with the newly introduced metric. Experiments show that our algorithm produces anisotropic meshes with much fewer obtuse triangles than using existing methods while maintaining mesh anisotropy. © 2011 Elsevier B.V. All rights reserved.
Connectivity editing for quad-dominant meshes
Peng, Chihan; Wonka, Peter
2013-01-01
and illustrate the advantages and disadvantages of different strategies for quad-dominant mesh design. © 2013 The Author(s) Computer Graphics Forum © 2013 The Eurographics Association and John Wiley & Sons Ltd.
Shape space exploration of constrained meshes
Yang, Yongliang
2011-12-12
We present a general computational framework to locally characterize any shape space of meshes implicitly prescribed by a collection of non-linear constraints. We computationally access such manifolds, typically of high dimension and co-dimension, through first and second order approximants, namely tangent spaces and quadratically parameterized osculant surfaces. Exploration and navigation of desirable subspaces of the shape space with regard to application specific quality measures are enabled using approximants that are intrinsic to the underlying manifold and directly computable in the parameter space of the osculant surface. We demonstrate our framework on shape spaces of planar quad (PQ) meshes, where each mesh face is constrained to be (nearly) planar, and circular meshes, where each face has a circumcircle. We evaluate our framework for navigation and design exploration on a variety of inputs, while keeping context specific properties such as fairness, proximity to a reference surface, etc. © 2011 ACM.
Shape space exploration of constrained meshes
Yang, Yongliang; Yang, Yijun; Pottmann, Helmut; Mitra, Niloy J.
2011-01-01
We present a general computational framework to locally characterize any shape space of meshes implicitly prescribed by a collection of non-linear constraints. We computationally access such manifolds, typically of high dimension and co-dimension, through first and second order approximants, namely tangent spaces and quadratically parameterized osculant surfaces. Exploration and navigation of desirable subspaces of the shape space with regard to application specific quality measures are enabled using approximants that are intrinsic to the underlying manifold and directly computable in the parameter space of the osculant surface. We demonstrate our framework on shape spaces of planar quad (PQ) meshes, where each mesh face is constrained to be (nearly) planar, and circular meshes, where each face has a circumcircle. We evaluate our framework for navigation and design exploration on a variety of inputs, while keeping context specific properties such as fairness, proximity to a reference surface, etc. © 2011 ACM.
Obtuse triangle suppression in anisotropic meshes
Sun, Feng
2011-12-01
Anisotropic triangle meshes are used for efficient approximation of surfaces and flow data in finite element analysis, and in these applications it is desirable to have as few obtuse triangles as possible to reduce the discretization error. We present a variational approach to suppressing obtuse triangles in anisotropic meshes. Specifically, we introduce a hexagonal Minkowski metric, which is sensitive to triangle orientation, to give a new formulation of the centroidal Voronoi tessellation (CVT) method. Furthermore, we prove several relevant properties of the CVT method with the newly introduced metric. Experiments show that our algorithm produces anisotropic meshes with much fewer obtuse triangles than using existing methods while maintaining mesh anisotropy. © 2011 Elsevier B.V. All rights reserved.
Mesh Processing in Medical Image Analysis
DEFF Research Database (Denmark)
The following topics are dealt with: mesh processing; medical image analysis; interactive freeform modeling; statistical shape analysis; clinical CT images; statistical surface recovery; automated segmentation; cerebral aneurysms; and real-time particle-based representation....
Capacity Analysis of Wireless Mesh Networks
Directory of Open Access Journals (Sweden)
M. I. Gumel
2012-06-01
Full Text Available The next generation wireless networks experienced a great development with emergence of wireless mesh networks (WMNs, which can be regarded as a realistic solution that provides wireless broadband access. The limited available bandwidth makes capacity analysis of the network very essential. While the network offers broadband wireless access to community and enterprise users, the problems that limit the network capacity must be addressed to exploit the optimum network performance. The wireless mesh network capacity analysis shows that the throughput of each mesh node degrades in order of l/n with increasing number of nodes (n in a linear topology. The degradation is found to be higher in a fully mesh network as a result of increase in interference and MAC layer contention in the network.
Energy-efficient wireless mesh networks
CSIR Research Space (South Africa)
Ntlatlapa, N
2007-06-01
Full Text Available This paper outlines the objectives of a recently formed research group at Meraka Institute. The authors consider application of wireless mesh networks in rural infrastructure deficient parts of the African continent where nodes operate on batteries...
LR: Compact connectivity representation for triangle meshes
Energy Technology Data Exchange (ETDEWEB)
Gurung, T; Luffel, M; Lindstrom, P; Rossignac, J
2011-01-28
We propose LR (Laced Ring) - a simple data structure for representing the connectivity of manifold triangle meshes. LR provides the option to store on average either 1.08 references per triangle or 26.2 bits per triangle. Its construction, from an input mesh that supports constant-time adjacency queries, has linear space and time complexity, and involves ordering most vertices along a nearly-Hamiltonian cycle. LR is best suited for applications that process meshes with fixed connectivity, as any changes to the connectivity require the data structure to be rebuilt. We provide an implementation of the set of standard random-access, constant-time operators for traversing a mesh, and show that LR often saves both space and traversal time over competing representations.
Seeking new surgical predictors of mesh exposure after transvaginal mesh repair.
Wu, Pei-Ying; Chang, Chih-Hung; Shen, Meng-Ru; Chou, Cheng-Yang; Yang, Yi-Ching; Huang, Yu-Fang
2016-10-01
The purpose of this study was to explore new preventable risk factors for mesh exposure. A retrospective review of 92 consecutive patients treated with transvaginal mesh (TVM) in the urogynecological unit of our university hospital. An analysis of perioperative predictors was conducted in patients after vaginal repairs using a type 1 mesh. Mesh complications were recorded according to International Urogynecological Association (IUGA) definitions. Mesh-exposure-free durations were calculated by using the Kaplan-Meier method and compared between different closure techniques using log-rank test. Hazard ratios (HR) of predictors for mesh exposure were estimated by univariate and multivariate analyses using Cox proportional hazards regression models. The median surveillance interval was 24.1 months. Two late occurrences were found beyond 1 year post operation. No statistically significant correlation was observed between mesh exposure and concomitant hysterectomy. Exposure risks were significantly higher in patients with interrupted whole-layer closure in univariate analysis. In the multivariate analysis, hematoma [HR 5.42, 95 % confidence interval (CI) 1.26-23.35, P = 0.024), Prolift mesh (HR 5.52, 95 % CI 1.15-26.53, P = 0.033), and interrupted whole-layer closure (HR 7.02, 95 % CI 1.62-30.53, P = 0.009) were the strongest predictors of mesh exposure. Findings indicate the risks of mesh exposure and reoperation may be prevented by avoiding hematoma, large amount of mesh, or interrupted whole-layer closure in TVM surgeries. If these risk factors are prevented, hysterectomy may not be a relative contraindication for TVM use. We also provide evidence regarding mesh exposure and the necessity for more than 1 year of follow-up and preoperative counselling.
MHD simulations on an unstructured mesh
International Nuclear Information System (INIS)
Strauss, H.R.; Park, W.; Belova, E.; Fu, G.Y.; Sugiyama, L.E.
1998-01-01
Two reasons for using an unstructured computational mesh are adaptivity, and alignment with arbitrarily shaped boundaries. Two codes which use finite element discretization on an unstructured mesh are described. FEM3D solves 2D and 3D RMHD using an adaptive grid. MH3D++, which incorporates methods of FEM3D into the MH3D generalized MHD code, can be used with shaped boundaries, which might be 3D
Towards Blockchain-enabled Wireless Mesh Networks
Selimi, Mennan; Kabbinale, Aniruddh Rao; Ali, Anwaar; Navarro, Leandro; Sathiaseelan, Arjuna
2018-01-01
Recently, mesh networking and blockchain are two of the hottest technologies in the telecommunications industry. Combining both can reformulate internet access and make connecting to the Internet not only easy, but affordable too. Hyperledger Fabric (HLF) is a blockchain framework implementation and one of the Hyperledger projects hosted by The Linux Foundation. We evaluate HLF in a real production mesh network and in the laboratory, quantify its performance, bottlenecks and limitations of th...
Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation
International Nuclear Information System (INIS)
Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki
2012-01-01
Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.
Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation
Energy Technology Data Exchange (ETDEWEB)
Permana, Sidik; Suzuki, Mitsutoshi; Su' ud, Zaki [Department of Science and Technology for Nuclear Material Management (STNM), Japan Atomic Energy Agency (JAEA), 2-4 Shirane, Shirakata, Tokai Mura, Naka-gun, Ibaraki 319-1195 Nuclear Physics and Bio (Indonesia); Department of Science and Technology for Nuclear Material Management (STNM), Japan Atomic Energy Agency (JAEA), 2-4 Shirane, Shirakata, Tokai Mura, Naka-gun, Ibaraki 319-1195 (Japan); Nuclear Physics and Bio Physics Research Group, Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132 (Indonesia)
2012-06-06
Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.
Identifying the primitive path mesh in entangled polymer liquids
International Nuclear Information System (INIS)
Sukumaran, Sathish K.; Kremer, Kurt; Grest, Gary Stephen; Everaers, Ralf
2004-01-01
Similar to entangled ropes, polymer chains cannot slide through each other. These topological constraints, the so-called entanglements, dominate the viscoelastic behavior of high-molecular-weight polymeric liquids. Tube models of polymer dynamics and rheology are based on the idea that entanglements confine a chain to small fluctuations around a primitive path which follows the coarse-grained chain contour. To establish the microscopic foundation for these highly successful phenomenological models, we have recently introduced a method for identifying the primitive path mesh that characterizes the microscopic topological state of computer-generated conformations of long-chain polymer melts and solutions. Here we give a more detailed account of the algorithm and discuss several key aspects of the analysis that are pertinent for its successful use in analyzing the topology of the polymer configurations. We also present a slight modification of the algorithm that preserves the previously neglected self-entanglements and allows us to distinguish between local self-knots and entanglements between distant sections of the same chain. Our results indicate that the latter make a negligible contribution to the tube and that the contour length between local self-knots, N 1k is significantly larger than the entanglement length N e
Unstructured Mesh Movement and Viscous Mesh Generation for CFD-Based Design Optimization, Phase II
National Aeronautics and Space Administration — The innovations proposed are twofold: 1) a robust unstructured mesh movement method able to handle isotropic (Euler), anisotropic (viscous), mixed element (hybrid)...
MHD simulations on an unstructured mesh
International Nuclear Information System (INIS)
Strauss, H.R.; Park, W.
1996-01-01
We describe work on a full MHD code using an unstructured mesh. MH3D++ is an extension of the PPPL MH3D resistive full MHD code. MH3D++ replaces the structured mesh and finite difference / fourier discretization of MH3D with an unstructured mesh and finite element / fourier discretization. Low level routines which perform differential operations, solution of PDEs such as Poisson's equation, and graphics, are encapsulated in C++ objects to isolate the finite element operations from the higher level code. The high level code is the same, whether it is run in structured or unstructured mesh versions. This allows the unstructured mesh version to be benchmarked against the structured mesh version. As a preliminary example, disruptions in DIIID reverse shear equilibria are studied numerically with the MH3D++ code. Numerical equilibria were first produced starting with an EQDSK file containing equilibrium data of a DIII-D L-mode negative central shear discharge. Using these equilibria, the linearized equations are time advanced to get the toroidal mode number n = 1 linear growth rate and eigenmode, which is resistively unstable. The equilibrium and linear mode are used to initialize 3D nonlinear runs. An example shows poloidal slices of 3D pressure surfaces: initially, on the left, and at an intermediate time, on the right
How to model wireless mesh networks topology
International Nuclear Information System (INIS)
Sanni, M L; Hashim, A A; Anwar, F; Ali, S; Ahmed, G S M
2013-01-01
The specification of network connectivity model or topology is the beginning of design and analysis in Computer Network researches. Wireless Mesh Networks is an autonomic network that is dynamically self-organised, self-configured while the mesh nodes establish automatic connectivity with the adjacent nodes in the relay network of wireless backbone routers. Researches in Wireless Mesh Networks range from node deployment to internetworking issues with sensor, Internet and cellular networks. These researches require modelling of relationships and interactions among nodes including technical characteristics of the links while satisfying the architectural requirements of the physical network. However, the existing topology generators model geographic topologies which constitute different architectures, thus may not be suitable in Wireless Mesh Networks scenarios. The existing methods of topology generation are explored, analysed and parameters for their characterisation are identified. Furthermore, an algorithm for the design of Wireless Mesh Networks topology based on square grid model is proposed in this paper. The performance of the topology generated is also evaluated. This research is particularly important in the generation of a close-to-real topology for ensuring relevance of design to the intended network and validity of results obtained in Wireless Mesh Networks researches
[Implants for genital prolapse : Contra mesh surgery].
Hampel, C
2017-12-01
Alloplastic transvaginal meshes have become very popular in the surgery of pelvic organ prolapse (POP) as did alloplastic suburethral slings in female stress incontinence surgery, but without adequate supporting data. The simplicity of the mesh procedure facilitates its propagation with acceptance of higher revision and complication rates. Since attending physicians do more and more prolapse surgeries without practicing or teaching alternative techniques, expertise in these alternatives, which might be very useful in cases of recurrence, persistence or complications, is permanently lost. It is doubtful that proper and detailed information about alternatives, risks, and benefits of transvaginal alloplastic meshes is provided to every single prolapse patient according to the recommendations of the German POP guidelines, since the number of implanted meshes exceeds the number of properly indicated mesh candidates by far. Although there is no dissent internationally about the available mesh data, thousands of lawsuits in the USA, insolvency of companies due to claims for compensation and unambiguous warnings from foreign urological societies leave German urogynecologists still unimpressed. The existing literature in pelvic organ prolapse exclusively focusses on POP stage and improvement of that stage with surgical therapy. Instead, typical prolapse symptoms should trigger therapy and improvement of these symptoms should be the utmost treatment goal. It is strongly recommended for liability reasons to obtain specific written informed consent.
Coarsely resolved topography along protein folding pathways
Fernández, Ariel; Kostov, Konstantin S.; Berry, R. Stephen
2000-03-01
The kinetic data from the coarse representation of polypeptide torsional dynamics described in the preceding paper [Fernandez and Berry, J. Chem. Phys. 112, 5212 (2000), preceding paper] is inverted by using detailed balance to obtain a topographic description of the potential-energy surface (PES) along the dominant folding pathway of the bovine pancreatic trypsin inhibitor (BPTI). The topography is represented as a sequence of minima and effective saddle points. The dominant folding pathway displays an overall monotonic decrease in energy with a large number of staircaselike steps, a clear signature of a good structure-seeker. The diversity and availability of alternative folding pathways is analyzed in terms of the Shannon entropy σ(t) associated with the time-dependent probability distribution over the kinetic ensemble of contact patterns. Several stages in the folding process are evident. Initially misfolded states form and dismantle revealing no definite pattern in the topography and exhibiting high Shannon entropy. Passage down a sequence of staircase steps then leads to the formation of a nativelike intermediate, for which σ(t) is much lower and fairly constant. Finally, the structure of the intermediate is refined to produce the native state of BPTI. We also examine how different levels of tolerance to mismatches of side chain contacts influence the folding kinetics, the topography of the dominant folding pathway, and the Shannon entropy. This analysis yields upper and lower bounds of the frustration tolerance required for the expeditious and robust folding of BPTI.
Coarse-Grained Modeling of Polyelectrolyte Solutions
Denton, Alan R.; May, Sylvio
2014-03-01
Ionic mixtures, such as electrolyte and polyelectrolyte solutions, have attracted much attention recently for their rich and challenging combination of electrostatic and non-electrostatic interparticle forces and their practical importance, from battery technologies to biological systems. Hydration of ions in aqueous solutions is known to entail ion-specific effects, including variable solubility of organic molecules, as manifested in the classic Hofmeister series for salting-in and salting-out of proteins. The physical mechanism by which the solvent (water) mediates effective interactions between ions, however, is still poorly understood. Starting from a microscopic model of a polyelectrolyte solution, we apply a perturbation theory to derive a coarse-grained model of ions interacting through both long-range electrostatic and short-range solvent-induced pair potentials. Taking these effective interactions as input to molecular dynamics simulations, we calculate structural and thermodynamic properties of aqueous ionic solutions. This work was supported by the National Science Foundation under Grant No. DMR-1106331.
De Sitter stability and coarse graining
International Nuclear Information System (INIS)
Markkanen, T.
2018-01-01
We present a 4-dimensional back reaction analysis of de Sitter space for a conformally coupled scalar field in the presence of vacuum energy initialized in the Bunch-Davies vacuum. In contrast to the usual semi-classical prescription, as the source term in the Friedmann equations we use expectation values where the unobservable information hidden by the cosmological event horizon has been neglected i.e. coarse grained over. It is shown that in this approach the energy-momentum is precisely thermal with constant temperature despite the dilution from the expansion of space due to a flux of energy radiated from the horizon. This leads to a self-consistent solution for the Hubble rate, which is gradually evolving and at late times deviates significantly from de Sitter. Our results hence imply de Sitter space to be unstable in this prescription. The solution also suggests dynamical vacuum energy: the continuous flux of energy is balanced by the generation of negative vacuum energy, which accumulatively decreases the overall contribution. Finally, we show that our results admit a thermodynamic interpretation which provides a simple alternate derivation of the mechanism. For very long times the solutions coincide with flat space. (orig.)
De Sitter stability and coarse graining
Energy Technology Data Exchange (ETDEWEB)
Markkanen, T. [Imperial College London, Department of Physics, London (United Kingdom); King' s College London, Department of Physics, London (United Kingdom)
2018-02-15
We present a 4-dimensional back reaction analysis of de Sitter space for a conformally coupled scalar field in the presence of vacuum energy initialized in the Bunch-Davies vacuum. In contrast to the usual semi-classical prescription, as the source term in the Friedmann equations we use expectation values where the unobservable information hidden by the cosmological event horizon has been neglected i.e. coarse grained over. It is shown that in this approach the energy-momentum is precisely thermal with constant temperature despite the dilution from the expansion of space due to a flux of energy radiated from the horizon. This leads to a self-consistent solution for the Hubble rate, which is gradually evolving and at late times deviates significantly from de Sitter. Our results hence imply de Sitter space to be unstable in this prescription. The solution also suggests dynamical vacuum energy: the continuous flux of energy is balanced by the generation of negative vacuum energy, which accumulatively decreases the overall contribution. Finally, we show that our results admit a thermodynamic interpretation which provides a simple alternate derivation of the mechanism. For very long times the solutions coincide with flat space. (orig.)
Thermodynamic forces in coarse-grained simulations
Noid, William
Atomically detailed molecular dynamics simulations have profoundly advanced our understanding of the structure and interactions in soft condensed phases. Nevertheless, despite dramatic advances in the methodology and resources for simulating atomically detailed models, low-resolution coarse-grained (CG) models play a central and rapidly growing role in science. CG models not only empower researchers to investigate phenomena beyond the scope of atomically detailed simulations, but also to precisely tailor models for specific phenomena. However, in contrast to atomically detailed simulations, which evolve on a potential energy surface, CG simulations should evolve on a free energy surface. Therefore, the forces in CG models should reflect the thermodynamic information that has been eliminated from the CG configuration space. As a consequence of these thermodynamic forces, CG models often demonstrate limited transferability and, moreover, rarely provide an accurate description of both structural and thermodynamic properties. In this talk, I will present a framework that clarifies the origin and impact of these thermodynamic forces. Additionally, I will present computational methods for quantifying these forces and incorporating their effects into CG MD simulations. As time allows, I will demonstrate applications of this framework for liquids, polymers, and interfaces. We gratefully acknowledge the support of the National Science Foundation via CHE 1565631.
OREST, LWR Burnup Simulation Using Program HAMMER and ORIGEN
International Nuclear Information System (INIS)
Hesse, Ulrich; Sieberer, Johann
2006-01-01
printer-output. 3 - Restrictions on the complexity of the problem: NEA version is limited for 100 loops, 1000 burnup time-steps and 10 post-irradiation steps. GRS recommends the use of LWR fuels based on oxygen and on the main HAMMER isotopes 235-U, 236-U, 238-U, 237-Np, 238-Pu, 239-Pu, 240-Pu, 241-Pu, 242-Pu, 241-Am and 243-Am. Gadolinium entries should be handled with care if singular positions of Gd-rods in real assemblies are found. Other mixture entries at start of calculation should only be impurities. Cladding should be Zr, Al or stainless steel. Special options for handling other materials can be found in the user description. Activation of structure materials is not calculated. Strong heterogeneous assembly problems outside of the input data processor should be pre-calculated by using more-dimensional codes to achieve a neutron spectra equivalent HAMMER lattice (FEC-method). Coolant pressure, coolant temperatures and coolant steam contents are assumed to be constant during burnup. During each program loop neutron spectra and cross sections are assumed to be constant
ASSIMILATION OF COARSE-SCALEDATAUSINGTHE ENSEMBLE KALMAN FILTER
Efendiev, Yalchin
2011-01-01
Reservoir data is usually scale dependent and exhibits multiscale features. In this paper we use the ensemble Kalman filter (EnKF) to integrate data at different spatial scales for estimating reservoir fine-scale characteristics. Relationships between the various scales is modeled via upscaling techniques. We propose two versions of the EnKF to assimilate the multiscale data, (i) where all the data are assimilated together and (ii) the data are assimilated sequentially in batches. Ensemble members obtained after assimilating one set of data are used as a prior to assimilate the next set of data. Both of these versions are easily implementable with any other upscaling which links the fine to the coarse scales. The numerical results with different methods are presented in a twin experiment setup using a two-dimensional, two-phase (oil and water) flow model. Results are shown with coarse-scale permeability and coarse-scale saturation data. They indicate that additional data provides better fine-scale estimates and fractional flow predictions. We observed that the two versions of the EnKF differed in their estimates when coarse-scale permeability is provided, whereas their results are similar when coarse-scale saturation is used. This behavior is thought to be due to the nonlinearity of the upscaling operator in the case of the former data. We also tested our procedures with various precisions of the coarse-scale data to account for the inexact relationship between the fine and coarse scale data. As expected, the results show that higher precision in the coarse-scale data yielded improved estimates. With better coarse-scale modeling and inversion techniques as more data at multiple coarse scales is made available, the proposed modification to the EnKF could be relevant in future studies.
Fire performance of basalt FRP mesh reinforced HPC thin plates
DEFF Research Database (Denmark)
Hulin, Thomas; Hodicky, Kamil; Schmidt, Jacob Wittrup
2013-01-01
An experimental program was carried out to investigate the influence of basalt FRP (BFRP) reinforcing mesh on the fire behaviour of thin high performance concrete (HPC) plates applied to sandwich elements. Samples with BFRP mesh were compared to samples with no mesh, samples with steel mesh...
Prolapse Recurrence after Transvaginal Mesh Removal.
Rawlings, Tanner; Lavelle, Rebecca S; Coskun, Burhan; Alhalabi, Feras; Zimmern, Philippe E
2015-11-01
We determined the rate of pelvic organ prolapse recurrence after transvaginal mesh removal. Following institutional review board approval a longitudinally collected database of women undergoing transvaginal mesh removal for complications after transvaginal mesh placement with at least 1 year minimum followup was queried for pelvic organ prolapse recurrence. Recurrent prolapse was defined as greater than stage 1 on examination or the need for reoperation at the site of transvaginal mesh removal. Outcome measures were based on POP-Q (Pelvic Organ Prolapse Quantification System) at the last visit. Patients were grouped into 3 groups, including group 1--recurrent prolapse in the same compartment as transvaginal mesh removal, 2--persistent prolapse and 3--prolapse in a compartment different than transvaginal mesh removal. Of 73 women 52 met study inclusion criteria from 2007 to 2013, including 73% who presented with multiple indications for transvaginal mesh removal. The mean interval between insertion and removal was 45 months (range 10 to 165). Overall mean followup after transvaginal mesh removal was 30 months (range 12 to 84). In group 1 (recurrent prolapse) the rate was 15% (6 of 40 patients). Four women underwent surgery for recurrent prolapse at a mean 7 of months (range 5 to 10). Two patients elected observation. The rate of persistent prolapse (group 2) was 23% (12 of 52 patients). Three women underwent prolapse reoperation at a mean of 10 months (range 8 to 12). In group 3 (de novo/different compartment prolapse) the rate was 6% (3 of 52 patients). One woman underwent surgical repair at 52 months. At a mean 2.5-year followup 62% of patients (32 of 52) did not have recurrent or persistent prolapse after transvaginal mesh removal and 85% (44 of 52) did not undergo any further procedure for prolapse. Specifically for pelvic organ prolapse in the same compartment as transvaginal mesh removal 12% of patients had recurrence, of whom 8% underwent prolapse repair
International Nuclear Information System (INIS)
1987-12-01
This User's Guide for the LWR Assemblies data base system is part of the Characteristics Data Base being developed under the Waste Systems Data Development Program. The objective of the LWR Assemblies data base is to provide access at the personal computer level to information about fuel assemblies used in light-water reactors. The information available is physical descriptions of intact fuel assemblies and radiological descriptions of spent fuel disassembly hardware. The LWR Assemblies data base is a user-oriented menu driven system. Each menu is instructive about its use. Section 5 of this guide provides a sample session with the data base to assist the user
Discretization of the Joule heating term for plasma discharge fluid models in unstructured meshes
International Nuclear Information System (INIS)
Deconinck, T.; Mahadevan, S.; Raja, L.L.
2009-01-01
The fluid (continuum) approach is commonly used for simulation of plasma phenomena in electrical discharges at moderate to high pressures (>10's mTorr). The description comprises governing equations for charged and neutral species transport and energy equations for electrons and the heavy species, coupled to equations for the electromagnetic fields. The coupling of energy from the electrostatic field to the plasma species is modeled by the Joule heating term which appears in the electron and heavy species (ion) energy equations. Proper numerical discretization of this term is necessary for accurate description of discharge energetics; however, discretization of this term poses a special problem in the case of unstructured meshes owing to the arbitrary orientation of the faces enclosing each cell. We propose a method for the numerical discretization of the Joule heating term using a cell-centered finite volume approach on unstructured meshes with closed convex cells. The Joule heating term is computed by evaluating both the electric field and the species flux at the cell center. The dot product of these two vector quantities is computed to obtain the Joule heating source term. We compare two methods to evaluate the species flux at the cell center. One is based on reconstructing the fluxes at the cell centers from the fluxes at the face centers. The other recomputes the flux at the cell center using the common drift-diffusion approximation. The reconstructed flux scheme is the most stable method and yields reasonably accurate results on coarse meshes.
Feature-Sensitive Tetrahedral Mesh Generation with Guaranteed Quality
Wang, Jun; Yu, Zeyun
2012-01-01
Tetrahedral meshes are being extensively used in finite element methods (FEM). This paper proposes an algorithm to generate feature-sensitive and high-quality tetrahedral meshes from an arbitrary surface mesh model. A top-down octree subdivision is conducted on the surface mesh and a set of tetrahedra are constructed using adaptive body-centered cubic (BCC) lattices. Special treatments are given to the tetrahedra near the surface such that the quality of the resulting tetrahedral mesh is prov...
Non-Galerkin Coarse Grids for Algebraic Multigrid
Energy Technology Data Exchange (ETDEWEB)
Falgout, Robert D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Schroder, Jacob B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)
2014-06-26
Algebraic multigrid (AMG) is a popular and effective solver for systems of linear equations that arise from discretized partial differential equations. And while AMG has been effectively implemented on large scale parallel machines, challenges remain, especially when moving to exascale. Particularly, stencil sizes (the number of nonzeros in a row) tend to increase further down in the coarse grid hierarchy, and this growth leads to more communication. Therefore, as problem size increases and the number of levels in the hierarchy grows, the overall efficiency of the parallel AMG method decreases, sometimes dramatically. This growth in stencil size is due to the standard Galerkin coarse grid operator, $P^T A P$, where $P$ is the prolongation (i.e., interpolation) operator. For example, the coarse grid stencil size for a simple three-dimensional (3D) seven-point finite differencing approximation to diffusion can increase into the thousands on present day machines, causing an associated increase in communication costs. We therefore consider algebraically truncating coarse grid stencils to obtain a non-Galerkin coarse grid. First, the sparsity pattern of the non-Galerkin coarse grid is determined by employing a heuristic minimal “safe” pattern together with strength-of-connection ideas. Second, the nonzero entries are determined by collapsing the stencils in the Galerkin operator using traditional AMG techniques. The result is a reduction in coarse grid stencil size, overall operator complexity, and parallel AMG solve phase times.
Roth, Ted M; Reight, Ian
2012-07-01
Sacral colpopexy may be complicated by mesh exposure, and the surgical treatment of mesh exposure typically results in minor postoperative morbidity and few delayed complications. A 75-year-old woman presented 7 years after a laparoscopic sacral colpopexy, with Mersilene mesh, with an apical mesh exposure. She underwent an uncomplicated transvaginal excision and was asymptomatic until 8 months later when she presented with vaginal drainage and a sacral abscess. This was successfully treated with laparoscopic enterolysis, drainage of the abscess, and explantation of the remaining mesh. Incomplete excision of exposed colpopexy mesh can lead to ascending infection and sacral abscess. Laparoscopic drainage and mesh removal may be considered in these patients.
Ridgeway, Beri; Walters, Mark D; Paraiso, Marie Fidela R; Barber, Matthew D; McAchran, Sarah E; Goldman, Howard B; Jelovsek, J Eric
2008-12-01
The purpose of this study was to determine the complications, treatments, and outcomes in patients choosing to undergo removal of mesh previously placed with a mesh procedural kit. This was a retrospective review of all patients who underwent surgical removal of transvaginal mesh for mesh-related complications during a 3-year period at Cleveland Clinic. At last follow-up, patients reported degree of pain, level of improvement, sexual activity, and continued symptoms. Nineteen patients underwent removal of mesh during the study period. Indications for removal included chronic pain (6/19), dyspareunia (6/19), recurrent pelvic organ prolapse (8/19), mesh erosion (12/19), and vesicovaginal fistula (3/19), with most patients (16/19) citing more than 1 reason. There were few complications related to the mesh removal. Most patients reported significant relief of symptoms. Mesh removal can be technically difficult but appears to be safe with few complications and high relief of symptoms, although some symptoms can persist.
Cartesian anisotropic mesh adaptation for compressible flow
International Nuclear Information System (INIS)
Keats, W.A.; Lien, F.-S.
2004-01-01
Simulating transient compressible flows involving shock waves presents challenges to the CFD practitioner in terms of the mesh quality required to resolve discontinuities and prevent smearing. This paper discusses a novel two-dimensional Cartesian anisotropic mesh adaptation technique implemented for compressible flow. This technique, developed for laminar flow by Ham, Lien and Strong, is efficient because it refines and coarsens cells using criteria that consider the solution in each of the cardinal directions separately. In this paper the method will be applied to compressible flow. The procedure shows promise in its ability to deliver good quality solutions while achieving computational savings. The convection scheme used is the Advective Upstream Splitting Method (Plus), and the refinement/ coarsening criteria are based on work done by Ham et al. Transient shock wave diffraction over a backward step and shock reflection over a forward step are considered as test cases because they demonstrate that the quality of the solution can be maintained as the mesh is refined and coarsened in time. The data structure is explained in relation to the computational mesh, and the object-oriented design and implementation of the code is presented. Refinement and coarsening algorithms are outlined. Computational savings over uniform and isotropic mesh approaches are shown to be significant. (author)
Mesh networks: an optimum solution for AMR
Energy Technology Data Exchange (ETDEWEB)
Mimno, G.
2003-12-01
Characteristics of mesh networks and the advantage of using them in automatic meter reading equipment (AMR) are discussed. Mesh networks are defined as being similar to a fishing net made of knots and links. In mesh networks the knots represent meter sites and the links are the radio paths between the meter sites and the neighbourhood concentrator. In mesh networks any knot in the communications chain can link to any other and the optimum path is calculated by the network by hopping from meter to meter until the radio message reaches a concentrator. This mesh communications architecture is said to be vastly superior to many older types of radio-based meter reading technologies; its main advantage is that it not only significantly improves the economics of fixed network deployment, but also supports time-of-use metering, remote disconnect services and advanced features, such as real-time pricing, demand response, and other efficiency measures, providing a better return on investment and reliability.
Mellano, Erin M; Nakamura, Leah Y; Choi, Judy M; Kang, Diana C; Grisales, Tamara; Raz, Shlomo; Rodriguez, Larissa V
2016-01-01
Vaginal mesh complications necessitating excision are increasingly prevalent. We aim to study whether subclinical chronically infected mesh contributes to the development of delayed-onset mesh complications or recurrent urinary tract infections (UTIs). Women undergoing mesh removal from August 2013 through May 2014 were identified by surgical code for vaginal mesh removal. Only women undergoing removal of anti-incontinence mesh were included. Exclusion criteria included any women undergoing simultaneous prolapse mesh removal. We abstracted preoperative and postoperative information from the medical record and compared mesh culture results from patients with and without mesh extrusion, de novo recurrent UTIs, and delayed-onset pain. One hundred seven women with only anti-incontinence mesh removed were included in the analysis. Onset of complications after mesh placement was within the first 6 months in 70 (65%) of 107 and delayed (≥6 months) in 37 (35%) of 107. A positive culture from the explanted mesh was obtained from 82 (77%) of 107 patients, and 40 (37%) of 107 were positive with potential pathogens. There were no significant differences in culture results when comparing patients with delayed-onset versus immediate pain, extrusion with no extrusion, and de novo recurrent UTIs with no infections. In this large cohort of patients with mesh removed for a diverse array of complications, cultures of the explanted vaginal mesh demonstrate frequent low-density bacterial colonization. We found no differences in culture results from women with delayed-onset pain versus acute pain, vaginal mesh extrusions versus no extrusions, or recurrent UTIs using standard culture methods. Chronic prosthetic infections in other areas of medicine are associated with bacterial biofilms, which are resistant to typical culture techniques. Further studies using culture-independent methods are needed to investigate the potential role of chronic bacterial infections in delayed vaginal mesh
International Nuclear Information System (INIS)
Fujita, Reiko; Mizuguchi, Koji; Fuse, Kouki; Saso, Michitaka; Utsunomiya, Kazuhiro; Arie, Kazuo
2008-01-01
Toshiba has been proposing a new fuel cycle concept of a transition from LWR to FBR. The new fuel cycle concept has better economical process of the LWR spent fuel reprocessing than the present Purex Process and the proliferation resistance for FBR cycle of plutonium with minor actinides after 2040. Toshiba has been developing a new Advanced Hybrid Process with Solvent Extraction and Pyrochemical process of spent fuel reprocessing for LWR to FBR. The Advanced Hybrid Process combines the solvent extraction process of the LWR spent fuel in nitric acid with the recovery of high pure uranium for LWR fuel and the pyro-chemical process in molten salts of impure plutonium recovery with minor actinides for metallic FBR fuel, which is the FBR spent fuel recycle system after FBR age based on the electrorefining process in molten salts since 1988. The new Advanced Hybrid Process enables the decrease of the high-level waste and the secondary waste from the spent fuel reprocessing plants. The R and D costs in the new Advanced Hybrid Process might be reduced because of the mutual Pyro-chemical process in molten salts. This paper describes the new fuel cycle concept of a transition from LWR to FBR and the feasibility of the new Advanced Hybrid Process by fundamental experiments. (author)
Property A and Coarse Embedding for Locally Compact Groups
DEFF Research Database (Denmark)
Li, Kang
property A. In a joint work with Knudby, we characterize the connected simple Lie groups with the discrete topology that have different approximation properties (see Article B). Moreover, we give a contractive Schur multiplier characterization of locally compact groups coarsely embeddable into Hilbert......In the study of the Novikov conjecture, property A and coarse embedding of metric spaces were introduced by Yu and Gromov, respectively. The main topic of the thesis is property A and coarse embedding for locally compact second countable groups. We prove that many of the results that are known...... to hold in the discrete setting, hold also in the locally compact setting.In a joint work with Deprez, we show that property A is equivalent to amenability at infinity and the strong Novikov conjecture is true for every locally compact group that embeds coarsely into a Hilbert space (see Article A...
Comparison of coarse-grained (MARTINI) and atomistic molecular ...
Indian Academy of Sciences (India)
Rajat Desikan
as the root mean square deviation (RMSD) histograms and the inner pore radius profiles from ... ever coarse-grained simulations of membrane-proteins ..... from the MARTINI simulations show greater fluctuations than the all-atom simulations.
Information Theoretic Tools for Parameter Fitting in Coarse Grained Models
Kalligiannaki, Evangelia; Harmandaris, Vagelis; Katsoulakis, Markos A.; Plechac, Petr
2015-01-01
We study the application of information theoretic tools for model reduction in the case of systems driven by stochastic dynamics out of equilibrium. The model/dimension reduction is considered by proposing parametrized coarse grained dynamics
Recycled tires as coarse aggregate in concrete pavement mixtures.
2013-07-01
The reuse potential of tire chips as coarse aggregates in pavement concrete was examined in this research by : investigating the effects of low- and high-volume tire chips on fresh and hardened concrete properties. One concrete : control mixture was ...
Two-level method with coarse space size independent convergence
Energy Technology Data Exchange (ETDEWEB)
Vanek, P.; Brezina, M. [Univ. of Colorado, Denver, CO (United States); Tezaur, R.; Krizkova, J. [UWB, Plzen (Czech Republic)
1996-12-31
The basic disadvantage of the standard two-level method is the strong dependence of its convergence rate on the size of the coarse-level problem. In order to obtain the optimal convergence result, one is limited to using a coarse space which is only a few times smaller than the size of the fine-level one. Consequently, the asymptotic cost of the resulting method is the same as in the case of using a coarse-level solver for the original problem. Today`s two-level domain decomposition methods typically offer an improvement by yielding a rate of convergence which depends on the ratio of fine and coarse level only polylogarithmically. However, these methods require the use of local subdomain solvers for which straightforward application of iterative methods is problematic, while the usual application of direct solvers is expensive. We suggest a method diminishing significantly these difficulties.
Connectivity editing for quad-dominant meshes
Peng, Chihan
2013-08-01
We propose a connectivity editing framework for quad-dominant meshes. In our framework, the user can edit the mesh connectivity to control the location, type, and number of irregular vertices (with more or fewer than four neighbors) and irregular faces (non-quads). We provide a theoretical analysis of the problem, discuss what edits are possible and impossible, and describe how to implement an editing framework that realizes all possible editing operations. In the results, we show example edits and illustrate the advantages and disadvantages of different strategies for quad-dominant mesh design. © 2013 The Author(s) Computer Graphics Forum © 2013 The Eurographics Association and John Wiley & Sons Ltd.
ZONE: a finite element mesh generator
International Nuclear Information System (INIS)
Burger, M.J.
1976-05-01
The ZONE computer program is a finite-element mesh generator which produces the nodes and element description of any two-dimensional geometry. The geometry is subdivided into a mesh of quadrilateral and triangular zones arranged sequentially in an ordered march through the geometry. The order of march can be chosen so that the minimum bandwidth is obtained. The node points are defined in terms of the x and y coordinates in a global rectangular coordinate system. The zones generated are quadrilaterals or triangles defined by four node points in a counterclockwise sequence. Node points defining the outside boundary are generated to describe pressure boundary conditions. The mesh that is generated can be used as input to any two-dimensional as well as any axisymmetrical structure program. The output from ZONE is essentially the input file to NAOS, HONDO, and other axisymmetric finite element programs. 14 figures
Open preperitoneal groin hernia repair with mesh
DEFF Research Database (Denmark)
Andresen, Kristoffer; Rosenberg, Jacob
2017-01-01
Background For the repair of inguinal hernias, several surgical methods have been presented where the purpose is to place a mesh in the preperitoneal plane through an open access. The aim of this systematic review was to describe preperitoneal repairs with emphasis on the technique. Data sources...... A systematic review was conducted and reported according to the PRISMA statement. PubMed, Cochrane library and Embase were searched systematically. Studies were included if they provided clinical data with more than 30 days follow up following repair of an inguinal hernia with an open preperitoneal mesh......-analysis. Open preperitoneal techniques with placement of a mesh through an open approach seem promising compared with the standard anterior techniques. This systematic review provides an overview of these techniques together with a description of surgical methods and clinical outcomes....
Open preperitoneal groin hernia repair with mesh
DEFF Research Database (Denmark)
Andresen, Kristoffer; Rosenberg, Jacob
2017-01-01
BACKGROUND: For the repair of inguinal hernias, several surgical methods have been presented where the purpose is to place a mesh in the preperitoneal plane through an open access. The aim of this systematic review was to describe preperitoneal repairs with emphasis on the technique. DATA SOURCES......: A systematic review was conducted and reported according to the PRISMA statement. PubMed, Cochrane library and Embase were searched systematically. Studies were included if they provided clinical data with more than 30 days follow up following repair of an inguinal hernia with an open preperitoneal mesh......-analysis. Open preperitoneal techniques with placement of a mesh through an open approach seem promising compared with the standard anterior techniques. This systematic review provides an overview of these techniques together with a description of surgical methods and clinical outcomes....
Proceedings of the 2007 LWR Fuel Performance Meeting / TopFuel 2007 'Zero by 2010'
International Nuclear Information System (INIS)
2007-01-01
ANS, ENS, AESJ and KNS are jointly organizing the 2007 International LWR Fuel Performance Meeting following the successful ENS TopFuel meeting held during 22-26 October, 2006 in Salamaca, Spain. Merging three premier nuclear fuel design and performance meetings: the ANS LWR Fuel Performance Meeting, the ENS TopFuel and Asian Water Reactor Fuel Performance Meeting (WRFPM) created this international meeting. The meeting will be held annually on a tri-annual rotational basis in USA, Asia, and Europe. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as performance experience in commercial and test reactors. The meeting excludes front end and back end fuel issues, however, it covers all front and/or back issues that impact fuel designs and performance
Workshop on initiation of stress corrosion cracking under LWR conditions: Proceedings
International Nuclear Information System (INIS)
Nelson, J.L.; Cubicciotti, D.; Licina, G.J.
1988-05-01
A workshop titled ''Initiation of Stress Corrosion Cracking under LWR Conditions'' was held in Palo Alto, California on November 13, 1986, hosted by the Electric Power Research Institute. Participants were experts on the topic from nuclear steam supply and component manufacturers, public and private research laboratories, and university environments. Presentations included discussions on the definition of crack initiation, the effects of environmental and electrochemical variables on cracking susceptibility, and detection methods for the determination of crack initiation events and measurement of critical environmental and stress parameters. Examination of the questions related to crack initiation and its relative importance to the overall question of cracking of LWR materials from these perspectives provided inputs to EPRI project managers on the future direction of research efforts designed to prevent and control cracking. Thirteen reports have been cataloged separately
International Nuclear Information System (INIS)
Whitesides, G.H.; Stephens, M.E.
1984-01-01
During the past two years, a Working Group established by the Organization for Economic Co-Operation and Development's Nuclear Energy Agency (OECD-NEA) has been developing a set of criticality benchmark problems which could be used to help establish the validity of criticality safety computer programs and their associated nuclear data for calculation of ksub(eff) for spent light water reactor (LWR) fuel transport containers. The basic goal of this effort was to identify a set of actual critical experiments which would contain the various material and geometric properties present in spent LWR transport contrainers. These data, when used by the various computational methods, are intended to demonstrate the ability of each method to accurately reproduce the experimentally measured ksub(eff) for the parameters under consideration
Development and testing of standardized procedures and reference data for LWR surveillance
International Nuclear Information System (INIS)
McElroy, W.N.
1979-02-01
The resources and talents of many national and international organizations and laboratories, both governmental and industrial, are being used to establish analysis methods for predicting the embrittlement condition of light water reactor (LWR) primary systems. The exact interrelationships and responsibilities between those developing, understanding, combining, and applying state-of-the-art technology in dosimetry, metallurgy, and fracture mechanics for reactor systems analysis are being carefully reviewed and studied. This has resulted in a more comprehensive definition of the scope of new and updated ASTM standards required for the analysis and interpretation of LWR pressure vessel surveillance results. Fifteen new and updated ASTM standards have now been identified, together with a restructuring of the main interfaces between the individual standard practices, guides, and methods. The paper briefly discusses these standards and the initial results of multi-laboratory research work involved in their validation and calibration
Evaluation of methods for decladding LWR fuel for a pyroprocessing-based reprocessing plant
International Nuclear Information System (INIS)
Bond, W.D.; Mailen, J.C.; Michaels, G.E.
1992-10-01
The first step in reprocessing disassembled light-water reactor (LWR) spent fuel is to separate the zirconium-based cladding from the UO 2 fuel. A survey of decladding technologies has been performed to identify candidate decladding processes suitable for LWR fuel and compatible with downstream pyropr for separation of actinides and fission products. Technologies for the primary separation of Zircaloy cladding from oxide fuel and for secondary separations (in most cases, a further decontamination of the cladding) were reviewed. Because cutting of the fuel cladding is a necessary step in all flowsheet options, metal cutting technologies were also briefly evaluated. The assessment of decladding processes resulted in the identification of the three or four potentially attractive options that may warrant additional near-term evaluation. These options are summarized, and major strengths and issues of each option are discussed
Fission product release from high gap-inventory LWR fuel under LOCA conditions
International Nuclear Information System (INIS)
Lorenz, R.A.; Collins, J.L.; Osborne, M.F.; Malinauskas, A.P.
1980-01-01
Fission product release tests were performed with light water reactor (LWR) fuel rod segments containing large amounts of cesium and iodine in the pellet-to-cladding gap space in order to check the validity of the previously published Source Term Model for this type of fuel. The model describes the release of fission product cesium and iodine from LWR fuel rods for controlled loss-of-coolant accident (LOCA) transients in the temperature range 500 to 1200 0 C. The basis for the model was test data obtained with simulated fuel rods and commercial fuel irradiated to high burnup but containing relatively small amounts of cesium and iodine in the pellet-to-cladding gap space
FABRICATION AND MATERIAL ISSUES FOR THE APPLICATION OF SiC COMPOSITES TO LWR FUEL CLADDING
Directory of Open Access Journals (Sweden)
WEON-JU KIM
2013-08-01
Full Text Available The fabrication methods and requirements of the fiber, interphase, and matrix of nuclear grade SiCf/SiC composites are briefly reviewed. A CVI-processed SiCf/SiC composite with a PyC or (PyC-SiCn interphase utilizing Hi-Nicalon Type S or Tyranno SA3 fiber is currently the best combination in terms of the irradiation performance. We also describe important material issues for the application of SiC composites to LWR fuel cladding. The kinetics of the SiC corrosion under LWR conditions needs to be clarified to confirm the possibility of a burn-up extension and the cost-benefit effect of the SiC composite cladding. In addition, the development of end-plug joining technology and fission products retention capability of the ceramic composite tube would be key challenges for the successful application of SiC composite cladding.
EUV lithography for 30nm half pitch and beyond: exploring resolution, sensitivity, and LWR tradeoffs
Putna, E. Steve; Younkin, Todd R.; Chandhok, Manish; Frasure, Kent
2009-03-01
The International Technology Roadmap for Semiconductors (ITRS) denotes Extreme Ultraviolet (EUV) lithography as a leading technology option for realizing the 32nm half-pitch node and beyond. Readiness of EUV materials is currently one high risk area according to assessments made at the 2008 EUVL Symposium. The main development issue regarding EUV resist has been how to simultaneously achieve high sensitivity, high resolution, and low line width roughness (LWR). This paper describes the strategy and current status of EUV resist development at Intel Corporation. Data is presented utilizing Intel's Micro-Exposure Tool (MET) examining the feasibility of establishing a resist process that simultaneously exhibits <=30nm half-pitch (HP) L/S resolution at <=10mJ/cm2 with <=4nm LWR.
International Nuclear Information System (INIS)
1996-12-01
At the invitation of the Government of the Russian Federation, following a proposal of the International Working Group on Water Reactor Fuel Performance and Technology, the IAEA convened a Technical Committee Meeting on Behaviour of LWR Core Materials Under Accident Conditions from 9 to 13 October 1995 in Dimitrovgrad to analyze and evaluate the behaviour of LWR core materials under accident conditions with special emphasis on severe accidents. In-vessel severe accidents phenomena were considered in detail, but specialized thermal hydraulic aspects as well as ex-vessel phenomena were outside the scope of the meeting. Forty participants representing eight countries attended the meeting. Twenty-three papers were presented and discussed during five sessions. Refs, figs, tabs
Performance of artificially defected LWR fuel rods in an unlimited air dry storage atmosphere
International Nuclear Information System (INIS)
Einziger, R.E.; Knecht, R.L.; Cantley, D.A.; Cook, J.A.
1983-09-01
Thus far the tests are inconclusive as to whether breached LWR fuel can be stored at 230 0 C for long periods of time in air without fuel oxidation and dispersion. There is every indication, as expected, that there is no oxidation problem in an inert atmosphere. Only one of four defects exposed to unlimited air gave any indication of fuel oxidation. It has been suggested that this might be an incubation effect and continued operation would result in oxidation occurring at all four defects. As yet the destructive examination of the BWR rod has not been completed, so it is not possible to determine if cladding splitting was due to an anomoly in this test rod or something that can be expected in LWR rods in general. Thus far there is no indication of respirable particle dispersal even if fuel oxidation does occur
Nonlinear analysis of LWR components: areas of investigation/benefits/recommendations
Energy Technology Data Exchange (ETDEWEB)
Brown, S. J. [ed.
1980-04-01
The purpose of this study is to identify specific topics of investigation into design procedures, design concepts, methods of analysis, testing practices, and standards which are characterized by nonlinear behavior (both geometric and material) and which are considered to offer some economic and/or technical benefits to the LWR industry (excluding piping). In this study these topics were collected, compiled, and subjectively evaluated as to their potential benefit. The topics considered to have the greatest benefit/impact potential are discussed. The topics listed are based upon the experience of ODAI and also based upon a sampling of over 100 engineers/scientists in the LWR industry. The topics of investigation were found to fall basically into three areas: component, code interpretation, and load/failure mechanism. The topics are arbitrarily reorganized into six areas of investigation: Fracture, Fatigue, Vibration/Dynamic/Seismic, Plasticity, Component/Computational Considerations, and Code Interpretation.
International Nuclear Information System (INIS)
Kim, J. S.; Youm, H. K.; Jin, T. E.
1999-01-01
The fatigue lifetime of principal components in nuclear power plant is evaluated by using the design fatigue curves in ASME B and PV code during design process. However, it is inadequate to evaluate fatigue lifetime considering the LWR environmental effect by these design fatigue curves because these are presented only under atmosphere environment. Therefore, many studies are recently performed for the design fatigue curves considering LWR environmental effect and are presented that the design fatigue curves in ASME B and PV code can be non-conservative. In present paper, the limits and differences of the design fatigue curves considering environmental effect are presented. To investigate the change of fatigue lifetime according to each design fatigue curve, the CUFs for the pressurizer spray nozzle partly composed of austenitic stainless steel are calculated according to each one. Finally, if the evaluation result can not be satisfied with fatigue design requirement, the alternatives to reduce design cumulative usage factor are discussed. (author)
Light water reactors development in Japan. (1) Introduction of LWR technology (PWR)
International Nuclear Information System (INIS)
Yamada, Ichita; Suzuki, Shigemitsu
2008-01-01
Evolutionary progress of the LWR plants in the last half-century was reviewed in series. Introduction of LWR technology (PWR) in Japan was reviewed in this article. Kansai Electric Power imported the Mihama-1 - a 340 MWe PWR built by Westinghouse Corp. It began operating in 1970 to supply power to the World Exposition (EXPO70). There followed a period in which designs was purchased from US vendors and they were constructed with the co-operation of Mitsubishi Heavy Industry, who would then receive a license to build similar plants in Japan and develop the capacity to design and construct PWRs by itself. Progress of designs, fabrications, project management and construction of PWRs were reviewed from technology transfer to its autonomy age. (T. Tanaka)
Nonlinear analysis of LWR components: areas of investigation/benefits/recommendations
International Nuclear Information System (INIS)
Brown, S.J.
1980-04-01
The purpose of this study is to identify specific topics of investigation into design procedures, design concepts, methods of analysis, testing practices, and standards which are characterized by nonlinear behavior (both geometric and material) and which are considered to offer some economic and/or technical benefits to the LWR industry (excluding piping). In this study these topics were collected, compiled, and subjectively evaluated as to their potential benefit. The topics considered to have the greatest benefit/impact potential are discussed. The topics listed are based upon the experience of ODAI and also based upon a sampling of over 100 engineers/scientists in the LWR industry. The topics of investigation were found to fall basically into three areas: component, code interpretation, and load/failure mechanism. The topics are arbitrarily reorganized into six areas of investigation: Fracture, Fatigue, Vibration/Dynamic/Seismic, Plasticity, Component/Computational Considerations, and Code Interpretation
Evaluation of methods for decladding LWR fuel for a pyroprocessing-based reprocessing plant
Energy Technology Data Exchange (ETDEWEB)
Bond, W.D.; Mailen, J.C.; Michaels, G.E.
1992-10-01
The first step in reprocessing disassembled light-water reactor (LWR) spent fuel is to separate the zirconium-based cladding from the UO[sub 2] fuel. A survey of decladding technologies has been performed to identify candidate decladding processes suitable for LWR fuel and compatible with downstream pyropr for separation of actinides and fission products. Technologies for the primary separation of Zircaloy cladding from oxide fuel and for secondary separations (in most cases, a further decontamination of the cladding) were reviewed. Because cutting of the fuel cladding is a necessary step in all flowsheet options, metal cutting technologies were also briefly evaluated. The assessment of decladding processes resulted in the identification of the three or four potentially attractive options that may warrant additional near-term evaluation. These options are summarized, and major strengths and issues of each option are discussed.
Evaluation of methods for decladding LWR fuel for a pyroprocessing-based reprocessing plant
Energy Technology Data Exchange (ETDEWEB)
Bond, W.D.; Mailen, J.C.; Michaels, G.E.
1992-10-01
The first step in reprocessing disassembled light-water reactor (LWR) spent fuel is to separate the zirconium-based cladding from the UO{sub 2} fuel. A survey of decladding technologies has been performed to identify candidate decladding processes suitable for LWR fuel and compatible with downstream pyropr for separation of actinides and fission products. Technologies for the primary separation of Zircaloy cladding from oxide fuel and for secondary separations (in most cases, a further decontamination of the cladding) were reviewed. Because cutting of the fuel cladding is a necessary step in all flowsheet options, metal cutting technologies were also briefly evaluated. The assessment of decladding processes resulted in the identification of the three or four potentially attractive options that may warrant additional near-term evaluation. These options are summarized, and major strengths and issues of each option are discussed.
International Nuclear Information System (INIS)
Pavlidis, D.; Lathouwers, D.
2011-01-01
A computational fluid dynamics model with anisotropic mesh adaptivity is used to investigate coolant flow and heat transfer in pebble bed 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. (author)
Estimating the shear strength of concrete with coarse aggregate replacement
Folagbade Olusoga Peter ORIOLA; George MOSES; Jacob Oyeniyi AFOLAYAN; John Engbonye SANI
2017-01-01
For economic, environmental and practical reasons, it is desirable to replace the constituents of concrete with wastes and cheaper alternative materials. However, it is best when such replacements are done at optimum replacement levels. In view of this, a laboratory investigative test was carried out to evaluate the shear strength of concrete with coarse aggregate replacement by Coconut Shell and by Waste Rubber Tyre. The coarse aggregate replacement was done at recommended optimum proportion...
An Evaluation of Coarse-Grained Locking for Multicore Microkernels
Elphinstone, Kevin; Zarrabi, Amirreza; Danis, Adrian; Shen, Yanyan; Heiser, Gernot
2016-01-01
The trade-off between coarse- and fine-grained locking is a well understood issue in operating systems. Coarse-grained locking provides lower overhead under low contention, fine-grained locking provides higher scalability under contention, though at the expense of implementation complexity and re- duced best-case performance. We revisit this trade-off in the context of microkernels and tightly-coupled cores with shared caches and low inter-core migration latencies. We evaluate performance on ...
Li, Zuoping; Kindig, Matthew W; Subit, Damien; Kent, Richard W
2010-11-01
The purpose of this paper was to investigate the sensitivity of the structural responses and bone fractures of the ribs to mesh density, cortical thickness, and material properties so as to provide guidelines for the development of finite element (FE) thorax models used in impact biomechanics. Subject-specific FE models of the second, fourth, sixth and tenth ribs were developed to reproduce dynamic failure experiments. Sensitivity studies were then conducted to quantify the effects of variations in mesh density, cortical thickness, and material parameters on the model-predicted reaction force-displacement relationship, cortical strains, and bone fracture locations for all four ribs. Overall, it was demonstrated that rib FE models consisting of 2000-3000 trabecular hexahedral elements (weighted element length 2-3mm) and associated quadrilateral cortical shell elements with variable thickness more closely predicted the rib structural responses and bone fracture force-failure displacement relationships observed in the experiments (except the fracture locations), compared to models with constant cortical thickness. Further increases in mesh density increased computational cost but did not markedly improve model predictions. A ±30% change in the major material parameters of cortical bone lead to a -16.7 to 33.3% change in fracture displacement and -22.5 to +19.1% change in the fracture force. The results in this study suggest that human rib structural responses can be modeled in an accurate and computationally efficient way using (a) a coarse mesh of 2000-3000 solid elements, (b) cortical shells elements with variable thickness distribution and (c) a rate-dependent elastic-plastic material model. Copyright © 2010 IPEM. Published by Elsevier Ltd. All rights reserved.
International Nuclear Information System (INIS)
Poursalehi, N.; Zolfaghari, A.; Minuchehr, A.
2013-01-01
Highlights: ► A new adaptive h-refinement approach has been developed for a class of nodal method. ► The resulting system of nodal equations is more amenable to efficient numerical solution. ► The benefit of the approach is reducing computational efforts relative to the uniform fine mesh modeling. ► Spatially adaptive approach greatly enhances the accuracy of the solution. - Abstract: The aim of this work is to develop a spatially adaptive coarse mesh strategy that progressively refines the nodes in appropriate regions of domain to solve the neutron balance equation by zeroth order nodal expansion method. A flux gradient based a posteriori estimation scheme has been utilized for checking the approximate solutions for various nodes. The relative surface net leakage of nodes has been considered as an assessment criterion. In this approach, the core module is called in by adaptive mesh generator to determine gradients of node surfaces flux to explore the possibility of node refinements in appropriate regions and directions of the problem. The benefit of the approach is reducing computational efforts relative to the uniform fine mesh modeling. For this purpose, a computer program ANRNE-2D, Adaptive Node Refinement Nodal Expansion, has been developed to solve neutron diffusion equation using average current nodal expansion method for 2D rectangular geometries. Implementing the adaptive algorithm confirms its superiority in enhancing the accuracy of the solution without using fine nodes throughout the domain and increasing the number of unknown solution. Some well-known benchmarks have been investigated and improvements are reported
Draft report: a selection methodology for LWR safety R and D programs and proposals
Energy Technology Data Exchange (ETDEWEB)
Husseiny, A. A.; Ritzman, R. L.
1980-03-01
The results of work done to develop a methodology for selecting LWR safety R and D programs and proposals is described. A critical survey of relevant decision analysis methods is provided including the specifics of multiattribute utility theory. This latter method forms the basis of the developed selection methodology. Details of the methodology and its use are provided along with a sample illustration of its application.
EUR, an European utility requirements documents for future LWR power stations
International Nuclear Information System (INIS)
Berbey, Pierre; Lienard, Michel; Redon, Ramon; Essmann, Juergen; Taylor, David T.
2004-01-01
A group of the major European utilities are developing a common requirement document which will be used for the LWR nuclear power plants to be built in Europe from the beginning of the next century. This document provides harmonised policies and technical requirements that will allow the implementation of a design developed in one country into another one. The objectives and contents of the document, the organisation set up for its production and the main requirements are summarised in the paper. (author)
Draft report: a selection methodology for LWR safety R and D programs and proposals
International Nuclear Information System (INIS)
Husseiny, A.A.; Ritzman, R.L.
1980-03-01
The results of work done to develop a methodology for selecting LWR safety R and D programs and proposals is described. A critical survey of relevant decision analysis methods is provided including the specifics of multiattribute utility theory. This latter method forms the basis of the developed selection methodology. Details of the methodology and its use are provided along with a sample illustration of its application
Literature search on Light Water Reactor (LWR) fuel and absorber rod fabrication, 1960--1976
International Nuclear Information System (INIS)
Sample, C.R.
1977-02-01
A literature search was conducted to provide information supporting the design of a conceptual Light Water Reactor (LWR) Fuel Fabrication plant. Emphasis was placed on fuel processing and pin bundle fabrication, effects of fuel impurities and microstructure on performance and densification, quality assurance, absorber and poison rod fabrication, and fuel pin welding. All data have been taken from publicly available documents, journals, and books. This work was sponsored by the Finishing Processes-Mixed Oxide (MOX) Fuel Fabrication Studies program at HEDL
Rate Theory Modeling and Simulation of Silicide Fuel at LWR Conditions
Energy Technology Data Exchange (ETDEWEB)
Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Ye, Bei [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Hofman, Gerard [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Gamble, Kyle [Idaho National Lab. (INL), Idaho Falls, ID (United States). Fuel Modeling and Simulation; Mei, Zhi-Gang [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division
2016-08-29
As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U_{3}Si_{2}) at LWR conditions needs to be well understood. In this report, rate theory model was developed based on existing experimental data and density functional theory (DFT) calculations so as to predict the fission gas behavior in U_{3}Si_{2} at LWR conditions. The fission gas behavior of U_{3}Si_{2} can be divided into three temperature regimes. During steady-state operation, the majority of the fission gas stays in intragranular bubbles, whereas the dominance of intergranular bubbles and fission gas release only occurs beyond 1000 K. The steady-state rate theory model was also used as reference to establish a gaseous swelling correlation of U_{3}Si_{2} for the BISON code. Meanwhile, the overpressurized bubble model was also developed so that the fission gas behavior at LOCA can be simulated. LOCA simulation showed that intragranular bubbles are still dominant after a 70 second LOCA, resulting in a controllable gaseous swelling. The fission gas behavior of U_{3}Si_{2} at LWR conditions is benign according to the rate theory prediction at both steady-state and LOCA conditions, which provides important references to the qualification of U_{3}Si_{2} as a LWR fuel material with excellent fuel performance and enhanced accident tolerance.
Recycle of LWR [Light Water Reactor] actinides to an IFR [Integral Fast Reactor
International Nuclear Information System (INIS)
Pierce, R.D.; Ackerman, J.P.; Johnson, G.K.; Mulcahey, T.P.; Poa, D.S.
1991-01-01
A large quantity of actinide elements is present in irradiated Light Water Reactor (LWR) fuel that is stored throughout the world. Because of the high fission-to-capture ratio for the transuranium (TRU) elements with the high-energy neutrons in the metal-fueled Integral Fast Reactor (IFR), that reactor can consume these elements effectively. The stored fuel represents a valuable resource for an expanding application of fast power reactors. In addition, removal of the TRU elements from the spent LWR fuel has the potential for increasing the capacity of a high-level waste facility by reducing the heat loads and increasing the margin of safety in meeting licensing requirements. Argonne National Laboratory (ANL) is developing a pyrochemical process, which is compatible with the IFR fuel cycle, for the recovery of TRU elements from LWR fuel. The proposed product is a metallic actinide ingot, which can be introduced into the electrorefining step of the IFR process. The major objective of the LWR fuel recovery process is high TRU element recovery, with decontamination a secondary issue, because fission product removal is accomplished in the IFR process. The extensive pyrochemical processing studies of the 1960s and 1970s provide a basis for the design of possible processes. Two processes were selected for laboratory-scale investigation. One is based on the Salt Transport Process studied at ANL for mixed-oxide fast reactor fuel, and the other is based on the blanket processing studies done for ANL's second Experimental Breeder Reactor (EBR-2). This paper discusses the two processes and is a status report on the experimental studies. 5 refs., 2 figs., 2 tabs
Literature search on Light Water Reactor (LWR) fuel and absorber rod fabrication, 1960--1976
Energy Technology Data Exchange (ETDEWEB)
Sample, C R [comp.
1977-02-01
A literature search was conducted to provide information supporting the design of a conceptual Light Water Reactor (LWR) Fuel Fabrication plant. Emphasis was placed on fuel processing and pin bundle fabrication, effects of fuel impurities and microstructure on performance and densification, quality assurance, absorber and poison rod fabrication, and fuel pin welding. All data have been taken from publicly available documents, journals, and books. This work was sponsored by the Finishing Processes-Mixed Oxide (MOX) Fuel Fabrication Studies program at HEDL.
Comment: collection of assay data on isotopic composition in LWR spent fuel
Energy Technology Data Exchange (ETDEWEB)
Naito, Yoshitaka; Kurosawa, Masayoshi; Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1997-03-01
Many assay data of LWR spent fuels have been collected from reactors in the world and some of them are already stored in the database SFCOMPO which was constructed on a personal computer IBM PC/AT. On the other hand, Group constant libraries for burnup calculation code ORIGEN-II were generated from the nuclear data file JENDL3.2. These libraries were evaluated by using the assay data in SFCOMPO. (author)
New development in nondestructive measurement and verification of irradiated LWR fuels
International Nuclear Information System (INIS)
Lee, D.M.; Phillips, J.R.; Halbig, J.K.; Hsue, S.T.; Lindquist, L.O.; Ortega, E.M.; Caine, J.C.; Swansen, J.; Kaieda, K.; Dermendjiev, E.
1979-01-01
Nondestructive techniques for characterizing irradiated LWR fuel assemblies are discussed. This includes detection systems that measure the axial activity profile, neutron yield and gamma yield. A multi-element profile monitor has been developed that offers a significant improvement in speed and complexity over existing mechanical scanning systems. New portable detectors and electronics, applicable to safeguard inspection, are presented and results of gamma-ray and neutron measurements at commercial reactor facilities are given
International Nuclear Information System (INIS)
Plaschy, M.; Murphy, M.; Jatuff, F.; Seiler, R.; Chawla, R.
2006-01-01
The PROTEUS research reactor at the Paul Scherrer Inst. (PSI) has been operating since the sixties and has already permitted, due to its high flexibility, investigation of a large range of very different nuclear systems. Currently, the ongoing experimental programme is called LWR-PROTEUS. This programme was started in 1997 and concerns large-scale investigations of advanced light water reactors (LWR) fuels. Until now, the different LWR-PROTEUS phases have permitted to study more than fifteen different configurations, each of them having to be demonstrated to be operationally safe, in particular, for the Swiss safety authorities. In this context, recent developments of the PSI computer capabilities have made possible the use of full-scale SD-heterogeneous MCNPX models to calculate accurately different safety related parameters (e.g. the critical driver loading and the shutdown rod worth). The current paper presents the MCNPX predictions of these operational characteristics for seven different LWR-PROTEUS configurations using a large number of nuclear data libraries. More specifically, this significant benchmarking exercise is based on the ENDF/B6v2, ENDF/B6v8, JEF2.2, JEFF3.0, JENDL3.2, and JENDL3.3 libraries. The results highlight certain library specific trends in the prediction of the multiplication factor k eff (e.g. the systematically larger reactivity calculated with JEF2.2 and the smaller reactivity associated with JEFF3.0). They also confirm the satisfactory determination of reactivity variations by all calculational schemes, for instance, due to the introduction of a safety rod pair, these calculations having been compared with experiments. (authors)
Unstructured Adaptive Meshes: Bad for Your Memory?
Biswas, Rupak; Feng, Hui-Yu; VanderWijngaart, Rob
2003-01-01
This viewgraph presentation explores the need for a NASA Advanced Supercomputing (NAS) parallel benchmark for problems with irregular dynamical memory access. This benchmark is important and necessary because: 1) Problems with localized error source benefit from adaptive nonuniform meshes; 2) Certain machines perform poorly on such problems; 3) Parallel implementation may provide further performance improvement but is difficult. Some examples of problems which use irregular dynamical memory access include: 1) Heat transfer problem; 2) Heat source term; 3) Spectral element method; 4) Base functions; 5) Elemental discrete equations; 6) Global discrete equations. Nonconforming Mesh and Mortar Element Method are covered in greater detail in this presentation.
Local adaptive mesh refinement for shock hydrodynamics
International Nuclear Information System (INIS)
Berger, M.J.; Colella, P.; Lawrence Livermore Laboratory, Livermore, 94550 California)
1989-01-01
The aim of this work is the development of an automatic, adaptive mesh refinement strategy for solving hyperbolic conservation laws in two dimensions. There are two main difficulties in doing this. The first problem is due to the presence of discontinuities in the solution and the effect on them of discontinuities in the mesh. The second problem is how to organize the algorithm to minimize memory and CPU overhead. This is an important consideration and will continue to be important as more sophisticated algorithms that use data structures other than arrays are developed for use on vector and parallel computers. copyright 1989 Academic Press, Inc
Adaptive mesh refinement for storm surge
Mandli, Kyle T.; Dawson, Clint N.
2014-01-01
An approach to utilizing adaptive mesh refinement algorithms for storm surge modeling is proposed. Currently numerical models exist that can resolve the details of coastal regions but are often too costly to be run in an ensemble forecasting framework without significant computing resources. The application of adaptive mesh refinement algorithms substantially lowers the computational cost of a storm surge model run while retaining much of the desired coastal resolution. The approach presented is implemented in the GeoClaw framework and compared to ADCIRC for Hurricane Ike along with observed tide gauge data and the computational cost of each model run. © 2014 Elsevier Ltd.
Adaptive mesh refinement for storm surge
Mandli, Kyle T.
2014-03-01
An approach to utilizing adaptive mesh refinement algorithms for storm surge modeling is proposed. Currently numerical models exist that can resolve the details of coastal regions but are often too costly to be run in an ensemble forecasting framework without significant computing resources. The application of adaptive mesh refinement algorithms substantially lowers the computational cost of a storm surge model run while retaining much of the desired coastal resolution. The approach presented is implemented in the GeoClaw framework and compared to ADCIRC for Hurricane Ike along with observed tide gauge data and the computational cost of each model run. © 2014 Elsevier Ltd.
Mesh removal following transvaginal mesh placement: a case series of 104 operations.
Marcus-Braun, Naama; von Theobald, Peter
2010-04-01
The objective of the study was to reveal the way we treat vaginal mesh complications in a trained referral center. This is a retrospective review of all patients who underwent surgical removal of transvaginal mesh for mesh-related complications during a 5-year period. Eighty-three patients underwent 104 operations including 61 complete mesh removal, 14 partial excision, 15 section of sub-urethral sling, and five laparoscopies. Main indications were erosion, infection, granuloma, incomplete voiding, and pain. Fifty-eight removals occurred more than 2 years after the primary mesh placement. Mean operation time was 21 min, and there were two intraoperative and ten minor postoperative complications. Stress urinary incontinence (SUI) recurred in 38% and cystocele in 19% of patients. In a trained center, mesh removal was found to be a quick and safe procedure. Mesh-related complications may frequently occur more than 2 years after the primary operation. Recurrence was mostly associated with SUI and less with genital prolapse.
An overview of advanced high-strength nickel-base alloys for LWR applications
International Nuclear Information System (INIS)
Prybylowski, J.; Ballinger, R.G.
1989-01-01
This paper reviews our current understanding of the behavior of high strength nickel base alloys used in light water reactor (LWR) applications. Emphasis is placed on understanding the fundamental mechanisms controlling crack propagation in these environments. To provide a foundation for this survey, general mechanisms of stress corrosion cracking and hydrogen embrittlement are first reviewed. The behavior of high strength nickel base alloys in LWR environments, as well as in other relevant environments is then reviewed. Suggested mechanisms of crack propagation are discussed. Alternate alloys and microstructural modifications that may result in improved behavior are presented. It is now clear that, at temperatures near 100C, alloy X-750, the predominant high strength nickel base alloy used today in LWR applications, is susceptible to hydrogen embrittlement. A review of published data from hydrogen embrittlement studies of nickel base superalloys during electrolytic charging and in hydrogen sulfide/brine solutions suggests that other nickel base superalloys are available possessing resistance to hydrogen embrittlement superior to that of alloy X-750. Available results of tests in gaseous hydrogen suggest that reduced grain boundary precipitation and a fine distribution of intragranular precipitates that act as irreversible hydrogen traps is the optimum microstructure for hydrogen embrittlement resistance. 42 refs., 2 figs., 5 tabs
Minutes of the Twelfth LWR pressure vessel surveillance dosimtery improvement program meeting
International Nuclear Information System (INIS)
1989-01-01
The 1983 Twelfth Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program (LWR-PV-SDIP) Meeting, which was held October 24-28, 1983. Sections 1 through 14 of this report provide documentation of agreements, commitments, and reports that are subject to the approval and concurrence of the participating laboratories and supporting agencies and organizations. Attachment No. 1 provides information on the preparation of a number of NUREG publications that will document the results of various aspects of the LWR-PV-SDIP. For each NUREG publication, a tentative ''Table of Contents'' is provided in addition to suggested interlaboratory writing assignments and camera-ready copy contribution due dates, as appropriate. Attachment No. 2 provides information on planning for the Fifth ASTM-EURATOM Symposium. Attachment No. 3 provides information on an ASTM press release about an MPC-6 meeting and dpa and E > 1 MeV exposure parameters. Attachments No. 4 and 5 provide copies of two LWR-PV-SDIP related papers presented at the Eleventh WRSR Information Meeting, October 24-28, 1983
APEX nuclear fuel cycle for production of LWR fuel and elimination of radioactive waste
International Nuclear Information System (INIS)
Steinberg, M.; Powell, J.R.
1981-08-01
The development of a nuclear fission fuel cycle is proposed which eliminates all the radioactive fission product waste effluent and the need for geological-age high level waste storage and provides a long term supply of fissile fuel for an LWR power reactor economy. The fuel cycle consists of reprocessing LWR spent fuel (1 to 2 years old) to remove the stable nonradioactive (NRFP, e.g. lanthanides, etc.) and short-lived fission products SLFP e.g. half-lives of (1 to 2 years) and returning, in dilute form, the long-lived fission products, ((LLFPs, e.g. 30 y half-life Cs, Sr, and 10 y Kr, and 16 x 10 6 y I) and the transuranics (TUs, e.g. Pu, Am, Cm, and Np) to be refabricated into fresh fuel elements. Makeup fertile and fissile fuel are to be supplied through the use of a Spallator (linear accelerator spallation-target fuel-producer). The reprocessing of LWR fuel elements is to be performed by means of the Chelox process which consists of Airox treatment (air oxidation and hydrogen reduction) followed by chelation with an organic reagent (β-diketonate) and vapor distillation of the organometallic compounds for separation and partitioning of the fission products
Effects of LWR coolant environments on fatigue lives of austenitic stainless steels
International Nuclear Information System (INIS)
Chopra, O.K.; Gavenda, D.J.
1997-01-01
The ASME Boiler and Pressure Vessel Code fatigue design curves for structural materials do not explicitly address the effects of reactor coolant environments on fatigue life. Recent test data indicate a significant decrease in fatigue life of pressure vessel and piping materials in light water reactor (LWR) environments. Fatigue tests have been conducted on Types 304 and 316NG stainless steel in air and LWR environments to evaluate the effects of various material and loading variables, e.g., steel type, strain rate, dissolved oxygen (DO) in water, and strain range, on fatigue lives of these steels. The results confirm the significant decrease in fatigue life in water. The environmentally assisted decrease in fatigue life depends both on strain rate and DO content in water. A decrease in strain rate from 0.4 to 0.004%/s decreases fatigue life by a factor of ∼ 8. However, unlike carbon and low-alloy steels, environmental effects are more pronounced in low-DO than in high-DO water. At ∼ 0.004%/s strain rate, reduction in fatigue life in water containing <10 ppb D is greater by a factor of ∼ 2 than in water containing ≥ 200 ppb DO. Experimental results have been compared with estimates of fatigue life based on the statistical model. The formation and growth of fatigue cracks in austenitic stainless steels in air and LWR environments are discussed
Results of the LIRES Round Robin test on high temperature reference electrodes for LWR applications
Energy Technology Data Exchange (ETDEWEB)
Bosch, R.W. [SCK.CEN, Nuclear Research Centre Belgium, Boeretang 200, B-2400 Mol (Belgium); Nagy, G. [Magyar Tudomanyos Akademia KFKI Atomenergia Kutatointezet, AEKI, Konkoly Thege ut 29-33, 1121 Budapest (Hungary); Feron, D. [CEA Saclay, 91191 Gif-Sur-Yvette Cedex (France); Navas, M. [CIEMAT, Edificio 30, Dpto. Fision Nuclear, Avda. Complutense 22, 28040 Madrid, (Spain); Bogaerts, W. [KU Leuven, Kasteelpark Arenberg 31, B-3001 Leuven (Belgium); Karnik, D. [Nuclear Research Institute, NRI, Rez (Czech Republic); Dorsch, T. [Framatone ANP, Inc., Charlotte, North Carolina (United States); Molander, A. [Studsvik AB SE-611 82 Nykoeping (Sweden); Maekelae, K. [Materials and Structural Integrity, VTT Technical Research Centre of Finland, Kemistintie 3, P.O. Box 1704, FIN-02044 VTT (Finland)
2004-07-01
A European sponsored research project has been started on 1 October 2000 to develop high temperature reference electrodes that can be used for in-core electrochemical measurements in Light Water Reactors (LWR's). This LIRES-project (Development of Light Water Reactor Reference Electrodes) consists of 9 partners (SCK-CEN, AEKI, CEA, CIEMAT, KU Leuven, NRI Rez, Framatone ANP, Studsvik Nuclear and VTT) and will last for four years. The main objective of this LIRES project is to develop a reference electrode, which is robust enough to be used inside a LWR. Emphasize is put on the radiation hardness of both the mechanical design of the electrode as the proper functioning of the electrode. A four steps development trajectory is foreseen: (1) To set a testing standard for a Round Robin, (2) To develop different reference electrodes, (3) To perform a Round Robin test of these reference electrodes followed by selection of the best reference electrode(s), (4) To perform irradiation tests under appropriate LWR conditions in a Material Test Reactor (MTR). Four different high temperature reference electrodes have been developed and are being tested in a Round Robin test. These electrodes are: A Ceramic Membrane Electrode (CME), a Rhodium electrode, an external Ag/AgCl electrode and a Palladium electrode. The presentation will focus on the results obtained with the Round Robin test. (authors)
Pure transvaginal excision of mesh erosion involving the bladder.
Firoozi, Farzeen; Goldman, Howard B
2013-06-01
We present a pure transvaginal approach to the removal of eroded mesh involving the bladder secondary to placement of transvaginal mesh for management of pelvic organ prolapse (POP) using a mesh kit. Although technically challenging, we demonstrate the feasibility of a purely transvaginal approach, avoiding a potentially more morbid transabdominal approach. The video presents the surgical technique of pure transvaginal excision of mesh erosion involving the bladder after mesh placement using a prolapse kit was performed. This video shows that purely transvaginal removal of mesh erosion involving the bladder can be done safely and is feasible.
Seker, D; Oztuna, D; Kulacoglu, H; Genc, Y; Akcil, M
2013-04-01
Small mesh size has been recognized as one of the factors responsible for recurrence after Lichtenstein hernia repair due to insufficient coverage or mesh shrinkage. The Lichtenstein Hernia Institute recommends a 7 × 15 cm mesh that can be trimmed up to 2 cm from the lateral side. We performed a systematic review to determine surgeons' mesh size preference for the Lichtenstein hernia repair and made a meta-analysis to determine the effect of mesh size, mesh type, and length of follow-up time on recurrence. Two medical databases, PubMed and ISI Web of Science, were systematically searched using the key word "Lichtenstein repair." All full text papers were selected. Publications mentioning mesh size were brought for further analysis. A mesh surface area of 90 cm(2) was accepted as the threshold for defining the mesh as small or large. Also, a subgroup analysis for recurrence pooled proportion according to the mesh size, mesh type, and follow-up period was done. In total, 514 papers were obtained. There were no prospective or retrospective clinical studies comparing mesh size and clinical outcome. A total of 141 papers were duplicated in both databases. As a result, 373 papers were obtained. The full text was available in over 95 % of papers. Only 41 (11.2 %) papers discussed mesh size. In 29 studies, a mesh larger than 90 cm(2) was used. The most frequently preferred commercial mesh size was 7.5 × 15 cm. No papers mentioned the size of the mesh after trimming. There was no information about the relationship between mesh size and patient BMI. The pooled proportion in recurrence for small meshes was 0.0019 (95 % confidence interval: 0.007-0.0036), favoring large meshes to decrease the chance of recurrence. Recurrence becomes more marked when follow-up period is longer than 1 year (p < 0.001). Heavy meshes also decreased recurrence (p = 0.015). This systematic review demonstrates that the size of the mesh used in Lichtenstein hernia repair is rarely
Properties of meshes used in hernia repair: a comprehensive review of synthetic and biologic meshes.
Ibrahim, Ahmed M S; Vargas, Christina R; Colakoglu, Salih; Nguyen, John T; Lin, Samuel J; Lee, Bernard T
2015-02-01
Data on the mechanical properties of the adult human abdominal wall have been difficult to obtain rendering manufacture of the ideal mesh for ventral hernia repair a challenge. An ideal mesh would need to exhibit greater biomechanical strength and elasticity than that of the abdominal wall. The aim of this study is to quantitatively compare the biomechanical properties of the most commonly used synthetic and biologic meshes in ventral hernia repair and presents a comprehensive literature review. A narrative review of the literature was performed using the PubMed database spanning articles from 1982 to 2012 including a review of company Web sites to identify all available information relating to the biomechanical properties of various synthetic and biologic meshes used in ventral hernia repair. There exist differences in the mechanical properties and the chemical nature of different meshes. In general, most synthetic materials have greater stiffness and elasticity than what is required for abdominal wall reconstruction; however, each exhibits unique properties that may be beneficial for clinical use. On the contrary, biologic meshes are more elastic but less stiff and with a lower tensile strength than their synthetic counterparts. The current standard of practice for the treatment of ventral hernias is the use of permanent synthetic mesh material. Recently, biologic meshes have become more frequently used. Most meshes exhibit biomechanical properties over the known abdominal wall thresholds. Augmenting strength requires increasing amounts of material contributing to more stiffness and foreign body reaction, which is not necessarily an advantage. Thieme Medical Publishers 333 Seventh Avenue, New York, NY 10001, USA.
CONTEMPT, LWR Containment Pressure and Temperature Distribution in LOCA
International Nuclear Information System (INIS)
Hargroves, D.W.; Metcalfe, L.J.; Cheng, Teh-Chin; Wheat, L.L.; Mings, W.J.
1991-01-01
describing structure behavior are advanced using an implicit technique. The resulting heat transfer rates are used to correct the previous estimates of the water and energy storage in the containment volume, and the containment conditions are obtained by solving for the second time the containment balance equations. The pressure suppression routines use the conditions at the beginning of a time-step to calculate both the initial explosion of water from the vents and the flow through the vents. From the calculated flow rates, mass and energy are removed from the dry well and added to the wet well. 3 - Restrictions on the complexity of the problem - Maxima of: 20 heat conducting structures, 101 mesh points for each structure, 20 regions for each structure, 50 flow elements in one segment of the horizontal vent pressure suppression system, 10 horizontal vents (or branches) in a segment, 50 reductions within an input time-step. CONTEMPT-LT can be used for analyzing the transient containment behavior of boiling-water reactors (BWRs) including Mark I, Mark II, and Mark III systems; pressurized-water reactors (PWRs), and experimental water reactor simulators or related experiments
Highly Symmetric and Congruently Tiled Meshes for Shells and Domes
Rasheed, Muhibur; Bajaj, Chandrajit
2016-01-01
We describe the generation of all possible shell and dome shapes that can be uniquely meshed (tiled) using a single type of mesh face (tile), and following a single meshing (tiling) rule that governs the mesh (tile) arrangement with maximal vertex, edge and face symmetries. Such tiling arrangements or congruently tiled meshed shapes, are frequently found in chemical forms (fullerenes or Bucky balls, crystals, quasi-crystals, virus nano shells or capsids), and synthetic shapes (cages, sports domes, modern architectural facades). Congruently tiled meshes are both aesthetic and complete, as they support maximal mesh symmetries with minimal complexity and possess simple generation rules. Here, we generate congruent tilings and meshed shape layouts that satisfy these optimality conditions. Further, the congruent meshes are uniquely mappable to an almost regular 3D polyhedron (or its dual polyhedron) and which exhibits face-transitive (and edge-transitive) congruency with at most two types of vertices (each type transitive to the other). The family of all such congruently meshed polyhedra create a new class of meshed shapes, beyond the well-studied regular, semi-regular and quasi-regular classes, and their duals (platonic, Catalan and Johnson). While our new mesh class is infinite, we prove that there exists a unique mesh parametrization, where each member of the class can be represented by two integer lattice variables, and moreover efficiently constructable. PMID:27563368
Wijdeven, S.M.J.; Vaessen, O.H.B.; Hees, van A.F.M.; Olsthoorn, A.F.M.
2005-01-01
Dead wood is recognized as one of the key indicators for sustainable forest management and biodiversity. Accurate assessments of dead wood volume are thus necessary. In this study New volume models were designed based on actual volume measurements of coarse woody debris. The New generic model
Markov Random Fields on Triangle Meshes
DEFF Research Database (Denmark)
Andersen, Vedrana; Aanæs, Henrik; Bærentzen, Jakob Andreas
2010-01-01
In this paper we propose a novel anisotropic smoothing scheme based on Markov Random Fields (MRF). Our scheme is formulated as two coupled processes. A vertex process is used to smooth the mesh by displacing the vertices according to a MRF smoothness prior, while an independent edge process label...
Performance Evaluation of Coded Meshed Networks
DEFF Research Database (Denmark)
Krigslund, Jeppe; Hansen, Jonas; Pedersen, Morten Videbæk
2013-01-01
of the former to enhance the gains of the latter. We first motivate our work through measurements in WiFi mesh networks. Later, we compare state-of-the-art approaches, e.g., COPE, RLNC, to CORE. Our measurements show the higher reliability and throughput of CORE over other schemes, especially, for asymmetric...
Solid Mesh Registration for Radiotherapy Treatment Planning
DEFF Research Database (Denmark)
Noe, Karsten Østergaard; Sørensen, Thomas Sangild
2010-01-01
We present an algorithm for solid organ registration of pre-segmented data represented as tetrahedral meshes. Registration of the organ surface is driven by force terms based on a distance field representation of the source and reference shapes. Registration of internal morphology is achieved usi...
Direct numerical simulation of bubbles with parallelized adaptive mesh refinement
International Nuclear Information System (INIS)
Talpaert, A.
2015-01-01
The study of two-phase Thermal-Hydraulics is a major topic for Nuclear Engineering for both security and efficiency of nuclear facilities. In addition to experiments, numerical modeling helps to knowing precisely where bubbles appear and how they behave, in the core as well as in the steam generators. This work presents the finest scale of representation of two-phase flows, Direct Numerical Simulation of bubbles. We use the 'Di-phasic Low Mach Number' equation model. It is particularly adapted to low-Mach number flows, that is to say flows which velocity is much slower than the speed of sound; this is very typical of nuclear thermal-hydraulics conditions. Because we study bubbles, we capture the front between vapor and liquid phases thanks to a downward flux limiting numerical scheme. The specific discrete analysis technique this work introduces is well-balanced parallel Adaptive Mesh Refinement (AMR). With AMR, we refined the coarse grid on a batch of patches in order to locally increase precision in areas which matter more, and capture fine changes in the front location and its topology. We show that patch-based AMR is very adapted for parallel computing. We use a variety of physical examples: forced advection, heat transfer, phase changes represented by a Stefan model, as well as the combination of all those models. We will present the results of those numerical simulations, as well as the speed up compared to equivalent non-AMR simulation and to serial computation of the same problems. This document is made up of an abstract and the slides of the presentation. (author)
Vertex Normals and Face Curvatures of Triangle Meshes
Sun, Xiang; Jiang, Caigui; Wallner, Johannes; Pottmann, Helmut
2016-01-01
This study contributes to the discrete differential geometry of triangle meshes, in combination with discrete line congruences associated with such meshes. In particular we discuss when a congruence defined by linear interpolation of vertex normals
Recurrence and Pain after Mesh Repair of Inguinal Hernias
African Journals Online (AJOL)
Abstract. Background: Surgery for inguinal hernias has ... repair. Methods: The study was conducted on all inguinal hernia patients operated between 1st. October ... bilateral (1.6%). Only 101 .... Open Mesh Versus Laparoscopic Mesh. Repair ...
Surgical Management of Pelvic floor Prolapse in women using Mesh
African Journals Online (AJOL)
RAH
polytetrafluoroethylene) . This article reviews our experience with polypropylene mesh in pelvic floor repair at the. Southern General Hospital Glasgow. The objective was to determine the safety and effectiveness of the prolene mesh in the repair ...
VARIABLE MESH STIFFNESS OF SPUR GEAR TEETH USING ...
African Journals Online (AJOL)
gear engagement. A gear mesh kinematic simulation ... model is appropnate for VMS of a spur gear tooth. The assumptions for ... This process has been continued until one complete tooth meshing cycle is ..... Element Method. Using MATLAB,.
To mesh or not to mesh: a review of pelvic organ reconstructive surgery
Dällenbach, Patrick
2015-01-01
Pelvic organ prolapse (POP) is a major health issue with a lifetime risk of undergoing at least one surgical intervention estimated at close to 10%. In the 1990s, the risk of reoperation after primary standard vaginal procedure was estimated to be as high as 30% to 50%. In order to reduce the risk of relapse, gynecological surgeons started to use mesh implants in pelvic organ reconstructive surgery with the emergence of new complications. Recent studies have nevertheless shown that the risk of POP recurrence requiring reoperation is lower than previously estimated, being closer to 10% rather than 30%. The development of mesh surgery – actively promoted by the marketing industry – was tremendous during the past decade, and preceded any studies supporting its benefit for our patients. Randomized trials comparing the use of mesh to native tissue repair in POP surgery have now shown better anatomical but similar functional outcomes, and meshes are associated with more complications, in particular for transvaginal mesh implants. POP is not a life-threatening condition, but a functional problem that impairs quality of life for women. The old adage “primum non nocere” is particularly appropriate when dealing with this condition which requires no treatment when asymptomatic. It is currently admitted that a certain degree of POP is physiological with aging when situated above the landmark of the hymen. Treatment should be individualized and the use of mesh needs to be selective and appropriate. Mesh implants are probably an important tool in pelvic reconstructive surgery, but the ideal implant has yet to be found. The indications for its use still require caution and discernment. This review explores the reasons behind the introduction of mesh augmentation in POP surgery, and aims to clarify the risks, benefits, and the recognized indications for its use. PMID:25848324
McCoy, Olugbemisola; Vaughan, Taylor; Nickles, S Walker; Ashley, Matt; MacLachlan, Lara S; Ginsberg, David; Rovner, Eric
2016-08-01
We reviewed the outcomes of the autologous fascial pubovaginal sling as a salvage procedure for recurrent stress incontinence after intervention for polypropylene mesh erosion/exposure and/or bladder outlet obstruction in patients treated with prior transvaginal synthetic mesh for stress urinary incontinence. In a review of surgical databases at 2 institutions between January 2007 and June 2013 we identified 46 patients who underwent autologous fascial pubovaginal sling following removal of transvaginal synthetic mesh in simultaneous or staged fashion. This cohort of patients was evaluated for outcomes, including subjective and objective success, change in quality of life and complications between those who underwent staged vs concomitant synthetic mesh removal with autologous fascial pubovaginal sling placement. All 46 patients had received at least 1 prior mesh sling for incontinence and 8 (17%) had received prior transvaginal polypropylene mesh for pelvic organ prolapse repair. A total of 30 patients underwent concomitant mesh incision with or without partial excision and autologous sling placement while 16 underwent staged autologous sling placement. Mean followup was 16 months. Of the patients 22% required a mean of 1.8 subsequent interventions an average of 6.5 months after autologous sling placement with no difference in median quality of life at final followup. At last followup 42 of 46 patients (91%) and 35 of 46 (76%) had achieved objective and subjective success, respectively. There was no difference in subjective success between patients treated with a staged vs a concomitant approach (69% vs 80%, p = 0.48). Autologous fascial pubovaginal sling placement after synthetic mesh removal can be performed successfully in patients with stress urinary incontinence as a single or staged procedure. Copyright © 2016 American Urological Association Education and Research, Inc. Published by Elsevier Inc. All rights reserved.
Persistent pelvic pain following transvaginal mesh surgery: a cause for mesh removal.
Marcus-Braun, Naama; Bourret, Antoine; von Theobald, Peter
2012-06-01
Persistent pelvic pain after vaginal mesh surgery is an uncommon but serious complication that greatly affects women's quality of life. Our aim was to evaluate various procedures for mesh removal performed at a tertiary referral center in cases of persistent pelvic pain, and to evaluate the ensuing complications and outcomes. A retrospective study was conducted at the University Hospital of Caen, France, including all patients treated for removal or section of vaginal mesh due to pelvic pain as a primary cause, between January 2004 and September 2009. Ten patients met the inclusion criteria. Patients were diagnosed between 10 months and 3 years after their primary operation. Eight cases followed suburethral sling procedures and two followed mesh surgery for pelvic organ prolapse. Patients presented with obturator neuralgia (6), pudendal neuralgia (2), dyspareunia (1), and non-specific pain (1). The surgical treatment to release the mesh included: three cases of extra-peritoneal laparoscopy, four cases of complete vaginal mesh removal, one case of partial mesh removal and two cases of section of the suburethral sling. In all patients with obturator neuralgia, symptoms were resolved or improved, whereas in both cases of pudendal neuralgia the symptoms continued. There were no intra-operative complications. Post-operative Retzius hematoma was observed in one patient after laparoscopy. Mesh removal in a tertiary center is a safe procedure, necessary in some cases of persistent pelvic pain. Obturator neuralgia seems to be easier to treat than pudendal neuralgia. Early diagnosis is the key to success in prevention of chronic disease. Copyright © 2012 Elsevier Ireland Ltd. All rights reserved.
Laparoscopic removal of mesh used in pelvic floor surgery.
Khong, Su-Yen; Lam, Alan
2009-01-01
Various meshes are being used widely in clinical practice for pelvic reconstructive surgery despite the lack of evidence of their long-term safety and efficacy. Management of complications such as mesh erosion and dyspareunia can be challenging. Most mesh-related complications can probably be managed successfully via the transvaginal route; however, this may be impossible if surgical access is poor. This case report demonstrates the successful laparoscopic removal of mesh after several failed attempts via the vaginal route.
On Reducing Delay in Mesh-Based P2P Streaming: A Mesh-Push Approach
Liu, Zheng; Xue, Kaiping; Hong, Peilin
The peer-assisted streaming paradigm has been widely employed to distribute live video data on the internet recently. In general, the mesh-based pull approach is more robust and efficient than the tree-based push approach. However, pull protocol brings about longer streaming delay, which is caused by the handshaking process of advertising buffer map message, sending request message and scheduling of the data block. In this paper, we propose a new approach, mesh-push, to address this issue. Different from the traditional pull approach, mesh-push implements block scheduling algorithm at sender side, where the block transmission is initiated by the sender rather than by the receiver. We first formulate the optimal upload bandwidth utilization problem, then present the mesh-push approach, in which a token protocol is designed to avoid block redundancy; a min-cost flow model is employed to derive the optimal scheduling for the push peer; and a push peer selection algorithm is introduced to reduce control overhead. Finally, we evaluate mesh-push through simulation, the results of which show mesh-push outperforms the pull scheduling in streaming delay, and achieves comparable delivery ratio at the same time.
Multi-cell vortices observed in fine-mesh solutions to the incompressible Euler equations
International Nuclear Information System (INIS)
Rizzi, A.
1986-01-01
Results are presented for a three dimensional flow, containing a vortex sheet shed from a delta wing. The numerical solution indicates that the shearing caused by the trailing edge of the wing set up a torsional wave on the vortex core and produces a structure with multiple cells of vorticity. Although observed in coarse grid solutions too, this effect becomes better resolved with mesh refinement to 614 000 grid volumes. In comparison with a potential solution in which the vortex sheet is fitted as a discontinuity, the results are analyzed for the position of the vortex features captured in the Euler flow field, the accuracy of the pressure field, and for the diffusion of the vortex sheets
Shah, Ketul; Nikolavsky, Dmitriy; Gilsdorf, Daniel; Flynn, Brian J
2013-12-01
We present our management of lower urinary tract (LUT) mesh perforation after mid-urethral polypropylene mesh sling using a novel combination of surgical techniques including total or near total mesh excision, urinary tract reconstruction, and concomitant pubovaginal sling with autologous rectus fascia in a single operation. We retrospectively reviewed the medical records of 189 patients undergoing transvaginal removal of polypropylene mesh from the lower urinary tract or vagina. The focus of this study is 21 patients with LUT mesh perforation after mid-urethral polypropylene mesh sling. We excluded patients with LUT mesh perforation from prolapse kits (n = 4) or sutures (n = 11), or mesh that was removed because of isolated vaginal wall exposure without concomitant LUT perforation (n = 164). Twenty-one patients underwent surgical removal of mesh through a transvaginal approach or combined transvaginal/abdominal approaches. The location of the perforation was the urethra in 14 and the bladder in 7. The mean follow-up was 22 months. There were no major intraoperative complications. All patients had complete resolution of the mesh complication and the primary symptom. Of the patients with urethral perforation, continence was achieved in 10 out of 14 (71.5 %). Of the patients with bladder perforation, continence was achieved in all 7. Total or near total removal of lower urinary tract (LUT) mesh perforation after mid-urethral polypropylene mesh sling can completely resolve LUT mesh perforation in a single operation. A concomitant pubovaginal sling can be safely performed in efforts to treat existing SUI or avoid future surgery for SUI.
21 CFR 870.3650 - Pacemaker polymeric mesh bag.
2010-04-01
... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Pacemaker polymeric mesh bag. 870.3650 Section 870...) MEDICAL DEVICES CARDIOVASCULAR DEVICES Cardiovascular Prosthetic Devices § 870.3650 Pacemaker polymeric mesh bag. (a) Identification. A pacemaker polymeric mesh bag is an implanted device used to hold a...
Multiphase flow of immiscible fluids on unstructured moving meshes
DEFF Research Database (Denmark)
Misztal, Marek Krzysztof; Erleben, Kenny; Bargteil, Adam
2012-01-01
In this paper, we present a method for animating multiphase flow of immiscible fluids using unstructured moving meshes. Our underlying discretization is an unstructured tetrahedral mesh, the deformable simplicial complex (DSC), that moves with the flow in a Lagrangian manner. Mesh optimization op...
Multiphase Flow of Immiscible Fluids on Unstructured Moving Meshes
DEFF Research Database (Denmark)
Misztal, Marek Krzysztof; Erleben, Kenny; Bargteil, Adam
2013-01-01
In this paper, we present a method for animating multiphase flow of immiscible fluids using unstructured moving meshes. Our underlying discretization is an unstructured tetrahedral mesh, the deformable simplicial complex (DSC), that moves with the flow in a Lagrangian manner. Mesh optimization op...
Prosthetic Mesh Repair for Incarcerated Inguinal Hernia
Directory of Open Access Journals (Sweden)
Cihad Tatar
2016-08-01
Full Text Available Background: Incarcerated inguinal hernia is a commonly encountered urgent surgical condition, and tension-free repair is a well-established method for the treatment of noncomplicated cases. However, due to the risk of prosthetic material-related infections, the use of mesh in the repair of strangulated or incarcerated hernia has often been subject to debate. Recent studies have demonstrated that biomaterials represent suitable materials for performing urgent hernia repair. Certain studies recommend mesh repair only for cases where no bowel resection is required; other studies, however, recommend mesh repair for patients requiring bowel resection as well. Aim: The aim of this study was to compare the outcomes of different surgical techniques performed for strangulated hernia, and to evaluate the effect of mesh use on postoperative complications. Study Design: Retrospective cross-sectional study. Methods: This retrospective study was performed with 151 patients who had been admitted to our hospital’s emergency department to undergo surgery for a diagnosis of incarcerated inguinal hernia. The patients were divided into two groups based on the applied surgical technique. Group 1 consisted of 112 patients treated with mesh-based repair techniques, while Group 2 consisted of 39 patients treated with tissue repair techniques. Patients in Group 1 were further divided into two sub-groups: one consisting of patients undergoing bowel resection (Group 3, and the other consisting of patients not undergoing bowel resection (Group 4. Results: In Group 1, it was observed that eight (7.14% of the patients had wound infections, while two (1.78% had hematomas, four (3.57% had seromas, and one (0.89% had relapse. In Group 2, one (2.56% of the patients had a wound infection, while three (7.69% had hematomas, one (2.56% had seroma, and none had relapses. There were no statistically significant differences between the two groups with respect to wound infection
To mesh or not to mesh: a review of pelvic organ reconstructive surgery
Directory of Open Access Journals (Sweden)
Dällenbach P
2015-04-01
Full Text Available Patrick Dällenbach Department of Gynecology and Obstetrics, Division of Gynecology, Urogynecology Unit, Geneva University Hospitals, Geneva, Switzerland Abstract: Pelvic organ prolapse (POP is a major health issue with a lifetime risk of undergoing at least one surgical intervention estimated at close to 10%. In the 1990s, the risk of reoperation after primary standard vaginal procedure was estimated to be as high as 30% to 50%. In order to reduce the risk of relapse, gynecological surgeons started to use mesh implants in pelvic organ reconstructive surgery with the emergence of new complications. Recent studies have nevertheless shown that the risk of POP recurrence requiring reoperation is lower than previously estimated, being closer to 10% rather than 30%. The development of mesh surgery – actively promoted by the marketing industry – was tremendous during the past decade, and preceded any studies supporting its benefit for our patients. Randomized trials comparing the use of mesh to native tissue repair in POP surgery have now shown better anatomical but similar functional outcomes, and meshes are associated with more complications, in particular for transvaginal mesh implants. POP is not a life-threatening condition, but a functional problem that impairs quality of life for women. The old adage “primum non nocere” is particularly appropriate when dealing with this condition which requires no treatment when asymptomatic. It is currently admitted that a certain degree of POP is physiological with aging when situated above the landmark of the hymen. Treatment should be individualized and the use of mesh needs to be selective and appropriate. Mesh implants are probably an important tool in pelvic reconstructive surgery, but the ideal implant has yet to be found. The indications for its use still require caution and discernment. This review explores the reasons behind the introduction of mesh augmentation in POP surgery, and aims to
Optimal Design of Experiments by Combining Coarse and Fine Measurements
Lee, Alpha A.; Brenner, Michael P.; Colwell, Lucy J.
2017-11-01
In many contexts, it is extremely costly to perform enough high-quality experimental measurements to accurately parametrize a predictive quantitative model. However, it is often much easier to carry out large numbers of experiments that indicate whether each sample is above or below a given threshold. Can many such categorical or "coarse" measurements be combined with a much smaller number of high-resolution or "fine" measurements to yield accurate models? Here, we demonstrate an intuitive strategy, inspired by statistical physics, wherein the coarse measurements are used to identify the salient features of the data, while the fine measurements determine the relative importance of these features. A linear model is inferred from the fine measurements, augmented by a quadratic term that captures the correlation structure of the coarse data. We illustrate our strategy by considering the problems of predicting the antimalarial potency and aqueous solubility of small organic molecules from their 2D molecular structure.
Entropies from Coarse-graining: Convex Polytopes vs. Ellipsoids
Directory of Open Access Journals (Sweden)
Nikos Kalogeropoulos
2015-09-01
Full Text Available We examine the Boltzmann/Gibbs/Shannon SBGS and the non-additive Havrda-Charvát/Daróczy/Cressie-Read/Tsallis Sq and the Kaniadakis κ-entropy Sκ from the viewpoint of coarse-graining, symplectic capacities and convexity. We argue that the functional form of such entropies can be ascribed to a discordance in phase-space coarse-graining between two generally different approaches: the Euclidean/Riemannian metric one that reflects independence and picks cubes as the fundamental cells in coarse-graining and the symplectic/canonical one that picks spheres/ellipsoids for this role. Our discussion is motivated by and confined to the behaviour of Hamiltonian systems of many degrees of freedom. We see that Dvoretzky’s theorem provides asymptotic estimates for the minimal dimension beyond which these two approaches are close to each other. We state and speculate about the role that dualities may play in this viewpoint.
Renormalization and Coarse-graining of Loop Quantum Gravity
Charles, Christoph
2017-01-01
The continuum limit of loop quantum gravity is still an open problem. Indeed, no proper dynamics in known to start with and we still lack the mathematical tools to study its would-be continuum limit. In the present PhD dissertation, we will investigate some coarse-graining methods that should become helpful in this enterprise. We concentrate on two aspects of the theory's coarse-graining: finding natural large scale observables on one hand and studying how the dynamics of varying graphs could...
Design of low-power coarse-grained reconfigurable architectures
Kim, Yoonjin
2010-01-01
Coarse-grained reconfigurable architecture (CGRA) has emerged as a solution for flexible, application-specific optimization of embedded systems. Helping you understand the issues involved in designing and constructing embedded systems, Design of Low-Power Coarse-Grained Reconfigurable Architectures offers new frameworks for optimizing the architecture of components in embedded systems in order to decrease area and save power. Real application benchmarks and gate-level simulations substantiate these frameworks.The first half of the book explains how to reduce power in the configuration cache. T
Partitioning of unstructured meshes for load balancing
International Nuclear Information System (INIS)
Martin, O.C.; Otto, S.W.
1994-01-01
Many large-scale engineering and scientific calculations involve repeated updating of variables on an unstructured mesh. To do these types of computations on distributed memory parallel computers, it is necessary to partition the mesh among the processors so that the load balance is maximized and inter-processor communication time is minimized. This can be approximated by the problem, of partitioning a graph so as to obtain a minimum cut, a well-studied combinatorial optimization problem. Graph partitioning algorithms are discussed that give good but not necessarily optimum solutions. These algorithms include local search methods recursive spectral bisection, and more general purpose methods such as simulated annealing. It is shown that a general procedure enables to combine simulated annealing with Kernighan-Lin. The resulting algorithm is both very fast and extremely effective. (authors) 23 refs., 3 figs., 1 tab
Adaptive upscaling with the dual mesh method
Energy Technology Data Exchange (ETDEWEB)
Guerillot, D.; Verdiere, S.
1997-08-01
The objective of this paper is to demonstrate that upscaling should be calculated during the flow simulation instead of trying to enhance the a priori upscaling methods. Hence, counter-examples are given to motivate our approach, the so-called Dual Mesh Method. The main steps of this numerical algorithm are recalled. Applications illustrate the necessity to consider different average relative permeability values depending on the direction in space. Moreover, these values could be different for the same average saturation. This proves that an a priori upscaling cannot be the answer even in homogeneous cases because of the {open_quotes}dynamical heterogeneity{close_quotes} created by the saturation profile. Other examples show the efficiency of the Dual Mesh Method applied to heterogeneous medium and to an actual field case in South America.
Variational mesh segmentation via quadric surface fitting
Yan, Dongming
2012-11-01
We present a new variational method for mesh segmentation by fitting quadric surfaces. Each component of the resulting segmentation is represented by a general quadric surface (including plane as a special case). A novel energy function is defined to evaluate the quality of the segmentation, which combines both L2 and L2 ,1 metrics from a triangle to a quadric surface. The Lloyd iteration is used to minimize the energy function, which repeatedly interleaves between mesh partition and quadric surface fitting. We also integrate feature-based and simplification-based techniques in the segmentation framework, which greatly improve the performance. The advantages of our algorithm are demonstrated by comparing with the state-of-the-art methods. © 2012 Elsevier Ltd. All rights reserved.
Variational mesh segmentation via quadric surface fitting
Yan, Dongming; Wang, Wen Ping; Liu, Yang; Yang, Zhouwang
2012-01-01
We present a new variational method for mesh segmentation by fitting quadric surfaces. Each component of the resulting segmentation is represented by a general quadric surface (including plane as a special case). A novel energy function is defined to evaluate the quality of the segmentation, which combines both L2 and L2 ,1 metrics from a triangle to a quadric surface. The Lloyd iteration is used to minimize the energy function, which repeatedly interleaves between mesh partition and quadric surface fitting. We also integrate feature-based and simplification-based techniques in the segmentation framework, which greatly improve the performance. The advantages of our algorithm are demonstrated by comparing with the state-of-the-art methods. © 2012 Elsevier Ltd. All rights reserved.
Meshed split skin graft for extensive vitiligo
Directory of Open Access Journals (Sweden)
Srinivas C
2004-05-01
Full Text Available A 30 year old female presented with generalized stable vitiligo involving large areas of the body. Since large areas were to be treated it was decided to do meshed split skin graft. A phototoxic blister over recipient site was induced by applying 8 MOP solution followed by exposure to UVA. The split skin graft was harvested from donor area by Padgett dermatome which was meshed by an ampligreffe to increase the size of the graft by 4 times. Significant pigmentation of the depigmented skin was seen after 5 months. This procedure helps to cover large recipient areas, when pigmented donor skin is limited with minimal risk of scarring. Phototoxic blister enables easy separation of epidermis thus saving time required for dermabrasion from recipient site.
Energy-efficient wireless mesh infrastructures
Al-Hazmi, Y.; de Meer, Hermann; Hummel, Karin Anna; Meyer, Harald; Meo, Michela; Remondo Bueno, David
2011-01-01
The Internet comprises access segments with wired and wireless technologies. In the future, we can expect wireless mesh infrastructures (WMIs) to proliferate in this context. Due to the relatively low energy efficiency of wireless transmission, as compared to wired transmission, energy consumption of WMIs can represent a significant part of the energy consumption of the Internet as a whole. We explore different approaches to reduce energy consumption in WMIs, taking into accoun...
MESHREF, Finite Elements Mesh Combination with Renumbering
International Nuclear Information System (INIS)
1973-01-01
1 - Nature of physical problem solved: The program can assemble different meshes stored on tape or cards. Renumbering is performed in order to keep band width low. Voids and/ or local refinement are possible. 2 - Method of solution: Topology and geometry are read according to input specifications. Abundant nodes and elements are eliminated. The new topology and geometry are stored on tape. 3 - Restrictions on the complexity of the problem: Maximum number of nodes = 2000. Maximum number of elements = 1500
Wireless experiments on a Motorola mesh testbed.
Energy Technology Data Exchange (ETDEWEB)
Riblett, Loren E., Jr.; Wiseman, James M.; Witzke, Edward L.
2010-06-01
Motomesh is a Motorola product that performs mesh networking at both the client and access point levels and allows broadband mobile data connections with or between clients moving at vehicular speeds. Sandia National aboratories has extensive experience with this product and its predecessors in infrastructure-less mobile environments. This report documents experiments, which characterize certain aspects of how the Motomesh network performs when obile units are added to a fixed network infrastructure.
Total sensitivity and uncertainty analysis for LWR pin-cells with improved UNICORN code
International Nuclear Information System (INIS)
Wan, Chenghui; Cao, Liangzhi; Wu, Hongchun; Shen, Wei
2017-01-01
Highlights: • A new model is established for the total sensitivity and uncertainty analysis. • The NR approximation applied in S&U analysis can be avoided by the new model. • Sensitivity and uncertainty analysis is performed to PWR pin-cells by the new model. • The effects of the NR approximation for the PWR pin-cells are quantified. - Abstract: In this paper, improvements to the multigroup cross-section perturbation model have been proposed and applied in the self-developed UNICORN code, which is capable of performing the total sensitivity and total uncertainty analysis for the neutron-physics calculations by applying the direct numerical perturbation method and the statistical sampling method respectively. The narrow resonance (NR) approximation was applied in the multigroup cross-section perturbation model, implemented in UNICORN. As improvements to the NR approximation to refine the multigroup cross-section perturbation model, an ultrafine-group cross-section perturbation model has been established, in which the actual perturbations are applied to the ultrafine-group cross-section library and the reconstructions of the resonance cross sections are performed by solving the neutron slowing-down equation. The total sensitivity and total uncertainty analysis were then applied to the LWR pin-cells, using both the multigroup and the ultrafine-group cross-section perturbation models. The numerical results show that the NR approximation overestimates the relative sensitivity coefficients and the corresponding uncertainty results for the LWR pin-cells, and the effects of the NR approximation are significant for σ_(_n_,_γ_) and σ_(_n_,_e_l_a_s_) of "2"3"8U. Therefore, the effects of the NR approximation applied in the total sensitivity and total uncertainty analysis for the neutron-physics calculations of LWR should be taken into account.
International Nuclear Information System (INIS)
Palau, J.M.; Cathalau, S.; Hudelot, J.P.; Barran, F.; Bellanger, V.; Magnaud, C.; Moreau, F.
2011-01-01
Burnable poisons are extensively used by Light Water Reactor designers in order to preserve the fuel reactivity potential and increase the cycle length (without increasing the uranium enrichment). In the industrial two-steps (assembly 2D transport-core 3D diffusion) calculation schemes these heterogeneities yield to strong flux and cross-sections perturbations that have to be taken into account in the final 3D burn-up calculations. This paper presents the application of an enhanced cross-section interpolation model (implemented in the French CRONOS2 code) to LWR (highly poisoned) depleted core calculations. The principle is to use the absorbers (or actinide) concentrations as the new interpolation parameters instead of the standard local burnup/fluence parameters. It is shown by comparing the standard (burnup/fluence) and new (concentration) interpolation models and using the lattice transport code APOLLO2 as a numerical reference that reactivity and local reaction rate prediction of a 2x2 LWR assembly configuration (slab geometry) is significantly improved with the concentration interpolation model. Gains on reactivity and local power predictions (resp. more than 1000 pcm and 20 % discrepancy reduction compared to the reference APOLLO2 scheme) are obtained by using this model. In particular, when epithermal absorbers are inserted close to thermal poison the 'shadowing' ('screening') spectral effects occurring during control operations are much more correctly modeled by concentration parameters. Through this outstanding example it is highlighted that attention has to be paid to the choice of cross-section interpolation parameters (burnup 'indicator') in core calculations with few energy groups and variable geometries all along the irradiation cycle. Actually, this new model could be advantageously applied to steady-state and transient LWR heterogeneous core computational analysis dealing with strong spectral-history variations under
A simplified geometrical model for transient corium propagation in core for LWR with heavy reflector
Directory of Open Access Journals (Sweden)
Saas Laurent
2017-01-01
Full Text Available In the context of the simulation of the Severe Accidents (SA in Light Water Reactors (LWR, we are interested on the in-core corium pool propagation transient in order to evaluate the corium relocation in the vessel lower head. The goal is to characterize the corium and debris flows from the core to accurately evaluate the corium pool propagation transient in the lower head and so the associated risk of vessel failure. In the case of LWR with heavy reflector, to evaluate the corium relocation into the lower head, we have to study the risk associated with focusing effect and the possibility to stabilize laterally the corium in core with a flooded down-comer. It is necessary to characterize the core degradation and the stratification of the corium pool that is formed in core. We assume that the core degradation until the corium pool formation and the corium pool propagation could be modeled separately. In this document, we present a simplified geometrical model (0D model for the in-core corium propagation transient. A degraded core with a formed corium pool is used as an initial state. This state can be obtained from a simulation computed with an integral code. This model does not use a grid for the core as integral codes do. Geometrical shapes and 0D models are associated with the corium pool and the other components of the degraded core (debris, heavy reflector, core plate…. During the transient, these shapes evolve taking into account the thermal and stratification behavior of the corium pool and the melting of the core surrounding components. Some results corresponding to the corium pool propagation in core transients obtained with this model on a LWR with a heavy reflector are given and compared to grid approach of the integral codes MAAP4.
Current situation of transvaginal mesh repair for pelvic organ prolapse.
Zhu, Lan; Zhang, Lei
2014-09-01
Surgical mesh is a metallic or polymeric screen intended to be implanted to reinforce soft tissue or bone where weakness exists. Surgical mesh has been used since the 1950s to repair abdominal hernias. In the 1970s, gynecologists began using surgical mesh products to indicate the repair of pelvic organ prolapse (POP), and in the 1990s, gynecologists began using surgical mesh for POP. Then the U.S. Food and Drug Administration (FDA) approved the first surgical mesh product specifically for use in POP. Surgical mesh materials can be divided into several categories. Most surgical mesh devices cleared for POP procedures are composed of non-absorbable synthetic polypropylene. Mesh can be placed in the anterior vaginal wall to aid in the correction of cystocele (anterior repair), in the posterior vaginal wall to aid in correction of rectocele (posterior repair), or attached to the top of the vagina to correct uterine prolapse or vaginal apical prolapse (apical repair). Over the past decades, surgical mesh products for transvaginal POP repair became incorporated into "kits" that included tools to aid in the delivery and insertion of the mesh. Surgical mesh kits continue to evolve, adding new insertion tools, tissue fixation anchors, surgical techniques, and ab- sorbable and biological materials. This procedure has been performed popularly. It was also performed increased in China. But this new technique met some trouble recently and let shake in urogynecology.
Topological patterns of mesh textures in serpentinites
Miyazawa, M.; Suzuki, A.; Shimizu, H.; Okamoto, A.; Hiraoka, Y.; Obayashi, I.; Tsuji, T.; Ito, T.
2017-12-01
Serpentinization is a hydration process that forms serpentine minerals and magnetite within the oceanic lithosphere. Microfractures crosscut these minerals during the reactions, and the structures look like mesh textures. It has been known that the patterns of microfractures and the system evolutions are affected by the hydration reaction and fluid transport in fractures and within matrices. This study aims at quantifying the topological patterns of the mesh textures and understanding possible conditions of fluid transport and reaction during serpentinization in the oceanic lithosphere. Two-dimensional simulation by the distinct element method (DEM) generates fracture patterns due to serpentinization. The microfracture patterns are evaluated by persistent homology, which measures features of connected components of a topological space and encodes multi-scale topological features in the persistence diagrams. The persistence diagrams of the different mesh textures are evaluated by principal component analysis to bring out the strong patterns of persistence diagrams. This approach help extract feature values of fracture patterns from high-dimensional and complex datasets.
Improved Mesh_Based Image Morphing
Directory of Open Access Journals (Sweden)
Mohammed Abdullah Taha
2017-11-01
Full Text Available Image morphing is a multi-step process that generates a sequence of transitions between two images. The thought is to get a ₔgrouping of middle pictures which, when ₔassembled with the first pictures would represent the change from one picture to the other. The process of morphing requires time and attention to detail in order to get good results. Morphing image requires at least two processes warping and cross dissolve. Warping is the process of geometric transformation of images. The cross dissolve is the process interpolation of color of eachₔ pixel from the first image value to theₔ corresponding second imageₔ value over the time. Image morphing techniques differ from in the approach of image warping procedure. This work presents a survey of different techniques to construct morphing images by review the different warping techniques. One of the predominant approaches of warping process is mesh warping which suffers from some problems including ghosting. This work proposed and implements an improved mesh warping technique to construct morphing images. The results show that the proposed approach can overcome the problems of the traditional mesh technique
Cu mesh for flexible transparent conductive electrodes.
Kim, Won-Kyung; Lee, Seunghun; Hee Lee, Duck; Hee Park, In; Seong Bae, Jong; Woo Lee, Tae; Kim, Ji-Young; Hun Park, Ji; Chan Cho, Yong; Ryong Cho, Chae; Jeong, Se-Young
2015-06-03
Copper electrodes with a micromesh/nanomesh structure were fabricated on a polyimide substrate using UV lithography and wet etching to produce flexible transparent conducting electrodes (TCEs). Well-defined mesh electrodes were realized through the use of high-quality Cu thin films. The films were fabricated using radio-frequency (RF) sputtering with a single-crystal Cu target--a simple but innovative approach that overcame the low oxidation resistance of ordinary Cu. Hybrid Cu mesh electrodes were fabricated by adding a capping layer of either ZnO or Al-doped ZnO. The sheet resistance and the transmittance of the electrode with an Al-doped ZnO capping layer were 6.197 ohm/sq and 90.657%, respectively, and the figure of merit was 60.502 × 10(-3)/ohm, which remained relatively unchanged after thermal annealing at 200 °C and 1,000 cycles of bending. This fabrication technique enables the mass production of large-area flexible TCEs, and the stability and high performance of Cu mesh hybrid electrodes in harsh environments suggests they have strong potential for application in smart displays and solar cells.
Numerical Investigation of Corrugated Wire Mesh Laminate
Directory of Open Access Journals (Sweden)
Jeongho Choi
2013-01-01
Full Text Available The aim of this work is to develop a numerical model of Corrugated Wire Mesh Laminate (CWML capturing all its complexities such as nonlinear material properties, nonlinear geometry and large deformation behaviour, and frictional behaviour. Development of such a model will facilitate numerical simulation of the mechanical behaviour of the wire mesh structure under various types of loading as well as the variation of the CWML configuration parameters to tailor its mechanical properties to suit the intended application. Starting with a single strand truss model consisting of four waves with a bilinear stress-strain model to represent the plastic behaviour of stainless steel, the finite element model is gradually built up to study single-layer structures with 18 strands of corrugated wire meshes consistency and double- and quadruple-layered laminates with alternating crossply orientations. The compressive behaviour of the CWML model is simulated using contact elements to model friction and is compared to the load-deflection behaviour determined experimentally in uniaxial compression tests. The numerical model of the CWML is then employed to conduct the aim of establishing the upper and lower bounds of stiffness and load capacity achievable by such structures.
MeSH Now: automatic MeSH indexing at PubMed scale via learning to rank.
Mao, Yuqing; Lu, Zhiyong
2017-04-17
MeSH indexing is the task of assigning relevant MeSH terms based on a manual reading of scholarly publications by human indexers. The task is highly important for improving literature retrieval and many other scientific investigations in biomedical research. Unfortunately, given its manual nature, the process of MeSH indexing is both time-consuming (new articles are not immediately indexed until 2 or 3 months later) and costly (approximately ten dollars per article). In response, automatic indexing by computers has been previously proposed and attempted but remains challenging. In order to advance the state of the art in automatic MeSH indexing, a community-wide shared task called BioASQ was recently organized. We propose MeSH Now, an integrated approach that first uses multiple strategies to generate a combined list of candidate MeSH terms for a target article. Through a novel learning-to-rank framework, MeSH Now then ranks the list of candidate terms based on their relevance to the target article. Finally, MeSH Now selects the highest-ranked MeSH terms via a post-processing module. We assessed MeSH Now on two separate benchmarking datasets using traditional precision, recall and F 1 -score metrics. In both evaluations, MeSH Now consistently achieved over 0.60 in F-score, ranging from 0.610 to 0.612. Furthermore, additional experiments show that MeSH Now can be optimized by parallel computing in order to process MEDLINE documents on a large scale. We conclude that MeSH Now is a robust approach with state-of-the-art performance for automatic MeSH indexing and that MeSH Now is capable of processing PubMed scale documents within a reasonable time frame. http://www.ncbi.nlm.nih.gov/CBBresearch/Lu/Demo/MeSHNow/ .
Characterization of coarse particulate matter in school gyms
Energy Technology Data Exchange (ETDEWEB)
Branis, Martin, E-mail: branis@natur.cuni.cz [Charles University in Prague, Faculty of Science, Institute for Environmental Studies, Prague (Czech Republic); Safranek, Jiri [Charles University in Prague, Faculty of Physical Education, Department of Outdoor Sports, Prague (Czech Republic)
2011-05-15
We investigated the mass concentration, mineral composition and morphology of particles resuspended by children during scheduled physical education in urban, suburban and rural elementary school gyms in Prague (Czech Republic). Cascade impactors were deployed to sample the particulate matter. Two fractions of coarse particulate matter (PM{sub 10-2.5} and PM{sub 2.5-1.0}) were characterized by gravimetry, energy dispersive X-ray spectrometry and scanning electron microscopy. Two indicators of human activity, the number of exercising children and the number of physical education hours, were also recorded. Lower mass concentrations of coarse particulate matter were recorded outdoors (average PM{sub 10-2.5} 4.1-7.4 {mu}g m{sup -3} and PM{sub 2.5-1.0} 2.0-3.3 {mu}g m{sup -3}) than indoors (average PM{sub 10-2.5} 13.6-26.7 {mu}g m{sup -3} and PM{sub 2.5-1.0} 3.7-7.4 {mu}g m{sup -3}). The indoor concentrations of coarse aerosol were elevated during days with scheduled physical education with an average indoor-outdoor (I/O) ratio of 2.5-16.3 for the PM{sub 10-2.5} and 1.4-4.8 for the PM{sub 2.5-1.0} values. Under extreme conditions, the I/O ratios reached 180 (PM{sub 10-2.5}) and 19.1 (PM{sub 2.5-1.0}). The multiple regression analysis based on the number of students and outdoor coarse PM as independent variables showed that the main predictor of the indoor coarse PM concentrations is the number of students in the gym. The effect of outdoor coarse PM was weak and inconsistent. The regression models for the three schools explained 60-70% of the particular dataset variability. X-ray spectrometry revealed 6 main groups of minerals contributing to resuspended indoor dust. The most abundant particles were those of crustal origin composed of Si, Al, O and Ca. Scanning electron microscopy showed that, in addition to numerous inorganic particles, various types of fibers and particularly skin scales make up the main part of the resuspended dust in the gyms. In conclusion, school
Characterization of coarse particulate matter in school gyms
International Nuclear Information System (INIS)
Branis, Martin; Safranek, Jiri
2011-01-01
We investigated the mass concentration, mineral composition and morphology of particles resuspended by children during scheduled physical education in urban, suburban and rural elementary school gyms in Prague (Czech Republic). Cascade impactors were deployed to sample the particulate matter. Two fractions of coarse particulate matter (PM 10-2.5 and PM 2.5-1.0 ) were characterized by gravimetry, energy dispersive X-ray spectrometry and scanning electron microscopy. Two indicators of human activity, the number of exercising children and the number of physical education hours, were also recorded. Lower mass concentrations of coarse particulate matter were recorded outdoors (average PM 10-2.5 4.1-7.4 μg m -3 and PM 2.5-1.0 2.0-3.3 μg m -3 ) than indoors (average PM 10-2.5 13.6-26.7 μg m -3 and PM 2.5-1.0 3.7-7.4 μg m -3 ). The indoor concentrations of coarse aerosol were elevated during days with scheduled physical education with an average indoor-outdoor (I/O) ratio of 2.5-16.3 for the PM 10-2.5 and 1.4-4.8 for the PM 2.5-1.0 values. Under extreme conditions, the I/O ratios reached 180 (PM 10-2.5 ) and 19.1 (PM 2.5-1.0 ). The multiple regression analysis based on the number of students and outdoor coarse PM as independent variables showed that the main predictor of the indoor coarse PM concentrations is the number of students in the gym. The effect of outdoor coarse PM was weak and inconsistent. The regression models for the three schools explained 60-70% of the particular dataset variability. X-ray spectrometry revealed 6 main groups of minerals contributing to resuspended indoor dust. The most abundant particles were those of crustal origin composed of Si, Al, O and Ca. Scanning electron microscopy showed that, in addition to numerous inorganic particles, various types of fibers and particularly skin scales make up the main part of the resuspended dust in the gyms. In conclusion, school gyms were found to be indoor microenvironments with high concentrations of
Data-Parallel Mesh Connected Components Labeling and Analysis
Energy Technology Data Exchange (ETDEWEB)
Harrison, Cyrus; Childs, Hank; Gaither, Kelly
2011-04-10
We present a data-parallel algorithm for identifying and labeling the connected sub-meshes within a domain-decomposed 3D mesh. The identification task is challenging in a distributed-memory parallel setting because connectivity is transitive and the cells composing each sub-mesh may span many or all processors. Our algorithm employs a multi-stage application of the Union-find algorithm and a spatial partitioning scheme to efficiently merge information across processors and produce a global labeling of connected sub-meshes. Marking each vertex with its corresponding sub-mesh label allows us to isolate mesh features based on topology, enabling new analysis capabilities. We briefly discuss two specific applications of the algorithm and present results from a weak scaling study. We demonstrate the algorithm at concurrency levels up to 2197 cores and analyze meshes containing up to 68 billion cells.
Evaluating the loss of a LWR spent fuel or plutonium shipping package into the sea
International Nuclear Information System (INIS)
Heaberlin, S.W.; Baker, D.A.
1976-06-01
As the nations of the world turn to nuclear power for an energy source, commerce in nuclear fuel cycle materials will increase. Some of this commerce will be transported by sea. Such shipments give rise to the possibility of loss of these materials into the sea. This paper discusses the postulated accidental loss of two materials, light water reactor (LWR) spent fuel and plutonium, at sea. The losses considered are that of a single shipping package which is either undamaged or damaged by fire prior to the loss. The containment failure of the package in the sea,
Safety-related investigations on power distribution in MOX fuel elements in LWR cores
International Nuclear Information System (INIS)
Kramer, E.; Langenbuch, S.
1991-01-01
For the concept of thermal recycling various fuel assembly designs have been developped during the last years. An overview is given describing the present status of MOX-fuel assembly design for PWR and BWR. The local power distribution within the MOX-fuel assembly and influences between neighbouring MOX- and Uranium fuel assemblies have been analyzed by own calculations. These investigations are limited to specific aspects of the spatial power distribution, which are related to the use of MOX-fuel assemblies within the reactor core of LWR. (orig.) [de
A Stochastic LWR Model with Consideration of the Driver's Individual Property
International Nuclear Information System (INIS)
Tang Tieqiao; Wang Yunpeng; Yu Guizhen; Huang Haijun
2012-01-01
In this paper, we develop a stochastic LWR model based on the influences of the driver's individual property on his/her perceived density and speed deviation. The numerical results show that the driver's individual property has great effects on traffic flow only when the initial density is moderate, i.e., at this time, oscillating traffic flow will occur and the oscillating phenomena in the traffic system consisting of the conservative and aggressive drivers is more serious than that in the traffic system consisting of the conservative (aggressive) drivers.
Residual life assessment of major LWR components: NPAR approach and results
International Nuclear Information System (INIS)
Shah, V.N.; Weidenhamer, G.H.; Vora, J.P.
1991-01-01
The nuclear plant aging research (NPAR) program is systematically addressing the technical issues associated with understanding and managing aging of major LWR components. Twenty-one major components have been identified and prioritized according to their relevance to plant safety. Qualitative aging assessment has identified pertinent design features, materials, stressors, environments, aging mechanisms. and failure modes for each of the components. Emerging inspection, surveillance, and monitoring methods to characterize aging damage and mitigation methods to reduce the damage are currently being assessed. The results of all these assessments are used to develop life-assessment procedures for the components and are included in appropriate documents supporting the regulatory requirements for license renewal. (author)
Prioritization of tasks in the draft LWR safety technology program plan. Final report
International Nuclear Information System (INIS)
Lim, E.Y.; Miller, W.J.; Parkinson, W.J.; Ritzman, R.L.; vonHerrmann, J.L.; Wood, P.J.
1980-05-01
The purpose of this report is to describe both the approach taken and the results produced in the SAI effort to prioritize the tasks in the Sandia draft LWR Safety Technology Program Plan. This work used the description of important safety issues developed in the Reactor Safety Study (2) to quantify the effect of safety improvements resulting from a research and development program on the risk from nuclear power plants. Costs of implementation of these safety improvements were also estimated to allow a presentation of the final results in a value (i.e., risk reduction) vs. impact (i.e., implementation costs) matrix
Characterization and chemistry of fission products released from LWR fuel under accident conditions
International Nuclear Information System (INIS)
Norwood, K.S.; Collins, J.L.; Osborne, M.F.; Lorenz, R.A.; Wichner, R.P.
1984-01-01
Segments from commercial LWR fuel rods have been tested at temperatures between 1400 and 2000 0 C in a flowing steam-helium atmosphere to simulate severe accident conditions. The primary goals of the tests were to determine the rate of fission product release and to characterize the chemical behavior. This paper is concerned primarily with the identification and chemical behavior of the released fission products with emphasis on antimony, cesium, iodine, and silver. The iodine appeared to behave primarily as cesium iodide and the antimony and silver as elements, while cesium behavior was much more complex. 17 refs., 7 figs., 1 tab
Review of tellurium release rates from LWR fuel elements under accident conditions
International Nuclear Information System (INIS)
Lorenz, R.A.; Beahm, E.C.; Wichner, R.P.
1983-01-01
Although fission product tellurium presents a potentially significant radiohazard, its release and transport in source-term experiments is frequently overlooked because it does not possess a readily measurable, gamma emission; moreover, a recent study emphasized noble gas, iodine and cesium release from LWR fuel elements because of the large data base that exists for these materials. Some new tests show that in some cases tellurium may be held up in core material to a greater degree than previously assumed - an observation that prompts a careful reappraisal of the existing tellurium-release data and its chemical foundation
International Nuclear Information System (INIS)
Tri-Yulianto
1996-01-01
Based on TOR BATAN for PELITA VI. On of BATAN program in the fuel element production technology section is the acquisition of the fuel element fabrication technology for research reactor as well as power reactor. The acquisition can be achieved using different strategies, e.g. by utilizing the facility owned for research and development of the technology desired or by transferring the technology directly from the source. With regards to the above, PEBN through its facility in BEBE has started the acquisition of the fuel element fabrication technology for power reactor by developing the existing equipment initially designed to fabricate HWR Cinere fuel element. The development, by way of modifying the equipment, is intended for the production of HWR (Candu) and LWR (PWR and BWR) fuel elements. To achieve above objective, at the early stage of activity, an assesment on the fabrication equipment for pelletizing, component production and assembly. The assesment was made by comparing the shape and the size of the existing fuel element with those used in the operating reactors such as Candu reactors, PWR and BWR. Equipment having the potential to be modified for the production of HWR fuel elements are as followed: For the pelletizing equipment, the punch and dies can be used of the pressing machine for making green pellet can be modified so that different sizes of punch and dies can be used, depending upon the size of the HWR and LWR pellets. The equipment for component production has good potential for modification to produce the HWR Candu fuel element, which has similar shape and size with those of the existing fuel element, while the possibility of producing the LWR fuel element component is small because only a limited number of the required component can be made with the existing equipment. The assembly equipment has similar situation whit that of the component production, that is, to assemble the HWR fuel element modification of few assembly units very probable
Perspective on US NRC Policy Issues Concerning Use of Risk Insights for Non-LWR
International Nuclear Information System (INIS)
Ha, Jun Su; Kim, In Goo; Huh, Chang Wook; Kim, Kyun Tae
2011-01-01
Since the PRA Implementation plan of US NRC (1994), PRA has been applied to all NPPs in USA and risk insights have been used for the regulation as a complement of the deterministic approaches. RIRIP (Risk-Informed Regulation Implementation Plan, 2000) and RPP (Risk-Informed and Performance-Based Plan, 2007) were announced by US NRC thereafter, which recommended enhanced use of risk insights. In the meantime, there have been lots of policy issues concerning use of risk insights for licensing Non-LWR designs, which will be discussed in this paper to understand the stream of perspectives on US NRC's approach
Validation of the LWR-EIR methods for the evaluation of compact beds
International Nuclear Information System (INIS)
Foskolos, K.; Grimm, P.; Maeder, C.; Paratte, J.M.
1983-10-01
The EIR code system for the calculation of light water reactors is presented and the methods used are briefly described. The application of the system on various types of critical experiments and benchmark problems proves its good precision, even for heterogeneous configurations with strong neutron absorbers like Boral. As the accuracy of the multiplication factor ksub(eff) is always better than 0.5% for normal LWR configurations, this code system is validated for the calculation of such configurations with a safety margin of 1.5% on ksub(eff). (Auth.)
Risk management: integration of social and technical risk variables into safety assessments of LWR'S
International Nuclear Information System (INIS)
Turnage, J.J.; Husseiny, A.A.
1980-01-01
A risk management methodology is developed here to formalize the acceptability levels of commercial LWR power plants via the estimation of risk levels acceptable to the public and the integration of such estimates into risk-benefit analysis. Utility theory is used for developing preference models based on value trade-offs among multiple objectives and uncertainties about the impact of alternatives. The method involves reducing the various variables affecting safety acceptability decisions to a single function that provides a metric for acceptability levels. The function accomondates for technical criteria related to design and licensing decisions, as well as public reactions to certain choices
Advantages of retrofitting high velocity separators to LWR turbines; experience in VVR NPP Loviisa
International Nuclear Information System (INIS)
Dueymes, E.; Peyrelongue, J.P.
1992-01-01
Erosion-corrosion by wet steam is a concern for VVER operators and also, in numerous LWR power plants of western technology. The backfitting of moisture separators at the HP Turbine outlets is a way to avoid maintenance costs, repairs, replacement of pipes or equipments. Installation of HVS at LOVIISA confirms that this device, whose installation work is reduced to a minimum, is able to remove quite all the water from the steam just a few meters downstream the HP cylinder. A long term operation can be expected for carbon steel equipments, even those previously damaged by erosion-corrosion. (authors). 6 figs., 2 tabs
Uncertainties in criticality analysis which affect the storage and transportation of LWR fuel
International Nuclear Information System (INIS)
Napolitani, D.G.
1989-01-01
Satisfying the design criteria for subcriticality with uncertainties affects: the capacity of LWR storage arrays, maximum allowable enrichment, minimum allowable burnup and economics of various storage options. There are uncertainties due to: calculational method, data libraries, geometric limitations, modelling bias, the number and quality of benchmarks performed and mechanical uncertainties in the array. Yankee Atomic Electric Co. (YAEC) has developed and benchmarked methods to handle: high density storage rack designs, pin consolidation, low density moderation and burnup credit. The uncertainties associated with such criticality analysis are quantified on the basis of clean criticals, power reactor criticals and intercomparison of independent analysis methods
Oral, intestinal, and skin bacteria in ventral hernia mesh implants
Directory of Open Access Journals (Sweden)
Odd Langbach
2016-07-01
Full Text Available Background: In ventral hernia surgery, mesh implants are used to reduce recurrence. Infection after mesh implantation can be a problem and rates around 6–10% have been reported. Bacterial colonization of mesh implants in patients without clinical signs of infection has not been thoroughly investigated. Molecular techniques have proven effective in demonstrating bacterial diversity in various environments and are able to identify bacteria on a gene-specific level. Objective: The purpose of this study was to detect bacterial biofilm in mesh implants, analyze its bacterial diversity, and look for possible resemblance with bacterial biofilm from the periodontal pocket. Methods: Thirty patients referred to our hospital for recurrence after former ventral hernia mesh repair, were examined for periodontitis in advance of new surgical hernia repair. Oral examination included periapical radiographs, periodontal probing, and subgingival plaque collection. A piece of mesh (1×1 cm from the abdominal wall was harvested during the new surgical hernia repair and analyzed for bacteria by PCR and 16S rRNA gene sequencing. From patients with positive PCR mesh samples, subgingival plaque samples were analyzed with the same techniques. Results: A great variety of taxa were detected in 20 (66.7% mesh samples, including typical oral commensals and periodontopathogens, enterics, and skin bacteria. Mesh and periodontal bacteria were further analyzed for similarity in 16S rRNA gene sequences. In 17 sequences, the level of resemblance between mesh and subgingival bacterial colonization was 98–100% suggesting, but not proving, a transfer of oral bacteria to the mesh. Conclusion: The results show great bacterial diversity on mesh implants from the anterior abdominal wall including oral commensals and periodontopathogens. Mesh can be reached by bacteria in several ways including hematogenous spread from an oral site. However, other sites such as gut and skin may also
Short Communication on "In-situ TEM ion irradiation investigations on U3Si2 at LWR temperatures"
Miao, Yinbin; Harp, Jason; Mo, Kun; Bhattacharya, Sumit; Baldo, Peter; Yacout, Abdellatif M.
2017-02-01
The radiation-induced amorphization of U3Si2 was investigated by in-situ transmission electron microscopy using 1 MeV Kr ion irradiation. Both arc-melted and sintered U3Si2 specimens were irradiated at room temperature to confirm the similarity in their responses to radiation. The sintered specimens were then irradiated at 350 °C and 550 °C up to 7.2 × 1015 ions/cm2 to examine their amorphization behavior under light water reactor (LWR) conditions. U3Si2 remains crystalline under irradiation at LWR temperatures. Oxidation of the material was observed at high irradiation doses.
International Nuclear Information System (INIS)
Raju, P.P.
1980-05-01
This report summarizes the results of the study program to assess the benefits of nonlinear analysis methods in Light Water Reactor (LWR) component designs. The current study reveals that despite its increased cost and other complexities, nonlinear analysis is a practical and valuable tool for the design of LWR components, especially under ASME Level D service conditions (faulted conditions) and it will greatly assist in the evaluation of ductile fracture potential of pressure boundary components. Since the nonlinear behavior is generally a local phenomenon, the design of complex components can be accomplished through substructuring isolated localized regions and evaluating them in detail using nonlinear analysis methods
LWR mox fuel experience in Belgium and France with special emphasis on results obtained in BR3
International Nuclear Information System (INIS)
Bairiot, H.; Haas, D.; Lippens, M.; Motte, F.; Lebastard, G.; Marin, J.F.
1986-09-01
The course of the paper reflects two main topics: LWR MOX fuel experience in Belgium and France, summarizing the fabrication techniques, the references, the underlying MOX fuel technology and the current R and D programs for expanding the data base; behaviour of MOX fuel rods irradiated under steady state and transient operating conditions, focusing on MOX fuel technology features acquired through the irradiations performed in the BR3 PWR, supplemented by tests in the BR2 MTR. This paper focuses on the thermomechanical behaviour of LWR MOX fuel rods, which is intimately related to the fabrication technique and vice-versa. 22 refs
Coarse woody debris metrics in a California oak woodland
William D. Tietje; Michael A. Hardy; Christopher C. Yim
2015-01-01
Little information is available on the metrics of coarse woody debris (CWD) in California oak woodland, most notably at the scale of the stand and woodland type. In a remote part of the National Guard Post, Camp Roberts, that has not burned in over a half century, we tallied 314 pieces of CWD in a blue oak (Quercus douglasii)-coast live oak (
Free-energy coarse-grained potential for C60
International Nuclear Information System (INIS)
Edmunds, D. M.; Tangney, P.; Vvedensky, D. D.; Foulkes, W. M. C.
2015-01-01
We propose a new deformable free energy method for generating a free-energy coarse-graining potential for C 60 . Potentials generated from this approach exhibit a strong temperature dependence and produce excellent agreement with benchmark fully atomistic molecular dynamics simulations. Parameter sets for analytical fits to this potential are provided at four different temperatures
Coarse particles-water mixtures flow in pipes
Czech Academy of Sciences Publication Activity Database
Vlasák, Pavel
2017-01-01
Roč. 225, č. 2017 (2017), s. 338-341 ISSN 2411-3336 R&D Projects: GA ČR GAP105/10/1574 Institutional support: RVO:67985874 Keywords : hydrotransport * coarse particles pipeline installation * pressure drop * pipe inclination Subject RIV: BK - Fluid Dynamics OBOR OECD: Fluids and plasma physics (including surface physics)
Coarse graining of atactic polystyrene and its derivatives
Agrawal, Anupriya; Perahia, Dvora; Grest, Gary S.
2014-03-01
Capturing large length scales in polymers and soft matter while retaining atomistic properties is imperative to computational studies of dynamic systems. Here we present a new methodology developing coarse-grain model based on atomistic simulation of atactic polystyrene (PS). Similar to previous work by Fritz et al., each monomer is described by two coarse grained beads. In contrast to this earlier work where intramolecular potentials were based on Monte Carlo simulation of both isotactic and syndiotactic single PS molecule to capture stereochemistry, we obtained intramolecular interactions from a single molecular dynamics simulation of an all-atom atactic PS melts. The non-bonded interactions are obtained using the iterative Boltzmann inversion (IBI) scheme. This methodology has been extended to coarse graining of poly-(t-butyl-styrene) (PtBS). An additional coarse-grained bead is used to describe the t-butyl group. Similar to the process for PS, the intramolecular interactions are obtained from a single all atom atactic melt simulation. Starting from the non-bonded interactions for PS, we show that the IBI method for the non-bonded interactions of PtBS converges relatively fast. A generalized scheme for substituted PS is currently in development. We would like to acknowledge Prof. Kurt Kremer for helpful discussions during this work.
Coarse-graining stochastic biochemical networks: adiabaticity and fast simulations
Energy Technology Data Exchange (ETDEWEB)
Nemenman, Ilya [Los Alamos National Laboratory; Sinitsyn, Nikolai [Los Alamos National Laboratory; Hengartner, Nick [Los Alamos National Laboratory
2008-01-01
We propose a universal approach for analysis and fast simulations of stiff stochastic biochemical kinetics networks, which rests on elimination of fast chemical species without a loss of information about mesoscoplc, non-Poissonian fluctuations of the slow ones. Our approach, which is similar to the Born-Oppenhelmer approximation in quantum mechanics, follows from the stochastic path Integral representation of the cumulant generating function of reaction events. In applications with a small number of chemIcal reactions, It produces analytical expressions for cumulants of chemical fluxes between the slow variables. This allows for a low-dimensional, Interpretable representation and can be used for coarse-grained numerical simulation schemes with a small computational complexity and yet high accuracy. As an example, we derive the coarse-grained description for a chain of biochemical reactions, and show that the coarse-grained and the microscopic simulations are in an agreement, but the coarse-gralned simulations are three orders of magnitude faster.
Variational approach to coarse-graining of generalized gradient flows
Duong, M.H.; Lamacz, A.; Peletier, M.A.; Sharma, U.
2017-01-01
In this paper we present a variational technique that handles coarse-graining and passing to a limit in a unified manner. The technique is based on a duality structure, which is present in many gradient flows and other variational evolutions, and which often arises from a large-deviations principle.
Managing coarse woody debris in forests of the Rocky Mountains
Russell T. Graham; Alan E. Harvey; Martin F. Jurgensen; Theresa B. Jain; Jonalea R. Tonn; Deborah S. Page-Dumroese
1994-01-01
Recommendations for managing coarse woody debris after timber harvest were developed for 14 habitat types, ranging from ponderosa pine (Pinus ponderosa) habitat types of Arizona to subalpine fir (Abies lasiocarpa) habitat types of western Montana. Ectomycorrhizae were used as a bioindicator of healthy, productive forest soils....
Terrain aided navigation for autonomous underwater vehicles with coarse maps
International Nuclear Information System (INIS)
Zhou, Ling; Cheng, Xianghong; Zhu, Yixian
2016-01-01
Terrain aided navigation (TAN) is a form of geophysical localization technique for autonomous underwater vehicles (AUVs) operating in GPS-denied environments. TAN performance on sensor-rich AUVs has been evaluated in sea trials. However, many challenges remain before TAN can be successfully implemented on sensor-limited AUVs, especially with coarse maps. To improve TAN performance over coarse maps, a Gaussian process (GP) is proposed for the modeling of bathymetric terrain and integrated into the particle filter (GP-PF). GP is applied to provide not only the bathymetric value prediction through learning a set of bathymetric data from coarse maps but also the variance of the prediction. As a measurement update, calculated on bathymetric deviation is performed through the PF to obtain absolute and bounded positioning accuracy. Through the analysis of TAN performance on experimental data for two different terrains with map resolutions of 10–50 m, both the ability of the proposed model to represent the actual bathymetric terrain with accuracy and the effect of the GP-PF for TAN on sensor-limited systems in suited terrain are demonstrated. The experiment results further verify that there is an inverse relationship between the coarseness of the map and the overall TAN accuracy in rough terrains, but there is hardly any relationship between them in relatively flat terrains. (paper)
Loop overhead reduction techniques for coarse grained reconfigurable architectures
Vadivel, K.; Wijtvliet, M.; Jordans, R.; Corporaal, H.
2017-01-01
Due to their flexibility and high performance, Coarse Grained Reconfigurable Array (CGRA) are a topic of increasing research interest. However, CGRAs also have the potential to achieve very high energy efficiency in comparison to other reconfigurable architectures when hardware optimizations are
Time evolution as refining, coarse graining and entangling
International Nuclear Information System (INIS)
Dittrich, Bianca; Steinhaus, Sebastian
2014-01-01
We argue that refining, coarse graining and entangling operators can be obtained from time evolution operators. This applies in particular to geometric theories, such as spin foams. We point out that this provides a construction principle for the physical vacuum in quantum gravity theories and more generally allows construction of a (cylindrically) consistent continuum limit of the theory. (paper)
Time evolution as refining, coarse graining and entangling
Dittrich, Bianca; Steinhaus, Sebastian
2014-12-01
We argue that refining, coarse graining and entangling operators can be obtained from time evolution operators. This applies in particular to geometric theories, such as spin foams. We point out that this provides a construction principle for the physical vacuum in quantum gravity theories and more generally allows construction of a (cylindrically) consistent continuum limit of the theory.
A distance limited method for sampling downed coarse woody debris
Jeffrey H. Gove; Mark J. Ducey; Harry T. Valentine; Michael S. Williams
2012-01-01
A new sampling method for down coarse woody debris is proposed based on limiting the perpendicular distance from individual pieces to a randomly chosen sample point. Two approaches are presented that allow different protocols to be used to determine field measurements; estimators for each protocol are also developed. Both protocols are compared via simulation against...
Mineral Elements Content of some Coarse Grains used as staple ...
African Journals Online (AJOL)
Analysis of mineral elements were carried out on some coarse grains used as staple food in Kano metropolis. The levels of Magnesium, Calcium, Manganese, Iron, Copper and Zinc were determined using atomic absorption spectrophotometer (AAS), and that of Sodium and Potassium were obtained using flame photometer ...
Non-Steady Oscillatory Flow in Coarse Granular Materials
DEFF Research Database (Denmark)
Andersen, O. H.; Gent, M. R. A. van; Meer, J. W. van der
1992-01-01
Stationary and oscillatory flow through coarse granular materials have been investigated experimentally at Delft Hydraulics in their oscillating water tunnel with the objective of determining the coefficients of the extended Forchheimer equation. Cylinders, spheres and different types of rock have....... Further, for the non-stationary term, the virtual mass coefficient will be derived....
The MARTINI force field : Coarse grained model for biomolecular simulations
Marrink, Siewert J.; Risselada, H. Jelger; Yefimov, Serge; Tieleman, D. Peter; de Vries, Alex H.
2007-01-01
We present an improved and extended version of our coarse grained lipid model. The new version, coined the MARTINI force field, is parametrized in a systematic way, based on the reproduction of partitioning free energies between polar and apolar phases of a large number of chemical compounds. To
Martini Coarse-Grained Force Field : Extension to RNA
Uusitalo, Jaakko J.; Ingolfsson, Helgi I.; Marrink, Siewert J.; Faustino, Ignacio
2017-01-01
RNA has an important role not only as the messenger of genetic information but also as a regulator of gene expression. Given its central role in cell biology, there is significant interest in studying the structural and dynamic behavior of RNA in relation to other biomolecules. Coarse-grain
Martini Coarse-Grained Force Field : Extension to Carbohydrates
Lopez, Cesar A.; Rzepiela, Andrzej J.; de Vries, Alex H.; Dijkhuizen, Lubbert; Huenenberger, Philippe H.; Marrink, Siewert J.
2009-01-01
We present an extension of the Martini coarse-grained force field to carbohydrates. The parametrization follows the same philosophy as was used previously for lipids and proteins, focusing on the reproduction of partitioning free energies of small compounds between polar and nonpolar phases. The
Coarse-grain modelling of protein-protein interactions
Baaden, Marc; Marrink, Siewert J.
2013-01-01
Here, we review recent advances towards the modelling of protein-protein interactions (PPI) at the coarse-grained (CG) level, a technique that is now widely used to understand protein affinity, aggregation and self-assembly behaviour. PPI models of soluble proteins and membrane proteins are