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Sample records for nodal neutron kinetics

  1. Nodal approximations in space and time for neutron kinetics

    International Nuclear Information System (INIS)

    Grossman, L.M.; Hennart, J.P.

    2005-01-01

    A general formalism is described of the nodal type in time and space for the neutron kinetics equations. In space, several nodal methods are given of the Raviart-Thomas type (RT0 and RT1), of the Brezzi-Douglas-Marini type (BDM0 and BDM1) and of the Brezzi-Douglas-Fortin-Marini type (BDFM 1). In time, polynomial and analytical approximations are derived. In the analytical case, they are based on the inclusion of an exponential term in the basis function. They can be continuous or discontinuous in time, leading in particular to the well-known Crank-Nicolson, Backward Euler and θ schemes

  2. A three-dimensional nodal neutron kinetics capability for relaps

    International Nuclear Information System (INIS)

    Judd, J.L.; Weaver, W.L.

    1996-01-01

    The incorporation of a three-dimensional neutron kinetics capability into the DOE version of the RELAP5/MOD3.2 reactor safety code is discussed. A brief discussion of the kinetics method is given along with a discussion of the cross section parameterization models available in RELAP5/MOD3.2. The RELAP5/MOD3.2 code is then used to perform calculations of the NEACRP rod ejection and rod withdrawal benchmarks, and results are presented

  3. Improvement of neutron kinetics module in TRAC-BF1code: one-dimensional nodal collocation method

    Energy Technology Data Exchange (ETDEWEB)

    Jambrina, Ana; Barrachina, Teresa; Miro, Rafael; Verdu, Gumersindo, E-mail: ajambrina@iqn.upv.es, E-mail: tbarrachina@iqn.upv.es, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidade Politecnica de Valencia (UPV), Valencia (Spain); Soler, Amparo, E-mail: asoler@iberdrola.es [SEA Propulsion S.L., Madrid (Spain); Concejal, Alberto, E-mail: acbe@iberdrola.es [Iberdrola Ingenieria y Construcion S.A.U., Madrid (Spain)

    2013-07-01

    The TRAC-BF1 one-dimensional kinetic model is a formulation of the neutron diffusion equation in the two energy groups' approximation, based on the analytical nodal method (ANM). The advantage compared with a zero-dimensional kinetic model is that the axial power profile may vary with time due to thermal-hydraulic parameter changes and/or actions of the control systems but at has the disadvantages that in unusual situations it fails to converge. The nodal collocation method developed for the neutron diffusion equation and applied to the kinetics resolution of TRAC-BF1 thermal-hydraulics, is an adaptation of the traditional collocation methods for the discretization of partial differential equations, based on the development of the solution as a linear combination of analytical functions. It has chosen to use a nodal collocation method based on a development of Legendre polynomials of neutron fluxes in each cell. The qualification is carried out by the analysis of the turbine trip transient from the NEA benchmark in Peach Bottom NPP using both the original 1D kinetics implemented in TRAC-BF1 and the 1D nodal collocation method. (author)

  4. NESTLE: A nodal kinetics code

    International Nuclear Information System (INIS)

    Al-Chalabi, R.M.; Turinsky, P.J.; Faure, F.-X.; Sarsour, H.N.; Engrand, P.R.

    1993-01-01

    The NESTLE nodal kinetics code has been developed for utilization as a stand-alone code for steady-state and transient reactor neutronic analysis and for incorporation into system transient codes, such as TRAC and RELAP. The latter is desirable to increase the simulation fidelity over that obtained from currently employed zero- and one-dimensional neutronic models and now feasible due to advances in computer performance and efficiency of nodal methods. As a stand-alone code, requirements are that it operate on a range of computing platforms from memory-limited personal computers (PCs) to supercomputers with vector processors. This paper summarizes the features of NESTLE that reflect the utilization and requirements just noted

  5. One-dimensional nodal neutronics routines for the TRAC-BD1 thermal-hydraulics program

    International Nuclear Information System (INIS)

    Nigg, D.W.

    1983-09-01

    Nuclear reactor core transient neutronic behavior is currently modeled in the TRAC-BD1 code using a point-reactor kinetics formulation. This report describes a set of subroutines based on the Analytic Nodal Method that were written to provide TRAC-BD1 with a one-dimensional space-dependent neutronics capability. Use of the routines is illustrated with several test problems. The results of these problems show that the Analytic Nodal neutronics routines have desirable accuracy and computing time characteristics and should be a useful addition to TRAC-BD1

  6. Fundamentals of 3-D Neutron Kinetics and Current Status

    International Nuclear Information System (INIS)

    Aragones, J.M.

    2008-01-01

    This lecture includes the following topics: 1) A summary of the cell and lattice calculations used to generate the neutron reaction data for neutron kinetics, including the spectral and burnup calculations of LWR cells and fuel assembly lattices, and the main nodal kinetics parameters: mean neutron generation time and delayed neutron fraction; 2) the features of the advanced nodal methods for 3-D LWR core physics, including the treatment of partially inserted control rods, fuel assembly grids, fuel burnup and xenon and samarium transients, and excore detector responses, that are essential for core surveillance, axial offset control and operating transient analysis; 3) the advanced nodal methods for 3-D LWR core neutron kinetics (best estimate safety analysis, real-time simulation); and 4) example applications to 3-D neutron kinetics problems in transient analysis of PWR cores, including model, benchmark and operational transients without, or with simple, thermal-hydraulics feedback.

  7. Discontinuous nodal schemes applied to the bidimensional neutron transport equation

    International Nuclear Information System (INIS)

    Delfin L, A.; Valle G, E. Del; Hennart B, J.P.

    1996-01-01

    In this paper several strong discontinuous nodal schemes are described, starting from the one that has only two interpolation parameters per cell to the one having ten. Their application to the spatial discretization of the neutron transport equation in X-Y geometry is also described, giving, for each one of the nodal schemes, the approximation for the angular neutron flux that includes the set of interpolation parameters and the corresponding polynomial space. Numerical results were obtained for several test problems presenting here the problem with the highest degree of difficulty and their comparison with published results 1,2 . (Author)

  8. Super-nodal methods for space-time kinetics

    Science.gov (United States)

    Mertyurek, Ugur

    The purpose of this research has been to develop an advanced Super-Nodal method to reduce the run time of 3-D core neutronics models, such as in the NESTLE reactor core simulator and FORMOSA nuclear fuel management optimization codes. Computational performance of the neutronics model is increased by reducing the number of spatial nodes used in the core modeling. However, as the number of spatial nodes decreases, the error in the solution increases. The Super-Nodal method reduces the error associated with the use of coarse nodes in the analyses by providing a new set of cross sections and ADFs (Assembly Discontinuity Factors) for the new nodalization. These so called homogenization parameters are obtained by employing consistent collapsing technique. During this research a new type of singularity, namely "fundamental mode singularity", is addressed in the ANM (Analytical Nodal Method) solution. The "Coordinate Shifting" approach is developed as a method to address this singularity. Also, the "Buckling Shifting" approach is developed as an alternative and more accurate method to address the zero buckling singularity, which is a more common and well known singularity problem in the ANM solution. In the course of addressing the treatment of these singularities, an effort was made to provide better and more robust results from the Super-Nodal method by developing several new methods for determining the transverse leakage and collapsed diffusion coefficient, which generally are the two main approximations in the ANM methodology. Unfortunately, the proposed new transverse leakage and diffusion coefficient approximations failed to provide a consistent improvement to the current methodology. However, improvement in the Super-Nodal solution is achieved by updating the homogenization parameters at several time points during a transient. The update is achieved by employing a refinement technique similar to pin-power reconstruction. A simple error analysis based on the relative

  9. Nodal methods for problems in fluid mechanics and neutron transport

    International Nuclear Information System (INIS)

    Azmy, Y.Y.

    1985-01-01

    A new high-accuracy, coarse-mesh, nodal integral approach is developed for the efficient numerical solution of linear partial differential equations. It is shown that various special cases of this general nodal integral approach correspond to several high efficiency nodal methods developed recently for the numerical solution of neutron diffusion and neutron transport problems. The new approach is extended to the nonlinear Navier-Stokes equations of fluid mechanics; its extension to these equations leads to a new computational method, the nodal integral method which is implemented for the numerical solution of these equations. Application to several test problems demonstrates the superior computational efficiency of this new method over previously developed methods. The solutions obtained for several driven cavity problems are compared with the available experimental data and are shown to be in very good agreement with experiment. Additional comparisons also show that the coarse-mesh, nodal integral method results agree very well with the results of definitive ultra-fine-mesh, finite-difference calculations for the driven cavity problem up to fairly high Reynolds numbers

  10. A comparison of Nodal methods in neutron diffusion calculations

    Energy Technology Data Exchange (ETDEWEB)

    Tavron, Barak [Israel Electric Company, Haifa (Israel) Nuclear Engineering Dept. Research and Development Div.

    1996-12-01

    The nuclear engineering department at IEC uses in the reactor analysis three neutron diffusion codes based on nodal methods. The codes, GNOMERl, ADMARC2 and NOXER3 solve the neutron diffusion equation to obtain flux and power distributions in the core. The resulting flux distributions are used for the furl cycle analysis and for fuel reload optimization. This work presents a comparison of the various nodal methods employed in the above codes. Nodal methods (also called Coarse-mesh methods) have been designed to solve problems that contain relatively coarse areas of homogeneous composition. In the nodal method parts of the equation that present the state in the homogeneous area are solved analytically while, according to various assumptions and continuity requirements, a general solution is sought out. Thus efficiency of the method for this kind of problems, is very high compared with the finite element and finite difference methods. On the other hand, using this method one can get only approximate information about the node vicinity (or coarse-mesh area, usually a feel assembly of a 20 cm size). These characteristics of the nodal method make it suitable for feel cycle analysis and reload optimization. This analysis requires many subsequent calculations of the flux and power distributions for the feel assemblies while there is no need for detailed distribution within the assembly. For obtaining detailed distribution within the assembly methods of power reconstruction may be applied. However homogenization of feel assembly properties, required for the nodal method, may cause difficulties when applied to fuel assemblies with many absorber rods, due to exciting strong neutron properties heterogeneity within the assembly. (author).

  11. Nodal spectrum method for solving neutron diffusion equation

    International Nuclear Information System (INIS)

    Sanchez, D.; Garcia, C. R.; Barros, R. C. de; Milian, D.E.

    1999-01-01

    Presented here is a new numerical nodal method for solving static multidimensional neutron diffusion equation in rectangular geometry. Our method is based on a spectral analysis of the nodal diffusion equations. These equations are obtained by integrating the diffusion equation in X, Y directions and then considering flat approximations for the current. These flat approximations are the only approximations that are considered in this method, as a result the numerical solutions are completely free from truncation errors. We show numerical results to illustrate the methods accuracy for coarse mesh calculations

  12. Numerical nodal simulation of the axial power distribution within nuclear reactors using a kinetics diffusion model. I

    International Nuclear Information System (INIS)

    Barros, R.C. de.

    1992-05-01

    Presented here is a new numerical nodal method for the simulation of the axial power distribution within nuclear reactors using the one-dimensional one speed kinetics diffusion model with one group of delayed neutron precursors. Our method is based on a spectral analysis of the nodal kinetics equations. These equations are obtained by integrating the original kinetics equations separately over a time step and over a spatial node, and then considering flat approximations for the forward difference terms. These flat approximations are the only approximations that are considered in the method. As a result, the spectral nodal method for space - time reactor kinetics generates numerical solutions for space independent problems or for time independent problems that are completely free from truncation errors. We show numerical results to illustrate the method's accuracy for coarse mesh calculations. (author)

  13. Solution and study of nodal neutron transport equation applying the LTSN-DiagExp method

    International Nuclear Information System (INIS)

    Hauser, Eliete Biasotto; Pazos, Ruben Panta; Vilhena, Marco Tullio de; Barros, Ricardo Carvalho de

    2003-01-01

    In this paper we report advances about the three-dimensional nodal discrete-ordinates approximations of neutron transport equation for Cartesian geometry. We use the combined collocation method of the angular variables and nodal approach for the spatial variables. By nodal approach we mean the iterated transverse integration of the S N equations. This procedure leads to the set of one-dimensional averages angular fluxes in each spatial variable. The resulting system of equations is solved with the LTS N method, first applying the Laplace transform to the set of the nodal S N equations and then obtained the solution by symbolic computation. We include the LTS N method by diagonalization to solve the nodal neutron transport equation and then we outline the convergence of these nodal-LTS N approximations with the help of a norm associated to the quadrature formula used to approximate the integral term of the neutron transport equation. (author)

  14. Comparison of neutronic transport equation resolution nodal methods

    International Nuclear Information System (INIS)

    Zamonsky, O.M.; Gho, C.J.

    1990-01-01

    In this work, some transport equation resolution nodal methods are comparatively studied: the constant-constant (CC), linear-nodal (LN) and the constant-quadratic (CQ). A nodal scheme equivalent to finite differences has been used for its programming, permitting its inclusion in existing codes. Some bidimensional problems have been solved, showing that linear-nodal (LN) are, in general, obtained with accuracy in CPU shorter times. (Author) [es

  15. Numerical solution of the 1D kinetics equations using a cubic reduced nodal scheme

    International Nuclear Information System (INIS)

    Gomez T, A.M.; Valle G, E. del; Delfin L, A.; Alonso V, G.

    2003-01-01

    In this work a finite differences technique centered in mesh based on a cubic reduced nodal scheme type finite element to solve the equations of the kinetics 1 D that include the equations corresponding to the concentrations of precursors of delayed neutrons is described. The technique of finite elements used is that of Galerkin where so much the neutron flux as the concentrations of precursors its are spatially approached by means of a three grade polynomial. The matrices of rigidity and of mass that arise during this discretization process are numerically evaluated using the open quadrature non standard of Newton-Cotes and that of Radau respectively. The purpose of the application of these quadratures is the one of to eliminate in the global matrices the couplings among the values of the flow in points of the discretization with the consequent advantages as for the reduction of the order of the matrix associated to the discreet problem that is to solve. As for the time dependent part the classical integration scheme known as Θ scheme is applied. After carrying out the one reordering of unknown and equations it arrives to a reduced system that it can be solved but quickly. With the McKin compute program developed its were solved three benchmark problems and those results are shown for the relative powers. (Author)

  16. Advances in the solution of three-dimensional nodal neutron transport equation

    International Nuclear Information System (INIS)

    Pazos, Ruben Panta; Hauser, Eliete Biasotto; Vilhena, Marco Tullio de

    2003-01-01

    In this paper we study the three-dimensional nodal discrete-ordinates approximations of neutron transport equation in a convex domain with piecewise smooth boundaries. We use the combined collocation method of the angular variables and nodal approach for the spatial variables. By nodal approach we mean the iterated transverse integration of the S N equations. This procedure leads to the set of one-dimensional averages angular fluxes in each spatial variable. The resulting system of equations is solved with the LTS N method, first applying the Laplace transform to the set of the nodal S N equations and then obtaining the solution by symbolic computation. We include the LTS N method by diagonalization to solve the nodal neutron transport equation and then we outline the convergence of these nodal-LTS N approximations with the help of a norm associated to the quadrature formula used to approximate the integral term of the neutron transport equation. We give numerical results obtained with an algebraic computer system (for N up to 8) and with a code for higher values of N. We compare our results for the geometry of a box with a source in a vertex and a leakage zone in the opposite with others techniques used in this problem. (author)

  17. Nodal integral method for the neutron diffusion equation in cylindrical geometry

    International Nuclear Information System (INIS)

    Azmy, Y.Y.

    1987-01-01

    The nodal methodology is based on retaining a higher a higher degree of analyticity in the process of deriving the discrete-variable equations compared to conventional numerical methods. As a result, extensive numerical testing of nodal methods developed for a wide variety of partial differential equations and comparison of the results to conventional methods have established the superior accuracy of nodal methods on coarse meshes. Moreover, these tests have shown that nodal methods are more computationally efficient than finite difference and finite-element methods in the sense that they require shorter CPU times to achieve comparable accuracy in the solutions. However, nodal formalisms and the final discrete-variable equations they produce are, in general, more complicated than their conventional counterparts. This, together with anticipated difficulties in applying the transverse-averaging procedure in curvilinear coordinates, has limited the applications of nodal methods, so far, to Cartesian geometry, and with additional approximations to hexagonal geometry. In this paper the authors report recent progress in deriving and numerically implementing a nodal integral method (NIM) for solving the neutron diffusion equation in cylindrical r-z geometry. Also, presented are comparisons of numerical solutions to two test problems with those obtained by the Exterminator-2 code, which indicate the superior accuracy of the nodal integral method solutions on much coarser meshes

  18. Solution and Study of the Two-Dimensional Nodal Neutron Transport Equation

    International Nuclear Information System (INIS)

    Panta Pazos, Ruben; Biasotto Hauser, Eliete; Tullio de Vilhena, Marco

    2002-01-01

    In the last decade Vilhena and coworkers reported an analytical solution to the two-dimensional nodal discrete-ordinates approximations of the neutron transport equation in a convex domain. The key feature of these works was the application of the combined collocation method of the angular variable and nodal approach in the spatial variables. By nodal approach we mean the transverse integration of the SN equations. This procedure leads to a set of one-dimensional S N equations for the average angular fluxes in the variables x and y. These equations were solved by the old version of the LTS N method, which consists in the application of the Laplace transform to the set of nodal S N equations and solution of the resulting linear system by symbolic computation. It is important to recall that this procedure allow us to increase N the order of S N up to 16. To overcome this drawback we step forward performing a spectral painstaking analysis of the nodal S N equations for N up to 16 and we begin the convergence of the S N nodal equations defining an error for the angular flux and estimating the error in terms of the truncation error of the quadrature approximations of the integral term. Furthermore, we compare numerical results of this approach with those of other techniques used to solve the two-dimensional discrete approximations of the neutron transport equation. (authors)

  19. A polygonal nodal SP3 method for whole core Pin-by-Pin neutronics calculation

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yunzhao; Wu, Hongchun; Cao, Liangzhi, E-mail: xjtulyz@gmail.com, E-mail: hongchun@mail.xjtu.edu.cn, E-mail: caolz@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi' an Jiaotong University, Shaanxi (China)

    2011-07-01

    In this polygonal nodal-SP3 method, neutron transport equation is transformed by employing an isotropic SP3 method into two coupled equations that are both in the same mathematic form with the diffusion equation, and then a polygonal nodal method is proposed to solve the two coupled equations. In the polygonal nodal method, adjacent nodes are coupled through partial currents, and a nodal response matrix between incoming and outgoing currents is obtained by expanding detailed nodal flux distribution into a sum of exponential functions. This method avoids the transverse integral technique, which is widely used in regular nodal method and can not be used in triangular geometry because of the mathematical singularity. It is demonstrated by the numerical results of the test problems that the k{sub eff} and power distribution agree well with other codes, the triangular nodal-SP3 method appears faster, and that whole core pin-by-pin transport calculation with fine meshes is feasible after parallelization and acceleration. (author)

  20. A new nodal kinetics method for analyzing fast control rod motions in nuclear reactor cores

    International Nuclear Information System (INIS)

    Kaya, S.; Yavuz, H.

    2001-01-01

    A new nodal kinetics approach is developed for analyzing large reactivity accidents in nuclear reactor cores. This method shows promising that it has capability of inspecting promt criticality transients and it gives comparable results with respect to those of other techniques. (orig.)

  1. Higher order polynomial expansion nodal method for hexagonal core neutronics analysis

    International Nuclear Information System (INIS)

    Jin, Young Cho; Chang, Hyo Kim

    1998-01-01

    A higher-order polynomial expansion nodal(PEN) method is newly formulated as a means to improve the accuracy of the conventional PEN method solutions to multi-group diffusion equations in hexagonal core geometry. The new method is applied to solving various hexagonal core neutronics benchmark problems. The computational accuracy of the higher order PEN method is then compared with that of the conventional PEN method, the analytic function expansion nodal (AFEN) method, and the ANC-H method. It is demonstrated that the higher order PEN method improves the accuracy of the conventional PEN method and that it compares very well with the other nodal methods like the AFEN and ANC-H methods in accuracy

  2. Development and verification of an efficient spatial neutron kinetics method for reactivity-initiated event analyses

    International Nuclear Information System (INIS)

    Ikeda, Hideaki; Takeda, Toshikazu

    2001-01-01

    A space/time nodal diffusion code based on the nodal expansion method (NEM), EPISODE, was developed in order to evaluate transient neutron behavior in light water reactor cores. The present code employs the improved quasistatic (IQS) method for spatial neutron kinetics, and neutron flux distribution is numerically obtained by solving the neutron diffusion equation with the nonlinear iteration scheme to achieve fast computation. A predictor-corrector (PC) method developed in the present study enabled to apply a coarse time mesh to the transient spatial neutron calculation than that applicable in the conventional IQS model, which improved computational efficiency further. Its computational advantage was demonstrated by applying to the numerical benchmark problems that simulate reactivity-initiated events, showing reduction of computational times up to a factor of three than the conventional IQS. The thermohydraulics model was also incorporated in EPISODE, and the capability of realistic reactivity event analyses was verified using the SPERT-III/E-Core experimental data. (author)

  3. Spatial neutron kinetic module of ROSA code

    International Nuclear Information System (INIS)

    Cherezov, A.L.; Shchukin, N.V.

    2009-01-01

    A spatial neutron kinetic module was developed for computer code ROSA. The paper describes a numerical scheme used in the module for resolving neutron kinetic equations. Analytical integration for delayed neutrons emitters method and direct numerical integration method (Gear's method) were analyzed. The two methods were compared on their efficiency and accuracy. Both methods were verified with test problems. The results obtained in the verification studies were presented [ru

  4. An analytical approach for a nodal scheme of two-dimensional neutron transport problems

    International Nuclear Information System (INIS)

    Barichello, L.B.; Cabrera, L.C.; Prolo Filho, J.F.

    2011-01-01

    Research highlights: → Nodal equations for a two-dimensional neutron transport problem. → Analytical Discrete Ordinates Method. → Numerical results compared with the literature. - Abstract: In this work, a solution for a two-dimensional neutron transport problem, in cartesian geometry, is proposed, on the basis of nodal schemes. In this context, one-dimensional equations are generated by an integration process of the multidimensional problem. Here, the integration is performed for the whole domain such that no iterative procedure between nodes is needed. The ADO method is used to develop analytical discrete ordinates solution for the one-dimensional integrated equations, such that final solutions are analytical in terms of the spatial variables. The ADO approach along with a level symmetric quadrature scheme, lead to a significant order reduction of the associated eigenvalues problems. Relations between the averaged fluxes and the unknown fluxes at the boundary are introduced as the usually needed, in nodal schemes, auxiliary equations. Numerical results are presented and compared with test problems.

  5. Numerical solution of the Neutron Transport Equation using discontinuous nodal methods at X-Y geometry

    International Nuclear Information System (INIS)

    Delfin L, A.

    1996-01-01

    The purpose of this work is to solve the neutron transport equation in discrete-ordinates and X-Y geometry by developing and using the strong discontinuous and strong modified discontinuous nodal finite element schemes. The strong discontinuous and modified strong discontinuous nodal finite element schemes go from two to ten interpolation parameters per cell. They are describing giving a set D c and polynomial space S c corresponding for each scheme BDMO, RTO, BL, BDM1, HdV, BDFM1, RT1, BQ and BDM2. The solution is obtained solving the neutron transport equation moments for each nodal scheme by developing the basis functions defined by Pascal triangle and the Legendre moments giving in the polynomial space S c and, finally, looking for the non singularity of the resulting linear system. The linear system is numerically solved using a computer program for each scheme mentioned . It uses the LU method and forward and backward substitution and makes a partition of the domain in cells. The source terms and angular flux are calculated, using the directions and weights associated to the S N approximation and solving the angular flux moments to find the effective multiplication constant. The programs are written in Fortran language, using the dynamic allocation of memory to increase efficiently the available memory of the computing equipment. (Author)

  6. Nodal kinetics model upgrade in the Penn State coupled TRAC/NEM codes

    International Nuclear Information System (INIS)

    Beam, Tara M.; Ivanov, Kostadin N.; Baratta, Anthony J.; Finnemann, Herbert

    1999-01-01

    The Pennsylvania State University currently maintains and does development and verification work for its own versions of the coupled three-dimensional kinetics/thermal-hydraulics codes TRAC-PF1/NEM and TRAC-BF1/NEM. The subject of this paper is nodal model enhancements in the above mentioned codes. Because of the numerous validation studies that have been performed on almost every aspect of these codes, this upgrade is done without a major code rewrite. The upgrade consists of four steps. The first two steps are designed to improve the accuracy of the kinetics model, based on the nodal expansion method. The polynomial expansion solution of 1D transverse integrated diffusion equation is replaced with a solution, which uses a semi-analytic expansion. Further the standard parabolic polynomial representation of the transverse leakage in the above 1D equations is replaced with an improved approximation. The last two steps of the upgrade address the code efficiency by improving the solution of the time-dependent NEM equations and implementing a multi-grid solver. These four improvements are implemented into the standalone NEM kinetics code. Verification of this code was accomplished based on the original verification studies. The results show that the new methods improve the accuracy and efficiency of the code. The verification of the upgraded NEM model in the TRAC-PF1/NEM and TRAC-BF1/NEM coupled codes is underway

  7. An iterative algorithm for solving the multidimensional neutron diffusion nodal method equations on parallel computers

    International Nuclear Information System (INIS)

    Kirk, B.L.; Azmy, Y.Y.

    1992-01-01

    In this paper the one-group, steady-state neutron diffusion equation in two-dimensional Cartesian geometry is solved using the nodal integral method. The discrete variable equations comprise loosely coupled sets of equations representing the nodal balance of neutrons, as well as neutron current continuity along rows or columns of computational cells. An iterative algorithm that is more suitable for solving large problems concurrently is derived based on the decomposition of the spatial domain and is accelerated using successive overrelaxation. This algorithm is very well suited for parallel computers, especially since the spatial domain decomposition occurs naturally, so that the number of iterations required for convergence does not depend on the number of processors participating in the calculation. Implementation of the authors' algorithm on the Intel iPSC/2 hypercube and Sequent Balance 8000 parallel computer is presented, and measured speedup and efficiency for test problems are reported. The results suggest that the efficiency of the hypercube quickly deteriorates when many processors are used, while the Sequent Balance retains very high efficiency for a comparable number of participating processors. This leads to the conjecture that message-passing parallel computers are not as well suited for this algorithm as shared-memory machines

  8. Solution and study of nodal neutron transport equation applying the LTS{sub N}-DiagExp method

    Energy Technology Data Exchange (ETDEWEB)

    Hauser, Eliete Biasotto; Pazos, Ruben Panta [Pontificia Univ. Catolica do Rio Grande do Sul, Porto Alegre, RS (Brazil). Faculdade de Matematica]. E-mail: eliete@pucrs.br; rpp@mat.pucrs.br; Vilhena, Marco Tullio de [Pontificia Univ. Catolica do Rio Grande do Sul, Porto Alegre, RS (Brazil). Instituto de Matematica]. E-mail: vilhena@mat.ufrgs.br; Barros, Ricardo Carvalho de [Universidade do Estado, Nova Friburgo, RJ (Brazil). Instituto Politecnico]. E-mail: ricardo@iprj.uerj.br

    2003-07-01

    In this paper we report advances about the three-dimensional nodal discrete-ordinates approximations of neutron transport equation for Cartesian geometry. We use the combined collocation method of the angular variables and nodal approach for the spatial variables. By nodal approach we mean the iterated transverse integration of the S{sub N} equations. This procedure leads to the set of one-dimensional averages angular fluxes in each spatial variable. The resulting system of equations is solved with the LTS{sub N} method, first applying the Laplace transform to the set of the nodal S{sub N} equations and then obtained the solution by symbolic computation. We include the LTS{sub N} method by diagonalization to solve the nodal neutron transport equation and then we outline the convergence of these nodal-LTS{sub N} approximations with the help of a norm associated to the quadrature formula used to approximate the integral term of the neutron transport equation. (author)

  9. A semi-experimental nodal synthesis method for the on-line reconstruction of three-dimensional neutron flux-shapes and reactivity

    International Nuclear Information System (INIS)

    Jacqmin, R.P.

    1991-01-01

    The safety and optimal performance of large, commercial, light-water reactors require the knowledge at all time of the neutron-flux distribution in the core. In principle, this information can be obtained by solving the time-dependent neutron diffusion equations. However, this approach is complicated and very expensive. Sufficiently accurate, real-time calculations (time scale of approximately one second) are not yet possible on desktop computers, even with fast-running, nodal kinetics codes. A semi-experimental, nodal synthesis method which avoids the solution of the time-dependent, neutron diffusion equations is described. The essential idea of this method is to approximate instantaneous nodal group-fluxes by a linear combination of K, precomputed, three-dimensional, static expansion-functions. The time-dependent coefficients of the combination are found from the requirement that the reconstructed flux-distribution agree in a least-squares sense with the readings of J (≥K) fixed, prompt-responding neutron-detectors. Possible numerical difficulties with the least-squares solution of the ill-conditioned, J-by-K system of equations are brought under complete control by the use of a singular-value-decomposition technique. This procedure amounts to the rearrangement of the original, linear combination of K expansion functions into an equivalent more convenient, linear combination of R (≤K) orthogonalized ''modes'' of decreasing magnitude. Exceedingly small modes are zeroed to eliminate any risk of roundoff-error amplification, and to assure consistency with the limited accuracy of the data. Additional modes are zeroed when it is desirable to limit the sensitivity of the results to measurement noise

  10. A semi-experimental nodal synthesis method for the on-line reconstruction of three-dimensional neutron flux-shapes and reactivity. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Jacqmin, Robert P. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    1991-12-10

    The safety and optimal performance of large, commercial, light-water reactors require the knowledge at all time of the neutron-flux distribution in the core. In principle, this information can be obtained by solving the time-dependent neutron diffusion equations. However, this approach is complicated and very expensive. Sufficiently accurate, real-time calculations (time scale of approximately one second) are not yet possible on desktop computers, even with fast-running, nodal kinetics codes. A semi-experimental, nodal synthesis method which avoids the solution of the time-dependent, neutron diffusion equations is described. The essential idea of this method is to approximate instantaneous nodal group-fluxes by a linear combination of K, precomputed, three-dimensional, static expansion-functions. The time-dependent coefficients of the combination are found from the requirement that the reconstructed flux-distribution agree in a least-squares sense with the readings of J (≥K) fixed, prompt-responding neutron-detectors. Possible numerical difficulties with the least-squares solution of the ill-conditioned, J-by-K system of equations are brought under complete control by the use of a singular-value-decomposition technique. This procedure amounts to the rearrangement of the original, linear combination of K expansion functions into an equivalent more convenient, linear combination of R (≤K) orthogonalized ``modes`` of decreasing magnitude. Exceedingly small modes are zeroed to eliminate any risk of roundoff-error amplification, and to assure consistency with the limited accuracy of the data. Additional modes are zeroed when it is desirable to limit the sensitivity of the results to measurement noise.

  11. A semi-experimental nodal synthesis method for the on-line reconstruction of three-dimensional neutron flux-shapes and reactivity

    Energy Technology Data Exchange (ETDEWEB)

    Jacqmin, R.P.

    1991-12-10

    The safety and optimal performance of large, commercial, light-water reactors require the knowledge at all time of the neutron-flux distribution in the core. In principle, this information can be obtained by solving the time-dependent neutron diffusion equations. However, this approach is complicated and very expensive. Sufficiently accurate, real-time calculations (time scale of approximately one second) are not yet possible on desktop computers, even with fast-running, nodal kinetics codes. A semi-experimental, nodal synthesis method which avoids the solution of the time-dependent, neutron diffusion equations is described. The essential idea of this method is to approximate instantaneous nodal group-fluxes by a linear combination of K, precomputed, three-dimensional, static expansion-functions. The time-dependent coefficients of the combination are found from the requirement that the reconstructed flux-distribution agree in a least-squares sense with the readings of J ({ge}K) fixed, prompt-responding neutron-detectors. Possible numerical difficulties with the least-squares solution of the ill-conditioned, J-by-K system of equations are brought under complete control by the use of a singular-value-decomposition technique. This procedure amounts to the rearrangement of the original, linear combination of K expansion functions into an equivalent more convenient, linear combination of R ({le}K) orthogonalized modes'' of decreasing magnitude. Exceedingly small modes are zeroed to eliminate any risk of roundoff-error amplification, and to assure consistency with the limited accuracy of the data. Additional modes are zeroed when it is desirable to limit the sensitivity of the results to measurement noise.

  12. The Verification of Coupled Neutronics Thermal-Hydraulics Code NODAL3 in the PWR Rod Ejection Benchmark

    Directory of Open Access Journals (Sweden)

    Surian Pinem

    2014-01-01

    Full Text Available A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the few-group neutron diffusion equation in 3-dimensional geometry for typical PWR static and transient analyses. The spatial variables are treated by using a polynomial nodal method while for the neutron dynamic solver the adiabatic and improved quasistatic methods are adopted. In this paper we report the benchmark calculation results of the code against the OECD/NEA CRP PWR rod ejection cases. The objective of this work is to determine the accuracy of NODAL3 code in analysing the reactivity initiated accident due to the control rod ejection. The NEACRP PWR rod ejection cases are chosen since many organizations participated in the NEA project using various methods as well as approximations, so that, in addition to the reference solutions, the calculation results of NODAL3 code can also be compared to other codes’ results. The transient parameters to be verified are time of power peak, power peak, final power, final average Doppler temperature, maximum fuel temperature, and final coolant temperature. The results of NODAL3 code agree well with the PHANTHER reference solutions in 1993 and 1997 (revised. Comparison with other validated codes, DYN3D/R and ANCK, shows also a satisfactory agreement.

  13. New aspects in the implementation of the quasi-static method for the solution of neutron diffusion problems in the framework of a nodal method

    International Nuclear Information System (INIS)

    Caron, D.; Dulla, S.; Ravetto, P.

    2016-01-01

    Highlights: • The implementation of the quasi-static method in 3D nodal diffusion theory model in hexagonal-z geometry is described. • Different formulations of the quasi-static technique are discussed. • The results presented illustrate the features of the various formulations, highlighting advantages and drawbacks. • A novel adaptive procedure for the selection of the time interval between shape recalculations is presented. - Abstract: The ability to accurately model the dynamic behaviour of the neutron distribution in a nuclear system is a fundamental aspect of reactor design and safety assessment. Due to the heavy computational burden associated to the direct time inversion of the full model, the quasi-static method has become a standard approach to the numerical solution of the nuclear reactor dynamic equations on the full phase space. The present paper is opened by an introductory critical review of the basics of the quasi-static scheme for the general neutron kinetic problem. Afterwards, the implementation of the quasi-static method in the context of a three-dimensional nodal diffusion theory model in hexagonal-z geometry is described, including some peculiar aspects of the adjoint nodal equations and the explicit formulation of the quasi-static nodal equations. The presentation includes the discussion of different formulations of the quasi-static technique. The results presented illustrate the features of the various formulations, highlighting the corresponding advantages and drawbacks. An adaptive procedure for the selection of the time interval between shape recalculations is also presented, showing its usefulness in practical applications.

  14. Solution of the Neutron transport equation in hexagonal geometry using strongly discontinuous nodal schemes

    International Nuclear Information System (INIS)

    Mugica R, C.A.; Valle G, E. del

    2005-01-01

    In 2002, E. del Valle and Ernest H. Mund developed a technique to solve numerically the Neutron transport equations in discrete ordinates and hexagonal geometry using two nodal schemes type finite element weakly discontinuous denominated WD 5,3 and WD 12,8 (of their initials in english Weakly Discontinuous). The technique consists on representing each hexagon in the union of three rhombuses each one of which it is transformed in a square in the one that the methods WD 5,3 and WD 12,8 were applied. In this work they are solved the mentioned equations of transport using the same discretization technique by hexagon but using two nodal schemes type finite element strongly discontinuous denominated SD 3 and SD 8 (of their initials in english Strongly Discontinuous). The application in each case as well as a reference problem for those that results are provided for the effective multiplication factor is described. It is carried out a comparison with the obtained results by del Valle and Mund for different discretization meshes so much angular as spatial. (Author)

  15. Pellet by pellet neutron flux calculations coupled with nodal expansion method

    International Nuclear Information System (INIS)

    Aldo, Dall'Osso

    2003-01-01

    We present a technique whose aim is to replace 2-dimensional pin by pin de-homogenization, currently done in core reactor calculations with the nodal expansion method (NEM), by a 3-dimensional finite difference diffusion calculation. This fine calculation is performed as a zoom in each node taking as boundary conditions the results of the NEM calculations. The size of fine mesh is of the order of a fuel pellet. The coupling between fine and NEM calculations is realised by an albedo like boundary condition. Some examples are presented showing fine neutron flux shape near control rods or assembly grids. Other fine flux behaviour as the thermal flux rise in the fuel near the reflector is emphasised. In general the results show the interest of the method in conditions where the separability of radial and axial directions is not granted. (author)

  16. An analytical discrete ordinates solution for a nodal model of a two-dimensional neutron transport problem

    International Nuclear Information System (INIS)

    Filho, J. F. P.; Barichello, L. B.

    2013-01-01

    In this work, an analytical discrete ordinates method is used to solve a nodal formulation of a neutron transport problem in x, y-geometry. The proposed approach leads to an important reduction in the order of the associated eigenvalue systems, when combined with the classical level symmetric quadrature scheme. Auxiliary equations are proposed, as usually required for nodal methods, to express the unknown fluxes at the boundary introduced as additional unknowns in the integrated equations. Numerical results, for the problem defined by a two-dimensional region with a spatially constant and isotropically emitting source, are presented and compared with those available in the literature. (authors)

  17. Numerical solution of the 1D kinetics equations using a cubic reduced nodal scheme; Solucion numerica de las ecuaciones de la cinetica 1D usando un esquema nodal cubico reducido

    Energy Technology Data Exchange (ETDEWEB)

    Gomez T, A.M.; Valle G, E. del [IPN-ESFM, 07738 Mexico D.F. (Mexico); Delfin L, A.; Alonso V, G. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)] e-mail: armagotorres@aol.com

    2003-07-01

    In this work a finite differences technique centered in mesh based on a cubic reduced nodal scheme type finite element to solve the equations of the kinetics 1 D that include the equations corresponding to the concentrations of precursors of delayed neutrons is described. The technique of finite elements used is that of Galerkin where so much the neutron flux as the concentrations of precursors its are spatially approached by means of a three grade polynomial. The matrices of rigidity and of mass that arise during this discretization process are numerically evaluated using the open quadrature non standard of Newton-Cotes and that of Radau respectively. The purpose of the application of these quadratures is the one of to eliminate in the global matrices the couplings among the values of the flow in points of the discretization with the consequent advantages as for the reduction of the order of the matrix associated to the discreet problem that is to solve. As for the time dependent part the classical integration scheme known as {theta} scheme is applied. After carrying out the one reordering of unknown and equations it arrives to a reduced system that it can be solved but quickly. With the McKin compute program developed its were solved three benchmark problems and those results are shown for the relative powers. (Author)

  18. DIF3D nodal neutronics option for two- and three-dimensional diffusion theory calculations in hexagonal geometry. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Lawrence, R.D.

    1983-03-01

    A nodal method is developed for the solution of the neutron-diffusion equation in two- and three-dimensional hexagonal geometries. The nodal scheme has been incorporated as an option in the finite-difference diffusion-theory code DIF3D, and is intended for use in the analysis of current LMFBR designs. The nodal equations are derived using higher-order polynomial approximations to the spatial dependence of the flux within the hexagonal-z node. The final equations, which are cast in the form of inhomogeneous response-matrix equations for each energy group, involved spatial moments of the node-interior flux distribution plus surface-averaged partial currents across the faces of the node. These equations are solved using a conventional fission-source iteration accelerated by coarse-mesh rebalance and asymptotic source extrapolation. This report describes the mathematical development and numerical solution of the nodal equations, as well as the use of the nodal option and details concerning its programming structure. This latter information is intended to supplement the information provided in the separate documentation of the DIF3D code.

  19. DIF3D nodal neutronics option for two- and three-dimensional diffusion theory calculations in hexagonal geometry

    International Nuclear Information System (INIS)

    Lawrence, R.D.

    1983-03-01

    A nodal method is developed for the solution of the neutron-diffusion equation in two- and three-dimensional hexagonal geometries. The nodal scheme has been incorporated as an option in the finite-difference diffusion-theory code DIF3D, and is intended for use in the analysis of current LMFBR designs. The nodal equations are derived using higher-order polynomial approximations to the spatial dependence of the flux within the hexagonal-z node. The final equations, which are cast in the form of inhomogeneous response-matrix equations for each energy group, involved spatial moments of the node-interior flux distribution plus surface-averaged partial currents across the faces of the node. These equations are solved using a conventional fission-source iteration accelerated by coarse-mesh rebalance and asymptotic source extrapolation. This report describes the mathematical development and numerical solution of the nodal equations, as well as the use of the nodal option and details concerning its programming structure. This latter information is intended to supplement the information provided in the separate documentation of the DIF3D code

  20. Neutron inverse kinetics via Gaussian Processes

    International Nuclear Information System (INIS)

    Picca, Paolo; Furfaro, Roberto

    2012-01-01

    Highlights: ► A novel technique for the interpretation of experiments in ADS is presented. ► The technique is based on Bayesian regression, implemented via Gaussian Processes. ► GPs overcome the limits of classical methods, based on PK approximation. ► Results compares GPs and ANN performance, underlining similarities and differences. - Abstract: The paper introduces the application of Gaussian Processes (GPs) to determine the subcriticality level in accelerator-driven systems (ADSs) through the interpretation of pulsed experiment data. ADSs have peculiar kinetic properties due to their special core design. For this reason, classical – inversion techniques based on point kinetic (PK) generally fail to generate an accurate estimate of reactor subcriticality. Similarly to Artificial Neural Networks (ANNs), Gaussian Processes can be successfully trained to learn the underlying inverse neutron kinetic model and, as such, they are not limited to the model choice. Importantly, GPs are strongly rooted into the Bayes’ theorem which makes them a powerful tool for statistical inference. Here, GPs have been designed and trained on a set of kinetics models (e.g. point kinetics and multi-point kinetics) for homogeneous and heterogeneous settings. The results presented in the paper show that GPs are very efficient and accurate in predicting the reactivity for ADS-like systems. The variance computed via GPs may provide an indication on how to generate additional data as function of the desired accuracy.

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

    Energy Technology Data Exchange (ETDEWEB)

    Duerigen, Susan

    2013-05-15

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

  2. Evaluation of the use of nodal methods for MTR neutronic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Reitsma, F.; Mueller, E.Z.

    1997-08-01

    Although modern nodal methods are used extensively in the nuclear power industry, their use for research reactor analysis has been very limited. The suitability of nodal methods for material testing reactor analysis is investigated with the emphasis on the modelling of the core region (fuel assemblies). The nodal approach`s performance is compared with that of the traditional finite-difference fine mesh approach. The advantages of using nodal methods coupled with integrated cross section generation systems are highlighted, especially with respect to data preparation, simplicity of use and the possibility of performing a great variety of reactor calculations subject to strict time limitations such as are required for the RERTR program.

  3. Numerical solution of the Neutron Transport Equation using discontinuous nodal methods at X-Y geometry; Solucion numerica de la ecuacion de transporte de neutrones usando metodos nodales discontinuos en geometria X-Y

    Energy Technology Data Exchange (ETDEWEB)

    Delfin L, A

    1997-12-31

    The purpose of this work is to solve the neutron transport equation in discrete-ordinates and X-Y geometry by developing and using the strong discontinuous and strong modified discontinuous nodal finite element schemes. The strong discontinuous and modified strong discontinuous nodal finite element schemes go from two to ten interpolation parameters per cell. They are describing giving a set D{sub c} and polynomial space S{sub c} corresponding for each scheme BDMO, RTO, BL, BDM1, HdV, BDFM1, RT1, BQ and BDM2. The solution is obtained solving the neutron transport equation moments for each nodal scheme by developing the basis functions defined by Pascal triangle and the Legendre moments giving in the polynomial space S{sub c} and, finally, looking for the non singularity of the resulting linear system. The linear system is numerically solved using a computer program for each scheme mentioned . It uses the LU method and forward and backward substitution and makes a partition of the domain in cells. The source terms and angular flux are calculated, using the directions and weights associated to the S{sub N} approximation and solving the angular flux moments to find the effective multiplication constant. The programs are written in Fortran language, using the dynamic allocation of memory to increase efficiently the available memory of the computing equipment. (Author).

  4. Development of 3-dimensional neutronics kinetics analysis code for CANDU-PHWR

    International Nuclear Information System (INIS)

    Kim, M. W.; Kim, C. H.; Hong, I. S.

    2005-02-01

    The followings are the major contents and scope of the research : development of kinetics power calculation module, formulation of space-dependent neutron transient analysis - implementation of 3-D and 2-G unified nodal method, verification of the kinetics module by benchmark problem - 3-D PHWR kinetics benchmark problem suggested by AECL, reactor trip simulation by shutdown system 1 in Wolsong unit 2. Development of a dynamic linked library code, SCAN D LL, for the coupled calculation with RELAP-CANDU : modeling of shutdown system 1, development of automatic shutdown module - automatic trip module based on rate log power control logic, automatic insertion of shutdown system 1. Development of a link code for coupled calculation - development of SCAN D LL(windows version), verification of coupled code by - 40% reactor inlet header break LOCA power pulse, 100% reactor outlet header break LOCA power pulse, 50% pump suction break LOCA power pulse

  5. Introduction to the neutron kinetics of nuclear power reactors

    CERN Document Server

    Tyror, J G; Grant, P J

    2013-01-01

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

  6. A one-dimensional, one-group absorption-production nodal method for neutron flux and power distributions calculations

    International Nuclear Information System (INIS)

    Ferreira, C.R.

    1984-01-01

    It is presented the absorption-production nodal method for steady and dynamical calculations in one-dimension and one group energy. It was elaborated the NOD1D computer code (in FORTRAN-IV language). Calculations of neutron flux and power distributions, burnup, effective multiplication factors and critical boron concentration were made with the NOD1D code and compared with results obtained through the CITATION code, which uses the finite difference method. The nuclear constants were produced by the LEOPARD code. (M.C.K.) [pt

  7. On the non-uniqueness of the nodal mathematical adjoint

    International Nuclear Information System (INIS)

    Müller, Erwin

    2014-01-01

    Highlights: • We evaluate three CMFD schemes for computing the nodal mathematical adjoint. • The nodal mathematical adjoint is not unique and can be non-positive (nonphysical). • Adjoint and forward eigenmodes are compatible if produced by the same CMFD method. • In nodal applications the excited eigenmodes are purely mathematical entities. - Abstract: Computation of the neutron adjoint flux within the framework of modern nodal diffusion methods is often facilitated by reducing the nodal equation system for the forward flux into a simpler coarse-mesh finite-difference form and then transposing the resultant matrix equations. The solution to the transposed problem is known as the nodal mathematical adjoint. Since the coarse-mesh finite-difference reduction of a given nodal formulation can be obtained in a number of ways, different nodal mathematical adjoint solutions can be computed. This non-uniqueness of the nodal mathematical adjoint challenges the credibility of the reduction strategy and demands a verdict as to its suitability in practical applications. This is the matter under consideration in this paper. A selected number of coarse-mesh finite-difference reduction schemes are described and compared. Numerical calculations are utilised to illustrate the differences in the adjoint solutions as well as to appraise the impact on such common applications as the computation of core point kinetics parameters. Recommendations are made for the proper application of the coarse-mesh finite-difference reduction approach to the nodal mathematical adjoint problem

  8. A new communication scheme for the neutron diffusion nodal method in a distributed computing environment

    International Nuclear Information System (INIS)

    Kirk, B.L.; Azmy, Y.

    1994-01-01

    A modified scheme is developed for solving the two-dimensional nodal diffusion equations on distributed memory computers. The scheme is aimed at minimizing the volume of communication among processors while maximizing the tasks in parallel. Results show a significant improvement in parallel efficiency on the Intel iPSC/860 hypercube compared to previous algorithms

  9. Development of LMR basic design technology - Development of 3-D multi-group nodal kinetics code for liquid metal reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myung Hyun [Kyunghee University, Seoul (Korea, Republic of)

    1996-07-01

    A development project of 3-dimensional kinetics code for ALMR has three level of works. In the first level, a multi-group, nodal kinetics code for the HEX-Z geometry has been developed. A code showed very good results for the static analysis as well as for the kinetics problems. At the second level, a core thermal-hydraulic analysis code was developed for the temperature feedback calculation in ALMR transients analysis. This code is coupled with kinetics code. A sodium property table was programmed and tested to the KAERI data and thermal feedback model was developed and coupled in code. Benchmarking of T/H calculation has been performed and showed fairly good results. At the third level of research work, reactivity feedback model for structure thermal expansion is developed and added to the code. At present, basic model was studied. However, code development in now on going. Benchmarking of this model developed can not be done because of lack of data. 31 refs., 17 tabs., 38 figs. (author)

  10. Fractional neutron point kinetics equations for nuclear reactor dynamics

    International Nuclear Information System (INIS)

    Espinosa-Paredes, Gilberto; Polo-Labarrios, Marco-A.; Espinosa-Martinez, Erick-G.; Valle-Gallegos, Edmundo del

    2011-01-01

    The fractional point-neutron kinetics model for the dynamic behavior in a nuclear reactor is derived and analyzed in this paper. The fractional model retains the main dynamic characteristics of the neutron motion in which the relaxation time associated with a rapid variation in the neutron flux contains a fractional order, acting as exponent of the relaxation time, to obtain the best representation of a nuclear reactor dynamics. The physical interpretation of the fractional order is related with non-Fickian effects from the neutron diffusion equation point of view. The numerical approximation to the solution of the fractional neutron point kinetics model, which can be represented as a multi-term high-order linear fractional differential equation, is calculated by reducing the problem to a system of ordinary and fractional differential equations. The numerical stability of the fractional scheme is investigated in this work. Results for neutron dynamic behavior for both positive and negative reactivity and for different values of fractional order are shown and compared with the classic neutron point kinetic equations. Additionally, a related review with the neutron point kinetics equations is presented, which encompasses papers written in English about this research topic (as well as some books and technical reports) published since 1940 up to 2010.

  11. Solution of the Neutron transport equation in hexagonal geometry using strongly discontinuous nodal schemes; Solucion de la Ecuacion de transporte de neutrones en geometria hexagonal usando esquemas nodales fuertemente discontinuos

    Energy Technology Data Exchange (ETDEWEB)

    Mugica R, C.A.; Valle G, E. del [IPN, ESFM, Departamento de Ingenieria Nuclear, 07738 Mexico D.F. (Mexico)]. e-mail: cmugica@ipn.mx

    2005-07-01

    In 2002, E. del Valle and Ernest H. Mund developed a technique to solve numerically the Neutron transport equations in discrete ordinates and hexagonal geometry using two nodal schemes type finite element weakly discontinuous denominated WD{sub 5,3} and WD{sub 12,8} (of their initials in english Weakly Discontinuous). The technique consists on representing each hexagon in the union of three rhombuses each one of which it is transformed in a square in the one that the methods WD{sub 5,3} and WD{sub 12,8} were applied. In this work they are solved the mentioned equations of transport using the same discretization technique by hexagon but using two nodal schemes type finite element strongly discontinuous denominated SD{sub 3} and SD{sub 8} (of their initials in english Strongly Discontinuous). The application in each case as well as a reference problem for those that results are provided for the effective multiplication factor is described. It is carried out a comparison with the obtained results by del Valle and Mund for different discretization meshes so much angular as spatial. (Author)

  12. Development of a computer code for neutronic calculations of a hexagonal lattice of nuclear reactor using the flux expansion nodal method

    Directory of Open Access Journals (Sweden)

    Mohammadnia Meysam

    2013-01-01

    Full Text Available The flux expansion nodal method is a suitable method for considering nodalization effects in node corners. In this paper we used this method to solve the intra-nodal flux analytically. Then, a computer code, named MA.CODE, was developed using the C# programming language. The code is capable of reactor core calculations for hexagonal geometries in two energy groups and three dimensions. The MA.CODE imports two group constants from the WIMS code and calculates the effective multiplication factor, thermal and fast neutron flux in three dimensions, power density, reactivity, and the power peaking factor of each fuel assembly. Some of the code's merits are low calculation time and a user friendly interface. MA.CODE results showed good agreement with IAEA benchmarks, i. e. AER-FCM-101 and AER-FCM-001.

  13. Performance of neutron kinetics models for ADS transient analyses

    International Nuclear Information System (INIS)

    Rineiski, A.; Maschek, W.; Rimpault, G.

    2002-01-01

    Within the framework of the SIMMER code development, neutron kinetics models for simulating transients and hypothetical accidents in advanced reactor systems, in particular in Accelerator Driven Systems (ADSs), have been developed at FZK/IKET in cooperation with CE Cadarache. SIMMER is a fluid-dynamics/thermal-hydraulics code, coupled with a structure model and a space-, time- and energy-dependent neutronics module for analyzing transients and accidents. The advanced kinetics models have also been implemented into KIN3D, a module of the VARIANT/TGV code (stand-alone neutron kinetics) for broadening application and for testing and benchmarking. In the paper, a short review of the SIMMER and KIN3D neutron kinetics models is given. Some typical transients related to ADS perturbations are analyzed. The general models of SIMMER and KIN3D are compared with more simple techniques developed in the context of this work to get a better understanding of the specifics of transients in subcritical systems and to estimate the performance of different kinetics options. These comparisons may also help in elaborating new kinetics models and extending existing computation tools for ADS transient analyses. The traditional point-kinetics model may give rather inaccurate transient reaction rate distributions in an ADS even if the material configuration does not change significantly. This inaccuracy is not related to the problem of choosing a 'right' weighting function: the point-kinetics model with any weighting function cannot take into account pronounced flux shape variations related to possible significant changes in the criticality level or to fast beam trips. To improve the accuracy of the point-kinetics option for slow transients, we have introduced a correction factor technique. The related analyses give a better understanding of 'long-timescale' kinetics phenomena in the subcritical domain and help to evaluate the performance of the quasi-static scheme in a particular case. One

  14. Performance of non-conventional factorization approaches for neutron kinetics

    International Nuclear Information System (INIS)

    Bulla, S.; Nervo, M.

    2013-01-01

    The use of factorization techniques provides a interesting option for the simulation of the time-dependent behavior of nuclear systems with a reduced computational effort. While point kinetics neglects all spatial and spectral effects, quasi-statics and multipoint kinetics allow to produce results with a higher accuracy for transients involving relevant modifications of the neutron distribution. However, in some conditions these methods can not work efficiently. In this paper, we discuss some possible alternative formulations for the factorization process for neutron kinetics, leading to mathematical models of reduced complications that can allow an accurate simulation of transients involving spatial and spectral effects. The performance of these innovative approaches are compared to standard techniques for some test cases, showing the benefits and shortcomings of the method proposed. (authors)

  15. A variational synthesis nodal discrete ordinates method

    International Nuclear Information System (INIS)

    Favorite, J.A.; Stacey, W.M.

    1999-01-01

    A self-consistent nodal approximation method for computing discrete ordinates neutron flux distributions has been developed from a variational functional for neutron transport theory. The advantage of the new nodal method formulation is that it is self-consistent in its definition of the homogenized nodal parameters, the construction of the global nodal equations, and the reconstruction of the detailed flux distribution. The efficacy of the method is demonstrated by two-dimensional test problems

  16. NESTLE: Few-group neutron diffusion equation solver utilizing the nodal expansion method for eigenvalue, adjoint, fixed-source steady-state and transient problems

    International Nuclear Information System (INIS)

    Turinsky, P.J.; Al-Chalabi, R.M.K.; Engrand, P.; Sarsour, H.N.; Faure, F.X.; Guo, W.

    1994-06-01

    NESTLE is a FORTRAN77 code that solves the few-group neutron diffusion equation utilizing the Nodal Expansion Method (NEM). NESTLE can solve the eigenvalue (criticality); eigenvalue adjoint; external fixed-source steady-state; or external fixed-source. or eigenvalue initiated transient problems. The code name NESTLE originates from the multi-problem solution capability, abbreviating Nodal Eigenvalue, Steady-state, Transient, Le core Evaluator. The eigenvalue problem allows criticality searches to be completed, and the external fixed-source steady-state problem can search to achieve a specified power level. Transient problems model delayed neutrons via precursor groups. Several core properties can be input as time dependent. Two or four energy groups can be utilized, with all energy groups being thermal groups (i.e. upscatter exits) if desired. Core geometries modelled include Cartesian and Hexagonal. Three, two and one dimensional models can be utilized with various symmetries. The non-linear iterative strategy associated with the NEM method is employed. An advantage of the non-linear iterative strategy is that NSTLE can be utilized to solve either the nodal or Finite Difference Method representation of the few-group neutron diffusion equation

  17. Spectral nodal methodology for multigroup slab-geometry discrete ordinates neutron transport problems with linearly anisotropic scattering

    Energy Technology Data Exchange (ETDEWEB)

    Oliva, Amaury M.; Filho, Hermes A.; Silva, Davi M.; Garcia, Carlos R., E-mail: aoliva@iprj.uerj.br, E-mail: halves@iprj.uerj.br, E-mail: davijmsilva@yahoo.com.br, E-mail: cgh@instec.cu [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Instituto Politecnico. Departamento de Modelagem Computacional; Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba)

    2017-07-01

    In this paper, we propose a numerical methodology for the development of a method of the spectral nodal class that will generate numerical solutions free from spatial truncation errors. This method, denominated Spectral Deterministic Method (SDM), is tested as an initial study of the solutions (spectral analysis) of neutron transport equations in the discrete ordinates (S{sub N}) formulation, in one-dimensional slab geometry, multigroup approximation, with linearly anisotropic scattering, considering homogeneous and heterogeneous domains with fixed source. The unknowns in the methodology are the cell-edge, and cell average angular fluxes, the numerical values calculated for these quantities coincide with the analytic solution of the equations. These numerical results are shown and compared with the traditional ne- mesh method Diamond Difference (DD) and the coarse-mesh method spectral Green's function (SGF) to illustrate the method's accuracy and stability. The solution algorithms problems are implemented in a computer simulator made in C++ language, the same that was used to generate the results of the reference work. (author)

  18. A highly efficient parallel algorithm for solving the neutron diffusion nodal equations on shared-memory computers

    International Nuclear Information System (INIS)

    Azmy, Y.Y.; Kirk, B.L.

    1990-01-01

    Modern parallel computer architectures offer an enormous potential for reducing CPU and wall-clock execution times of large-scale computations commonly performed in various applications in science and engineering. Recently, several authors have reported their efforts in developing and implementing parallel algorithms for solving the neutron diffusion equation on a variety of shared- and distributed-memory parallel computers. Testing of these algorithms for a variety of two- and three-dimensional meshes showed significant speedup of the computation. Even for very large problems (i.e., three-dimensional fine meshes) executed concurrently on a few nodes in serial (nonvector) mode, however, the measured computational efficiency is very low (40 to 86%). In this paper, the authors present a highly efficient (∼85 to 99.9%) algorithm for solving the two-dimensional nodal diffusion equations on the Sequent Balance 8000 parallel computer. Also presented is a model for the performance, represented by the efficiency, as a function of problem size and the number of participating processors. The model is validated through several tests and then extrapolated to larger problems and more processors to predict the performance of the algorithm in more computationally demanding situations

  19. Nodal deterministic simulation for problems of neutron shielding in multigroup formulation

    International Nuclear Information System (INIS)

    Baptista, Josue Costa; Heringer, Juan Diego dos Santos; Santos, Luiz Fernando Trindade; Alves Filho, Hermes

    2013-01-01

    In this paper, we propose the use of some computational tools, with the implementation of numerical methods SGF (Spectral Green's Function), making use of a deterministic model of transport of neutral particles in the study and analysis of a known and simplified problem of nuclear engineering, known in the literature as a problem of neutron shielding, considering the model with two energy groups. These simulations are performed in MatLab platform, version 7.0, and are presented and developed with the help of a Computer Simulator providing a friendly computer application for their utilities

  20. An accurate solution of point reactor neutron kinetics equations of multi-group of delayed neutrons

    International Nuclear Information System (INIS)

    Yamoah, S.; Akaho, E.H.K.; Nyarko, B.J.B.

    2013-01-01

    Highlights: ► Analytical solution is proposed to solve the point reactor kinetics equations (PRKE). ► The method is based on formulating a coefficient matrix of the PRKE. ► The method was applied to solve the PRKE for six groups of delayed neutrons. ► Results shows good agreement with other traditional methods in literature. ► The method is accurate and efficient for solving the point reactor kinetics equations. - Abstract: The understanding of the time-dependent behaviour of the neutron population in a nuclear reactor in response to either a planned or unplanned change in the reactor conditions is of great importance to the safe and reliable operation of the reactor. In this study, an accurate analytical solution of point reactor kinetics equations with multi-group of delayed neutrons for specified reactivity changes is proposed to calculate the change in neutron density. The method is based on formulating a coefficient matrix of the homogenous differential equations of the point reactor kinetics equations and calculating the eigenvalues and the corresponding eigenvectors of the coefficient matrix. A small time interval is chosen within which reactivity relatively stays constant. The analytical method was applied to solve the point reactor kinetics equations with six-groups delayed neutrons for a representative thermal reactor. The problems of step, ramp and temperature feedback reactivities are computed and the results compared with other traditional methods. The comparison shows that the method presented in this study is accurate and efficient for solving the point reactor kinetics equations of multi-group of delayed neutrons

  1. Inverse kinetics for subcritical systems with external neutron source

    International Nuclear Information System (INIS)

    Carvalho Gonçalves, Wemerson de; Martinez, Aquilino Senra; Carvalho da Silva, Fernando

    2017-01-01

    Highlights: • It was developed formalism for reactivity calculation. • The importance function is related to the system subcriticality. • The importance function is also related with the value of the external source. • The equations were analyzed for seven different levels of sub criticality. • The results are physically consistent with others formalism discussed in the paper. - Abstract: Nuclear reactor reactivity is one of the most important properties since it is directly related to the reactor control during the power operation. This reactivity is influenced by the neutron behavior in the reactor core. The time-dependent neutrons behavior in response to any change in material composition is important for the reactor operation safety. Transient changes may occur during the reactor startup or shutdown and due to accidental disturbances of the reactor operation. Therefore, it is very important to predict the time-dependent neutron behavior population induced by changes in neutron multiplication. Reactivity determination in subcritical systems driven by an external neutron source can be obtained through the solution of the inverse kinetics equation for subcritical nuclear reactors. The main purpose of this paper is to find the solution of the inverse kinetics equation the main purpose of this paper is to device the inverse kinetics equations for subcritical systems based in a previous paper published by the authors (Gonçalves et al., 2015) and by (Gandini and Salvatores, 2002; Dulla et al., 2006). The solutions of those equations were also obtained. Formulations presented in this paper were tested for seven different values of k eff with external neutrons source constant in time and for a powers ratio varying exponentially over time.

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

    International Nuclear Information System (INIS)

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

    2015-09-01

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

  3. Neutron detection using a current biased kinetic inductance detector

    Energy Technology Data Exchange (ETDEWEB)

    Shishido, Hiroaki, E-mail: shishido@pe.osakafu-u.ac.jp; Miyajima, Shigeyuki; Ishida, Takekazu [Department of Physics and Electronics, Graduate School of Engineering, Osaka Prefecture University, Sakai, Osaka 599-8531 (Japan); Institute for Nanofabrication Research, Osaka Prefecture University, Sakai, Osaka 599-8531 (Japan); Narukami, Yoshito [Department of Physics and Electronics, Graduate School of Engineering, Osaka Prefecture University, Sakai, Osaka 599-8531 (Japan); Oikawa, Kenichi; Harada, Masahide; Oku, Takayuki; Arai, Masatoshi [Materials and Life Science Division, J-PARC Center, Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Hidaka, Mutsuo [National Institute of Advanced Industrial Science and Technology, Tsukuba, Ibaraki 305-8568 (Japan); Fujimaki, Akira [Department of Quantum Engineering, Nagoya University, Nagoya, Aichi 464-8603 (Japan)

    2015-12-07

    We demonstrate neutron detection using a solid state superconducting current biased kinetic inductance detector (CB-KID), which consists of a superconducting Nb meander line of 1 μm width and 40 nm thickness. {sup 10}B-enriched neutron absorber layer of 150 nm thickness is placed on top of the CB-KID. Our neutron detectors are able to operate in a wide superconducting region in the bias current–temperature diagram. This is in sharp contrast with our preceding current-biased transition edge detector, which can operate only in a narrow range just below the superconducting critical temperature. The full width at half maximum of the signals remains of the order of a few tens of ns, which confirms the high speed operation of our detectors.

  4. Neutron detection using a current biased kinetic inductance detector

    International Nuclear Information System (INIS)

    Shishido, Hiroaki; Miyajima, Shigeyuki; Ishida, Takekazu; Narukami, Yoshito; Oikawa, Kenichi; Harada, Masahide; Oku, Takayuki; Arai, Masatoshi; Hidaka, Mutsuo; Fujimaki, Akira

    2015-01-01

    We demonstrate neutron detection using a solid state superconducting current biased kinetic inductance detector (CB-KID), which consists of a superconducting Nb meander line of 1 μm width and 40 nm thickness. 10 B-enriched neutron absorber layer of 150 nm thickness is placed on top of the CB-KID. Our neutron detectors are able to operate in a wide superconducting region in the bias current–temperature diagram. This is in sharp contrast with our preceding current-biased transition edge detector, which can operate only in a narrow range just below the superconducting critical temperature. The full width at half maximum of the signals remains of the order of a few tens of ns, which confirms the high speed operation of our detectors

  5. Neutron kinetics in moderators and SNM detection through epithermal-neutron-induced fissions

    Energy Technology Data Exchange (ETDEWEB)

    Gozani, Tsahi, E-mail: tgmaven@gmail.com [1050 Harriet St., Palo Alto, CA 94301 (United States); King, Michael J. [Rapiscan Laboratories Inc., 520 Almanor Ave., Sunnyvale, CA 94085 (United States)

    2016-01-01

    Extension of the well-established Differential Die Away Analysis (DDAA) into a faster time domain, where more penetrating epithermal neutrons induce fissions, is proposed and demonstrated via simulations and experiments. In the proposed method the fissions stimulated by thermal, epithermal and even higher-energy neutrons are measured after injection of a narrow pulse of high-energy 14 MeV (d,T) or 2.5 MeV (d,D) source neutrons, appropriately moderated. The ability to measure these fissions stems from the inherent correlation of neutron energy and time (“E–T” correlation) during the process of slowing down of high-energy source neutrons in common moderating materials such as hydrogenous compounds (e.g., polyethylene), heavy water, beryllium and graphite. The kinetic behavior following injection of a delta-function-shaped pulse (in time) of 14 MeV neutrons into such moderators is studied employing MCNPX simulations and, when applicable, some simple “one-group” models. These calculations served as a guide for the design of a source moderator which was used in experiments. Qualitative relationships between slowing-down time after the pulse and the prevailing neutron energy are discussed. A laboratory system consisting of a 14 MeV neutron generator, a polyethylene-reflected Be moderator, a liquid scintillator with pulse-shape discrimination (PSD) and a two-parameter E–T data acquisition system was set up to measure prompt neutron and delayed gamma-ray fission signatures in a 19.5% enriched LEU sample. The measured time behavior of thermal and epithermal neutron fission signals agreed well with the detailed simulations. The laboratory system can readily be redesigned and deployed as a mobile inspection system for SNM in, e.g., cars and vans. A strong pulsed neutron generator with narrow pulse (<75 ns) at a reasonably high pulse frequency could make the high-energy neutron induced fission modality a realizable SNM detection technique.

  6. An analytical approach for a nodal formulation of a two-dimensional fixed-source neutron transport problem in heterogeneous medium

    Energy Technology Data Exchange (ETDEWEB)

    Basso Barichello, Liliane; Dias da Cunha, Rudnei [Universidade Federal do Rio Grande do Sul, Porto Alegre, RS (Brazil). Inst. de Matematica; Becker Picoloto, Camila [Universidade Federal do Rio Grande do Sul, Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica; Tres, Anderson [Universidade Federal do Rio Grande do Sul, Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Matematica Aplicada

    2015-05-15

    A nodal formulation of a fixed-source two-dimensional neutron transport problem, in Cartesian geometry, defined in a heterogeneous medium, is solved by an analytical approach. Explicit expressions, in terms of the spatial variables, are derived for averaged fluxes in each region in which the domain is subdivided. The procedure is an extension of an analytical discrete ordinates method, the ADO method, for the solution of the two-dimensional homogeneous medium case. The scheme is developed from the discrete ordinates version of the two-dimensional transport equation along with the level symmetric quadrature scheme. As usual for nodal schemes, relations between the averaged fluxes and the unknown angular fluxes at the contours are introduced as auxiliary equations. Numerical results are in agreement with results available in the literature.

  7. A procedure for solving the neutron diffusion equation on a parallel micro-processor; modifications to the nodal expansion codes RECNEC and HEXNEC to implement the procedure

    International Nuclear Information System (INIS)

    Putney, J.M.

    1983-05-01

    The characteristics of a simple parallel micro-processor (PMP) are reviewed and its software requirements discussed. One of the more immediate applications is the multi-spatial simulation of a nuclear reactor station. This is of particular interest because 3D reactor simulation might then be possible as part of operating procedure for PFR and CDFR. A major part of a multi-spatial reactor simulator is the solution of the neutron diffusion equation. A procedure is described for solving the equation on a PMP, which is applied to the nodal expansion method with modifications to the nodal expansion codes RECNEC and HEXNEC. Estimations of the micro-processor requirements for the simulation of both PFR and CDFR are given. (U.K.)

  8. Study of the funtionalization of nodal cross sections in multigrupos for neutronics-thermohydraulic PWR core 3D calculations

    International Nuclear Information System (INIS)

    Sanchez-Cervera, S.; Hueso, C.; Herrero, J. J.

    2011-01-01

    This paper contains the work developed to study the dependencies of the nodal parameters with local variables. After entering the parameter space of operation, are obtained constants homogenized through calculations with deterministic code of transport NEWT with SCALE system codes.

  9. TUTANK a two-dimensional neutron kinetics code

    International Nuclear Information System (INIS)

    Watts, M.G.; Halsall, M.J.; Fayers, F.J.

    1975-04-01

    TUTANK is a two-dimensional neutron kinetics code which treats two neutron energy groups and up to six groups of delayed neutron precursors. A 'theta differencing' method is used to integrate the time dependence of the equations. A position dependent exponential transformation on the time variable is available as an option, which in many circumstances can remove much of the time dependence, and thereby allow longer time steps to be taken. A further manipulation is made to separate the solutions of the neutron fluxes and the precursor concentrations. The spatial equations are based on standard diffusion theory, and their solution is obtained from alternating direction sweeps with a transverse buckling - the so-called ADI-B 2 method. Other features of the code include an elementary temperature feedback and heat removal treatment, automatic time step adjustment, a flexible method of specifying cross-section and heat transfer coefficient variations during a transient, and a restart facility which requires a minimal data specification. Full details of the code input are given. An example of the solution of a NEACRP benchmark for an LWR control rod withdrawal is given. (author)

  10. A 3D nodal mixed dual method for nuclear reactor kinetics with improved quasistatic model and a semi-implicit scheme to solve the precursor equations

    International Nuclear Information System (INIS)

    Dahmani, M.; Baudron, A.M.; Lautard, J.J.; Erradi, L.

    2001-01-01

    The mixed dual nodal method MINOS is used to solve the reactor kinetics equations with improved quasistatic IQS model and the θ method is used to solve the precursor equations. The speed of calculation which is the main advantage of the MINOS method and the possibility to use the large time step for shape flux calculation permitted by the IQS method, allow us to reduce considerably the computing time. The IQS/MINOS method is implemented in CRONOS 3D reactor code. Numerical tests on different transient benchmarks show that the results obtained with the IQS/MINOS method and the direct numerical method used to solve the kinetics equations, are very close and the total computing time is largely reduced

  11. The validation of neutron kinetic calculations of CEGB reactors

    International Nuclear Information System (INIS)

    Emmett, J.C.A.; Hutt, P.K.; Nunn, D.L.; Waterson, R.H.

    1982-01-01

    Reactor kinetic calculations are required by the CEGB to predict space and time varying neutron fluxes through the course of various hypothesized core transients. These transients arise through flow or reactivity perturbations occurring in a part of the core. A description is given of the results of dual programmes of work undertaken at BNL to validate such calculations. Firstly, analyses have been carried out to establish how data for these calculations should best be derived. Secondly, experimental measurements have been compared against the predictions of such calculations with data derived in the recommended way. (author)

  12. Application of the nodal method RTN-0 for the solution of the neutron diffusion equation dependent of time in hexagonal-Z geometry; Aplicacion del metodo nodal RTN-0 para la solucion de la ecuacion de difusion de neutrones dependiente del tiempo en geometria hexagonal-Z

    Energy Technology Data Exchange (ETDEWEB)

    Esquivel E, J.; Alonso V, G. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E., E-mail: jaime.esquivel@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, 07738 Ciudad de Mexico (Mexico)

    2015-09-15

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

  13. Point Genetics: A New Concept to Assess Neutron Kinetics

    International Nuclear Information System (INIS)

    Klein Meulekamp, R.; Kuijper, J.C.; Schikorr, M.

    2005-01-01

    Point genetic equations are introduced. These equations are similar to the well-known point kinetic equations but characterize and couple individual fission generations in subcritical systems. Point genetic equations are able to describe dynamic behavior of source-driven subcritical systems on shorter timescales than is possible with point kinetic equations. Point genetic parameters can be used as a first-order characterization of the system and can be calculated using standard Monte Carlo techniques; the implementation in other calculational schemes seems straightforward. A Godiva sphere is considered to show the applicability of the point genetic equations in describing a detector response on short timescales. For this system the point genetic parameters are calculated and compared with reference calculations. Typical dynamic source behavior is considered by studying a transient in which the neutron source energy decreases from 20 to 1 MeV. For all cases studied, the point genetic equations are compared to full space-time kinetic solutions, and it is shown that point genetics performs well

  14. A Point Kinetics Model for Estimating Neutron Multiplication of Bare Uranium Metal in Tagged Neutron Measurements

    International Nuclear Information System (INIS)

    Tweardy, Matthew C.; McConchie, Seth; Hayward, Jason P.

    2017-01-01

    An extension of the point kinetics model is developed in this paper to describe the neutron multiplicity response of a bare uranium object under interrogation by an associated particle imaging deuterium-tritium (D-T) measurement system. This extended model is used to estimate the total neutron multiplication of the uranium. Both MCNPX-PoliMi simulations and data from active interrogation measurements of highly enriched and depleted uranium geometries are used to evaluate the potential of this method and to identify the sources of systematic error. The detection efficiency correction for measured coincidence response is identified as a large source of systematic error. If the detection process is not considered, results suggest that the method can estimate total multiplication to within 13% of the simulated value. Values for multiplicity constants in the point kinetics equations are sensitive to enrichment due to (n, xn) interactions by D-T neutrons and can introduce another significant source of systematic bias. This can theoretically be corrected if isotopic composition is known a priori. Finally, the spatial dependence of multiplication is also suspected of introducing further systematic bias for high multiplication uranium objects.

  15. A Point Kinetics Model for Estimating Neutron Multiplication of Bare Uranium Metal in Tagged Neutron Measurements

    Science.gov (United States)

    Tweardy, Matthew C.; McConchie, Seth; Hayward, Jason P.

    2017-07-01

    An extension of the point kinetics model is developed to describe the neutron multiplicity response of a bare uranium object under interrogation by an associated particle imaging deuterium-tritium (D-T) measurement system. This extended model is used to estimate the total neutron multiplication of the uranium. Both MCNPX-PoliMi simulations and data from active interrogation measurements of highly enriched and depleted uranium geometries are used to evaluate the potential of this method and to identify the sources of systematic error. The detection efficiency correction for measured coincidence response is identified as a large source of systematic error. If the detection process is not considered, results suggest that the method can estimate total multiplication to within 13% of the simulated value. Values for multiplicity constants in the point kinetics equations are sensitive to enrichment due to (n, xn) interactions by D-T neutrons and can introduce another significant source of systematic bias. This can theoretically be corrected if isotopic composition is known a priori. The spatial dependence of multiplication is also suspected of introducing further systematic bias for high multiplication uranium objects.

  16. An Explicit Finite Difference scheme for numerical solution of fractional neutron point kinetic equation

    International Nuclear Information System (INIS)

    Saha Ray, S.; Patra, A.

    2012-01-01

    Highlights: ► In this paper fractional neutron point kinetic equation has been analyzed. ► The numerical solution for fractional neutron point kinetic equation is obtained. ► Explicit Finite Difference Method has been applied. ► Supercritical reactivity, critical reactivity and subcritical reactivity analyzed. ► Comparison between fractional and classical neutron density is presented. - Abstract: In the present article, a numerical procedure to efficiently calculate the solution for fractional point kinetics equation in nuclear reactor dynamics is investigated. The Explicit Finite Difference Method is applied to solve the fractional neutron point kinetic equation with the Grunwald–Letnikov (GL) definition (). Fractional Neutron Point Kinetic Model has been analyzed for the dynamic behavior of the neutron motion in which the relaxation time associated with a variation in the neutron flux involves a fractional order acting as exponent of the relaxation time, to obtain the best operation of a nuclear reactor dynamics. Results for neutron dynamic behavior for subcritical reactivity, supercritical reactivity and critical reactivity and also for different values of fractional order have been presented and compared with the classical neutron point kinetic (NPK) equation as well as the results obtained by the learned researchers .

  17. Nodal methods in numerical reactor calculations

    International Nuclear Information System (INIS)

    Hennart, J.P.; Valle, E. del

    2004-01-01

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

  18. Nodal methods in numerical reactor calculations

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

  19. Application of the fractional neutron point kinetic equation: Start-up of a nuclear reactor

    International Nuclear Information System (INIS)

    Polo-Labarrios, M.-A.; Espinosa-Paredes, G.

    2012-01-01

    Highlights: ► Neutron density behavior at reactor start up with fractional neutron point kinetics. ► There is a relaxation time associated with a rapid variation in the neutron flux. ► Physical interpretation of the fractional order is related with non-Fickian effects. ► Effect of the anomalous diffusion coefficient and the relaxation time is analyzed. ► Neutron density is related with speed and duration of the control rods lifting. - Abstract: In this paper we present the behavior of the variation of neutron density when the nuclear reactor power is increased using the fractional neutron point kinetic (FNPK) equation with a single-group of delayed neutron precursor. It is considered that there is a relaxation time associated with a rapid variation in the neutron flux and its physical interpretation of the fractional order is related with non-Fickian effects from the neutron diffusion equation point of view. We analyzed the case of increase the nuclear reactor power when reactor is cold start-up which is a process of inserting reactivity by lifting control rods discontinuously. The results show that for short time scales of the start-up the neutronic density behavior with FNPK shows sub-diffusive effects whose absorption are government by control rods velocity. For large times scale, the results shows that the classical equation of the neutron point kinetics over predicted the neutron density regarding to FNPK.

  20. A New Coupled CFD/Neutron Kinetics System for High Fidelity Simulations of LWR Core Phenomena: Proof of Concept

    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.

  1. Effects of Neutron Emission on Fragment Mass and Kinetic Energy Distribution from Thermal Neutron-Induced Fission of 235U

    International Nuclear Information System (INIS)

    Montoya, M.; Rojas, J.; Saetone, E.

    2007-01-01

    The mass and kinetic energy distribution of nuclear fragments from thermal neutron-induced fission of 235 U(n th ,f) have been studied using a Monte-Carlo simulation. Besides reproducing the pronounced broadening in the standard deviation of the kinetic energy at the final fragment mass number around m = 109, our simulation also produces a second broadening around m = 125. These results are in good agreement with the experimental data obtained by Belhafaf et al. and other results on yield of mass. We conclude that the obtained results are a consequence of the characteristics of the neutron emission, the sharp variation in the primary fragment kinetic energy and mass yield curves. We show that because neutron emission is hazardous to make any conclusion on primary quantities distribution of fragments from experimental results on final quantities distributions

  2. An object-oriented 3D nodal finite element solver for neutron transport calculations in the Descartes project

    Energy Technology Data Exchange (ETDEWEB)

    Akherraz, B.; Lautard, J.J. [CEA Saclay, Dept. Modelisation de Systemes et Structures, Serv. d' Etudes des Reacteurs et de Modelisation Avancee (DMSS/SERMA), 91 - Gif sur Yvette (France); Erhard, P. [Electricite de France (EDF), Dir. de Recherche et Developpement, Dept. Sinetics, 92 - Clamart (France)

    2003-07-01

    In this paper we present two applications of the Nodal finite elements developed by Hennart and del Valle, first to three-dimensional Cartesian meshes and then to two-dimensional Hexagonal meshes. This work has been achieved within the framework of the DESCARTES project, which is a co-development effort by the 'Commissariat a l'Energie Atomique' (CEA) and 'Electricite de France' (EDF) for the development of a toolbox for reactor core calculations based on object oriented programming. The general structure of this project is based on the object oriented method. By using a mapping technique proposed in Schneider's thesis and del Valle, Mund, we show how this structuration allows us an easy implementation of the hexagonal case from the Cartesian case. The main attractiveness of this methodology is the possibility of a pin-by-pin representation by division of each lozenge into smaller ones. Furthermore, we will explore the use of non structured quadrangles to treat the circular geometry within a hexagon. It remains nevertheless, in the hexagonal case, the implementation of the acceleration of the internal iterations by the DSA (Diffusion Synthetic Acceleration) or the TSA. (authors)

  3. The solution of the neutron point kinetics equation with stochastic extension: an analysis of two moments

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Milena Wollmann da; Vilhena, Marco Tullio M.B.; Bodmann, Bardo Ernst J.; Vasques, Richard, E-mail: milena.wollmann@ufrgs.br, E-mail: vilhena@mat.ufrgs.br, E-mail: bardobodmann@ufrgs.br, E-mail: richard.vasques@fulbrightmail.org [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica

    2015-07-01

    The neutron point kinetics equation, which models the time-dependent behavior of nuclear reactors, is often used to understand the dynamics of nuclear reactor operations. It consists of a system of coupled differential equations that models the interaction between (i) the neutron population; and (II) the concentration of the delayed neutron precursors, which are radioactive isotopes formed in the fission process that decay through neutron emission. These equations are deterministic in nature, and therefore can provide only average values of the modeled populations. However, the actual dynamical process is stochastic: the neutron density and the delayed neutron precursor concentrations vary randomly with time. To address this stochastic behavior, Hayes and Allen have generalized the standard deterministic point kinetics equation. They derived a system of stochastic differential equations that can accurately model the random behavior of the neutron density and the precursor concentrations in a point reactor. Due to the stiffness of these equations, this system was numerically implemented using a stochastic piecewise constant approximation method (Stochastic PCA). Here, we present a study of the influence of stochastic fluctuations on the results of the neutron point kinetics equation. We reproduce the stochastic formulation introduced by Hayes and Allen and compute Monte Carlo numerical results for examples with constant and time-dependent reactivity, comparing these results with stochastic and deterministic methods found in the literature. Moreover, we introduce a modified version of the stochastic method to obtain a non-stiff solution, analogue to a previously derived deterministic approach. (author)

  4. The solution of the neutron point kinetics equation with stochastic extension: an analysis of two moments

    International Nuclear Information System (INIS)

    Silva, Milena Wollmann da; Vilhena, Marco Tullio M.B.; Bodmann, Bardo Ernst J.; Vasques, Richard

    2015-01-01

    The neutron point kinetics equation, which models the time-dependent behavior of nuclear reactors, is often used to understand the dynamics of nuclear reactor operations. It consists of a system of coupled differential equations that models the interaction between (i) the neutron population; and (II) the concentration of the delayed neutron precursors, which are radioactive isotopes formed in the fission process that decay through neutron emission. These equations are deterministic in nature, and therefore can provide only average values of the modeled populations. However, the actual dynamical process is stochastic: the neutron density and the delayed neutron precursor concentrations vary randomly with time. To address this stochastic behavior, Hayes and Allen have generalized the standard deterministic point kinetics equation. They derived a system of stochastic differential equations that can accurately model the random behavior of the neutron density and the precursor concentrations in a point reactor. Due to the stiffness of these equations, this system was numerically implemented using a stochastic piecewise constant approximation method (Stochastic PCA). Here, we present a study of the influence of stochastic fluctuations on the results of the neutron point kinetics equation. We reproduce the stochastic formulation introduced by Hayes and Allen and compute Monte Carlo numerical results for examples with constant and time-dependent reactivity, comparing these results with stochastic and deterministic methods found in the literature. Moreover, we introduce a modified version of the stochastic method to obtain a non-stiff solution, analogue to a previously derived deterministic approach. (author)

  5. The total kinetic energy release in the fast neutron-induced fission of {sup 232}Th

    Energy Technology Data Exchange (ETDEWEB)

    King, Jonathan; Yanez, Ricardo; Loveland, Walter; Barrett, J. Spencer; Oscar, Breland [Oregon State University, Dept. of Chemistry, Corvallis, OR (United States); Fotiades, Nikolaos; Tovesson, Fredrik; Young Lee, Hye [Los Alamos National Laboratory, Physics Division, Los Alamos, NM (United States)

    2017-12-15

    The post-emission total kinetic energy release (TKE) in the neutron-induced fission of {sup 232}Th was measured (using white spectrum neutrons from LANSCE) for neutron energies from E{sub n} = 3 to 91 MeV. In this energy range the average post-neutron total kinetic energy release decreases from 162.3 ± 0.3 at E{sub n} = 3 MeV to 154.9 ± 0.3 MeV at E{sub n} = 91 MeV. Analysis of the fission mass distributions indicates that the decrease in TKE with increasing neutron energy is a combination of increasing yields of symmetric fission (which has a lower associated TKE) and a decrease in the TKE release in asymmetric fission. (orig.)

  6. Kinetic simulation of neutron production in a deuterium z-pinch

    International Nuclear Information System (INIS)

    Mostrom, C.; Stygar, William A.; Thoma, Carsten; Welch, Dale Robert; Clark, R.E.; Leeper, Ramon Joe; Rose, David V.

    2010-01-01

    We have found computationally that, at sufficiently high currents, half of the neutrons produced by a deuterium z pinch are thermonuclear in origin. Early experiments below 1-MA current found that essentially all of the neutrons produced by a deuterium pinch are not thermonuclear, but are initiated by an instability that creates beam-target neutrons. Many subsequent authors have supported this result while others have claimed that pinch neutrons are thermonuclear. To resolve this issue, we have conducted fully kinetic, collisional, and electromagnetic simulations of the complete time evolution of a deuterium pinch. We find that at 1-MA pinch currents, most of the neutrons are, indeed, beam-target in origin. At much higher current, half of the neutrons are thermonuclear and half are beam-target driven by instabilities that produce a power law fall off in the ion energy distribution function at large energy. The implications for fusion energy production with such pinches are discussed.

  7. Space-time reactor kinetics for heterogeneous reactor structure

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1969-11-15

    An attempt is made to formulate time dependent diffusion equation based on Feinberg-Galanin theory in the from analogue to the classical reactor kinetic equation. Parameters of these equations could be calculated using the existing codes for static reactor calculation based on the heterogeneous reactor theory. The obtained kinetic equation could be analogues in form to the nodal kinetic equation. Space-time distribution of neutron flux in the reactor can be obtained by solving these equations using standard methods.

  8. An analytical solution for the two-group kinetic neutron diffusion equation in cylindrical geometry

    International Nuclear Information System (INIS)

    Fernandes, Julio Cesar L.; Vilhena, Marco Tullio; Bodmann, Bardo Ernst

    2011-01-01

    Recently the two-group Kinetic Neutron Diffusion Equation with six groups of delay neutron precursor in a rectangle was solved by the Laplace Transform Technique. In this work, we report on an analytical solution for this sort of problem but in cylindrical geometry, assuming a homogeneous and infinite height cylinder. The solution is obtained applying the Hankel Transform to the Kinetic Diffusion equation and solving the transformed problem by the same procedure used in the rectangle. We also present numerical simulations and comparisons against results available in literature. (author)

  9. Solution for the multigroup neutron space kinetics equations by the modified Picard algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Tavares, Matheus G.; Petersen, Claudio Z., E-mail: matheus.gulartetavares@gmail.com [Universidade Federal de Pelotas (UFPEL), Capao do Leao, RS (Brazil). Departamento de Matematica e Estatistica; Schramm, Marcelo, E-mail: schrammmarcelo@gmail.com [Universidade Federal de Pelotas (UFPEL), RS (Brazil). Centro de Engenharias; Zanette, Rodrigo, E-mail: rodrigozanette@hotmail.com [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Instituto de Matematica e Estatistica

    2017-07-01

    In this work, we used a modified Picards method to solve the Multigroup Neutron Space Kinetics Equations (MNSKE) in Cartesian geometry. The method consists in assuming an initial guess for the neutron flux and using it to calculate a fictitious source term in the MNSKE. A new source term is calculated applying its solution, and so on, iteratively, until a stop criterion is satisfied. For the solution of the fast and thermal neutron fluxes equations, the Laplace Transform technique is used in time variable resulting in a rst order linear differential matrix equation, which are solved by classical methods in the literature. After each iteration, the scalar neutron flux and the delayed neutron precursors are reconstructed by polynomial interpolation. We obtain the fluxes and precursors through Numerical Inverse Laplace Transform using the Stehfest method. We present numerical simulations and comparisons with available results in literature. (author)

  10. Solution for the multigroup neutron space kinetics equations by the modified Picard algorithm

    International Nuclear Information System (INIS)

    Tavares, Matheus G.; Petersen, Claudio Z.; Schramm, Marcelo; Zanette, Rodrigo

    2017-01-01

    In this work, we used a modified Picards method to solve the Multigroup Neutron Space Kinetics Equations (MNSKE) in Cartesian geometry. The method consists in assuming an initial guess for the neutron flux and using it to calculate a fictitious source term in the MNSKE. A new source term is calculated applying its solution, and so on, iteratively, until a stop criterion is satisfied. For the solution of the fast and thermal neutron fluxes equations, the Laplace Transform technique is used in time variable resulting in a rst order linear differential matrix equation, which are solved by classical methods in the literature. After each iteration, the scalar neutron flux and the delayed neutron precursors are reconstructed by polynomial interpolation. We obtain the fluxes and precursors through Numerical Inverse Laplace Transform using the Stehfest method. We present numerical simulations and comparisons with available results in literature. (author)

  11. An In-Pile Kinetic Method for Determining the Delayed Neutron Fraction βeff

    International Nuclear Information System (INIS)

    Gilad, E.; Rivin, O.; Ettedgui, H.; Yaar, I.; Geslot, B.; Pepino, A.; Di Salvo, J.; Gruel, A.; Blaise, P.

    2014-01-01

    Delayed neutrons are of fundamental importance in the field of nuclear reactor dynamics and control. Although only a small fraction of the neutrons emitted by fission are not prompt, the knowledge of the delayed neutrons parameters is essential for transient analysis, such as startup or shutdown of the reactor, as well as for accidents analysis and control system design [1]. One of the main delayed neutron parameters used in the point reactor model equations is the effective delayed neutron fraction, which incorporates both delayed neutron spectral properties and core geometrical configuration [1,2]. Additional delayed neutron parameters include the fraction of fission neutrons emitted in each delayed group, and the delayed neutron precursors decay constants . Experimental efforts aimed at determining the value ofβ, which provide experimental support for the evaluation of delayed neutron parameters, are extremely valuable. This is due to the fact that unlike other fields in reactor physics, e.g. criticality safety or shielding, the availability of experimental data and benchmark problems for validating delayed neutron parameters and its implementation in different models is highly limited. Furthermore, the existing experimental data exhibit significant discrepancies between the different sets of parameter, which lead to substantial disparity in the analysis of kinetic experiments and reactor dynamic behavior]. In this work, a method for determining the effective delayed neutron fraction using in-pile reactivity oscillation and Fourier analysis is presented. The method is based on measurements of the reactor's power response to small periodic in-pile reactivity perturbations and utilizes Fourier analysis for reconstruction of the reactor zero power transfer function. Knowledge of the reactor transfer function enables the estimation of theβ value using multi-parameter nonlinear fit. The method accounts for higher harmonics, which are excited by the trapezoidal

  12. A calculational study on neutron kinetics and thermodynamics of a gaseous core fission reactor

    International Nuclear Information System (INIS)

    Kuijper, J.C.

    1992-06-01

    A numerical and analytical study of the static and dynamic properties of a GCFR with oscillating fuel gas (uranium and carbon fluorides) is presented. Neutron kinetics parts of combined GCFR models are introduced. Thermodynamic properties of the GCFR and of the fuel gas are treated. (HP)

  13. Delayed neutron kinetic functions for /sup 232/Th and /sup 238/U mixtures

    Energy Technology Data Exchange (ETDEWEB)

    Ganich, P P; Goshovskij, M V; Lendel, A I; Lomonosov, V I; Sikora, D I; Sychev, S I

    1984-11-01

    In order to investigate the applicability of the method based on using kinetic functions, describing the emission of delayed neutrons by samples for determination of the content of fissionable nuclides in binary mixtures, the /sup 232/Th+/sup 238/U mixtures have been analyzed with the M-30 microtron. Fresh samples containing ThO/sub 2/, U/sub 3/O/sub 8/ and their mixtures are irradiated by bremstrahlung at the 15.5 MeV energy of accelerated electrons and 9 ..mu..A average current. The mass of samples is about 6 g. To determine the kinetic functions, temporal distributions of delayed neutron pulses are used, their maximum number for different samples being (1.7-3.0) x 10/sup 4/. In processing the data obtained two methods of normalization of the delayed neutron number in the kinetic functions are used: to the total yield of delayed neutrons and to the yield of /sup 133/I ..gamma..-quanta. The conclusion is drawn that the method investigated permits to determine relative /sup 238/U concentrations in the mixtures considered with 0.06-0.2 errors. Error reduction is achieved during the normalization of the number of delayed neutrons to the yield of /sup 130/I ..gamma..-quanta.

  14. Neutron emission effects on fragment mass and kinetic energy distribution from fission of 239Pu induced by thermal neutrons

    International Nuclear Information System (INIS)

    Montoya, M.; Rojas, J.; Lobato, I.

    2010-01-01

    The average of fragment kinetic energy (E-bar sign*) and the multiplicity of prompt neutrons (ν(bar sign)) as a function of fragment mass (m*), as well as the fragment mass yield (Y(m*)) from thermal neutron-induced fission of 239 Pu have been measured by Tsuchiya et al.. In that work the mass and kinetic energy are calculated from the measured kinetic energy of one fragment and the difference of time of flight of the two complementary fragments. However they do not present their results about the standard deviation σ E *(m*). In this work we have made a numerical simulation of that experiment which reproduces its results, assuming an initial distribution of the primary fragment kinetic energy (E(A)) with a constant value of the standard deviation as function of fragment mass (σ E (A)). As a result of the simulation we obtain the dependence σ E *(m*) which presents an enhancement between m* = 92 and m* = 110, and a peak at m* = 121.

  15. On the exact solution for the multi-group kinetic neutron diffusion equation in a rectangle

    International Nuclear Information System (INIS)

    Petersen, C.Z.; Vilhena, M.T.M.B. de; Bodmann, B.E.J.

    2011-01-01

    In this work we consider the two-group bi-dimensional kinetic neutron diffusion equation. The solution procedure formalism is general with respect to the number of energy groups, neutron precursor families and regions with different chemical compositions. The fast and thermal flux and the delayed neutron precursor yields are expanded in a truncated double series in terms of eigenfunctions that, upon insertion into the kinetic equation and upon taking moments, results in a first order linear differential matrix equation with source terms. We split the matrix appearing in the transformed problem into a sum of a diagonal matrix plus the matrix containing the remaining terms and recast the transformed problem into a form that can be solved in the spirit of Adomian's recursive decomposition formalism. Convergence of the solution is guaranteed by the Cardinal Interpolation Theorem. We give numerical simulations and comparisons with available results in the literature. (author)

  16. Application of Extreme Learning Machines to inverse neutron kinetics

    International Nuclear Information System (INIS)

    Picca, Paolo; Furfaro, Roberto

    2017-01-01

    Highlights: • The paper applies the Extreme Learning Machines (ELMs) to inverse reactor problems. • Multi-group transport model is used for the inversion as opposed to point kinetics. • ELMs are compared against Artificial Neural Networks (ANNs). • Various options are tested to improve the reliability of the estimation. • Results highlight the potential of the ELM approach. - Abstract: The paper presents the application of Extreme Leaning Machines (ELMs) for inverse reactor kinetic applications. ELMs were proposed by Huang and co-workers (2004, 2006a,b, 2015), which showed their enhances capabilities in terms of training speed and generalization with respect to classical Artificial Neural Networks (ANNs). ELMs are here implemented for reactivity determination as an alternative to ANNs (e.g. Picca et al. (2008)) and Gaussian Processes (Picca and Furfaro, 2012). After a review of the main features of ELMs, their application to inverse kinetic problems is proposed. The ELMs performance is tested on a typical accelerator drive system configuration (Yalina reactor) and the inversion is carried out on an accurate kinetic model (multi-group transport).

  17. An analytical solution of the one-dimensional neutron diffusion kinetic equation in cartesian geometry

    International Nuclear Information System (INIS)

    Ceolin, Celina; Vilhena, Marco T.; Petersen, Claudio Z.

    2009-01-01

    In this work we report an analytical solution for the monoenergetic neutron diffusion kinetic equation in cartesian geometry. Bearing in mind that the equation for the delayed neutron precursor concentration is a first order linear differential equation in the time variable, to make possible the application of the GITT approach to the kinetic equation, we introduce a fictitious diffusion term multiplied by a positive small value ε. By this procedure, we are able to solve this set of equations. Indeed, applying the GITT technique to the modified diffusion kinetic equation, we come out with a matrix differential equation which has a well known analytical solution when ε goes to zero. We report numerical simulations as well study of numerical convergence of the results attained. (author)

  18. Neutron kinetics of fluid-fuel systems by the quasi-static method

    International Nuclear Information System (INIS)

    Dulla, S.; Ravetto, P.; Rostagno, M.M.

    2004-01-01

    The quasi-static method for the neutron kinetics of nuclear reactors is generalized for application to neutron multiplying systems fueled by a fluid multiplying material, typically a mixture of fissile molten salts. The method is derived by the application of factorization formulae for both the neutron density and the delayed precursor concentrations and the projection of the balance equations upon a weighting function. A physically meaningful weight can be assumed as the solution of the adjoint model, which is constructed for the situation considered, including delayed neutrons. The quasi-static scheme is then applied to calculations of some transients for a typical configuration of a molten-salt reactor, in a multigroup diffusion model with a one-dimensional slug-flow velocity field. The physical features associated to the motion of the fissile material are highlighted

  19. Theory of stochastic space-dependent neutron kinetics with a Gaussian parametric excitation

    International Nuclear Information System (INIS)

    Saito, K.

    1980-01-01

    Neutron kinetics and statics in a multiplying medium with a statistically fluctuating reactivity are unified and systematically studied by applying the Novikov-Furutsu formula. The parametric or multiplicative noise is spatially distributed and of Gaussian nature with an arbitrary spectral profile. It is found that the noise introduces a new definite production term into the conventional balance equation for the mean neutron number. The term is characterized by the magnitude and the correlation function of the random excitation. Its relaxation phenomena bring forth a non-Markoffian or a memory effect, which is conceptualised by introducing 'pseudo-precursors' or 'pseudo-delayed neutrons'. By using the concept, some typical reactor physical problems are solved; they are (1) reactivity and flux perturbation originating from the random dispersal of core materials and (2) analysis of neutron decay mode and it relaxation constant, and derivation of the corresponding new inhour equation. (author)

  20. Dark Kinetic Heating of Neutron Stars and an Infrared Window on WIMPs, SIMPs, and Pure Higgsinos.

    Science.gov (United States)

    Baryakhtar, Masha; Bramante, Joseph; Li, Shirley Weishi; Linden, Tim; Raj, Nirmal

    2017-09-29

    We identify a largely model-independent signature of dark matter (DM) interactions with nucleons and electrons. DM in the local galactic halo, gravitationally accelerated to over half the speed of light, scatters against and deposits kinetic energy into neutron stars, heating them to infrared blackbody temperatures. The resulting radiation could potentially be detected by the James Webb Space Telescope, the Thirty Meter Telescope, or the European Extremely Large Telescope. This mechanism also produces optical emission from neutron stars in the galactic bulge, and x-ray emission near the galactic center because dark matter is denser in these regions. For GeV-PeV mass dark matter, dark kinetic heating would initially unmask any spin-independent or spin-dependent dark matter-nucleon cross sections exceeding 2×10^{-45}  cm^{2}, with improved sensitivity after more telescope exposure. For lighter-than-GeV dark matter, cross-section sensitivity scales inversely with dark matter mass because of Pauli blocking; for heavier-than-PeV dark matter, it scales linearly with mass as a result of needing multiple scatters for capture. Future observations of dark sector-warmed neutron stars could determine whether dark matter annihilates in or only kinetically heats neutron stars. Because inelastic interstate transitions of up to a few GeV would occur in relativistic scattering against nucleons, elusive inelastic dark matter like pure Higgsinos can also be discovered.

  1. Determination of the kinetic parameters of the CALIBAN metallic core reactor from stochastic neutron measurements

    Energy Technology Data Exchange (ETDEWEB)

    Casoli, P.; Authier, N.; Chapelle, A. [Commissariat a l' Energie Atomique et Aux Energies Alternatives, CEA, DAM, F-21120 Is sur Tille (France)

    2012-07-01

    Several experimental devices are operated by the Criticality and Neutron Science Research Dept. of the CEA Valduc Laboratory. One of these is the Caliban metallic core reactor. The purpose of this study is to develop and perform experiments allowing to determinate some of fundamental kinetic parameters of the reactor. The prompt neutron decay constant and particularly its value at criticality can be measured with reactor noise techniques such as Rossi-{alpha} and Feynman variance-to-mean methods. Subcritical, critical, and even supercritical experiments were performed. Fission chambers detectors were put nearby the core and measurements were analyzed with the Rossi-{alpha} technique. A new value of the prompt neutron decay constant at criticality was determined, which allows, using the Nelson number method, new evaluations of the effective delayed neutron fraction and the in core neutron lifetime. As an introduction of this paper, some motivations of this work are given in part 1. In part 2, principles of the noise measurements experiments performed at the CEA Valduc Laboratory are reminded. The Caliban reactor is described in part 3. Stochastic neutron measurements analysis techniques used in this study are then presented in part 4. Results of fission chamber experiments are summarized in part 5. Part 6 is devoted to the current work, improvement of the experimental device using He 3 neutron detectors and first results obtained with it. Finally, conclusions and perspectives are given in part 7. (authors)

  2. Kinetics, dynamics and neutron noise in Molten Salt Reactors

    International Nuclear Information System (INIS)

    Pazsit, Imre

    2013-01-01

    Reactor kinetic and dynamic properties of Molten Salt Reactors (MSR) are investigated in a simple model, which allows closed compact analytical solutions to be obtained. The goal is to gain insight, rather than to produce high-quality quantitative data. Through an interpretation of the different terms in the basic equations, and by means of analytical solutions, various approximations are introduced and their validity discussed. The dynamical behaviour of MSRs and their response to small stationary perturbations is described and discussed in comparison with traditional systems. (author)

  3. Development of a neutronics code based on analytic function expansion nodal method for pebble-type High Temperature Gas-cooled Reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Nam Zin; Lee, Joo Hee; Lee, Jae Jun; Yu, Hui; Lee, Gil Soo [Korea Advanced Institute of Science and Tehcnology, Daejeon (Korea, Republic of)

    2006-03-15

    There is growing interest in developing Pebble Bed Reactors(PBRs) as a candidate of Very High Temperature gas-cooled Reactors(VHTRs). Until now, most existing methods of nuclear design analysis for this type of reactors are base on old finite-difference solvers or on statistical methods. And other existing nodal cannot be adapted for this kind of reactors because of transverse integration problem. In this project, we developed the TOPS code in three dimensional cylindrical geometry based on Analytic Function Expansion Nodal (AFEN) method developed at KAIST. The TOPS code showed better results in computing time than FDM and MCNP. Also TOPS showed very accurate results in reactor analysis.

  4. Development of a neutronics code based on analytic function expansion nodal method for pebble-type High Temperature Gas-cooled Reactor design

    International Nuclear Information System (INIS)

    Cho, Nam Zin; Lee, Joo Hee; Lee, Jae Jun; Yu, Hui; Lee, Gil Soo

    2006-03-01

    There is growing interest in developing Pebble Bed Reactors(PBRs) as a candidate of Very High Temperature gas-cooled Reactors(VHTRs). Until now, most existing methods of nuclear design analysis for this type of reactors are base on old finite-difference solvers or on statistical methods. And other existing nodal cannot be adapted for this kind of reactors because of transverse integration problem. In this project, we developed the TOPS code in three dimensional cylindrical geometry based on Analytic Function Expansion Nodal (AFEN) method developed at KAIST. The TOPS code showed better results in computing time than FDM and MCNP. Also TOPS showed very accurate results in reactor analysis

  5. KEK NODAL user's guide

    International Nuclear Information System (INIS)

    Akiyama, Atsuyoshi; Katoh, Tadahiko; Kikutani, Eiji; Koiso, Haruyo; Kurokawa, Shin-ichi; Oide, Katsunobu.

    1984-06-01

    NODAL is an interpreter language for accelerator control developed at CERN SPS and has been used successfully since 1974. At present NODAL or NODAL-like languages are used at DESY PETRA and CERN CPS. At KEK, we have also adopted NODAL for the control of TRISTAN, a 30 GeV x 30 GeV electron-positron colliding beam facility. The KEK version of NODAL has the following improvements on the SPS NODAL: (1) the fast execution speed due to the compiler-interpreter scheme, and (2) the full-screen editing facility. This manual explains how to use the KEK NODAL. It is based on the manual of the SPS NODAL, THE NODAL SYSTEM FOR THE SPS, by M.C. Crowley-Milling and G.C. Shering, CERN 78-07. We have made some additions and modifications to make the manual more appropriate for the KEK NODAL system, paying attention to retaining the good features of the original SPS NODAL manual. We acknowledge Professor M.C. Crowley-Milling, Dr G.C. Shering and CERN for their kind permission for this modification. (author)

  6. KEK NODAL system

    International Nuclear Information System (INIS)

    Kurokawa, S.; Abe, K.; Akiyama, A.; Katoh, T.; Kikutani, E.; Koiso, H.; Kurihara, N.; Oide, K.; Shinomoto, M.

    1985-01-01

    The KEK NODAL system, which is based on the NODAL devised at the CERN SPS, works on an optical-fiber token ring network of twenty-four minicomputers (Hitachi HIDIC 80's) to control the TRISTAN accelerator complex, now being constructed at KEK. KEK NODAL retains main features of the original NODAL: the interpreting scheme, the multi-computer programming facility, and the data-module concept. In addition, it has the following characteristics: fast execution due to the compiler-interpreter method, a multicomputer file system, a full-screen editing facility, and a dynamic linkage scheme of data modules and NODAL functions. The structure of the KEK NODAL system under PMS, a real-time multitasking operating system of HIDIC 80, is described; the NODAL file system is also explained

  7. A new integral method for solving the point reactor neutron kinetics equations

    International Nuclear Information System (INIS)

    Li Haofeng; Chen Wenzhen; Luo Lei; Zhu Qian

    2009-01-01

    A numerical integral method that efficiently provides the solution of the point kinetics equations by using the better basis function (BBF) for the approximation of the neutron density in one time step integrations is described and investigated. The approach is based on an exact analytic integration of the neutron density equation, where the stiffness of the equations is overcome by the fully implicit formulation. The procedure is tested by using a variety of reactivity functions, including step reactivity insertion, ramp input and oscillatory reactivity changes. The solution of the better basis function method is compared to other analytical and numerical solutions of the point reactor kinetics equations. The results show that selecting a better basis function can improve the efficiency and accuracy of this integral method. The better basis function method can be used in real time forecasting for power reactors in order to prevent reactivity accidents.

  8. A benchmark for coupled thermohydraulics system/three-dimensional neutron kinetics core models

    International Nuclear Information System (INIS)

    Kliem, S.

    1999-01-01

    During the last years 3D neutron kinetics core models have been coupled to advanced thermohydraulics system codes. These coupled codes can be used for the analysis of the whole reactor system. Although the stand-alone versions of the 3D neutron kinetics core models and of the thermohydraulics system codes generally have a good verification and validation basis, there is a need for additional validation work. This especially concerns the interaction between the reactor core and the other components of a nuclear power plant (NPP). In the framework of the international 'Atomic Energy Research' (AER) association on VVER Reactor Physics and Reactor Safety, a benchmark for these code systems was defined. (orig.)

  9. One dimensional neutron kinetics in the TRAC-BF1 code

    International Nuclear Information System (INIS)

    Weaver, W.L. III; Wagner, K.C.

    1987-01-01

    The TRAC-BWR code development program at the Idaho National Engineering Laboratory is developing a version of the TRAC code for the U.S. Nuclear Regulatory Commission (USNRC) to provide a best-estimate analysis capability for the simulation of postulated accidents in boiling water reactor (BWR) power systems and related experimental facilities. Recent development efforts in the TRAC-BWR program have focused on improving the computational efficiency through the incorporation of a hybrid Courant- limit-violating numerical solution scheme in the one-dimensional component models and on improving code accuracy through the development of a one-dimensional neutron kinetics model. Many other improvements have been incorporated into TRAC-BWR to improve code portability, accuracy, efficiency, and maintainability. This paper will describe the one- dimensional neutron kinetics model, the generation of the required input data for this model, and present results of the first calculations using the model

  10. Polynomial approach method to solve the neutron point kinetics equations with use of the analytic continuation

    Energy Technology Data Exchange (ETDEWEB)

    Tumelero, Fernanda; Petersen, Claudio Zen; Goncalves, Glenio Aguiar [Universidade Federal de Pelotas, Capao do Leao, RS (Brazil). Programa de Pos Graduacao em Modelagem Matematica; Schramm, Marcelo [Universidade Federal do Rio Grande do Sul, Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica

    2016-12-15

    In this work, we report a solution to solve the Neutron Point Kinetics Equations applying the Polynomial Approach Method. The main idea is to expand the neutron density and delayed neutron precursors as a power series considering the reactivity as an arbitrary function of the time in a relatively short time interval around an ordinary point. In the first interval one applies the initial conditions and the analytical continuation is used to determine the solutions of the next intervals. A genuine error control is developed based on an analogy with the Rest Theorem. For illustration, we also report simulations for different approaches types (linear, quadratic and cubic). The results obtained by numerical simulations for linear approximation are compared with results in the literature.

  11. On a closed form approach to the fractional neutron point kinetics equation with temperature feedback

    International Nuclear Information System (INIS)

    Schramm, Marcelo; Bodmann, Bardo E.J.; Vilhena, Marco T.M.B.; Petersen, Claudio Z.; Alvim, Antonio C.M.

    2013-01-01

    Following the quest to find analytical solutions, we extend the methodology applied successfully to timely fractional neutron point kinetics (FNPK) equations by adding the effects of temperature. The FNPK equations with temperature feedback correspond to a nonlinear system and “stiff” type for the neutron density and the concentration of delayed neutron precursors. These variables determine the behavior of a nuclear reactor power with time and are influenced by the position of control rods, for example. The solutions of kinetics equations provide time information about the dynamics in a nuclear reactor in operation and are useful, for example, to understand the power fluctuations with time that occur during startup or shutdown of the reactor, due to adjustments of the control rods. The inclusion of temperature feedback in the model introduces an estimate of the transient behavior of the power and other variables, which are strongly coupled. Normally, a single value of reactivity is used across the energy spectrum. Especially in case of power change, the neutron energy spectrum changes as well as physical parameters such as the average cross sections. However, even knowing the importance of temperature effects on the control of the reactor power, the character of the set of nonlinear equations governing this system makes it difficult to obtain a purely analytical solution. Studies have been published in this sense, using numerical approaches. Here the idea is to consider temperature effects to make the model more realistic and thus solve it in a semi-analytical way. Therefore, the main objective of this paper is to obtain an analytical representation of fractional neutron point kinetics equations with temperature feedback, without having to resort to approximations inherent in numerical methods. To this end, we will use the decomposition method, which has been successfully used by the authors to solve neutron point kinetics problems. The results obtained will

  12. On a closed form approach to the fractional neutron point kinetics equation with temperature feedback

    Energy Technology Data Exchange (ETDEWEB)

    Schramm, Marcelo; Bodmann, Bardo E.J.; Vilhena, Marco T.M.B., E-mail: marceloschramm@hotmail.com, E-mail: bardo.bodmann@ufrgs.br, E-mail: mtmbvilhena@gmail.com [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Departamento de Engenharia Mecanica; Petersen, Claudio Z., E-mail: claudiopetersen@yahoo.com.br [Universidade Federal de Pelotas (UFPel), RS (Brazil). Departamento de Matematica; Alvim, Antonio C.M., E-mail: alvim@nuclear.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Instituto Alberto Luiz Coimbra de Pos-Graduacao e Pesquisa em Engenharia

    2013-07-01

    Following the quest to find analytical solutions, we extend the methodology applied successfully to timely fractional neutron point kinetics (FNPK) equations by adding the effects of temperature. The FNPK equations with temperature feedback correspond to a nonlinear system and “stiff” type for the neutron density and the concentration of delayed neutron precursors. These variables determine the behavior of a nuclear reactor power with time and are influenced by the position of control rods, for example. The solutions of kinetics equations provide time information about the dynamics in a nuclear reactor in operation and are useful, for example, to understand the power fluctuations with time that occur during startup or shutdown of the reactor, due to adjustments of the control rods. The inclusion of temperature feedback in the model introduces an estimate of the transient behavior of the power and other variables, which are strongly coupled. Normally, a single value of reactivity is used across the energy spectrum. Especially in case of power change, the neutron energy spectrum changes as well as physical parameters such as the average cross sections. However, even knowing the importance of temperature effects on the control of the reactor power, the character of the set of nonlinear equations governing this system makes it difficult to obtain a purely analytical solution. Studies have been published in this sense, using numerical approaches. Here the idea is to consider temperature effects to make the model more realistic and thus solve it in a semi-analytical way. Therefore, the main objective of this paper is to obtain an analytical representation of fractional neutron point kinetics equations with temperature feedback, without having to resort to approximations inherent in numerical methods. To this end, we will use the decomposition method, which has been successfully used by the authors to solve neutron point kinetics problems. The results obtained will

  13. Solution of the neutron point kinetics equations with temperature feedback effects applying the polynomial approach method

    Energy Technology Data Exchange (ETDEWEB)

    Tumelero, Fernanda, E-mail: fernanda.tumelero@yahoo.com.br [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica; Petersen, Claudio Z.; Goncalves, Glenio A.; Lazzari, Luana, E-mail: claudiopeteren@yahoo.com.br, E-mail: gleniogoncalves@yahoo.com.br, E-mail: luana-lazzari@hotmail.com [Universidade Federal de Pelotas (DME/UFPEL), Capao do Leao, RS (Brazil). Instituto de Fisica e Matematica

    2015-07-01

    In this work, we present a solution of the Neutron Point Kinetics Equations with temperature feedback effects applying the Polynomial Approach Method. For the solution, we consider one and six groups of delayed neutrons precursors with temperature feedback effects and constant reactivity. The main idea is to expand the neutron density, delayed neutron precursors and temperature as a power series considering the reactivity as an arbitrary function of the time in a relatively short time interval around an ordinary point. In the first interval one applies the initial conditions of the problem and the analytical continuation is used to determine the solutions of the next intervals. With the application of the Polynomial Approximation Method it is possible to overcome the stiffness problem of the equations. In such a way, one varies the time step size of the Polynomial Approach Method and performs an analysis about the precision and computational time. Moreover, we compare the method with different types of approaches (linear, quadratic and cubic) of the power series. The answer of neutron density and temperature obtained by numerical simulations with linear approximation are compared with results in the literature. (author)

  14. Solution of the neutron point kinetics equations with temperature feedback effects applying the polynomial approach method

    International Nuclear Information System (INIS)

    Tumelero, Fernanda; Petersen, Claudio Z.; Goncalves, Glenio A.; Lazzari, Luana

    2015-01-01

    In this work, we present a solution of the Neutron Point Kinetics Equations with temperature feedback effects applying the Polynomial Approach Method. For the solution, we consider one and six groups of delayed neutrons precursors with temperature feedback effects and constant reactivity. The main idea is to expand the neutron density, delayed neutron precursors and temperature as a power series considering the reactivity as an arbitrary function of the time in a relatively short time interval around an ordinary point. In the first interval one applies the initial conditions of the problem and the analytical continuation is used to determine the solutions of the next intervals. With the application of the Polynomial Approximation Method it is possible to overcome the stiffness problem of the equations. In such a way, one varies the time step size of the Polynomial Approach Method and performs an analysis about the precision and computational time. Moreover, we compare the method with different types of approaches (linear, quadratic and cubic) of the power series. The answer of neutron density and temperature obtained by numerical simulations with linear approximation are compared with results in the literature. (author)

  15. Calculation of kinetic parameters of Caliban metallic core experimental reactor from stochastic neutron measurements

    Energy Technology Data Exchange (ETDEWEB)

    Casoli, P.; Authier, N.; Baud, J. [Commissariat a l' energie Atomique, Centre de Valduc, 21120 Is-sur-Tille (France)

    2009-07-01

    Several experimental devices are operated by the Criticality and Neutron Science Research Department of the CEA Valduc Laboratory. One of these is the metallic core reactor Caliban. The knowledge of the fundamental kinetic parameters of the reactor is very useful, indeed necessary, to the operator. The purpose of this study was to develop and perform experiments allowing to determinate some of these parameters. The prompt neutron decay constant and particularly its value at criticality can be measured with reactor noise techniques such as the interval-distribution, the Feynman variance-to-mean, and the Rossi-{alpha} methods. By introducing the Nelson number, the effective delayed neutron fraction and the average neutron lifetime can also be calculated with the Rossi-{alpha} method. Subcritical, critical, and even supercritical experiments were performed. With the Rossi-{alpha} technique, it was found that the prompt neutron decay constant at criticality was (6.02*10{sup 5} {+-} 9%). Experiments also brought out the limitations of the used experimental parameters. (authors)

  16. Implosion anisotropy of neutron kinetic energy distributions as measured with the neutron time-of-flight diagnostics at the National Ignition Facility

    Science.gov (United States)

    Hartouni, Edward; Eckart, Mark; Field, John; Grim, Gary; Hatarik, Robert; Moore, Alastair; Munro, David; Sayer, Daniel; Schlossberg, David

    2017-10-01

    Neutron kinetic energy distributions from fusion reactions are characterized predominantly by the excess energy, Q, of the fusion reaction and the variance of kinetic energy which is related to the thermal temperature of the plasma as shown by e.g. Brysk. High statistics, high quality neutron time-of-flight spectra obtained at the National Ignition Facility provide a means of measuring small changes to the neutron kinetic energy due to the spatial and temporal distribution of plasma temperature, density and velocity. The modifications to the neutron kinetic energy distribution as described by Munro include plasma velocity terms with spatial orientation, suggesting that the neutron kinetic energy distributions could be anisotropic when viewed by multiple lines-of-sight. These anisotropies provide a diagnostic of burn averaged plasma velocity distributions. We present the results of measurements made for a variety of DT implosions and discuss their possible physical interpretations. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under contract DE-AC52-07NA27344. Lawrence Livermore National Security, LLC.

  17. Reactivity and kinetic parameters determination in a multiplicative non-stationary system

    International Nuclear Information System (INIS)

    Minguez, E.

    1982-01-01

    A revision of several methods used for solving kinetic equations of a neutronic system is considered. Firstly, kinetic equations in general form are analized, before to revise more important aproximations: point-kinetic method; adiabatic; cuasistatic; eigenvalue equations; nodal, modal and systhesis methods; and variational principles for obtaining kinetic equations. Perturbation theory is used to obtain these parameters, with differents eigenvalue equations representatives of the parameter to be calculated. Also, experimental methods have been included in this work, because of importance the parameters can be measured, and related with those obtained by calculations. Finally, adjoint kinetic equations are resolved to obtain the importance function used in weighted reactivity and kinetic parameters determinations. (author)

  18. Basis for calculating boron dilution scenarios in PWR by 3D neutron kinetics

    Energy Technology Data Exchange (ETDEWEB)

    Pla, P., E-mail: patricia_pla@hotmail.com [Univ. of Pisa, San Piero a Grado Nuclear Research Group (GRNSPG), Pisa (Italy); Tech. Univ. of Catalonia, Barcelona (Spain); Parisi, C., E-mail: c.parisi@ing.unipi.it [Univ. of Pisa, San Piero a Grado Nuclear Research Group (GRNSPG), Pisa (Italy); Galetti, R., E-mail: regina@cnen.gov.br [National Commission for Nuclear Energy (CNEN), Rio de Janeiro (Brazil); D' Auria, F.; Galassi, G., E-mail: f.dauria@ing.unipi.it, E-mail: g.galassi@ing.unipi.it [Univ. of Pisa, San Piero a Grado Nuclear Research Group (GRNSPG), Pisa (Italy); Reventos, F., E-mail: francesc.reventos@upc.edu [Tech. Univ. of Catalonia, Barcelona (Spain)

    2011-07-01

    The origin of the performed study was the analysis of 20 cm{sup 2} small break LOCA in the lower plenum in a four-loop PWR nuclear reactor by Relap5 code stand-alone (0DNK) in which boron dilution was observed in more than one loop seal. In order to have a more precise result of the boron dilution NK feedback effect, the original nodalization was refined axially in the core area to couple with PARCS v.2.7 code (3DNK). The neutron macroscopic XSec database was generated by the lattice transport code HELIOS. Before using the new model to predict boron dilution transients, a necessary activity is the qualification of the model (the boron feedback calculated by the Neutronic Cross Sections) against boron changes, so a group of sensitivity calculations injecting more or less borated water in the cold leg were performed either with Relap5 code stand-alone (0DNK) and with Relap5 coupled with PARCS v.2.7 (3DNK) code in order to analyze the reactor power response to the boron injection and the differences using a 0DNK or a coupled 3DNK nodalization. To complete the study a benchmark calculation was performed considering a 20 cm{sup 2} break in the lower plenum, in which the reactor trip by control rods has been disabled and boron injection was simulated in the cold leg. This calculation utilized the Relap5 code stand-alone (0DNK) and the Relap5 coupled with PARCS v.2.7 (3DNK) code, in order to see the differences using a 0DNK or a coupled 3DNK model. Non negligible differences have been found in all cases in the comparison of 0DNK and coupled 3DNK results analyzed, in relation to the core power. These results challenge the evaluation of the uncertainties in case of coupled thermalhydraulic-3DNK calculations. A comprehensive evaluation of the relevant uncertainties of the 3D NK TH coupled calculations is needed. (author)

  19. Basis for calculating boron dilution scenarios in PWR by 3D neutron kinetics

    International Nuclear Information System (INIS)

    Pla, P.; Parisi, C.; Galetti, R.; D'Auria, F.; Galassi, G.; Reventos, F.

    2011-01-01

    The origin of the performed study was the analysis of 20 cm 2 small break LOCA in the lower plenum in a four-loop PWR nuclear reactor by Relap5 code stand-alone (0DNK) in which boron dilution was observed in more than one loop seal. In order to have a more precise result of the boron dilution NK feedback effect, the original nodalization was refined axially in the core area to couple with PARCS v.2.7 code (3DNK). The neutron macroscopic XSec database was generated by the lattice transport code HELIOS. Before using the new model to predict boron dilution transients, a necessary activity is the qualification of the model (the boron feedback calculated by the Neutronic Cross Sections) against boron changes, so a group of sensitivity calculations injecting more or less borated water in the cold leg were performed either with Relap5 code stand-alone (0DNK) and with Relap5 coupled with PARCS v.2.7 (3DNK) code in order to analyze the reactor power response to the boron injection and the differences using a 0DNK or a coupled 3DNK nodalization. To complete the study a benchmark calculation was performed considering a 20 cm 2 break in the lower plenum, in which the reactor trip by control rods has been disabled and boron injection was simulated in the cold leg. This calculation utilized the Relap5 code stand-alone (0DNK) and the Relap5 coupled with PARCS v.2.7 (3DNK) code, in order to see the differences using a 0DNK or a coupled 3DNK model. Non negligible differences have been found in all cases in the comparison of 0DNK and coupled 3DNK results analyzed, in relation to the core power. These results challenge the evaluation of the uncertainties in case of coupled thermalhydraulic-3DNK calculations. A comprehensive evaluation of the relevant uncertainties of the 3D NK TH coupled calculations is needed. (author)

  20. Monte Carlo simulation for fragment mass and kinetic energy distributions from the neutron-induced fission of 235U

    International Nuclear Information System (INIS)

    Montoya, M.; Rojas, J.; Saettone, E.

    2007-01-01

    The mass and kinetic energy distribution of nuclear fragments from the thermal neutron-induced fission of 235 U have been studied using a Monte Carlo simulation. Besides reproducing the pronounced broadening on the standard deviation of the final fragment kinetic energy distribution (σ e (m)) around the mass number m = 109, our simulation also produces a second broadening around m = 125 that is in agreement with the experimental data obtained by Belhafaf et al. These results are a consequence of the characteristics of the neutron emission, the variation in the primary fragment mean kinetic energy, and the yield as a function of the mass. (Author)

  1. Monte Carlo simulation for fragment mass and kinetic energy distributions from the neutron-induced fission of {sup 235}U

    Energy Technology Data Exchange (ETDEWEB)

    Montoya, M.; Rojas, J. [Instituto Peruano de Energia Nuclear, Av. Canada 1470, Lima 41 (Peru); Saettone, E. [Facultad de Ciencias, Universidad Nacional de lngenieria, Av. Tupac Amaru 210, Apartado 31-139, Lima (Peru)

    2007-07-01

    The mass and kinetic energy distribution of nuclear fragments from the thermal neutron-induced fission of {sup 235}U have been studied using a Monte Carlo simulation. Besides reproducing the pronounced broadening on the standard deviation of the final fragment kinetic energy distribution ({sigma}{sub e}(m)) around the mass number m = 109, our simulation also produces a second broadening around m = 125 that is in agreement with the experimental data obtained by Belhafaf et al. These results are a consequence of the characteristics of the neutron emission, the variation in the primary fragment mean kinetic energy, and the yield as a function of the mass. (Author)

  2. Modeling Taylor series approximations for prompt neutron kinetics with lab view simulations

    International Nuclear Information System (INIS)

    Adzri, E. P.

    2012-09-01

    The reactor point kinetics equations have been subjected to intense research in an effort to find simple yet accurate numerical solutions methods. The equations are very stiff numerically, meaning that there is a wide variation in the decay constants, so that using a particular time step in the numerical solution may provide sufficient accuracy for the group, but not for another. Several solutions techniques have been presented on the point kinetics equations with varying degrees of complexity. These include Power Series Solutions, CORE, PCA, Genapol and Taylor series methods. In this research, algorithms were developed based on the first and second order Taylor series expansion and simulated in LabVIEW to solve the Reactor Point Kinetics equations using block diagram nodes implemented within stacked sequences. The algorithms developed were fast,accurate and simple to code. Several reactivity insertions were used to simulate the change in neutron population with time. The LabVIEW- Taylor series solutions were compared with other solution techniques such as Power Series Solutions, CORE, PCA, Genapol and McMahon and Pierson's Taylor series approximation. The results of LabVIEW-Taylor series technique used by McMahon and Pearson The LabVIEW-implemented techniques were found to agree very well with these other methods. At 1x10 -8 s the neutron population was 1.000220 neutrons / cm 3 , at 1 x 10 -2 s it was 2.007681 neutrons / cm 3 and at 1x10 -1 s it was 2.075317 neutrons / cm 3 ; same results reported by Genapol for a fast reactor, it produced good and accurate results and compared very favorably with other methods found in the literature. Using much smaller time steps to the order or 10 -8 s commensurate with fast reactor parameters also produced very satisfactory results, indicating that the LabVIEW-based Taylor series technique is suitable for simulating the kinetics of fast reactors as well as thermal reactors. Algorithms developed that included second order terms

  3. Neutron lifetimes behavior analysis considering the two-region kinetic model in the IPEN/MB-01 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gonnelli, Eduardo; Diniz, Ricardo [Instituto de Pesquisas Energéticas e Nucleares - IPEN/CNEN-SP Travessa R-400, 05508-900, Cidade Universitária, São Paulo (Brazil)

    2014-11-11

    This is a complementary work about the behavior analysis of the neutron lifetimes that was developed in the IPEN/MB-01 nuclear reactor facility. The macroscopic neutron noise technique was experimentally employed using pulse mode detectors for two stages of control rods insertion, where a total of twenty levels of subcriticality have been carried out. It was also considered that the neutron reflector density was treated as an additional group of delayed neutrons, being a sophisticated approach in the two-region kinetic theoretical model.

  4. Neutron lifetimes behavior analysis considering the two-region kinetic model in the IPEN/MB-01 reactor

    International Nuclear Information System (INIS)

    Gonnelli, Eduardo; Diniz, Ricardo

    2014-01-01

    This is a complementary work about the behavior analysis of the neutron lifetimes that was developed in the IPEN/MB-01 nuclear reactor facility. The macroscopic neutron noise technique was experimentally employed using pulse mode detectors for two stages of control rods insertion, where a total of twenty levels of subcriticality have been carried out. It was also considered that the neutron reflector density was treated as an additional group of delayed neutrons, being a sophisticated approach in the two-region kinetic theoretical model

  5. Three-dimensional neutron kinetics-thermal-hydraulics VVER 1000 main steam line break analysis by RELAP5-3D code

    International Nuclear Information System (INIS)

    Frisani, A.; Parisi, C.; D'Auria, F.

    2007-01-01

    After the development and the assessment of Three-Dimensional (3D) Neutron Kinetics (NK) - 1D Thermal-Hydraulics (TH) coupled codes analyses methods, deterministic nuclear safety technology is nowadays producing noticeable efforts for the validation of 3D NK - 3D TH coupled codes analyses methods too. Thus, the purpose of this work was to address the capability of the RELAP5-3D 3D NK-3D TH code to reproduce VVER 1000 Nuclear Power Plant (NPP) core dynamic in simulating the mixing effects that could happen in the vessel downcomer and lower plenum during some scenarios. The work was developed in three steps. The first step dealt with the 3D TH modeling of the Kozloduy-6 VVER 1000 reactor pressure vessel. Then this model was validated following a Steam Generator Isolation transient. The second step has been the development of a 3D NK nodalization for the reactor core region. Then the 3D NK model was directly coupled with the previously developed 3D TH model. The third step was the calculation of a Main Steam Line Break (MSLB) transient. The 3D NK global nuclear parameters were then compared with the 0-D results showing a good agreement; nevertheless only the 3D NK- 3D TH model allowed the calculation of each single assembly power trend for this strong NK-TH asymmetric transient. (author)

  6. Numerical Solution of Fractional Neutron Point Kinetics Model in Nuclear Reactor

    Directory of Open Access Journals (Sweden)

    Nowak Tomasz Karol

    2014-06-01

    Full Text Available This paper presents results concerning solutions of the fractional neutron point kinetics model for a nuclear reactor. Proposed model consists of a bilinear system of fractional and ordinary differential equations. Three methods to solve the model are presented and compared. The first one entails application of discrete Grünwald-Letnikov definition of the fractional derivative in the model. Second involves building an analog scheme in the FOMCON Toolbox in MATLAB environment. Third is the method proposed by Edwards. The impact of selected parameters on the model’s response was examined. The results for typical input were discussed and compared.

  7. Neutron scattering from myelin revisited: bilayer asymmetry and water-exchange kinetics

    Energy Technology Data Exchange (ETDEWEB)

    Denninger, Andrew R. [Boston College, Chestnut Hill, MA 02467 (United States); Demé, Bruno; Cristiglio, Viviana [Institut Laue–Langevin (ILL), CS 20156, F-38042 Grenoble CEDEX 9 (France); LeDuc, Géraldine [European Synchrotron Radiation Facility (ESRF), CS 40220, F-38043 Grenoble CEDEX 9 (France); Feller, W. Bruce [NOVA Scientific Inc., Sturbridge, MA 01566 (United States); Kirschner, Daniel A., E-mail: kirschnd@bc.edu [Boston College, Chestnut Hill, MA 02467 (United States)

    2014-12-01

    The structure of internodal myelin in the rodent central and peripheral nervous systems has been determined using neutron diffraction. The kinetics of water exchange in these tissues is also described. Rapid nerve conduction in the central and peripheral nervous systems (CNS and PNS, respectively) of higher vertebrates is brought about by the ensheathment of axons with myelin, a lipid-rich, multilamellar assembly of membranes. The ability of myelin to electrically insulate depends on the regular stacking of these plasma membranes and on the presence of a number of specialized membrane-protein assemblies in the sheath, including the radial component, Schmidt–Lanterman incisures and the axo–glial junctions of the paranodal loops. The disruption of this fine-structure is the basis for many demyelinating neuropathies in the CNS and PNS. Understanding the processes that govern myelin biogenesis, maintenance and destabilization requires knowledge of myelin structure; however, the tight packing of internodal myelin and the complexity of its junctional specializations make myelin a challenging target for comprehensive structural analysis. This paper describes an examination of myelin from the CNS and PNS using neutron diffraction. This investigation revealed the dimensions of the bilayers and aqueous spaces of myelin, asymmetry between the cytoplasmic and extracellular leaflets of the membrane, and the distribution of water and exchangeable hydrogen in internodal multilamellar myelin. It also uncovered differences between CNS and PNS myelin in their water-exchange kinetics.

  8. On the Neutron Kinetics and Control of Accelerator-Driven Systems

    International Nuclear Information System (INIS)

    Cacuci, D.G.

    2004-01-01

    This work addresses fundamental aspects of the time- and space-dependent behavior of an Accelerator-Driven Subcritical Core System (ADS) and presents a paradigm ADS neutron kinetics model that is solved exactly. Thus, this paradigm model can serve for benchmarking two- and/or three-dimensional computational tools. Furthermore, this work also proposes a global optimal control theory framework for the operation and control of an ADS. This framework encompasses conceptually the time- and space-dependent behavior of the ADS coupled neutron kinetics/thermal-hydraulic balance equations and aims at the optimal control of ADS operational objectives, which would include minimization of local flux disturbances, load and source following, etc. Importantly, this new conceptual framework makes no use of a 'fictitious ADS steady state' and yields the correct and complete (i.e., including sources) adjoint equations, without leaving any room for ambiguities. Thus, this new conceptual framework provides a natural basis for developing new computational methods and corresponding verification experiments specifically tailored for the control and operation of ADS

  9. Thermal hydraulic and neutron kinetic coupled simulation of the IPR-R1 Triga reactor

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Silva, Clarysson A.M. da; Veloso, Maria Auxiliadora F.; Soares, Humbero V., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: clarysson@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: betovitor@ig.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq Rede), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    The nuclear industry and the scientific community have turned the attention for the development of coupled 3D neutron kinetics (NK) and thermal-hydraulic (TH) system codes to investigate specific nuclear reactor transients. Improving in theoretical investigations of complex phenomena in nuclear reactor technology have been increased thanks to numerical methods and computational resources incorporated in nuclear codes. This paper presents a model for the IPR-R1 TRIGA research reactor using the RELAP5-3D 3.0 code. The development and the assessment of the thermal-hydraulic RELAP5 code model for the IPR-R1 have been validated for steady state and transient situations and the results were published in preceding works. Results of RELAP5-3D steady state and a transient case presented in this paper show good agreement with experimental data, validating then this model for point kinetic calculations. To supply adequate cross sections to the NK code, the WIMSD5 is being used. First results of steady state calculation using the 3D neutron modeling are being presented in this paper. (author)

  10. Kinetic studies of methane-ethane mixed gas hydrates by neutron diffraction and Raman spectroscopy.

    Science.gov (United States)

    Murshed, M Mangir; Kuhs, Werner F

    2009-04-16

    In situ formations of CH(4)-C(2)H(6) mixed gas hydrates were made using high flux neutron diffraction at 270 K and 5 MPa. For this purpose, a feed gas composition of CH(4) and C(2)H(6) (95 mol% CH(4)) was employed. The rates of transformation of spherical grains of deuterated ice Ih into hydrates were measured by time-resolved neutron powder diffraction on D20 at ILL, Grenoble. Phase fractions of the crystalline constituents were obtained from Rietveld refinements. A concomitant formation of structure type I (sI) and structure type II (sII) hydrates were observed soon after the gas pressure was applied. The initial fast formation of sII hydrate reached its maximum volume and started declining very slowly. The formation of sI hydrate followed a sigmoid growth kinetics that slowed down due to diffusion limitation. This observation has been interpreted in terms of a kinetically favored nucleation of the sII hydrate along with a slow transformation into sI. Both powder diffraction and Raman spectroscopic results suggest that a C(2)H(6)-rich sII hydrate was formed at the early part of the clathration, which slowly decreased to approximately 3% after a reaction of 158 days as confirmed by synchrotron XRD. The final persistence of a small portion of sII hydrate points to a miscibility gap between CH(4)-rich sI and C(2)H(6)-rich sII hydrates.

  11. Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

    Energy Technology Data Exchange (ETDEWEB)

    Aji, Indarta Kuncoro, E-mail: indartaaji@s.itb.ac.id [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10 Bandung 40132 (Indonesia); Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Permana, Sidik [Nuclear Physics & Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10 Bandung 40132 (Indonesia)

    2015-09-30

    Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF{sub 2}-ThF{sub 4}-{sup 233}UF{sub 4} respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.

  12. Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

    International Nuclear Information System (INIS)

    Aji, Indarta Kuncoro; Waris, Abdul; Permana, Sidik

    2015-01-01

    Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF 2 -ThF 4 - 233 UF 4 respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155

  13. Coupled 3D neutron kinetics and thermalhydraulic characteristics of the Canadian supercritical water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hummel, David William, E-mail: hummeld@mcmaster.ca; Novog, David Raymond

    2016-03-15

    Highlights: • A coupled spatial kinetics and thermalhydraulics model of the PT-SCWR was created. • Positive power excursions were demonstrated during accident-like transients. • The reactor will inherently self-shutdown in such transients with some delay. • A fast-acting shutdown system would limit the consequences of the power pulse. - Abstract: The Canadian Supercritical Water-cooled Reactor concept, as an evolution of the CANada Deuterium Uranium (CANDU) reactor, includes both pressure tubes and a low temperature heavy water moderator. The current Pressure Tube type SCWR (PT-SCWR) concept features 64-element fuel assemblies placed within High Efficiency Re-entrant Channels (HERCs) that connect to core inlet and outlet plena. Among current SCWR concepts the PT-SCWR is unique in that the HERC separates multiple coolant and moderator regions, giving rise to coupled neutronic-thermalhydraulic feedbacks beyond those present in CANDU or contemporary Light Water Reactors. The objective of this work was thus to model the coupled neutronic-thermal hydraulic properties of the PT-SCWR to establish the impact of these multiple regions on the core's transient behavior. To that end, the features of the PT-SCWR were first modeled with the neutron transport code DRAGON to create a database of homogenized and condensed cross-sections and thermalhydraulic feedback coefficients. These were used as input to a core-level neutron diffusion model created with the code DONJON. The behavior of the primary heat transport system was modeled with the thermalhydraulic system code CATHENA. A procedure was developed to couple the outputs of DONJON and CATHENA, facilitating three-dimensional spatial neutron kinetics and coupled thermalhydraulic analysis of the PT-SCWR core. Several postulated transients were initiated within the coupled model by changing the core inlet and outlet boundary conditions. Decreasing coolant density around the fuel was demonstrated to produce positive

  14. Study on coupling of three-dimension space time neutron kinetics model and RELAP5 and improvement of RELAP5

    International Nuclear Information System (INIS)

    Gui Xuewen; Cai Qi; Luo Bangqi

    2007-01-01

    A two-group three-dimension space-time neutron kinetics model is applied to the RELAP5 code, which replaces the point reactor kinetics model. A visual operation interface is designed to convenience interactive operation between operator and computer. The calculation results and practical applications indicate that the functions and precision of improved RELAP5 are enhanced and can be easily used. The improved RELAP5 has a good application perspective in nuclear power plant simulation. (authors)

  15. Kinetic energy spectrum and polarization of neutrons from the reaction 12C(p,n)X at 590 MeV

    International Nuclear Information System (INIS)

    Arnold, J.

    1998-01-01

    The kinetic energy spectrum and the polarization of the PSI neutron beam produced in the reaction 12 C(p,n)X at 0 with 590 MeV polarized protons were investigated. A strong energy dependence of the neutron beam polarization is observed which was not expected at the time the neutron beam was built. (orig.)

  16. Extension of the analytic nodal method to four energy groups

    International Nuclear Information System (INIS)

    Parsons, D.K.; Nigg, D.W.

    1985-01-01

    The Analytic Nodal Method is one of several recently-developed coarse mesh numerical methods for efficiently and accurately solving the multidimensional static and transient neutron diffusion equations. This summary describes a mathematically rigorous extension of the Analytic Nodal Method to the frequently more physically realistic four-group case. A few general theoretical considerations are discussed, followed by some calculated results for a typical steady-state two-dimensional PWR quarter core application. 8 refs

  17. Algorithm development and verification of UASCM for multi-dimension and multi-group neutron kinetics model

    International Nuclear Information System (INIS)

    Si, S.

    2012-01-01

    The Universal Algorithm of Stiffness Confinement Method (UASCM) for neutron kinetics model of multi-dimensional and multi-group transport equations or diffusion equations has been developed. The numerical experiments based on transport theory code MGSNM and diffusion theory code MGNEM have demonstrated that the algorithm has sufficient accuracy and stability. (authors)

  18. A Hennart nodal method for the diffusion equation

    International Nuclear Information System (INIS)

    Lesaint, P.; Noceir, S.; Verwaerde, D.

    1995-01-01

    A modification of the Hennart nodal method for neutron diffusion problems is presented. The final system of equations obtained by this method is not positive definite. However, a flux elimination technique leads to a simple positive definite system, which can be solved by the traditional iterative methods. Calculations of a two-dimensional International Atomic Energy Agency benchmark problem are performed and compared with results of the original Hennart nodal method and some finite element methods. The high computational efficiency of this modified nodal method is clearly demonstrated

  19. A diffuse neutron scattering study of clustering kinetics in Cu-Ni alloys

    International Nuclear Information System (INIS)

    Vrijen, J.; Radelaar, S.; Schwahn, D.

    1977-01-01

    Diffuse scattering of thermal neutrons was used to investigate the kinetics of clustering in Cu-Ni alloys. In order to optimize the experimental conditions the isotopes 65 Cu and 62 Ni were alloyed. The time evolution of the diffuse scattered intensity at 400 0 C has been measured for eight Cu-Ni alloys, varying in composition between 30 and 80 at. pour cent Ni. The relaxation of the so called null matrix, containing 56.5 at. pour cent Ni has also been investigated at 320, 340, 425 and 450 0 C. Using Cook's model from all these measurements information has been deduced about diffusion at low temperatures and about thermodynamic properties of the Cu-Ni system. It turns out that Cook's model is not sufficiently detailed for an accurate description of the initial stages of these relaxations

  20. Tritium release kinetics in lithium orthosilicate ceramic pebbles irradiated with low thermal-neutron fluence

    International Nuclear Information System (INIS)

    Xiao, Chengjian; Gao, Xiaoling; Kobayashi, Makoto; Kawasaki, Kiyotaka; Uchimura, Hiromichi; Toda, Kensuke; Kang, Chunmei; Chen, Xiaojun; Wang, Heyi; Peng, Shuming; Wang, Xiaolin; Oya, Yasuhisa; Okuno, Kenji

    2013-01-01

    Tritium release kinetics in lithium orthosilicate (Li 4 SiO 4 ) ceramic pebbles irradiated with low thermal-neutron fluence was studied by out-of-pile annealing experiments. It was found that the tritium produced in Li 4 SiO 4 pebbles was mainly released as tritiated water vapor (HTO). The apparent desorption activation energy of tritium on the pebble surface was consistent with the diffusion activation energy of tritium in the crystal grains, indicating that tritium release was mainly controlled by diffusion process. The diffusion coefficients of tritium in the crystal grains at temperatures ranging from 450 K to 600 K were obtained by isothermal annealing tests, and the Arrhenius relation was determined to be D = 1 × 10 −7.0 exp (−40.3 × 10 3 /RT) cm 2 s −1

  1. Tritium release kinetics in lithium orthosilicate ceramic pebbles irradiated with low thermal-neutron fluence

    Energy Technology Data Exchange (ETDEWEB)

    Xiao, Chengjian; Gao, Xiaoling [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Box 919-214, Mian Yang 621900 (China); Kobayashi, Makoto; Kawasaki, Kiyotaka; Uchimura, Hiromichi; Toda, Kensuke [China Academy of Engineering Physics, Box 919-1, Mian Yang 621900 (China); Kang, Chunmei; Chen, Xiaojun; Wang, Heyi; Peng, Shuming [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Box 919-214, Mian Yang 621900 (China); Wang, Xiaolin, E-mail: xlwang@caep.ac.cn [China Academy of Engineering Physics, Box 919-1, Mian Yang 621900 (China); Oya, Yasuhisa; Okuno, Kenji [Radiochemistry Research Laboratory, Faculty of Science, Shizuoka University, 836 Ohya, Shizuoka 422-8529 (Japan)

    2013-07-15

    Tritium release kinetics in lithium orthosilicate (Li{sub 4}SiO{sub 4}) ceramic pebbles irradiated with low thermal-neutron fluence was studied by out-of-pile annealing experiments. It was found that the tritium produced in Li{sub 4}SiO{sub 4} pebbles was mainly released as tritiated water vapor (HTO). The apparent desorption activation energy of tritium on the pebble surface was consistent with the diffusion activation energy of tritium in the crystal grains, indicating that tritium release was mainly controlled by diffusion process. The diffusion coefficients of tritium in the crystal grains at temperatures ranging from 450 K to 600 K were obtained by isothermal annealing tests, and the Arrhenius relation was determined to be D = 1 × 10{sup −7.0} exp (−40.3 × 10{sup 3}/RT) cm{sup 2} s{sup −1}.

  2. Complex models of nodal nuclear data

    International Nuclear Information System (INIS)

    Dufek, Jan

    2011-01-01

    During the core simulations, nuclear data are required at various nodal thermal-hydraulic and fuel burnup conditions. The nodal data are also partially affected by thermal-hydraulic and fuel burnup conditions in surrounding nodes as these change the neutron energy spectrum in the node. Therefore, the nodal data are functions of many parameters (state variables), and the more state variables are considered by the nodal data models the more accurate and flexible the models get. The existing table and polynomial regression models, however, cannot reflect the data dependences on many state variables. As for the table models, the number of mesh points (and necessary lattice calculations) grows exponentially with the number of variables. As for the polynomial regression models, the number of possible multivariate polynomials exceeds the limits of existing selection algorithms that should identify a few dozens of the most important polynomials. Also, the standard scheme of lattice calculations is not convenient for modelling the data dependences on various burnup conditions since it performs only a single or few burnup calculations at fixed nominal conditions. We suggest a new efficient algorithm for selecting the most important multivariate polynomials for the polynomial regression models so that dependences on many state variables can be considered. We also present a new scheme for lattice calculations where a large number of burnup histories are accomplished at varied nodal conditions. The number of lattice calculations being performed and the number of polynomials being analysed are controlled and minimised while building the nodal data models of a required accuracy. (author)

  3. Nanostructure evolution of neutron-irradiated reactor pressure vessel steels: Revised Object kinetic Monte Carlo model

    Energy Technology Data Exchange (ETDEWEB)

    Chiapetto, M., E-mail: mchiapet@sckcen.be [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium); Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Messina, L. [DEN-Service de Recherches de Métallurgie Physique, CEA, Université Paris-Saclay, F-91191 Gif-sur-Yvette (France); KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-114 21 Stockholm (Sweden); Becquart, C.S. [Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Olsson, P. [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-114 21 Stockholm (Sweden); Malerba, L. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium)

    2017-02-15

    This work presents a revised set of parameters to be used in an Object kinetic Monte Carlo model to simulate the microstructure evolution under neutron irradiation of reactor pressure vessel steels at the operational temperature of light water reactors (∼300 °C). Within a “grey-alloy” approach, a more physical description than in a previous work is used to translate the effect of Mn and Ni solute atoms on the defect cluster diffusivity reduction. The slowing down of self-interstitial clusters, due to the interaction between solutes and crowdions in Fe is now parameterized using binding energies from the latest DFT calculations and the solute concentration in the matrix from atom-probe experiments. The mobility of vacancy clusters in the presence of Mn and Ni solute atoms was also modified on the basis of recent DFT results, thereby removing some previous approximations. The same set of parameters was seen to predict the correct microstructure evolution for two different types of alloys, under very different irradiation conditions: an Fe-C-MnNi model alloy, neutron irradiated at a relatively high flux, and a high-Mn, high-Ni RPV steel from the Swedish Ringhals reactor surveillance program. In both cases, the predicted self-interstitial loop density matches the experimental solute cluster density, further corroborating the surmise that the MnNi-rich nanofeatures form by solute enrichment of immobilized small interstitial loops, which are invisible to the electron microscope.

  4. On an analytical formulation for the mono-energetic neutron space-kinetic equation in full cylinder symmetry

    International Nuclear Information System (INIS)

    Oliveira, F.R.; Bodmann, B.E.J.; Vilhena, M.T.; Carvalho, F.

    2017-01-01

    Highlights: • The present work presents an exact solution to neutron spatial kinetic equation. • It is an exact solution in a heterogeneous cylinder with temporal dependence. • The solution was constructed through the separation of variables method. - Abstract: In the present work we discuss a system of partial differential equations that model neutron space-kinetics in cylindrical geometry and are defined by two sectionally homogeneous cylinder cells, mono-energetic neutrons and one group of delayed neutron precursors. The solution is determined using the technique of variable separation. The associated complete spectra with respect to each variable separation are analysed and truncated such as to allow a parameterized global solution. For the obtained solution we present some numerical results for the scalar neutron flux and its time dependence and projection on the cylinder axis z and the radial and cylinder axis projection. As a case study we consider an insertion of an absorbing medium in the upper cylinder cell. Continuity of the scalar flux at the interface between the two cylinder elements and conserved current density is explained and related to scale invariance of the partial differential equation system together with the initial and boundary conditions. Some numerical results for the scalar angular neutron flux and associated current densities are shown.

  5. Coupling of unidimensional neutron kinetics to thermal hydraulics in parallel channels

    International Nuclear Information System (INIS)

    Cecenas F, M.; Campos G, R.M.

    2003-01-01

    In this work the dynamic behavior of a consistent system in fifteen channels in parallel that represent the reactor core of a BWR type, coupled of a kinetic neutronic model in one dimension is studied by means of time series. The arrangement of channels is obtained collapsing the assemblies that it consists the core to an arrangement of channels prepared in straight lines, and it is coupled to the unidimensional solution of the neutron diffusion equation. This solution represents the radial power distribution, and initially the static solution is obtained to verify that the one modeling core is critic. The coupled set nuclear-thermal hydraulics it is solved numerically by means of a net of CPUs working in the outline teacher-slave by means of Parallel Virtual Machine (PVM), subject to the restriction that the pressure drop is equal for each channel, which is executed iterating on the refrigerant distribution. The channels are dimensioned according to the one Stability Benchmark of the Ringhals swedish plant, organized by the Nuclear Energy Agency in 1994. From the information of this benchmark it is obtained the axial power profile for each channel, which is assumed as invariant in the time. To obtain the time series, the system gets excited with white noise (sequence that statistically obeys to a normal distribution with zero media), so that the power generated in each channel it possesses the same ones characteristics of a typical signal obtained by means of the acquisition of those signals of neutron flux in a BWR reactor. (Author)

  6. Kinetics that govern the release of tritium from neutron-irradiated lithium oxide

    International Nuclear Information System (INIS)

    Bertone, P.C.

    1986-01-01

    The Lithium Blanket Module (LBM) program being conducted at the Princeton Plasma Physics Laboratory requires that tritium concentrations as low as 0.1 nCi/g, bred in both LBM lithium oxide pellets and gram-size lithium samples, be measured with an uncertainty not exceeding +/-6%. This thesis reports two satisfactory methods of assaying LBM pellets and one satisfactory method of assaying lithium samples. Results of a fundamental kinetic investigation are also reported. The thermally driven release of tritium from neutron-irradiated lithium oxide pellets is studied between the temperatures of 300 and 400 0 C. The observed release clearly obeys first-order kinetics, and the governing activation energy appears to be 28.4 kcal/mole. Finally, a model is presented that may explain the thermally driven release of tritium from a lithium oxide crystal and assemblies thereof. It predicts that under most circumstances the release is controlled by either the diffusion of a tritiated species through the crystal, or by the desorption of tritiated water from it

  7. Neutron emission effects on final fragments mass and kinetic energy distribution from low energy fission of 234U

    International Nuclear Information System (INIS)

    Montoya, M.; Rojas, J.; Lobato, I.

    2008-01-01

    The standard deviation of the final kinetic energy distribution (σ e ) as a function of mass of final fragments (m) from low energy fission of 234 U, measured with the Lohengrin spectrometer by Belhafaf et al., presents a peak around m = 109 and another around m = 122. The authors attribute the first peak to the evaporation of a large number of neutrons around the corresponding mass number, i.e. there is no peak on the standard deviation of the primary kinetic energy distribution (σ E ) as a function of primary fragment mass (A). The second peak is attributed to a real peak on σ E (A). However, theoretical calculations related to primary distributions made by H.R. Faust and Z. Bao do not suggest any peak on σ E (A). In order to clarify this apparent controversy, we have made a numerical experiment in which the masses and the kinetic energy of final fragments are calculated, assuming an initial distribution of the kinetic energy without structures on the standard deviation as function of fragment mass. As a result we obtain a pronounced peak on σ e (m) curve around m = 109, a depletion from m = 121 to m = 129, and an small peak around m = 122, which is not as great as that measured by Belhafaf et al. Our simulation also reproduces the experimental results on the yield of the final mass Y(m), the average number of emitted neutrons as a function of the provisional mass (calculated from the values of the final kinetic energy of the complementary fragments) and the average value of fragment kinetic energy as a function of the final mass. From our results we conclude that there are no peaks on the σ E (A) curve, and the observed peaks on σ e (m) are due to the emitted neutron multiplicity and the variation of the average fragment kinetic energy as a function of primary fragment mass. (Author)

  8. Nodal-chain metals.

    Science.gov (United States)

    Bzdušek, Tomáš; Wu, QuanSheng; Rüegg, Andreas; Sigrist, Manfred; Soluyanov, Alexey A

    2016-10-06

    The band theory of solids is arguably the most successful theory of condensed-matter physics, providing a description of the electronic energy levels in various materials. Electronic wavefunctions obtained from the band theory enable a topological characterization of metals for which the electronic spectrum may host robust, topologically protected, fermionic quasiparticles. Many of these quasiparticles are analogues of the elementary particles of the Standard Model, but others do not have a counterpart in relativistic high-energy theories. A complete list of possible quasiparticles in solids is lacking, even in the non-interacting case. Here we describe the possible existence of a hitherto unrecognized type of fermionic excitation in metals. This excitation forms a nodal chain-a chain of connected loops in momentum space-along which conduction and valence bands touch. We prove that the nodal chain is topologically distinct from previously reported excitations. We discuss the symmetry requirements for the appearance of this excitation and predict that it is realized in an existing material, iridium tetrafluoride (IrF 4 ), as well as in other compounds of this class of materials. Using IrF 4 as an example, we provide a discussion of the topological surface states associated with the nodal chain. We argue that the presence of the nodal-chain fermions will result in anomalous magnetotransport properties, distinct from those of materials exhibiting previously known excitations.

  9. A nodalization study of steam separator in real time simulation

    International Nuclear Information System (INIS)

    Horugshyang, Lein; Luh, R.T.J.; Zen-Yow, Wang

    1999-01-01

    The motive of this paper is to investigate the influence of steam separator nodalization on reactor thermohydraulics in terms of stability and level response. Three different nodalizations of steam separator are studied by using THEATRE and REMARK Code in a BWR simulator. The first nodalization is the traditional one with two nodes for steam separator. In this nodalization, the steam separation is modeled in the outer node, i.e., upper downcomer. Separated steam enters the Steen dome node and the liquid goes to the feedwater node. The second nodalization is similar to the first one with the steam separation modeled in the inner node. There is one additional junction connecting steam dome node and the inner node. The liquid fallback junction connects the inner node and feedwater node. The third nodalization is a combination of the former two with an integrated node for steam separator. Boundary conditions in this study are provided by a simplified feedwater and main steam driver. For comparison purpose, three tests including full power steady state initialisation, recirculation pumps runback and reactor scram are conducted. Major parameters such as reactor pressure, reactor level, void fractions, neutronic power and junction flows are recorded for analysis. Test results clearly show that the first nodalization is stable for steady state initialisation. However it has too responsive level performance in core flow reduction transients. The second nodalization is the closest representation of real plant structure, but not the performance. Test results show that an instability occurs in the separator region for both steady state initialisation and transients. This instability is caused by an unbalanced momentum in the dual loop configuration. The magnitude of the oscillation reduces as the power decreases. No superiority to the other nodalizations is shown in the test results. The third nodalization shows both stability and responsiveness in the tests. (author)

  10. Investigation of digestion Kinetics in commercial starches using in-situ small-angle neutron scattering

    International Nuclear Information System (INIS)

    Blazek, Jaroslav; Gilbert, Elliot Paul

    2009-01-01

    Full text: The digestion of starch has been the subject of many investigations, mostly involving in vitro measurement of the susceptibility of starches to attack by different enzymes, rather than measuring actual digestibility in vivo. The rate and extent of amylolytic hydrolysis of granular starches is known to vary according to botanical origin. Granule characteristics considered to influence susceptibility to attack by alpha-amylase include crystallinity, granule size and available specific surface, amylose content, porosity, structural inhomogeneities and degree of integrity. Most in-vitro studies of granular starch digestion have been limited to samples for which aliquots have been removed from the reaction mixture at various time intervals and freeze-dried to be subsequently characterized using a range of techniques. It remains unclear whether sample preparation creates artefacts in the samples. In this study, we have studied the kinetics of starch digestion of several commercial granular starches by time-resolved small-angle neutron scattering using an in-situ digestion chamber allowing, for the first time, to follow structural changes of starch in the course of digestion directly in the digestion mixture. Additionally, samples before and after digestion were studied by x-ray diffraction, small-angle x-ray scattering, differential scanning calorimetry and microscopy. Microscopy revealed that studied starches, which varied in their amylose content and digestion kinetics, followed different modes of attack The multidisciplinary approach allowed the nanostructural changes detected by small-angle scattering in the course of enzymic breakdown to be correlated with changes in crystallinity and functional properties.

  11. Preliminary results of total kinetic energy modelling for neutron-induced fission

    International Nuclear Information System (INIS)

    Visan, I.; Giubega, G.; Tudora, A.

    2015-01-01

    The total kinetic energy as a function of fission fragments mass TKE(A) is an important quantity entering in prompt emission calculations. The experimentally distributions of TKE(A) are referring to a limited number of fission systems and incident energies. In the present paper, a preliminary model for TKE calculation in neutron induced fission system is presented. The range of fission fragments is chosen as in the Point by Point treatment. The model needs as input only mass excesses and deformation parameters taken from available nuclear databases being based on the following approximations: total excitation energy of fully accelerated fission fragments TXE is calculated from energy balance of neutron-induced fission systems as sum of the total excitation energy at scission E*sciss and deformation energy Edef. The deformation energy at scission is given by minimizing the potential energy at the scission configuration. At the scission point, the fission system is described by two spheroidal fragments nearly touching by a pre-scission distance or neck caused by the nuclear forces between fragments. Therefore, the Columbian repulsion depending on neck and, consequently, on the fragments deformation at scission, is essentially in TKE determination. An approximation is made based on the fission modes. For the very symmetric fission, the dominant super long channel is characterized by long distance between fragments leading to low TKE values. Due to magic and double-magic shells closure, the dominant S1 fission mode for pairs with heavy fragment mass AH around 130-134 is characterized by spherical heavy fragment shape and easily deformed light fragment. The nearly spherical shape of the complementary fragments are characterized by minimum distance, and consequently to maximum TKE values. The results obtained for TKE(A) are in good agreement with existing experimental data for many neutron induced fission systems, e.g. ''2''3''3&apos

  12. Benchmarking with high-order nodal diffusion methods

    International Nuclear Information System (INIS)

    Tomasevic, D.; Larsen, E.W.

    1993-01-01

    Significant progress in the solution of multidimensional neutron diffusion problems was made in the late 1970s with the introduction of nodal methods. Modern nodal reactor analysis codes provide significant improvements in both accuracy and computing speed over earlier codes based on fine-mesh finite difference methods. In the past, the performance of advanced nodal methods was determined by comparisons with fine-mesh finite difference codes. More recently, the excellent spatial convergence of nodal methods has permitted their use in establishing reference solutions for some important bench-mark problems. The recent development of the self-consistent high-order nodal diffusion method and its subsequent variational formulation has permitted the calculation of reference solutions with one node per assembly mesh size. In this paper, we compare results for four selected benchmark problems to those obtained by high-order response matrix methods and by two well-known state-of-the-art nodal methods (the open-quotes analyticalclose quotes and open-quotes nodal expansionclose quotes methods)

  13. Experimental discovery of nodal chains

    Science.gov (United States)

    Yan, Qinghui; Liu, Rongjuan; Yan, Zhongbo; Liu, Boyuan; Chen, Hongsheng; Wang, Zhong; Lu, Ling

    2018-05-01

    Three-dimensional Weyl and Dirac nodal points1 have attracted widespread interest across multiple disciplines and in many platforms but allow for few structural variations. In contrast, nodal lines2-4 can have numerous topological configurations in momentum space, forming nodal rings5-9, nodal chains10-15, nodal links16-20 and nodal knots21,22. However, nodal lines are much less explored because of the lack of an ideal experimental realization23-25. For example, in condensed-matter systems, nodal lines are often fragile to spin-orbit coupling, located away from the Fermi level, coexist with energy-degenerate trivial bands or have a degeneracy line that disperses strongly in energy. Here, overcoming all these difficulties, we theoretically predict and experimentally observe nodal chains in a metallic-mesh photonic crystal having frequency-isolated linear band-touching rings chained across the entire Brillouin zone. These nodal chains are protected by mirror symmetry and have a frequency variation of less than 1%. We use angle-resolved transmission measurements to probe the projected bulk dispersion and perform Fourier-transformed field scans to map out the dispersion of the drumhead surface state. Our results establish an ideal nodal-line material for further study of topological line degeneracies with non-trivial connectivity and consequent wave dynamics that are richer than those in Weyl and Dirac materials.

  14. Verification of a three-dimensional neutronics model based on multi-point kinetics equations for transient problems

    Energy Technology Data Exchange (ETDEWEB)

    Park, Kyung Seok; Kim, Hyun Dae; Yeom, Choong Sub [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    A computer code for solving the three-dimensional reactor neutronic transient problems utilizing multi-point reactor kinetics equations recently developed has been developed. For evaluating its applicability, the code has been tested with typical 3-D LWR and CANDU reactor transient problems. The performance of the method and code has been compared with the results by fine and coarse meshes computer codes employing the direct methods.

  15. Tritium breeding and release-rate kinetics from neutron-irradiated lithium oxide

    International Nuclear Information System (INIS)

    Quanci, J.F.

    1989-01-01

    The research encompasses the measurement of the tritium breeding and release-rate kinetics from lithium oxide, a ceramic tritium-breeding material. A thermal extraction apparatus which allows the accurate measurement of the total tritium inventory and release rate from lithium oxide samples under different temperatures, pressures and carrier-gas compositions with an uncertainty not exceeding 3% was developed. The goal of the Lithium Blanket Module program was to determine if advanced computer codes could accurately predict the tritium production in the lithium oxide blanket of a fusion power plant. A fusion blanket module prototype was built and irradiated with a deuterium-tritium fusion-neutron source. The tritium production throughout the module was modeled with the MCNP three dimensional Monte Carlo code and was compared to the assay of the tritium bred in the module. The MCNP code accurately predicted tritium-breeding trends but underestimated the overall tritium breeding by 30%. The release rate of tritium from small grain polycrystalline sintered lithium oxides with a helium carrier gas from 300 to 450 C was found to be controlled by the first order surface desorption of monotritiated water. When small amounts of hydrogen were added to the helium carrier gas, the first order rate constant increased from the isotopic exchange of hydrogen for tritium at the lithium oxide surface occurring in parallel with the first order desorption process. The isotopic-exchange first order rate constant temperature dependence and hydrogen partial pressure dependence were evaluated

  16. Reactor kinetics calculated in the summation method and key delayed-neutron data

    International Nuclear Information System (INIS)

    Oyamatsu, Kazuhiro

    2001-01-01

    The point-reactor kinetics after a step reactivity insertion to a critical condition is solved directly form fission-product (FP) data (fission yields and decay data) for the first time. Numerical calculations are performed with the FP data in ENDF/B-VI. The inhour equation obtained directly from the FP data shows a different behavior at long periods from the one obtained from Tuttle's six-group parameter sets. The behavior is quite similar to the one obtained from the six-group parameter sets in ENDF/B-VI, that were obtained from FP data in a preliminary version of ENDF/B-VI. To identify the erroneous FP data, we examine the asymptotic form of the inhour equation at an infinitely long period. It is found that the most important precursors for long reactor periods are found 137 I, 88 Br and 87 Br. They cover more than 60% of the reactivity. It is remarkable that 137 I alone covers 30-50% depending on the fissioning system. In addition to the three precursors, 136 Te is found a candidate precursor for the peculiarity from the time dependence of the delayed neutron activity. It is recommended that the precision of their Pn values should be improved experimentally. For 137 I, 88 Br, and 87 Br, the relative uncertainty, dPn/Pn, should be decreased down to 2% and for 136 Te to 5%. (author)

  17. TASK, 1-D Multigroup Diffusion or Transport Theory Reactor Kinetics with Delayed Neutron

    International Nuclear Information System (INIS)

    Buhl, A.R.; Hermann, O.W.; Hinton, R.J.; Dodds, H.L. Jr.; Robinson, J.C.; Lillie, R.A.

    1975-01-01

    1 - Description of problem or function: TASK solves the one-dimensional multigroup form of the reactor kinetics equations, using either transport or diffusion theory and allowing an arbitrary number of delayed neutron groups. The program can also be used to solve standard static problems efficiently such as eigenvalue problems, distributed source problems, and boundary source problems. Convergence problems associated with sources in highly multiplicative media are circumvented, and such problems are readily calculable. 2 - Method of solution: TASK employs a combination scattering and transfer matrix method to eliminate certain difficulties that arise in classical finite difference approximations. As such, within-group (inner) iterations are eliminated and solution convergence is independent of spatial mesh size. The time variable is removed by Laplace transformation. (A later version will permit direct time solutions.) The code can be run either in an outer iteration mode or in closed (non-iterative) form. The running mode is dictated by the number of groups times the number of angles, consistent with available storage. 3 - Restrictions on the complexity of the problem: The principal restrictions are available storage and computation time. Since the code is flexibly-dimensioned and has an outer iteration option there are no internal restrictions on group structure, quadrature, and number of ordinates. The flexible-dimensioning scheme allows optional use of core storage. The generalized cylindrical geometry option is not complete in Version I of the code. The feedback options and omega-mode search options are not included in Version I

  18. Advanced methodology to simulate boiling water reactor transient using coupled thermal-hydraulic/neutron-kinetic codes

    Energy Technology Data Exchange (ETDEWEB)

    Hartmann, Christoph Oliver

    2016-06-13

    Coupled Thermal-hydraulic/Neutron-kinetic (TH/NK) simulations of Boiling Water Reactor transients require well validated and accurate simulation tools. The generation of cross-section (XS) libraries, depending on the individual thermal-hydraulic state parameters, is of paramount importance for coupled simulations. Problem-dependent XS-sets for 3D core simulations are being generated mainly by well validated, fast running commercial and user-friendly lattice codes such as CASMO and HELIOS. In this dissertation a computational route, based on the lattice code SCALE6/TRITON, the cross-section interface GenPMAXS, the best-estimate thermal-hydraulic system code TRACE and the core simulator PARCS, for best-estimate simulations of Boiling Water (BWR) transients has been developed and validated. The computational route has been supplemented by a subsequent uncertainty and sensitivity study based on Monte Carlo sampling and propagation of the uncertainties of input parameters to the output (SUSA code). The analysis of a single BWR fuel assembly depletion problem with PARCS using SCALE/TRITON cross-sections has been shown a good agreement with the results obtained with CASMO cross-section sets. However, to compensate the deficiencies of the interface program GenPMAXS, PYTHON scripts had to be developed to incorporate missing data, as the yields of Iodine, Xenon and Promethium, into the cross-section-data sets (PMAXS-format) generated by GenPMAXS from the SCALE/TRITON output. The results of the depletion analysis of a full BWR core with PARCS have indicated the importance of considering history effects, adequate modeling of the reflector region and the control rods, as the PARCS simulations for depleted fuel and all control rods inserted (ARI) differs significantly at the fuel assembly top and bottom. Systematic investigations with the coupled codes TRACE/PARCS have been performed to analyse the core behaviour at different thermal conditions using nuclear data (XS

  19. Development of neutronic models for the thermal hydraulics coupling of the MSFR and the calculation of effective kinetic parameters

    International Nuclear Information System (INIS)

    Laureau, Axel

    2015-01-01

    In this PhD thesis, we describe the development of innovative neutronic models for their coupling with thermal hydraulics such that they combine precision and reasonable computational times. One of the main cases where this method is applied is the Molten Salt Fast Reactor (MSFR) whose combines a fast neutron spectrum with a thorium cycle. In this fourth generation reactor, the motion of the delayed neutron precursors and the associated phenomena have to be taken into account due to the liquid fuel circulation. The starting point for these developments was the preliminary design of this type of system where a dedicated multi-physical representation was needed to study the reactor performance in steady and transient conditions. As a first step, a stationary coupling was developed. A neutronic model based on a stochastic approach was associated to a CFD (Computational Fluid Dynamics) code to solve the Navier Stokes equations for turbulent flows and the transport of the delayed neutron precursors. The impact of this precursor motion is taken into account by reconstructing the prompt shower that they generate. This approach, called by shower, views the critical reactor as a prompt subcritical reactor that amplifies a source of delayed neutrons. A second step consisted in developing a neutronic model based on a time dependent version of the fission matrices (Transient Fission Matrix or TFM) so as to enable reactor transient studies. With the TFM model, an initial computation of the matrices with a stochastic code (MCNP, SERPENT) allows the characterization of the global spatial and time dependent neutronic response of the reactor with a precision close to that of a Monte Carlo calculation. The information thus obtained is then used to calculate transients, while retaining the advantage of reduced computational time. The TFM model, which can be used for various system concepts, also allows the evaluation of effective kinetic parameters such as the effective fraction of

  20. 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)

  1. 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)

  2. Point kinetics equations for subcritical systems based on the importance function associated to an external neutron source

    International Nuclear Information System (INIS)

    Carvalho Gonçalves, Wemerson de; Martinez, Aquilino Senra; Carvalho da Silva, Fernando

    2015-01-01

    Highlights: • We define the new function importance. • We calculate the kinetic parameters Λ, β, Γ and Q to: 0.95, 0.96, 0.97, 0.98 and 0.99. • We compared the results with those obtained by the main important functions. • We found that the calculated kinetic parameters are physically consistent. - Abstract: This paper aims to determine the parameters for a new set of equations of point kinetic subcritical systems, based on the concept of importance of Heuristic Generalized Perturbation Theory (HGPT). The importance function defined here is related to both the subcriticality and the external neutron source worth (which keeps the system at steady state). The kinetic parameters defined in this work are compared with the corresponding parameters when adopting the importance functions proposed by Gandini and Salvatores (2002), Dulla et al. (2006) and Nishihara et al. (2003). Furthermore, the point kinetics equations developed here are solved for two different transients, considering the parameters obtained with different importance functions. The results collected show that there is a similar behavior of the solution of the point kinetics equations, when used with the parameters obtained by the importance functions proposed by Gandini and Salvatores (2002) and Dulla et al. (2006), specially near the criticality. However, this is not verified as the system gets farther from criticality

  3. Modified mean generation time parameter in the neutron point kinetics equations

    Energy Technology Data Exchange (ETDEWEB)

    Diniz, Rodrigo C.; Gonçalves, Alessandro C.; Rosa, Felipe S.S., E-mail: alessandro@nuclear.ufrj.br, E-mail: frosa@if.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    This paper proposes an approximation for the modified point kinetics equations proposed by NUNES et. al, 2015, through the adjustment of a kinetic parameter. This approximation consists of analyzing the terms of the modified point kinetics equations in order to identify the least important ones for the solution, resulting in a modification of the mean generation time parameter that incorporates all influences of the additional terms of the modified kinetics. This approximation is applied on the inverse kinetics, to compare the results with the inverse kinetics from the modified kinetics in order to validate the proposed model. (author)

  4. Modified mean generation time parameter in the neutron point kinetics equations

    International Nuclear Information System (INIS)

    Diniz, Rodrigo C.; Gonçalves, Alessandro C.; Rosa, Felipe S.S.

    2017-01-01

    This paper proposes an approximation for the modified point kinetics equations proposed by NUNES et. al, 2015, through the adjustment of a kinetic parameter. This approximation consists of analyzing the terms of the modified point kinetics equations in order to identify the least important ones for the solution, resulting in a modification of the mean generation time parameter that incorporates all influences of the additional terms of the modified kinetics. This approximation is applied on the inverse kinetics, to compare the results with the inverse kinetics from the modified kinetics in order to validate the proposed model. (author)

  5. Avoided intersections of nodal lines

    International Nuclear Information System (INIS)

    Monastra, Alejandro G; Smilansky, Uzy; Gnutzmann, Sven

    2003-01-01

    We consider real eigenfunctions of the Schroedinger operator in 2D. The nodal lines of separable systems form a regular grid, and the number of nodal crossings equals the number of nodal domains. In contrast, for wavefunctions of non-integrable systems nodal intersections are rare, and for random waves, the expected number of intersections in any finite area vanishes. However, nodal lines display characteristic avoided crossings which we study in this work. We define a measure for the avoidance range and compute its distribution for the random wave ensemble. We show that the avoidance range distribution of wavefunctions of chaotic systems follows the expected random wave distributions, whereas for wavefunctions of classically integrable but quantum non-separable systems, the distribution is quite different. Thus, the study of the avoidance distribution provides more support to the conjecture that nodal structures of chaotic systems are reproduced by the predictions of the random wave ensemble

  6. Neutron emission effects on final fragments mass and kinetic energy distribution from low energy fission of {sup 234}U

    Energy Technology Data Exchange (ETDEWEB)

    Montoya, M.; Rojas, J. [Instituto Peruano de Energia Nuclear, Av. Canada 1470, Lima 41 (Peru); Lobato, I. [Facultad de Ciencias, Universidad Nacional de Ingenieria, Av. Tupac Amaru 210, Apartado Postal 31-139, Lima (Peru)]. e-mail: mmontoya@ipen.gob.pe

    2008-07-01

    The standard deviation of the final kinetic energy distribution ({sigma}{sub e}) as a function of mass of final fragments (m) from low energy fission of {sup 234}U, measured with the Lohengrin spectrometer by Belhafaf et al., presents a peak around m = 109 and another around m = 122. The authors attribute the first peak to the evaporation of a large number of neutrons around the corresponding mass number, i.e. there is no peak on the standard deviation of the primary kinetic energy distribution ({sigma}{sub E}) as a function of primary fragment mass (A). The second peak is attributed to a real peak on {sigma}{sub E}(A). However, theoretical calculations related to primary distributions made by H.R. Faust and Z. Bao do not suggest any peak on {sigma}{sub E}(A). In order to clarify this apparent controversy, we have made a numerical experiment in which the masses and the kinetic energy of final fragments are calculated, assuming an initial distribution of the kinetic energy without structures on the standard deviation as function of fragment mass. As a result we obtain a pronounced peak on {sigma}{sub e} (m) curve around m = 109, a depletion from m = 121 to m = 129, and an small peak around m = 122, which is not as great as that measured by Belhafaf et al. Our simulation also reproduces the experimental results on the yield of the final mass Y(m), the average number of emitted neutrons as a function of the provisional mass (calculated from the values of the final kinetic energy of the complementary fragments) and the average value of fragment kinetic energy as a function of the final mass. From our results we conclude that there are no peaks on the {sigma}{sub E} (A) curve, and the observed peaks on {sigma}{sub e} (m) are due to the emitted neutron multiplicity and the variation of the average fragment kinetic energy as a function of primary fragment mass. (Author)

  7. Tritium breeding and release-rate kinetics from neutron-irradiated lithium oxide

    International Nuclear Information System (INIS)

    Quanci, J.F.

    1989-01-01

    The research encompasses the measurement of the tritium breeding and release-rate kinetics from lithium oxide, a ceramic tritium-breeding material. A thermal extraction apparatus which allows the accurate measurement of the total tritium inventory and release rate from lithium oxide samples under different temperatures, pressures and carrier-gas compositions with an uncertainty not exceeding 3% was developed. The goal of the Lithium Blanket Module program was to determine if advanced computer codes could accurately predict the tritium production in the lithium oxide blanket of a fusion power plant. A fusion blanket module prototype, was built and irradiated with a deuterium-tritium fusion-neutron source. The tritium production throughout the module was modeled with the MCNP three dimensional Monte Carlo code and was compared to the assay of the tritium bred in the module. The MCNP code accurately predicted tritium-breeding trends but underestimated the overall tritium breeding by 30%. The release rate of tritium from small grain polycrystalline sintered lithium oxide with a helium carrier gas from 300 to 450 C was found to be controlled by the first order surface desorption of mono-tritiated water. When small amounts of hydrogen were added to the helium carrier gas, the first order rate constant increased from the isotopic exchange of hydrogen for tritium at the lithium oxide surface occurring in parallel with the first order desorption process. The isotopic-exchange first order rate constant temperature dependence and hydrogen partial pressure dependence were evaluated. Large single crystals of lithium oxide were fabricated by the vacuum fusion technique. The release rate of tritium from the large single crystals was found to be controlled by diffusion, and the mixed diffusion-desorption controlled release regime

  8. Disclosure of the oscillations in kinetics of the reactor pressure vessel steel damage at fast neutron intensity decreasing

    Science.gov (United States)

    Krasikov, E.; Nikolaenko, V.

    2017-01-01

    Fast neutron intensity influence on reactor materials radiation damage is a critically important question in the problem of the correct use of the accelerated irradiation tests data for substantiation of the materials workability in real irradiation conditions that is low neutron intensity. Investigations of the fast neutron intensity (flux) influence on radiation damage and experimental data scattering reveal the existence of non-monotonous sections in kinetics of the reactor pressure vessels (RPV) steel damage. Discovery of the oscillations as indicator of the self-organization processes presence give reasons for new ways searching on reactor pressure vessel (RPV) steel radiation stability increasing and attempt of the self-restoring metal elaboration. Revealing of the wavelike process in the form of non monotonous parts of the kinetics of radiation embrittlement testifies that periodic transformation of the structure take place. This fact actualizes the problem of more precise definition of the RPV materials radiation embrittlement mechanisms and gives reasons for search of the ways to manage the radiation stability (nanostructuring and so on to stimulate the radiation defects annihilation), development of the means for creating of more stableness self recovering smart materials.

  9. Bilinear nodal transport method in weighted diamond difference form

    International Nuclear Information System (INIS)

    Azmy, Y.Y.

    1987-01-01

    Nodal methods have been developed and implemented for the numerical solution of the discrete ordinates neutron transport equation. Numerical testing of these methods and comparison of their results to those obtained by conventional methods have established the high accuracy of nodal methods. Furthermore, it has been suggested that the linear-linear approximation is the most computationally efficient, practical nodal approximation. Indeed, this claim has been substantiated by comparing the accuracy in the solution, and the CPU time required to achieve convergence to that solution by several nodal approximations, as well as the diamond difference scheme. Two types of linear-linear nodal methods have been developed in the literature: analytic linear-linear (NLL) methods, in which the transverse-leakage terms are derived analytically, and approximate linear-linear (PLL) methods, in which these terms are approximated. In spite of their higher accuracy, NLL methods result in very complicated discrete-variable equations that exhibit a high degree of coupling, thus requiring special solution algorithms. On the other hand, the sacrificed accuracy in PLL methods is compensated for by the simple discrete-variable equations and diamond-difference-like solution algorithm. In this paper the authors outline the development of an NLL nodal method, the bilinear method, which can be written in a weighted diamond difference form with one spatial weight per dimension that is analytically derived rather than preassigned in an ad hoc fashion

  10. Five-point form of the nodal diffusion method and comparison with finite-difference

    International Nuclear Information System (INIS)

    Azmy, Y.Y.

    1988-01-01

    Nodal Methods have been derived, implemented and numerically tested for several problems in physics and engineering. In the field of nuclear engineering, many nodal formalisms have been used for the neutron diffusion equation, all yielding results which were far more computationally efficient than conventional Finite Difference (FD) and Finite Element (FE) methods. However, not much effort has been devoted to theoretically comparing nodal and FD methods in order to explain the very high accuracy of the former. In this summary we outline the derivation of a simple five-point form for the lowest order nodal method and compare it to the traditional five-point, edge-centered FD scheme. The effect of the observed differences on the accuracy of the respective methods is established by considering a simple test problem. It must be emphasized that the nodal five-point scheme derived here is mathematically equivalent to previously derived lowest order nodal methods. 7 refs., 1 tab

  11. The SINTRAN III NODAL system

    International Nuclear Information System (INIS)

    Skaali, T.B.

    1980-10-01

    NODAL is a high level programming language based on FOCAL and SNOBOL4, with some influence from BASIC. The language was developed to operate on the computer network controlling the SPS accelerator at CERN. NODAL is an interpretive language designed for interactive use. This is the most important aspect of the language, and is reflected in its structure. The interactive facilities make it possible to write, debug and modify programs much faster than with compiler based languages like FORTRAN and ALGOL. Apart from a few minor modifications, the basic part of the Oslo University NODAL system does not differ from the CERN version. However, the Oslo University implementation has been expanded with new functions which enable the user to execute many of the SINTRAN III monitor calls from the NODAL level. In particular the most important RT monitor calls have been implemented in this way, a property which renders possible the use of NODAL as a RT program administrator. (JIW)

  12. Performance of the coupled thermalhydraulics/neutron kinetics code R/P/C on workstation clusters and multiprocessor systems

    International Nuclear Information System (INIS)

    Hammer, C.; Paffrath, M.; Boeer, R.; Finnemann, H.; Jackson, C.J.

    1996-01-01

    The light water reactor core simulation code PANBOX has been coupled with the transient analysis code RELAP5 for the purpose of performing plant safety analyses with a three-dimensional (3-D) neutron kinetics model. The system has been parallelized to improve the computational efficiency. The paper describes the features of this system with emphasis on performance aspects. Performance results are given for different types of parallelization, i. e. for using an automatic parallelizing compiler, using the portable PVM platform on a workstation cluster, using PVM on a shared memory multiprocessor, and for using machine dependent interfaces. (author)

  13. Monte Carlo simulation of fission yields, kinetic energy, fission neutron spectrum and decay γ-ray spectrum for 232Th(n,f) reaction induced by 3H(d,n) 4He neutron source

    International Nuclear Information System (INIS)

    Zheng Wei; Zeen Yao; Changlin Lan; Yan Yan; Yunjian Shi; Siqi Yan; Jie Wang; Junrun Wang; Jingen Chen; Chinese Academy of Sciences, Shanghai

    2015-01-01

    Monte Carlo transport code Geant4 has been successfully utilised to study of neutron-induced fission reaction for 232 Th in the transport neutrons generated from 3 H(d,n) 4 He neutron source. The purpose of this work is to examine the applicability of Monte Carlo simulations for the computation of fission reaction process. For this, Monte Carlo simulates and calculates the characteristics of fission reaction process of 232 Th(n,f), such as the fission yields distribution, kinetic energy distribution, fission neutron spectrum and decay γ-ray spectrum. This is the first time to simulate the process of neutron-induced fission reaction using Geant4 code. Typical computational results of neutron-induced fission reaction of 232 Th(n,f) reaction are presented. The computational results are compared with the previous experimental data and evaluated nuclear data to confirm the certain physical process model in Geant4 of scientific rationality. (author)

  14. On the extension of the analytic nodal diffusion solver ANDES to sodium fast reactors

    International Nuclear Information System (INIS)

    Ochoa, R.; Herrero, J.J.; Garcia-Herranz, N.

    2011-01-01

    Within the framework of the Collaborative Project for a European Sodium Fast Reactor, the reactor physics group at UPM is working on the extension of its in-house multi-scale advanced deterministic code COBAYA3 to Sodium Fast Reactors (SFR). COBAYA3 is a 3D multigroup neutron kinetics diffusion code that can be used either as a pin-by-pin code or as a stand-alone nodal code by using the analytic nodal diffusion solver ANDES. It is coupled with thermal-hydraulics codes such as COBRA-TF and FLICA, allowing transient analysis of LWR at both fine-mesh and coarse-mesh scales. In order to enable also 3D pin-by-pin and nodal coupled NK-TH simulations of SFR, different developments are in progress. This paper presents the first steps towards the application of COBAYA3 to this type of reactors. ANDES solver, already extended to triangular-Z geometry, has been applied to fast reactor steady-state calculations. The required cross section libraries were generated with ERANOS code for several configurations. Here some of the limitations encountered when attempting to apply the Analytical Coarse Mesh Finite Difference (ACMFD) method - implemented inside ANDES - to fast reactor calculations are discussed and the sensitivity of the method to the energy-group structure is studied. In order to reinforce some of the conclusions obtained two calculations are presented. The first one involves a 3D mini-core model in 33 groups, where the ANDES solver presents several issues. And secondly, a benchmark from the NEA for a small 3D FBR in hexagonal-Z geometry in 4 energy groups is used to verify the good convergence of the code in a few-energy-group structure. (author)

  15. Time-resolved small angle neutron scattering measurements of asphaltene nanoparticle aggregation kinetics in incompatible crude oil mixtures

    International Nuclear Information System (INIS)

    Mason, Thomas G.; Lin, Min Y.

    2003-01-01

    We use time-resolved-small angle neutron scattering to study the kinetics of asphaltene nanoparticle aggregation in incompatible crude oil mixtures. We induce asphaltene aggregation by mixing asphaltene-rich Syrian crude oil (SACO) with a paraffinic British crude oil and observe the scattered neutron intensity, I, as a function of wave number, q, over times, t, ranging from twenty minutes to about a week. We observe a growth in I at low q as the nanoscale asphaltenes agglomerate into microscale aggregates and interpret this growth as an increase in surface scattering from the aggregates. We fit I(q,t) to an empirical model and measure the growth in the power-law exponent, α, associated with the low-q logarithmic slope of I(q). We define a time, τ α , associated with the first appearance of the aggregates when α>3; τ α increases as a function of the volume fraction, φ m , of SACO in the mixture. The surface scattering intensity initially increases and then saturates at long times when the aggregate structures no longer evolve at the length scales we probe. Based on this saturation, we define a time scale, τ I , which is larger than τ α but has essentially the same dependence on φ m . We interpret τ α (φ m ) and τ I (φ m ) in terms of a simple aggregation model based on diffusion-limited kinetics and a repulsive potential barrier that models the effective solvent quality

  16. Neutron kinetics developments of the SIMMER-III safety code for APS application

    International Nuclear Information System (INIS)

    Rineiski, A.; Kiefhaber, E.; Merk, B.; Maschek, W.

    2000-01-01

    Recent developments extending the capabilities of the SIMMER-III code for dealing with transients and accidents in an ADS are presented. The impact of weighting functions on the point-kinetics parameters at steady-state is shown. Some preliminary results of using a space-time kinetics model for beam-trip related transients are highlighted. (orig.)

  17. ANGRA-1 neutron kinetics model at BOL using WIMSD-5B and PARCS V2.7 codes

    International Nuclear Information System (INIS)

    Hamers, Adolfo R.; Reis, Patricia A.L.; Rodrigues, Thiago D.A.; Pereira, Claubia; Costa, Antonella L.

    2015-01-01

    A steady-state neutron kinetics model of the Angra-1 NPP at BOL (Beginning Of Life) has been developed with the PARCS V2.7 neutron diffusion code. The information of the burnable poison rods, fuel enrichments and control rod banks distributions within the core have been taken from the Angra-1 FSAR (Final Safety Analysis Report) and implemented in the model. The macroscopic cross sections for the fast and thermal neutron groups have been calculated with the WIMSD-5B lattice cell code. The cross sections were obtained for the rodded and unrodded cases for each composition in the core. In order to establish the initial steady-state, an eigenvalue was made with the PARCS V2.7 code for three steady-state scenario cases reported at the FSAR; a K eff of 1.0733 was obtained for the unrodded case, K eff of 1.0718 for a 24% of bank D inserted case and K eff of 0.8512 for the full rodded case. The normalized core power density distributions were obtained and compared with the corresponding FSAR case. (author)

  18. Thermal hydraulic and neutron kinetic simulation of the Angra 2 reactor using a RELAP5/PARCS coupled model

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Patricia A.L.; Costa, Antonella L.; Hamers, Adolfo R.; Pereira, Claubia; Rodrigues, Thiago D.A.; Mantecon, Javier G.; Veloso, Maria A.F., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: adolforomerohamers@hotmail.com, E-mail: claubia@nuclear.ufmg.br, E-mail: thiagodanielbh@gmail.com, E-mail: mantecon1987@gmail.com, E-mail: dora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Belo Horizonte (Brazil); Miro, Rafael; Verdu, Gumersindo, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidad Politecnica de Valencia (Spain). Departamento de Ingenieria Quimica y Nuclear

    2015-07-01

    The computational advances observed in the last two decades have been provided direct impact on the researches related to nuclear simulations, which use several types of computer codes, including coupled between them, allowing representing with very accuracy the behavior of nuclear plants. Studies of complex scenarios in nuclear reactors have been improved by the use of thermal-hydraulic (TH) and neutron kinetics (NK) coupled codes. This technique consists in incorporating three-dimensional (3D) neutron modeling of the reactor core into codes, mainly to simulate transients that involve asymmetric core spatial power distributions and strong feedback effects between neutronics and reactor thermal-hydraulics. Therefore, this work presents preliminary results of TH RELAP5 and the NK PARCS calculations applied to model of the Angra 2 reactor. The WIMSD-5B code has been used to generate the macroscopic cross sections used in the NK code. The results obtained are satisfactory and represent important part of the development of this methodology. The next step is to couple the codes. (author)

  19. Object kinetic Monte Carlo model for neutron and ion irradiation in tungsten: Impact of transmutation and carbon impurities

    Science.gov (United States)

    Castin, N.; Bonny, G.; Bakaev, A.; Ortiz, C. J.; Sand, A. E.; Terentyev, D.

    2018-03-01

    We upgrade our object kinetic Monte Carlo (OKMC) model, aimed at describing the microstructural evolution in tungsten (W) under neutron and ion irradiation. Two main improvements are proposed based on recently published atomistic data: (a) interstitial carbon impurities, and their interaction with radiation-induced defects (point defect clusters and loops), are more accurately parameterized thanks to ab initio findings; (b) W transmutation to rhenium (Re) upon neutron irradiation, impacting the diffusivity of radiation defects, is included, also relying on recent atomistic data. These essential amendments highly improve the portability of our OKMC model, providing a description for the formation of SIA-type loops under different irradiation conditions. The model is applied to simulate neutron and ion irradiation in pure W samples, in a wide range of fluxes and temperatures. We demonstrate that it performs a realistic prediction of the population of TEM-visible voids and loops, as compared to experimental evidence. The impact of the transmutation of W to Re, and of carbon trapping, is assessed.

  20. Results of comparative RBMK neutron computation using VNIIEF codes (cell computation, 3D statics, 3D kinetics). Final report

    Energy Technology Data Exchange (ETDEWEB)

    Grebennikov, A.N.; Zhitnik, A.K.; Zvenigorodskaya, O.A. [and others

    1995-12-31

    In conformity with the protocol of the Workshop under Contract {open_quotes}Assessment of RBMK reactor safety using modern Western Codes{close_quotes} VNIIEF performed a neutronics computation series to compare western and VNIIEF codes and assess whether VNIIEF codes are suitable for RBMK type reactor safety assessment computation. The work was carried out in close collaboration with M.I. Rozhdestvensky and L.M. Podlazov, NIKIET employees. The effort involved: (1) cell computations with the WIMS, EKRAN codes (improved modification of the LOMA code) and the S-90 code (VNIIEF Monte Carlo). Cell, polycell, burnup computation; (2) 3D computation of static states with the KORAT-3D and NEU codes and comparison with results of computation with the NESTLE code (USA). The computations were performed in the geometry and using the neutron constants presented by the American party; (3) 3D computation of neutron kinetics with the KORAT-3D and NEU codes. These computations were performed in two formulations, both being developed in collaboration with NIKIET. Formulation of the first problem maximally possibly agrees with one of NESTLE problems and imitates gas bubble travel through a core. The second problem is a model of the RBMK as a whole with imitation of control and protection system controls (CPS) movement in a core.

  1. Analytic function expansion nodal method for nuclear reactor core design

    International Nuclear Information System (INIS)

    Noh, Hae Man

    1995-02-01

    In most advanced nodal methods the transverse integration is commonly used to reduce the multi-dimensional diffusion equation into equivalent one- dimensional diffusion equations when derving the nodal coupling equations. But the use of the transverse integration results in some limitations. The first limitation is that the transverse leakage term which appears in the transverse integration procedure must be appropriately approximated. The second limitation is that the one-dimensional flux shapes in each spatial direction resulted from the nodal calculation are not accurate enough to be directly used in reconstructing the pinwise flux distributions. Finally the transverse leakage defined for a non-rectangular node such as a hexagonal node or a triangular node is too complicated to be easily handled and may contain non-physical singular terms of step-function and delta-function types. In this thesis, the Analytic Function Expansion Nodal (AFEN) method and its two variations : the Polynomial Expansion Nodal (PEN) method and the hybrid of the AFEN and PEN methods, have been developed to overcome the limitations of the transverse integration procedure. All of the methods solve the multidimensional diffusion equation without the transverse integration. The AFEN method which we believe is the major contribution of this study to the reactor core analysis expands the homogeneous flux distributions within a node in non-separable analytic basis functions satisfying the neutron diffusion equations at any point of the node and expresses the coefficients of the flux expansion in terms of the nodal unknowns which comprise a node-average flux, node-interface fluxes, and corner-point fluxes. Then, the nodal coupling equations composed of the neutron balance equations, the interface current continuity equations, and the corner-point leakage balance equations are solved iteratively to determine all the nodal unknowns. Since the AFEN method does not use the transverse integration in

  2. Fission Fragment Mass Distributions and Total Kinetic Energy Release of 235-Uranium and 238-Uranium in Neutron-Induced Fission at Intermediate and Fast Neutron Energies

    Energy Technology Data Exchange (ETDEWEB)

    Duke, Dana Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-11-12

    This Ph.D. dissertation describes a measurement of the change in mass distributions and average total kinetic energy (TKE) release with increasing incident neutron energy for fission of 235U and 238U. Although fission was discovered over seventy-five years ago, open questions remain about the physics of the fission process. The energy of the incident neutron, En, changes the division of energy release in the resulting fission fragments, however, the details of energy partitioning remain ambiguous because the nucleus is a many-body quantum system. Creating a full theoretical model is difficult and experimental data to validate existing models are lacking. Additional fission measurements will lead to higher-quality models of the fission process, therefore improving applications such as the development of next-generation nuclear reactors and defense. This work also paves the way for precision experiments such as the Time Projection Chamber (TPC) for fission cross section measurements and the Spectrometer for Ion Determination in Fission (SPIDER) for precision mass yields.

  3. Implementation into a CFD code of neutron kinetics and fuel pin models for nuclear reactor transient analyses

    International Nuclear Information System (INIS)

    Chen Zhao; Chen, Xue-Nong; Rineiski, Andrei; Zhao Pengcheng; Chen Hongli

    2014-01-01

    Safety analysis is an important tool for justifying the safety of nuclear reactors. The traditional method for nuclear reactor safety analysis is performed by means of system codes, which use one-dimensional lumped-parameter method to model real reactor systems. However, there are many multi-dimensional thermal-hydraulic phenomena cannot be predicated using traditional one-dimensional system codes. This problem is extremely important for pool-type nuclear systems. Computational fluid dynamics (CFD) codes are powerful numerical simulation tools to solve multi-dimensional thermal-hydraulics problems, which are widely used in industrial applications for single phase flows. In order to use general CFD codes to solve nuclear reactor transient problems, some additional models beyond general ones are required. Neutron kinetics model for power calculation and fuel pin model for fuel pin temperature calculation are two important models of these additional models. The motivation of this work is to develop an advance numerical simulation method for nuclear reactor safety analysis by implementing neutron kinetics model and fuel pin model into general CFD codes. In this paper, the Point Kinetics Model (PKM) and Fuel Pin Model (FPM) are implemented into a general CFD code FLUENT. The improved FLUENT was called as FLUENT/PK. The mathematical models and implementary method of FLUENT/PK are descripted and two demonstration application cases, e.g. the unprotected transient overpower (UTOP) accident of a Liquid Metal cooled Fast Reactor (LMFR) and the unprotected beam overpower (UBOP) accident of an Accelerator Driven System (ADS), are presented. (author)

  4. Solution of Point Reactor Neutron Kinetics Equations with Temperature Feedback by Singularly Perturbed Method

    Directory of Open Access Journals (Sweden)

    Wenzhen Chen

    2013-01-01

    Full Text Available The singularly perturbed method (SPM is proposed to obtain the analytical solution for the delayed supercritical process of nuclear reactor with temperature feedback and small step reactivity inserted. The relation between the reactivity and time is derived. Also, the neutron density (or power and the average density of delayed neutron precursors as the function of reactivity are presented. The variations of neutron density (or power and temperature with time are calculated and plotted and compared with those by accurate solution and other analytical methods. It is shown that the results by the SPM are valid and accurate in the large range and the SPM is simpler than those in the previous literature.

  5. Kinetic calculations for miniature neutron source reactor using analytical and numerical techniques

    International Nuclear Information System (INIS)

    Ampomah-Amoako, E.

    2008-06-01

    The analytical methods, step change in reactivity and ramp change in reactivity as well as numerical methods, fixed point iteration and Runge Kutta-gill were used to simulate the initial build up of neutrons in a miniature neutron source reactor with and without temperature feedback effect. The methods were modified to include photo neutron concentration. PARET 7.3 was used to simulate the transients behaviour of Ghana Research Reactor-1. The PARET code was capable of simulating the transients for 2.1 mk and 4 mk insertions of reactivity with peak powers of 49.87 kW and 92.34 kW, respectively. PARET code however failed to simulate 6.71 mk of reactivity which was predicted by Akaho et al through TEMPFED. (au)

  6. NODAL interpreter for CP/M

    International Nuclear Information System (INIS)

    Oide, Katsunobu.

    1982-11-01

    A NODAL interpreter which works under CP/M operating system is made for microcomputers. This interpreter language named NODAL-80 has a similar structure to the NODAL of SPS, but its commands, variables, and expressions are modified to increase the flexibility of programming. NODAL-80 also uses a simple intermediate code to make the execution speed fast without imposing any restriction on the dynamic feature of NODAL language. (author)

  7. A nodal model for the simulation of a PWR core

    International Nuclear Information System (INIS)

    Souza Pinto, R. de.

    1981-06-01

    A computer program FORTRAN language was developed to simulate the neutronic and thermal-hydraulic transient behaviour of a PWR reactor core. The reator power is calculated using a point kinectics model with six groups of delayed neutron precursors. The fission product decay heat was considered assuming three effective decay heat groups. A nodal model was employed for the treatment of heat transfer in the fuel rod, with integration of the heat equation by the lumped parameter technique. Axial conduction was neglected. A single-channel nodal model was developed for the thermo-hydrodynamic simulation using mass and energy conservation equations for the control volumes. The effect of the axial pressure variation was neglected. The computer program was tested, with good results, through the simulation of the transient behaviour of postulated accidents in a typical PWR. (Author) [pt

  8. On solution to the problem of reactor kinetics with delayed neutrons by Monte Carlo method

    International Nuclear Information System (INIS)

    Kyncl, Jan

    2013-07-01

    The initial value problem is addressed for the neutron transport equation and for the system of equations that describe the behaviour of emitters of delayed neutrons. Examination of the solution to this problem is based on several main assumptions concerning the behaviour of macroscopic effective cross-sections describing the reaction of the neutron with the medium, the temperature of medium and the remaining parameters of the equations. Formulation of these assumptions is adequately general and is in agreement with the properties of all known models of the physical quantities involved. Among others, the assumptions admit dependence of the macroscopic effective cross-sections and temperature on spatial coordinates and time that can be arbitrary to a great extent. The problem starts from a set of integro-differential equations. This problem is first transposed into the equivalent problem of solving a linear integral equation for neutron flux. This integral equation is solved by the method of successive iterations and its uniqueness is demonstrated. Numeric solution to the integral equation by Monte Carlo method consists in finding a functional of the exact solution. For this, a random process is set up and some random variables are proposed. Then it is demonstrated that each of these variables is an unbiased estimator of that functional. (author)

  9. Kinetic theory of neutron gases. Progress report, October 1, 1974--June 30, 1975

    International Nuclear Information System (INIS)

    Goldstein, H.; Cacuci, D.; Ku, L.P.; Ostrow, S.; Peng, W.

    1975-01-01

    Interim progress reports are presented on four investigations involving the transport of fast neutrons. The common theme tying together the four research projects is the manner in which details of microscopic cross sections affect the macroscopic transport behavior of neutron gases in material media. In the first, the classic problem of transport and slowing down in a non-hydrogenous medium with constant cross section is being reexamined, using Case's method of singular eigenfunctions. Both the asymptotic limit derived by Wick and the age-theory extreme have been recovered, and higher order corrections to these are being derived. Semi-numerical techniques have been developed which should permit calculation of the flux over wide ranges of distance and lethargy. The second investigation is the study of the effects of square-well minima in otherwise constant cross sections. A specialized moments-method code has been developed for the purpose, which gives the flux as far out as 80 to 100 mfp from the source. A simple but highly accurate model has been developed to include the effects of the minima, and it is now feasible to predict under what conditions a given minimum will significantly affect the penetration. In the third study Monte Carlo has been used to obtain the characteristics of neutron histories penetrating to various distances and energies in iron. The role of such cross section phenomena as inelastic scattering and cross section minima is being elucidated. In the last investigation, the discrete energy S/sub N/ method is being applied to various transport problems difficult of access by other methods. For example, an examination is under way of the dependence of fast-neutron albedo on the form of representation of the scattered neutron distribution and on the fine structure of the cross section. (U.S.)

  10. OBJECT KINETIC MONTE CARLO SIMULATIONS OF RADIATION DAMAGE IN TUNGSTEN SUBJECTED TO NEUTRON FLUX WITH PKA SPECTRUM CORRESPONDING TO THE HFIR

    Energy Technology Data Exchange (ETDEWEB)

    Nandipati, Giridhar; Setyawan, Wahyu; Heinisch, Howard L.; Roche, Kenneth J.; Kurtz, Richard J.; Wirth, Brian D.

    2015-12-31

    The objective of this work is to study the damage accumulation in pure tungsten (W) subjected to neutron bombardment with a primary knock-on atom (PKA) spectrum corresponding to the High Flux Isotope Reactor (HFIR), using the object kinetic Monte Carlo (OKMC) method.

  11. Kinetics of the urea-induced dissociation of human plasma alpha-2-macroglobulin as measured by small-angle neutron scattering

    DEFF Research Database (Denmark)

    Sjöberg, B.; Pap, S.; Järnberg, S.-E.

    1991-01-01

    The kinetics of the urea-induced dissociation of human plasma alpha-2-macroglobulin into two half-molecular fragments was investigated at 21.0-degrees-C by using small-angle neutron scattering. The relative change in molecular mass that occurs upon dissociation was monitored by recording the forw...

  12. Neutron diffraction study of the formation kinetics of ordered antiphase domains in titanium carbohydride TiCxHy

    International Nuclear Information System (INIS)

    Khidirov, I.

    2015-01-01

    The kinetics of formation and growth of ordered antiphase domains (APDs) in titanium carbohydride TiC 0.50 H 0.21 has been investigated by neutron diffraction. A model of ordered APDs is proposed. It is established that the pronounced ordering of interstitial atoms and APDs begin at 450°C. It is shown that the period of ordered APDs (P ≈ 10–12) is independent of the exposure time at a constant temperature. It is found that the temperature of ordered APDs, T OAPD , increases nonlinearly with an increase in the carbon concentration in the range 0.50 ≤ C/Ti ≤ 0.70. The formation temperature of ordered APDs is found to correlate with the concentration dependence of the order–disorder transition temperature and be 0.60 of the order–disorder transition temperature: T APD = 0.60Τ C

  13. Transient cases analyses of the TRIGA IPR-R1 using thermal hydraulic and neutron kinetic coupled codes

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Scari, Maria E., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: melizabethscari@yahoo.com [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Belo Horizonte (Brazil); Miro, Rafael; Verdu, Gumersindo, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidad Politecnica de Valencia (Spain). Departamento de Ingenieria Quimica y Nuclear

    2015-07-01

    Simulations and analyses of nuclear reactors have been improved by utilization of coupled thermal-hydraulic (TH) and neutron kinetics (NK) system codes especially to simulate transients that involve strong feedback effects between NK and TH. The TH-NK coupling technique was initially developed and used to simulate the behavior of power reactors; however, several coupling methodologies are now being applied for research reactors. This work presents the coupling methodology application between RELAP5 and PARCS codes using as a model the TRIGA IPR-R1 research reactor. Analyses of steady state and transient conditions and comparisons with results from simulations using only the RELAP5 code are being presented in this paper. (author)

  14. Relation between nonlinear or 'not-linear' characteristics in nuclear kinetics and noise analysis of neutron flux

    International Nuclear Information System (INIS)

    Kataoka, H.

    1975-01-01

    The 'not-linear' or '2nd-class-nonlinear' characteristics in nuclear reactor kinetics with the feedback effect in the high-power operation and induce the increase in the amplitude of the neutron flux noise, specially in the very low frequency region. The fundamental behaviour of 'not-linear' characteristics and its effect for the reactor noise was investigated. Application of the reactor noise analysis technique to power reactors has not been successful because of unknown large disagreement between the result of the conventional theoretical analysis and the experimental facts. When the cause of this discrepancy is clear, reactor noise analysis techniques can be effectively applied to instrumentation, control, monitoring and diagnosis of power reactors. (author)

  15. Investigation on generalized Variational Nodal Methods for heterogeneous nodes

    International Nuclear Information System (INIS)

    Wang, Yongping; Wu, Hongchun; Li, Yunzhao; Cao, Liangzhi; Shen, Wei

    2017-01-01

    Highlights: • We developed two heterogeneous nodal methods based on the Variational Nodal Method. • Four problems were solved to evaluate the two heterogeneous nodal methods. • The function expansion method is good at treating continuous-changing heterogeneity. • The finite sub-element method is good at treating discontinuous-changing heterogeneity. - Abstract: The Variational Nodal Method (VNM) is generalized for heterogeneous nodes and applied to four kinds of problems including Molten Salt Reactor (MSR) core problem with continuous cross section profile, Pressurized Water Reactor (PWR) control rod cusping effect problem, PWR whole-core pin-by-pin problem, and heterogeneous PWR core problem without fuel-coolant homogenization in each pin cell. Two approaches have been investigated for the treatment of the nodal heterogeneity in this paper. To concentrate on spatial heterogeneity, diffusion approximation was adopted for the angular variable in neutron transport equation. To provide demonstrative numerical results, the codes in this paper were developed in slab geometry. The first method, named as function expansion (FE) method, expands nodal flux by orthogonal polynomials and the nodal cross sections are also expressed as spatial depended functions. The second path, named as finite sub-element (FS) method, takes advantage of the finite-element method by dividing each node into numbers of homogeneous sub-elements and expanding nodal flux into the combination of linear sub-element trial functions. Numerical tests have been carried out to evaluate the ability of the two nodal (coarse-mesh) heterogeneous VNMs by comparing with the fine-mesh homogeneous VNM. It has been demonstrated that both heterogeneous approaches can handle heterogeneous nodes. The FE method is good at continuous-changing heterogeneity as in the MSR core problem, while the FS method is good at discontinuous-changing heterogeneity such as the PWR pin-by-pin problem and heterogeneous PWR core

  16. Comparison Of A Neutron Kinetics Parameter For A Polyethylene Moderated Highly Enriched Uranium System

    Energy Technology Data Exchange (ETDEWEB)

    McKenzie, IV, George Espy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Goda, Joetta Marie [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Grove, Travis Justin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Sanchez, Rene Gerardo [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-04-17

    This paper examines the comparison of MCNP® code’s capability to calculate kinetics parameters effectively for a thermal system containing highly enriched uranium (HEU). The Rossi-α parameter was chosen for this examination because it is relatively easy to measure as well as easy to calculate using MCNP®’s kopts card. The Rossi-α also incorporates many other parameters of interest in nuclear kinetics most of which are more difficult to precisely measure. The comparison looks at two different nuclear data libraries for comparison to the experimental data. These libraries are ENDF/BVI (.66c) and ENDF/BVII (.80c).

  17. Using nodal expansion method in calculation of reactor core with square fuel assemblies

    International Nuclear Information System (INIS)

    Abdollahzadeh, M. Y.; Boroushaki, M.

    2009-01-01

    A polynomial nodal method is developed to solve few-group neutron diffusion equations in cartesian geometry. In this article, the effective multiplication factor, group flux and power distribution based on the nodal polynomial expansion procedure is presented. In addition, by comparison of the results the superiority of nodal expansion method on finite-difference and finite-element are fully demonstrated. The comparison of the results obtained by these method with those of the well known benchmark problems have shown that they are in very good agreement.

  18. Prototype coupling of the CFD software ansys CFX with the 3D neutron kinetic core model DYN3D - 249

    International Nuclear Information System (INIS)

    Kliem, S.; Rohde, U.; Schutze, J.; Frank, Th.

    2010-01-01

    The CFD code ANSYS CFX has been coupled with the neutron-kinetic core model DYN3D. ANSYS CFX calculates the fluid dynamics and related transport phenomena in the reactor's coolant and provides the corresponding data to DYN3D. In the fluid flow simulation of the coolant, the core itself is modeled within the porous body approach. DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the prototype that is currently available, the coupling is restricted to single-phase flow problems. In the time domain an explicit coupling of the codes has been implemented so far. Steady-state and transient verification calculations for a small-size test problem confirm the correctness of the implementation of the prototype coupling. This test problem was a mini-core consisting of nine real-size fuel assemblies. Comparison was performed with the DYN3D standalone code. In the steady state, the effective multiplication factor obtained by the ANSYS CFX/DYN3D codes shows a deviation of 9.8 pcm from the DYN3D stand-alone solution. This difference can be attributed to the use of different water property packages in the two codes. The transient test case simulated the withdrawal of the control rod from the central fuel assembly at hot zero power. Power increase during the introduction of positive reactivity and power reduction due to fuel temperature increase are calculated in the same manner by the coupled and the stand-alone codes. The maximum values reached during the power rise differ by about 1 MW at a power level of 50 MW. Beside the different water property packages, these differences are caused by the use of different flow solvers. (authors)

  19. Yield of Prompt Gamma Radiation in Slow-Neutron Induced Fission of 235U as a Function of the Total Fragment Kinetic Energy

    Energy Technology Data Exchange (ETDEWEB)

    Albinsson, H [Chalmers Univ. of Technology, Goeteborg (SE)

    1971-07-01

    Fission gamma radiation yields as functions of the total fragment kinetic energy were obtained for 235U thermal-neutron induced fission. The fragments were detected with silicon surface-barrier detectors and the gamma radiation with a Nal(Tl) scintillator. In some of the measurements mass selection was used so that the gamma radiation could also be measured as a function of fragment mass. Time discrimination between the fission gammas and the prompt neutrons released in the fission process was employed to reduce the background. The gamma radiation emitted during different time intervals after the fission event was studied with the help of a collimator, the position of which was changed along the path of the fission fragments. Fission-neutron and gamma-ray data of previous experiments were used for comparisons of the yields, and estimates were made of the variation of the prompt gamma-ray energy with the total fragment kinetic energy

  20. Sensitivity analysis of the kinetic behaviour of a Gas Cooled Fast Reactor to variations of the delayed neutron parameters

    International Nuclear Information System (INIS)

    Van Rooijen, W. F. G.; Lathouwers, D.

    2007-01-01

    In advanced Generation IV (fast) reactors an integral fuel cycle is envisaged, where all Heavy Metal is recycled in the reactor. This leads to a nuclear fuel with a considerable content of Minor Actinides. For many of these isotopes the nuclear data is not very well known. In this paper the sensitivity of the kinetic behaviour of the reactor to the dynamic parameters λ k , β k and the delayed spectrum χ d,k is studied using first order perturbation theory. In the current study, feedback due to Doppler and/or thermohydraulic effects are not treated. The theoretical framework is applied to a Generation IV Gas Cooled Fast Reactor. The results indicate that the first-order approach is satisfactory for small variations of the data. Sensitivities to delayed neutron data are similar for increasing and decreasing transients. Sensitivities generally increase with reactivity for increasing transients. For decreasing transients, there are less clearly defined trends, although the sensitivity to the delayed neutron spectrum decreases with larger sub-criticality, as expected. For this research, an adjoint capable version of the time-dependent diffusion code DALTON is under development. (authors)

  1. Kinetic neutron diffraction and SANS studies of phase formation in bioactive machinable glass ceramics

    International Nuclear Information System (INIS)

    Bentley, P M; Kilcoyne, S H; Bubb, N L; Ritter, C; Dewhurst, C D; Wood, D J

    2007-01-01

    Bioactive fluormica-fluorapatite glass-ceramic materials offer a very encouraging solution to the problem of efficient restoration and reconstruction of hard tissues. To produce material with the desired crystalline phases, a five-stage heat treatment must be performed. This thermal processing has a large impact on the microstructure and ultimately the final mechanical properties of the materials. We have examined the thermal processing of one of our most promising machinable biomaterials, using time-resolved small angle neutron scattering and neutron diffraction to study the nucleation and growth of crystallites. The processing route had already been optimized by studying the properties of quenched samples using x-ray diffraction, mechanical measurements and differential thermal analysis. However these results show that the heat treatment can be further optimized in terms of crystal nucleation, and we show that these techniques are the only methods by which a truly optimized thermal processing route may be obtained

  2. Comparison of reactor RA-4 kinetics with simulations with Matlab-Simulink for one group and six groups of delayed neutrons

    International Nuclear Information System (INIS)

    Orso, J A

    2012-01-01

    The critical state of a nuclear reactor is an unstable equilibrium. The nuclear reactor can go from critical to subcritical state or can go from critical to hypercritical state. Although the evolution of the system in these cases is slow, it requires the intervention of an operator to correct deviations. For this reason an automatic control technique was designed, based on the kinetic point to a group of delayed neutrons, which corrects deviations automatically. In this paper we study the point kinetics models in a group and six groups of delayed neutrons for different values of reactivity using the simulations software MATLAB, Simulink. A comparison of two models with the reactor kinetic behavior is made (author)

  3. A practical implementation of the higher-order transverse-integrated nodal diffusion method

    International Nuclear Information System (INIS)

    Prinsloo, Rian H.; Tomašević, Djordje I.; Moraal, Harm

    2014-01-01

    Highlights: • A practical higher-order nodal method is developed for diffusion calculations. • The method resolves the issue of the transverse leakage approximation. • The method achieves much superior accuracy as compared to standard nodal methods. • The calculational cost is only about 50% greater than standard nodal methods. • The method is packaged in a module for connection to existing nodal codes. - Abstract: Transverse-integrated nodal diffusion methods currently represent the standard in full core neutronic simulation. The primary shortcoming of this approach is the utilization of the quadratic transverse leakage approximation. This approach, although proven to work well for typical LWR problems, is not consistent with the formulation of nodal methods and can cause accuracy and convergence problems. In this work, an improved, consistent quadratic leakage approximation is formulated, which derives from the class of higher-order nodal methods developed some years ago. Further, a number of iteration schemes are developed around this consistent quadratic leakage approximation which yields accurate node average results in much improved calculational times. The most promising of these iteration schemes results from utilizing the consistent leakage approximation as a correction method to the standard quadratic leakage approximation. Numerical results are demonstrated on a set of benchmark problems and further applied to a realistic reactor problem, particularly the SAFARI-1 reactor, operating at Necsa, South Africa. The final optimal solution strategy is packaged into a standalone module which may simply be coupled to existing nodal diffusion codes

  4. Coupled Model of channels in parallel and neutron kinetics in two dimensions

    International Nuclear Information System (INIS)

    Cecenas F, M.; Campos G, R.M.; Valle G, E. del

    2004-01-01

    In this work an arrangement of thermohydraulic channels is presented that represent those four quadrants of a nucleus of reactor type BWR. The channels are coupled to a model of neutronic in two dimensions that allow to generate the radial profile of power of the reactor. Nevertheless that the neutronic pattern is of two dimensions, it is supplemented with axial additional information when considering the axial profiles of power for each thermo hydraulic channel. The stationary state is obtained the one it imposes as frontier condition the same pressure drop for all the channels. This condition is satisfied to iterating on the flow of coolant in each channel to equal the pressure drop in all the channels. This stationary state is perturbed later on when modifying the values for the effective sections corresponding to an it assembles. The calculation in parallel of the neutronic and the thermo hydraulic is carried out with Vpm (Virtual parallel machine) by means of an outline teacher-slave in a local net of computers. (Author)

  5. Nodal metastasis in thyroid cancer

    International Nuclear Information System (INIS)

    Samuel, A.M.

    1999-01-01

    The biological behavior and hence the prognosis of thyroid cancer (TC) depends among other factors on the extent of spread of the disease outside the thyroid bed. This effect is controversial, especially for nodal metastasis of well differentiated thyroid carcinoma (WDC). Nodal metastasis at the time of initial diagnosis behaves differently depending on the histology, age of the patient, presence of extrathyroidal extension, and the sex of the individual. The type of the surgery, administration of 131 I and thyroxin suppression also to some extent influence the rate of recurrence and mortality. Experience has shown that it is not as innocuous as a small intrathyroidal tumor without any invasion outside the thyroid bed and due consideration should be accorded to the management strategies for handling patients with nodal metastasis

  6. The adjoint variational nodal method

    International Nuclear Information System (INIS)

    Laurin-Kovitz, K.; Lewis, E.E.

    1993-01-01

    The widespread use of nodal methods for reactor core calculations in both diffusion and transport approximations has created a demand for the corresponding adjoint solutions as a prerequisite for performing perturbation calculations. With some computational methods, however, the solution of the adjoint problem presents a difficulty; the physical adjoint obtained by discretizing the adjoint equation is not the same as the mathematical adjoint obtained by taking the transpose of the coefficient matrix, which results from the discretization of the forward equation. This difficulty arises, in particular, when interface current nodal methods based on quasi-one-dimensional solution of the diffusion or transport equation are employed. The mathematical adjoint is needed to perform perturbation calculations. The utilization of existing nodal computational algorithms, however, requires the physical adjoint. As a result, similarity transforms or related techniques must be utilized to relate physical and mathematical adjoints. Thus far, such techniques have been developed only for diffusion theory

  7. Nodal in computerized control systems of accelerators

    International Nuclear Information System (INIS)

    Kagarmanov, A.A.; Koval'tsov, V.I.; Korobov, S.A.

    1994-01-01

    Brief description of the Nodal language programming structure is presented. Its possibilities as high-level programming language for accelerator control systems are considered. The status of the Nodal language in the HEPI is discussed. 3 refs

  8. Determination of reactivity of multiplying systems filled with spherical HTGR-fuel elements using kinetic methods with regard to the pulsed-neutron method

    International Nuclear Information System (INIS)

    Drueke, V.

    1978-06-01

    At three critical or subcritical facilities - two of them filled with spherical HTGR-fuel elements - the reactivity is determined using kinetic methods. Besides the inverskinetic method the applicability of the pulsed-neutron method is investigated. The experimental results using the pulsed-neutron method are compared partly with the inverskinetic method and partly with diffusion-calculations. It is shown, that in the HTGR the space dependence of the reactivity in radial direction is not remarkable in spite of the 'kinetic distortion'; on the contrary in axial direction - the direction of the external neutron source - space dependent reactivity worths are measured. The results of the pulsed-neutron methods of Sjoestrand and Simmons-King are rather good applicable in all configurations. For the method of Sjoestrand it is necessary to select the detector positions, whereas for Simmons-King the calculated life-time determines the results. Therefore it is proposed to compare calculated and measured decay constants of the prompt neutron field in future. (orig.) [de

  9. Fission fragment yields and total kinetic energy release in neutron-induced fission of235,238U,and239Pu

    Science.gov (United States)

    Tovesson, F.; Duke, D.; Geppert-Kleinrath, V.; Manning, B.; Mayorov, D.; Mosby, S.; Schmitt, K.

    2018-03-01

    Different aspects of the nuclear fission process have been studied at Los Alamos Neutron Science Center (LANSCE) using various instruments and experimental techniques. Properties of the fragments emitted in fission have been investigated using Frisch-grid ionization chambers, a Time Projection Chamber (TPC), and the SPIDER instrument which employs the 2v-2E method. These instruments and experimental techniques have been used to determine fission product mass yields, the energy dependent total kinetic energy (TKE) release, and anisotropy in neutron-induced fission of U-235, U-238 and Pu-239.

  10. Uncertainty and sensitivity analysis for the modeling of transients with interaction of thermal hydraulics and neutron kinetics

    International Nuclear Information System (INIS)

    Soeren Kliem; Siegfried Mittag; Siegfried Langenbuch

    2005-01-01

    Full text of publication follows: The transition from the application of conservative models to the use of best-estimate models raises the question about the uncertainty of the obtained results. This question becomes especially important, if the best-estimate models should be used for safety analyses in the field of nuclear engineering. Different methodologies were developed to assess the uncertainty of the calculation results of computer simulation codes. One of them is the methodology developed by Gesellschaft fuer Anlagenund Reaktorsicherheit (GRS) which uses the statistical code package SUSA. In the past, this methodology was applied to the calculation results of the advanced thermal hydraulic system code ATHLET. In the frame of the recently finished EU FP5 funded research project VALCO, that methodology was extended and successfully applied to different coupled code systems, including the uncertainty analysis for neutronics. These code systems consist of a thermal hydraulic system code and a 3D neutron kinetic core model. One of the code systems applied was ATHLET coupled with the Rossendorf kinetics code DYN3D. Two real transients at NPPs with VVER-type reactors documented within the VALCO project were selected for analyses. One was the load drop of one of two turbines to house load level at the Loviisa-1 NPP (VVER-440), the second was a test with the switching-off of one of two main feed water pumps at the VVER-1000 Balakovo-4 NPP. The current paper is dedicated to the different steps of the use and implementation of the GRS methodology to coupled code systems and to the assessment of the results obtained by the DYN3D/ATHLET code. Based on the relevant physical processes in both transients, lists of possible sources of uncertainties were compiled. They are specific for the two transients. Besides control parameters like control rod movement and thermal hydraulic parameters like secondary side pressure, mass flow rates, pressurizer sprayer and heater

  11. Multi-scale method for the resolution of the neutronic kinetics equations

    International Nuclear Information System (INIS)

    Chauvet, St.

    2008-10-01

    In this PhD thesis and in order to improve the time/precision ratio of the numerical simulation calculations, we investigate multi-scale techniques for the resolution of the reactor kinetics equations. We choose to focus on the mixed dual diffusion approximation and the quasi-static methods. We introduce a space dependency for the amplitude function which only depends on the time variable in the standard quasi-static context. With this new factorization, we develop two mixed dual problems which can be solved with Cea's solver MINOS. An algorithm is implemented, performing the resolution of these problems defined on different scales (for time and space). We name this approach: the Local Quasi-Static method. We present here this new multi-scale approach and its implementation. The inherent details of amplitude and shape treatments are discussed and justified. Results and performances, compared to MINOS, are studied. They illustrate the improvement on the time/precision ratio for kinetics calculations. Furthermore, we open some new possibilities to parallelize computations with MINOS. For the future, we also introduce some improvement tracks with adaptive scales. (author)

  12. Kinetic study of human hand sodium using local in vivo neutron activation analysis

    International Nuclear Information System (INIS)

    Cohen Boulakia, Francine.

    1978-01-01

    Using local 'in vivo' activation analysis, turnover of human hand sodium is studied in 14 subjects, 7 controls and 7 decalcified osteoporotics patients. The hand of each subject is irradiated with neutrons emitted by 52 Cf sources; the equivalent dose delivered is 8 cGy. The 24 Na activity variation is plotted as function of time and the experimental curve so obtained is fitted to two exponentials. Two compartements are identified: a rapidly exchangeable one, with a half life of 1 h; an other, with a very slow turnover, the half lifes varying from 79 h to 35 h as the calcium concentration becomes sub-normal. The ratios calcium to slowly exchangeable sodium and rapidly to slowly exchangeable sodium appear to be promising for the evaluation of bone disease [fr

  13. Coupling of the computational fluid dynamics code ANSYS CFX with the 3D neutron kinetic core model DYN3D

    International Nuclear Information System (INIS)

    Kliem, S.; Grahn, A.; Rohde, U.; Schuetze, J.; Frank, Th.

    2010-01-01

    The computational fluid dynamics code ANSYS CFX has been coupled with the neutron-kinetic core model DYN3D. ANSYS CFX calculates the fluid dynamics and related transport phenomena in the reactors coolant and provides the corresponding data to DYN3D. In the fluid flow simulation of the coolant, the core itself is modeled within the porous body approach. DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the prototype that is currently available, the coupling is restricted to single-phase flow problems. In the time domain an explicit coupling of the codes has been implemented so far. Steady-state and transient verification calculations for two small-size test problems confirm the correctness of the implementation of the prototype coupling. The first test problem was a mini-core consisting of nine real-size fuel assemblies with quadratic cross section. Comparison was performed with the DYN3D stand-alone code. In the steady state, the effective multiplication factor obtained by the DYN3D/ANSYS CFX codes hows a deviation of 9.8 pcm from the DYN3D stand-alone solution. This difference can be attributed to the use of different water property packages in the two codes. The transient test case simulated the withdrawal of the control rod from the central fuel assembly at hot zero power in the same mini-core. Power increase during the introduction of positive reactivity and power reduction due to fuel temperature increase are calculated in the same manner by the coupled and the stand-alone codes. The maximum values reached during the power rise differ by about 1 MW at a power level of 50 MW. Beside the different water property packages, these differences are caused by the use of different flow solvers. The same calculations were carried for a mini-core with seven real-size fuel assemblies with hexagonal cross section in

  14. Coupling tests of parallel channels to modal neutronic kinetics; Pruebas de acoplamiento de canales paralelos a cinetica neutronica modal

    Energy Technology Data Exchange (ETDEWEB)

    Cecenas F, M.; Campos G, R.M. [IIE, 62490 Cuernavaca, Morelos (Mexico)]. e-mail: mcf@iie.org.mx

    2007-07-01

    In this work the initial results of the joining of an arrangement of 36 channels in parallel are studied to a modal neutronic kinetic model to represent the core of a BWR type reactor. The set of channels is obtained group the assemblies of that it consists the core in an arrangement of concentric rings, for later on to subdivide each ring in four parts to assign each segment to a quadrant of the core. The obtained channels are coupled to a modal kinetics model that considers the fundamental way and the first harmonic. The obtained solution represents the radial distribution and power azimuthal, the one which is feedback to the channels to update the thermohydraulic variables. The restriction that the pressure drop is same for each channel it is only imposed as initial condition, like part of the stationary state, and it is allowed that the pressure drop in the assemblies them it is different in each channel during a reactivity interference. For the tests to the system, it is convenient to select a relatively big core and that it operates near their stability frontier, for that the channels are dimensioning according to the case 9 of the Stability Benchmark of the Ringhals Swedish plant, organized by the Nuclear Energy Agency in 1994. In general, they reproduce the results of this benchmark, when reproducing oscillations outside of phase, with the additional results that the quadrants 1 and 2 of the core present oscillations of more width that the quadrants 3 and 4. The set group nuclear-thermohydraulics it is solved numerically by means of the outline master-slave of distributed calculation implemented by means of Parallel Virtual Machine (PVM). (Author)

  15. The implementation of a simplified spherical harmonics semi-analytic nodal method in PANTHER

    International Nuclear Information System (INIS)

    Hall, S.K.; Eaton, M.D.; Knight, M.P.

    2013-01-01

    Highlights: ► An SP N nodal method is proposed. ► Consistent CMFD derived and tested. ► Mark vacuum boundary conditions applied. ► Benchmarked against other diffusions and transport codes. - Abstract: In this paper an SP N nodal method is proposed which can utilise existing multi-group neutron diffusion solvers to obtain the solution. The semi-analytic nodal method is used in conjunction with a coarse mesh finite difference (CMFD) scheme to solve the resulting set of equations. This is compared against various nuclear benchmarks to show that the method is capable of computing an accurate solution for practical cases. A few different CMFD formulations are implemented and their performance compared. It is found that the effective diffusion coefficent (EDC) can provide additional stability and require less power iterations on a coarse mesh. A re-arrangement of the EDC is proposed that allows the iteration matrix to be computed at the beginning of a calculation. Successive nodal updates only modify the source term unlike existing CMFD methods which update the iteration matrix. A set of Mark vacuum boundary conditions are also derived which can be applied to the SP N nodal method extending its validity. This is possible due to a similarity transformation of the angular coupling matrix, which is used when applying the nodal method. It is found that the Marshak vacuum condition can also be derived, but would require the significant modification of existing neutron diffusion codes to implement it

  16. An alternative solver for the nodal expansion method equations - 106

    International Nuclear Information System (INIS)

    Carvalho da Silva, F.; Carlos Marques Alvim, A.; Senra Martinez, A.

    2010-01-01

    An automated procedure for nuclear reactor core design is accomplished by using a quick and accurate 3D nodal code, aiming at solving the diffusion equation, which describes the spatial neutron distribution in the reactor. This paper deals with an alternative solver for nodal expansion method (NEM), with only two inner iterations (mesh sweeps) per outer iteration, thus having the potential to reduce the time required to calculate the power distribution in nuclear reactors, but with accuracy similar to the ones found in conventional NEM. The proposed solver was implemented into a computational system which, besides solving the diffusion equation, also solves the burnup equations governing the gradual changes in material compositions of the core due to fuel depletion. Results confirm the effectiveness of the method for practical purposes. (authors)

  17. Kinetics of the stress induced phase transition in quartz by real-time neutron scattering

    International Nuclear Information System (INIS)

    Gibhardt, H.; Eckold, G.; Guethoff, F.

    1999-01-01

    Complete text of publication follows. The stability regime of the incommensurate phase of quartz is influenced by uniaxial stress. Hence, the phase transition can be induced under isothermal conditions by the application of external mechanical forces. Using real-time neutron scattering the time evolution of structural changes is investigated id detail during stress variations. The time dependent behaviour of the satellite reflection is compared with that one of the fundamental Bragg reflection which - via primary extinction - gives information about the perfection of the crystal. On increasing stress the perfection of the lattice is destroyed immediately while the modulated structure is built up with a delay of about 1 s. Decreasing the stress leads to a reverse behaviour. Moreover, there is evidence that under periodical load residual non-relaxed strain fields survive leading to a different temperature dependence as compared to static conditions. This finding is compatible with pronounced hysteresis effects observed under cycling stress. It is argued that these residual strains are associated with non-relaxed topological 4-line defects, that drive the structural changes in quartz (1). (author)

  18. RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - user's manual

    International Nuclear Information System (INIS)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

    1998-03-01

    This document is the User's Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code's capabilities and limitations; Chapter 2 describes the code's structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARC and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs

  19. Analytical solution of the multigroup neutron diffusion kinetic equation in one-dimensional cartesian geometry by the integral transform technique

    International Nuclear Information System (INIS)

    Ceolin, Celina

    2010-01-01

    The objective of this work is to obtain an analytical solution of the neutron diffusion kinetic equation in one-dimensional cartesian geometry, to monoenergetic and multigroup problems. These equations are of the type stiff, due to large differences in the orders of magnitude of the time scales of the physical phenomena involved, which make them difficult to solve. The basic idea of the proposed method is applying the spectral expansion in the scalar flux and in the precursor concentration, taking moments and solving the resulting matrix problem by the Laplace transform technique. Bearing in mind that the equation for the precursor concentration is a first order linear differential equation in the time variable, to enable the application of the spectral method we introduce a fictitious diffusion term multiplied by a positive value which tends to zero. This procedure opened the possibility to find an analytical solution to the problem studied. We report numerical simulations and analysis of the results obtained with the precision controlled by the truncation order of the series. (author)

  20. Coupling of the 3D neutron kinetic core model DYN3D with the CFD software ANSYS-CFX

    International Nuclear Information System (INIS)

    Grahn, Alexander; Kliem, Sören; Rohde, Ulrich

    2015-01-01

    Highlights: • Improved thermal hydraulic description of nuclear reactor cores. • Possibility of three-dimensional flow phenomena in the core, such as cross flow, flow reversal, flow around obstacles. • Simulation at higher spatial resolution as compared to system codes. - Abstract: This article presents the implementation of a coupling between the 3D neutron kinetic core model DYN3D and the commercial, general purpose computational fluid dynamics (CFD) software ANSYS-CFX. In the coupling approach, parts of the thermal hydraulic calculation are transferred to CFX for its better ability to simulate the three-dimensional coolant redistribution in the reactor core region. The calculation of the heat transfer from the fuel into the coolant remains with DYN3D, which incorporates well tested and validated heat transfer models for rod-type fuel elements. On the CFX side, the core region is modeled based on the porous body approach. The implementation of the code coupling is verified by comparing test case results with reference solutions of the DYN3D standalone version. Test cases cover mini and full core geometries, control rod movement and partial overcooling transients

  1. Status on development and verification of reactivity initiated accident analysis code for PWR (NODAL3)

    International Nuclear Information System (INIS)

    Peng Hong Liem; Surian Pinem; Tagor Malem Sembiring; Tran Hoai Nam

    2015-01-01

    A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the nodal few-group neutron diffusion theory in 3-dimensional Cartesian geometry for a typical pressurized water reactor (PWR) static and transient analyses, especially for reactivity initiated accidents (RIA). The spatial variables are treated by using a polynomial nodal method (PNM) while for the neutron dynamic solver the adiabatic and improved quasi-static methods are adopted. A simple single channel thermal-hydraulics module and its steam table is implemented into the code. Verification works on static and transient benchmarks are being conducting to assess the accuracy of the code. For the static benchmark verification, the IAEA-2D, IAEA-3D, BIBLIS and KOEBERG light water reactor (LWR) benchmark problems were selected, while for the transient benchmark verification, the OECD NEACRP 3-D LWR Core Transient Benchmark and NEA-NSC 3-D/1-D PWR Core Transient Benchmark (Uncontrolled Withdrawal of Control Rods at Zero Power). Excellent agreement of the NODAL3 results with the reference solutions and other validated nodal codes was confirmed. (author)

  2. The merits of cell kinetic parameters for the assessment of intrinsic cellular radiosensitivity to photon and high linear energy transfer neutron irradiation

    International Nuclear Information System (INIS)

    Theron, Therina; Slabbert, Jacobus; Serafin, Antonio; Boehm, Lothar

    1997-01-01

    Purpose: Differences in tumor response and intrinsic cellular radiosensitivity make the selection of patients for specific radiation modalities very difficult. The reasons for these differences are still unclear, but are thought to be due to genomic and cellular characteristics. Because radiosensitivities vary between cell cycle stages and because S phase cells are very radioresistant, cell cycle kinetic parameters could be a candidate for predicting intrinsic radiosensitivity. Methods and Materials: A panel of 15 tumor cell lines was analyzed for S phase content and potential doubling times (T pot ), and the influence of these parameters on the intrinsic radiosensitivity to 60 Coγ- and p(66)/Be neutron irradiation was assessed. Results: S phase content and T pot show a statistically significant correlation with the mean inactivation dose for photons. The correlation between cell kinetic parameters and the mean inactivation dose for neutrons showed the same trend as photon sensitivity but this was not found to be statistically significant. Conclusions: S phase content and T pot were identified as suitable criteria for predicting photon sensitivity. It is suggested that cell kinetic parameters could play a role in identifying neutron sensitive tumors if both tumor and normal cells are analyzed

  3. Nodal lymphomas of the abdomen

    International Nuclear Information System (INIS)

    Bruneton, J.N.; Caramella, E.; Manzino, J.J.

    1986-01-01

    Modern imaging modalities have greatly contributed to current knowledge about intra-abdominal nodal lymphomas. Since both intra and retroperitoneal node involvement can be demonstrated by computed tomography (CT) and ultrasonography, it seems legitimate to treat these two sites together in the same chapter, particularly since the older separation between intraperitoneal and retroperitoneal nodal disease was based to a large degree on the limitations of lymphography. Hodgkin's disease (HD) has benefited less from recent technological advances. The diversity in the incidence of nodal involvement between HD and NHL, the diagnostic capabilities of modern imaging techniques, and the histopathological features of lymphomatous non-Hodgkin and Hodgkin nodes, justify adoption of an investigatory approach which takes all of these factors into account. Details of this investigative strategy are discussed in this paper following a review of available imaging modalities. In current practice, the four main methods for the exploration of abdominal lymph nodes are lymphography, ultrasonography, CT, and radionuclide studies. The first three techniques are also utilized to guide biopsies for staging purposes and for the evaluation of response to treatment

  4. Time-dependent flux from pulsed neutrons revealed by superconducting Nb current-biased kinetic inductance detector with "1"0B converter operated at 4 K

    International Nuclear Information System (INIS)

    Miyajima, Shigeyuki; Narukami, Yoshito; Shishido, Hiroaki; Yoshioka, Naohito; Ishida, Takekazu; Fujimaki, Akira; Hidaka, Mutsuo; Oikawa, Kenichi; Harada, Masahide; Oku, Takayuki; Arai, Masatoshi

    2015-01-01

    We have demonstrated a new superconducting detector for a neutron based on Nb superconductor meanderline with a "1"0B conversion layer. We use a current-biased kinetic inductance detector (CB-KID), which is composed of a meanderline, for detection of a neutron with high spatial resolution and fast response. The thickness of Nb meanderlines is 40 nm and widths are 3 μm, 1 μm, and 0.6 μm. The CB-KIDs are fabricated at the center of the Si chip of the size 22 mm × 22 mm and the total area of CB-KIDs covers 8 mm × 8 mm. The chip was cooled to a temperature lower than 4 K below the transition temperature of Nb using a Gifford-McMahon (GM) cryocooler. The Nb CB-KIDs with a "1"0B conversion layer output the voltage by irradiating pulsed neutrons at the material life science experimental facility (MLF) of Japan Proton Accelerator Research Complex (J-PARC) center. The response time of CB-KIDs is about a few tens ns. We have also obtained the time dependence of neutron flux generated from pulsed neutrons using a CB-KID. Experimental results were in good agreement with the simulated results. (author)

  5. Study of the multiplication and kinetic effects of salt mixtures and salt blanket micromodels on thermal neutron spectra of heavy water MAKET facility

    International Nuclear Information System (INIS)

    Titarenko, Yu.E.; Batyaev, V.F.; Borovlev, S.P.; Gladkikh, N.G.; Igumnov, M.M.; Legostaev, V.O.; Karpikhin, E.I.; Konev, V.N.; Kushnerev, Yu.T.; Ryazhsky, V.I.; Spiridonov, V.G.; Chernyavsky, E.V.; Shvedov, O.V.

    2009-10-01

    The main goal of the Project is to study and evaluate nuclear characteristics of materials and isotopes involved in processes of irradiated nuclear fuel transmutation. This principal task is subdivided into 9 subtasks subject to the neutron or proton source used, the type of the nuclear process under study, isotope collection, characteristics of which are to be investigated, etc. In the presented extract of the Project Activity report the measurements there were used the MAKET zero-power heavy-water reactor in the measurements there was employed a large set of minor actinide samples highly enriched with the main isotope. The samples were obtained with mass-separator SM-2 (VNIIEF). At the heavy-water reactor MAKET (ITEP) there were measured multiplying and kinetic characteristics of salt mixtures basing on the spectra of fast and thermal neutrons. The salt mixtures of zirconium and sodium fluorides were available in salt blanket models (SBM) of cylindrical shape. There were measured the neutron spectra formed by this micro-model as well as the effective fission cross-sections of neptunium, plutonium, americium and curium isotopes caused by SBM neutrons. The neutron spectra in the measurement positions were determined from activation reaction rates. (author)

  6. A nodal method applied to a diffusion problem with generalized coefficients

    International Nuclear Information System (INIS)

    Laazizi, A.; Guessous, N.

    1999-01-01

    In this paper, we consider second order neutrons diffusion problem with coefficients in L ∞ (Ω). Nodal method of the lowest order is applied to approximate the problem's solution. The approximation uses special basis functions in which the coefficients appear. The rate of convergence obtained is O(h 2 ) in L 2 (Ω), with a free rectangular triangulation. (authors)

  7. Nodal collocation approximation for the multidimensional PL equations applied to transport source problems

    International Nuclear Information System (INIS)

    Verdu, G.; Capilla, M.; Talavera, C. F.; Ginestar, D.

    2012-01-01

    PL equations are classical high order approximations to the transport equations which are based on the expansion of the angular dependence of the angular neutron flux and the nuclear cross sections in terms of spherical harmonics. A nodal collocation method is used to discretize the PL equations associated with a neutron source transport problem. The performance of the method is tested solving two 1D problems with analytical solution for the transport equation and a classical 2D problem. (authors)

  8. Nodal collocation approximation for the multidimensional PL equations applied to transport source problems

    Energy Technology Data Exchange (ETDEWEB)

    Verdu, G. [Departamento de Ingenieria Quimica Y Nuclear, Universitat Politecnica de Valencia, Cami de Vera, 14, 46022. Valencia (Spain); Capilla, M.; Talavera, C. F.; Ginestar, D. [Dept. of Nuclear Engineering, Departamento de Matematica Aplicada, Universitat Politecnica de Valencia, Cami de Vera, 14, 46022. Valencia (Spain)

    2012-07-01

    PL equations are classical high order approximations to the transport equations which are based on the expansion of the angular dependence of the angular neutron flux and the nuclear cross sections in terms of spherical harmonics. A nodal collocation method is used to discretize the PL equations associated with a neutron source transport problem. The performance of the method is tested solving two 1D problems with analytical solution for the transport equation and a classical 2D problem. (authors)

  9. Predictive uncertainty reduction in coupled neutron-kinetics/thermal hydraulics modeling of the BWR-TT2 benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Badea, Aurelian F., E-mail: aurelian.badea@kit.edu [Karlsruhe Institute of Technology, Vincenz-Prießnitz-Str. 3, 76131 Karlsruhe (Germany); Cacuci, Dan G. [Center for Nuclear Science and Energy/Dept. of ME, University of South Carolina, 300 Main Street, Columbia, SC 29208 (United States)

    2017-03-15

    Highlights: • BWR Turbine Trip 2 (BWR-TT2) benchmark. • Substantial (up to 50%) reduction of uncertainties in the predicted transient power. • 6660 uncertain model parameters were calibrated. - Abstract: By applying a comprehensive predictive modeling methodology, this work demonstrates a substantial (up to 50%) reduction of uncertainties in the predicted total transient power in the BWR Turbine Trip 2 (BWR-TT2) benchmark while calibrating the numerical simulation of this benchmark, comprising 6090 macroscopic cross sections, and 570 thermal-hydraulics parameters involved in modeling the phase-slip correlation, transient outlet pressure, and total mass flow. The BWR-TT2 benchmark is based on an experiment that was carried out in 1977 in the NPP Peach Bottom 2, involving the closure of the turbine stop valve which caused a pressure wave that propagated with attenuation into the reactor core. The condensation of the steam in the reactor core caused by the pressure increase led to a positive reactivity insertion. The subsequent rise of power was limited by the feedback and the insertion of the control rods. The BWR-TT2 benchmark was modeled with the three-dimensional reactor physics code system DYN3D, by coupling neutron kinetics with two-phase thermal-hydraulics. All 6660 DYN3D model parameters were calibrated by applying a predictive modeling methodology that combines experimental and computational information to produce optimally predicted best-estimate results with reduced predicted uncertainties. Simultaneously, the predictive modeling methodology yields optimally predicted values for the BWR total transient power while reducing significantly the accompanying predicted standard deviations.

  10. Probing the hydrogen equilibrium and kinetics in zeolite imidazolate frameworks via molecular dynamics and quasi-elastic neutron scattering experiments.

    Science.gov (United States)

    Pantatosaki, Evangelia; Jobic, Hervé; Kolokolov, Daniil I; Karmakar, Shilpi; Biniwale, Rajesh; Papadopoulos, George K

    2013-01-21

    The problem of simulating processes involving equilibria and dynamics of guest sorbates within zeolitic imidazolate frameworks (ZIF) by means of molecular dynamics (MD) computer experiments is of growing importance because of the promising role of ZIFs as molecular "traps" for clean energy applications. A key issue for validating such an atomistic modeling attempt is the possibility of comparing the MD results, with real experiments being able to capture analogous space and time scales to the ones pertained to the computer experiments. In the present study, this prerequisite is fulfilled through the quasi-elastic neutron scattering technique (QENS) for measuring self-diffusivity, by elaborating the incoherent scattering signal of hydrogen nuclei. QENS and MD experiments were performed in parallel to probe the hydrogen motion, for the first time in ZIF members. The predicted and measured dynamics behaviors show considerable concentration variation of the hydrogen self-diffusion coefficient in the two topologically different ZIF pore networks of this study, the ZIF-3 and ZIF-8. Modeling options such as the flexibility of the entire matrix versus a rigid framework version, the mobility of the imidazolate ligand, and the inclusion of quantum mechanical effects in the potential functions were examined in detail for the sorption thermodynamics and kinetics of hydrogen and also of deuterium, by employing MD combined with Widom averaging towards studying phase equilibria. The latter methodology ensures a rigorous and efficient way for post-processing the dynamics trajectory, thereby avoiding stochastic moves via Monte Carlo simulation, over the large number of configurational degrees of freedom a nonrigid framework encompasses.

  11. Predictive uncertainty reduction in coupled neutron-kinetics/thermal hydraulics modeling of the BWR-TT2 benchmark

    International Nuclear Information System (INIS)

    Badea, Aurelian F.; Cacuci, Dan G.

    2017-01-01

    Highlights: • BWR Turbine Trip 2 (BWR-TT2) benchmark. • Substantial (up to 50%) reduction of uncertainties in the predicted transient power. • 6660 uncertain model parameters were calibrated. - Abstract: By applying a comprehensive predictive modeling methodology, this work demonstrates a substantial (up to 50%) reduction of uncertainties in the predicted total transient power in the BWR Turbine Trip 2 (BWR-TT2) benchmark while calibrating the numerical simulation of this benchmark, comprising 6090 macroscopic cross sections, and 570 thermal-hydraulics parameters involved in modeling the phase-slip correlation, transient outlet pressure, and total mass flow. The BWR-TT2 benchmark is based on an experiment that was carried out in 1977 in the NPP Peach Bottom 2, involving the closure of the turbine stop valve which caused a pressure wave that propagated with attenuation into the reactor core. The condensation of the steam in the reactor core caused by the pressure increase led to a positive reactivity insertion. The subsequent rise of power was limited by the feedback and the insertion of the control rods. The BWR-TT2 benchmark was modeled with the three-dimensional reactor physics code system DYN3D, by coupling neutron kinetics with two-phase thermal-hydraulics. All 6660 DYN3D model parameters were calibrated by applying a predictive modeling methodology that combines experimental and computational information to produce optimally predicted best-estimate results with reduced predicted uncertainties. Simultaneously, the predictive modeling methodology yields optimally predicted values for the BWR total transient power while reducing significantly the accompanying predicted standard deviations.

  12. Validation of full core geometry model of the NODAL3 code in the PWR transient Benchmark problems

    International Nuclear Information System (INIS)

    T-M Sembiring; S-Pinem; P-H Liem

    2015-01-01

    The coupled neutronic and thermal-hydraulic (T/H) code, NODAL3 code, has been validated in some PWR static benchmark and the NEACRP PWR transient benchmark cases. However, the NODAL3 code have not yet validated in the transient benchmark cases of a control rod assembly (CR) ejection at peripheral core using a full core geometry model, the C1 and C2 cases. By this research work, the accuracy of the NODAL3 code for one CR ejection or the unsymmetrical group of CRs ejection case can be validated. The calculations by the NODAL3 code have been carried out by the adiabatic method (AM) and the improved quasistatic method (IQS). All calculated transient parameters by the NODAL3 code were compared with the reference results by the PANTHER code. The maximum relative difference of 16 % occurs in the calculated time of power maximum parameter by using the IQS method, while the relative difference of the AM method is 4 % for C2 case. All calculation results by the NODAL3 code shows there is no systematic difference, it means the neutronic and T/H modules are adopted in the code are considered correct. Therefore, all calculation results by using the NODAL3 code are very good agreement with the reference results. (author)

  13. Development of nodal interface conditions for a PN approximation nodal model

    International Nuclear Information System (INIS)

    Feiz, M.

    1993-01-01

    A relation was developed for approximating higher order odd-moments from lower order odd-moments at the nodal interfaces of a Legendre polynomial nodal model. Two sample problems were tested using different order P N expansions in adjacent nodes. The developed relation proved to be adequate and matched the nodal interface flux accurately. The development allows the use of different order expansions in adjacent nodes, and will be used in a hybrid diffusion-transport nodal model. (author)

  14. Analytical representation for solution of the neutron point kinetics equation with time-dependent reactivity and free of the stiffness character

    International Nuclear Information System (INIS)

    Silva, Milena Wollmann da

    2013-01-01

    In this work, we report a genuine analytical representation for the solution of the neutron point kinetics equation free of the stiffness character, assuming that the reactivity is a continuous and sectionally continuous function of time. To this end, we initially cast the point kinetics equation in a first order linear differential equation. Next, we split the corresponding matrix as a sum of a diagonal matrix with a matrix, whose components contain the off-diagonal elements. Next, expanding the neutron density and the delayed neutron precursors concentrations in a truncated series, and replacing these expansions in the matrix equation, we come out with an equation, which allows to construct a recursive system, a first order matrix differential equation with source. The fundamental characteristic of this system relies on the fact that the corresponding matrix is diagonal, meanwhile the source term is written in terms of the matrix with the off-diagonal components. Further, the first equation of the recursive system has no source and satisfies the initial conditions. On the other hand, the remaining equations satisfy the null initial condition. Due to the diagonal feature of the matrix, we attain analytical solutions for these recursive equations. We also mention that we evaluate the results for any time value, without the analytical continuity because the purposed solution is free on the stiffness character. Finally, we present numerical simulations and comparisons against literature results, considering specific the applications for the following reactivity functions: constant, step, ramp, and sine. (author)

  15. Influence of reactor vessel nodalization in the coupled code analysis of Asymmetric Main Feedwater Isolation

    International Nuclear Information System (INIS)

    Bencik, V.; Feretic, D.; Grgic, D.

    2001-01-01

    Asymmetric Main Feedwater Isolation (AMFWI) transient in one Steam Generator (SG) for NPP Krsko using RELAP5 standalone code and coupled code RELAP5- QUABOX/CUBBOX (R5QC) was analyzed. In the RELAP5 standalone calculation, a point kinetics model was used, while in the coupled code a three-dimensional (3D) neutronics model of QUABOX with different RELAP5 nodalization schemes of reactor vessel was used. Both code versions use best-estimate thermal-hydraulic system code for all components in the plant and include realistic description of plant protection and control systems. Two different types of calculations were performed: with and without automatic control rod system available. The AMFWI transient causes the great asymmetry of the transferred heat in the SGs and subsequently the asymmetry of the power produced across the core due to different reactivity feedback resulting from the thermal-hydraulic channels assigned to different loops. The work presented in the paper is a part of validation of the 3D coupled code R5QC in the analysis of asymmetric transients.(author)

  16. A quasi-static polynomial nodal method for nuclear reactor analysis

    International Nuclear Information System (INIS)

    Gehin, J.C.

    1992-09-01

    Modern nodal methods are currently available which can accurately and efficiently solve the static and transient neutron diffusion equations. Most of the methods, however, are limited to two energy groups for practical application. The objective of this research is the development of a static and transient, multidimensional nodal method which allows more than two energy groups and uses a non-linear iterative method for efficient solution of the nodal equations. For both the static and transient methods, finite-difference equations which are corrected by the use of discontinuity factors are derived. The discontinuity factors are computed from a polynomial nodal method using a non-linear iteration technique. The polynomial nodal method is based upon a quartic approximation and utilizes a quadratic transverse-leakage approximation. The solution of the time-dependent equations is performed by the use of a quasi-static method in which the node-averaged fluxes are factored into shape and amplitude functions. The application of the quasi-static polynomial method to several benchmark problems demonstrates that the accuracy is consistent with that of other nodal methods. The use of the quasi-static method is shown to substantially reduce the computation time over the traditional fully-implicit time-integration method. Problems involving thermal-hydraulic feedback are accurately, and efficiently, solved by performing several reactivity/thermal-hydraulic updates per shape calculation

  17. A quasi-static polynomial nodal method for nuclear reactor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Gehin, Jess C. [Massachusetts Inst. of Tech., Cambridge, MA (United States)

    1992-09-01

    Modern nodal methods are currently available which can accurately and efficiently solve the static and transient neutron diffusion equations. Most of the methods, however, are limited to two energy groups for practical application. The objective of this research is the development of a static and transient, multidimensional nodal method which allows more than two energy groups and uses a non-linear iterative method for efficient solution of the nodal equations. For both the static and transient methods, finite-difference equations which are corrected by the use of discontinuity factors are derived. The discontinuity factors are computed from a polynomial nodal method using a non-linear iteration technique. The polynomial nodal method is based upon a quartic approximation and utilizes a quadratic transverse-leakage approximation. The solution of the time-dependent equations is performed by the use of a quasi-static method in which the node-averaged fluxes are factored into shape and amplitude functions. The application of the quasi-static polynomial method to several benchmark problems demonstrates that the accuracy is consistent with that of other nodal methods. The use of the quasi-static method is shown to substantially reduce the computation time over the traditional fully-implicit time-integration method. Problems involving thermal-hydraulic feedback are accurately, and efficiently, solved by performing several reactivity/thermal-hydraulic updates per shape calculation.

  18. Celerity, pulse and wavelength of De broglie for a nucleon (proton or neutron), according to its kinetic energy. - Formulas and curves; Vitesse, impulsion et longueur d'onde de de broglie d'un nucleon (proton ou neutron), en fonction de son energie cinetique. - Formules et courbes

    Energy Technology Data Exchange (ETDEWEB)

    Rogozmski, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1952-07-01

    The celerity curves, pulse and wavelength of De Broglie for a nucleon (proton or neutron) have been calculated and traced accordingly to its kinetic energy. (M.B.) [French] Les courbes de vitesses, d mpulsion et de longueur d nde de Broglie d n nucleon (proton ou neutron) ont ete calcule et trace en fonction de son energie cinetique. (M.B.)

  19. TP1 - A computer program for the calculation of reactivity and kinetic parameters by one-dimensional neutron transport perturbation theory

    International Nuclear Information System (INIS)

    Kobayashi, K.

    1979-03-01

    TP1, a FORTRAN-IV program based on transport theory, has been developed to determine reactivity effects and kinetic parameters such as effective delayed neutron fractions and mean generation time by applying the usual perturbation formalism for one-dimensional geometry. Direct and adjoint angular dependent neutron fluxes are read from an interface file prepared by using the one-dimensional Ssub(n)-code DTK which provides options for slab, cylindrical and spherical geometry. Multigroup cross sections which are equivalent to those of the DTK-calculations are supplied in the SIGM-block which is also read from an interface file. This block which is usually produced by the code GRUCAL should contain the necessary delayed neutron data, which can be added to the original SIGMN-block by using the code SIGMUT. Two perturbation options are included in TP1: a) the usual first oder perturbation theory can be applied to determine probe reactivities, b) assuming that there are available direct fluxes for the unperturbed reactor system and adjoint fluxes for the perturbed system, the exact reactivity effect induced by the perturbation can be determined by an exact perturbation calculation. According to the input specifications, the output lists the reactivity contributions for each neutron reaction process in the desired detailed spatial and energy group resolution. (orig./RW) [de

  20. The development of the measurement technique of the control rod worth with the inverse kinetics method considering the influence of the steady neutron source

    International Nuclear Information System (INIS)

    Takeuchi, Mitsuo; Wada, Shigeru; Takahashi, Hiroyuki; Hayashi, Kazuhiko; Murayama, Yoji

    2000-09-01

    At the research reactor such as JRR-3M, the operation management is carried out in order to ensure safe operation, for example, the excess reactivity is measured regularly and confirmed that it satisfies a safety condition. The excess reactivity is calculated using control rod position in criticality and control rod worth measured by a positive period method (P.P method), the conventional inverse kinetic method (IK method) and so on. The neutron source, however, influences measurement results and brings in a measurement error. A new IK method considering the influence of the steady neutron sources is proposed and applied to the JRR-3M. This report shows that the proposed IK method measures control rod worth more precisely than a conventional IK method. (author)

  1. The Nudo, Rollo, Melon codes and nodal correlations

    International Nuclear Information System (INIS)

    Perlado, J.M.; Aragones, J.M.; Minguez, E.; Pena, J.

    1975-01-01

    Analysis of nodal calculation and checking results by the reference reactor experimental data. Nudo code description, adapting experimental data to nodal calculations. Rollo, Melon codes as improvement in the cycle life calculations of albedos, mixing parameters and nodal correlations. (author)

  2. VALIDATION OF FULL CORE GEOMETRY MODEL OF THE NODAL3 CODE IN THE PWR TRANSIENT BENCHMARK PROBLEMS

    Directory of Open Access Journals (Sweden)

    Tagor Malem Sembiring

    2015-10-01

    Full Text Available ABSTRACT VALIDATION OF FULL CORE GEOMETRY MODEL OF THE NODAL3 CODE IN THE PWR TRANSIENT BENCHMARK PROBLEMS. The coupled neutronic and thermal-hydraulic (T/H code, NODAL3 code, has been validated in some PWR static benchmark and the NEACRP PWR transient benchmark cases. However, the NODAL3 code have not yet validated in the transient benchmark cases of a control rod assembly (CR ejection at peripheral core using a full core geometry model, the C1 and C2 cases.  By this research work, the accuracy of the NODAL3 code for one CR ejection or the unsymmetrical group of CRs ejection case can be validated. The calculations by the NODAL3 code have been carried out by the adiabatic method (AM and the improved quasistatic method (IQS. All calculated transient parameters by the NODAL3 code were compared with the reference results by the PANTHER code. The maximum relative difference of 16% occurs in the calculated time of power maximum parameter by using the IQS method, while the relative difference of the AM method is 4% for C2 case.  All calculation results by the NODAL3 code shows there is no systematic difference, it means the neutronic and T/H modules are adopted in the code are considered correct. Therefore, all calculation results by using the NODAL3 code are very good agreement with the reference results. Keywords: nodal method, coupled neutronic and thermal-hydraulic code, PWR, transient case, control rod ejection.   ABSTRAK VALIDASI MODEL GEOMETRI TERAS PENUH PAKET PROGRAM NODAL3 DALAM PROBLEM BENCHMARK GAYUT WAKTU PWR. Paket program kopel neutronik dan termohidraulika (T/H, NODAL3, telah divalidasi dengan beberapa kasus benchmark statis PWR dan kasus benchmark gayut waktu PWR NEACRP.  Akan tetapi, paket program NODAL3 belum divalidasi dalam kasus benchmark gayut waktu akibat penarikan sebuah perangkat batang kendali (CR di tepi teras menggunakan model geometri teras penuh, yaitu kasus C1 dan C2. Dengan penelitian ini, akurasi paket program

  3. Kinetics of methane-ethane gas replacement in clathrate-hydrates studied by time-resolved neutron diffraction and Raman spectroscopy.

    Science.gov (United States)

    Murshed, M Mangir; Schmidt, Burkhard C; Kuhs, Werner F

    2010-01-14

    The kinetics of CH(4)-C(2)H(6) replacement in gas hydrates has been studied by in situ neutron diffraction and Raman spectroscopy. Deuterated ethane structure type I (C(2)H(6) sI) hydrates were transformed in a closed volume into methane-ethane mixed structure type II (CH(4)-C(2)H(6) sII) hydrates at 5 MPa and various temperatures in the vicinity of 0 degrees C while followed by time-resolved neutron powder diffraction on D20 at ILL, Grenoble. The role of available surface area of the sI starting material on the formation kinetics of sII hydrates was studied. Ex situ Raman spectroscopic investigations were carried out to crosscheck the gas composition and the distribution of the gas species over the cages as a function of structure type and compared to the in situ neutron results. Raman micromapping on single hydrate grains showed compositional and structural gradients between the surface and core of the transformed hydrates. Moreover, the observed methane-ethane ratio is very far from the one expected for a formation from a constantly equilibrated gas phase. The results also prove that gas replacement in CH(4)-C(2)H(6) hydrates is a regrowth process involving the nucleation of new crystallites commencing at the surface of the parent C(2)H(6) sI hydrate with a progressively shrinking core of unreacted material. The time-resolved neutron diffraction results clearly indicate an increasing diffusion limitation of the exchange process. This diffusion limitation leads to a progressive slowing down of the exchange reaction and is likely to be responsible for the incomplete exchange of the gases.

  4. On the treatment of nonlinear local feedbacks within advanced nodal generalized perturbation theory

    International Nuclear Information System (INIS)

    Maldonado, G.I.; Turinsky, P.J.; Kropaczek, D.J.

    1993-01-01

    Recent efforts to upgrade the underlying neutronics formulations within the in-core nuclear fuel management optimization code FORMOSA (Ref. 1) have produced two important developments; first, a computationally efficient and second-order-accurate advanced nodal generalized perturbation theory (GPT) model [derived from the nonlinear iterative nodal expansion method (NEM)] for evaluating core attributes (i.e., k eff and power distribution versus cycle burnup), and second, an equally efficient and accurate treatment of local thermal-hydraulic and fission product feedbacks embedded within NEM GPT. The latter development is the focus of this paper

  5. Spectral nodal method for one-speed X,Y-geometry Eigenvalue diffusion problems

    International Nuclear Information System (INIS)

    Dominguez, Dany S.; Lorenzo, Daniel M.; Hernandez, Carlos G.; Barros, Ricardo C.; Silva, Fernando C. da

    2001-01-01

    Presented here is a new numerical nodal method for steady-state multidimensional neutron diffusion equation in rectangular geometry. Our method is based on a spectral analysis of the transverse-integrated nodal diffusion equations. These equations are obtained by integrating the diffusion equation in X and Y directions, and then considering flat approximations for the transverse leakage terms. These flat approximations are the only approximations that we consider in this method; as a result the numerical solutions are completely free from truncation errors in slab geometry. We show numerical results to illustrate the method's accuracy for coarse mesh calculations in a heterogeneous medium. (author)

  6. Heterogeneous treatment in the variational nodal method

    International Nuclear Information System (INIS)

    Fanning, T.H.

    1995-01-01

    The variational nodal transport method is reduced to its diffusion form and generalized for the treatment of heterogeneous nodes while maintaining nodal balances. Adapting variational methods to heterogeneous nodes requires the ability to integrate over a node with discontinuous cross sections. In this work, integrals are evaluated using composite gaussian quadrature rules, which permit accurate integration while minimizing computing time. Allowing structure within a nodal solution scheme avoids some of the necessity of cross section homogenization, and more accurately defines the intra-nodal flux shape. Ideally, any desired heterogeneity can be constructed within the node; but in reality, the finite set of basis functions limits the practical resolution to which fine detail can be defined within the node. Preliminary comparison tests show that the heterogeneous variational nodal method provides satisfactory results even if some improvements are needed for very difficult, configurations

  7. Stochastic kinetics

    International Nuclear Information System (INIS)

    Colombino, A.; Mosiello, R.; Norelli, F.; Jorio, V.M.; Pacilio, N.

    1975-01-01

    A nuclear system kinetics is formulated according to a stochastic approach. The detailed probability balance equations are written for the probability of finding the mixed population of neutrons and detected neutrons, i.e. detectrons, at a given level for a given instant of time. Equations are integrated in search of a probability profile: a series of cases is analyzed through a progressive criterium. It tends to take into account an increasing number of physical processes within the chosen model. The most important contribution is that solutions interpret analytically experimental conditions of equilibrium (moise analysis) and non equilibrium (pulsed neutron measurements, source drop technique, start up procedures)

  8. RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - models and correlations

    Energy Technology Data Exchange (ETDEWEB)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

    1998-03-01

    This document describes the major modifications and improvements made to the modeling of the RAMONA-3B/MOD0 code since 1981, when the code description and assessment report was completed. The new version of the code is RAMONA-4B. RAMONA-4B is a systems transient code for application to different versions of Boiling Water Reactors (BWR) such as the current BWR, the Advanced Boiling Water Reactor (ABWR), and the Simplified Boiling Water Reactor (SBWR). This code uses a three-dimensional neutron kinetics model coupled with a multichannel, non-equilibrium, drift-flux, two-phase flow formulation of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients and instability issues. Chapter 1 is an overview of the code`s capabilities and limitations; Chapter 2 discusses the neutron kinetics modeling and the implementation of reactivity edits. Chapter 3 is an overview of the heat conduction calculations. Chapter 4 presents modifications to the thermal-hydraulics model of the vessel, recirculation loop, steam separators, boron transport, and SBWR specific components. Chapter 5 describes modeling of the plant control and safety systems. Chapter 6 presents and modeling of Balance of Plant (BOP). Chapter 7 describes the mechanistic containment model in the code. The content of this report is complementary to the RAMONA-3B code description and assessment document. 53 refs., 81 figs., 13 tabs.

  9. Discrete rod burnup analysis capability in the Westinghouse advanced nodal code

    International Nuclear Information System (INIS)

    Buechel, R.J.; Fetterman, R.J.; Petrunyak, M.A.

    1992-01-01

    Core design analysis in the last several years has evolved toward the adoption of nodal-based methods to replace traditional fine-mesh models as the standard neutronic tool for first core and reload design applications throughout the nuclear industry. The accuracy, speed, and reduction in computation requirements associated with the nodal methods have made three-dimensional modeling the preferred approach to obtain the most realistic core model. These methods incorporate detailed rod power reconstruction as well. Certain design applications such as confirmation of fuel rod design limits and fuel reconstitution considerations, for example, require knowledge of the rodwise burnup distribution to avoid unnecessary conservatism in design analyses. The Westinghouse Advanced Nodal Code (ANC) incorporates the capability to generate the intra-assembly pin burnup distribution using an efficient algorithm

  10. Analysis of NEA-NSC PWR Uncontrolled Control Rod Withdrawal at Zero Power Benchmark Cases with NODAL3 Code

    Directory of Open Access Journals (Sweden)

    Tagor Malem Sembiring

    2017-01-01

    Full Text Available The in-house coupled neutronic and thermal-hydraulic (N/T-H code of BATAN (National Nuclear Energy Agency of Indonesia, NODAL3, based on the few-group neutron diffusion equation in 3-dimensional geometry using the polynomial nodal method, has been verified with static and transient PWR benchmark cases. This paper reports the verification of NODAL3 code in the NEA-NSC PWR uncontrolled control rods withdrawal at zero power benchmark. The objective of this paper is to determine the accuracy of NODAL3 code in solving the continuously slow and fast reactivity insertions due to single and group of control rod bank withdrawn while the power and temperature increment are limited by the Doppler coefficient. The benchmark is chosen since many organizations participated using various methods and approximations, so the calculation results of NODAL3 can be compared to other codes’ results. The calculated parameters are performed for the steady-state, transient core averaged, and transient hot pellet results. The influence of radial and axial nodes number was investigated for all cases. The results of NODAL3 code are in very good agreement with the reference solutions if the radial and axial nodes number is 2 × 2 and 2 × 18 (total axial layers, respectively.

  11. Nodal pricing in a coupled electricity market

    OpenAIRE

    Bjørndal, Endre; Bjørndal, Mette; Cai, Hong

    2014-01-01

    This paper investigates a pricing model for an electricity market with a hybrid congestion management method, i.e. part of the system applies a nodal pricing scheme and the rest applies a zonal pricing scheme. The model clears the zonal and nodal pricing areas simultaneously. The nodal pricing area is affected by the changes in the zonal pricing area since it is directly connected to the zonal pricing area by commercial trading. The model is tested on a 13-node power system. Within the area t...

  12. Hybrid nodal methods in the solution of the diffusion equations in X Y geometry; Metodos nodales hibridos en la solucion de las ecuaciones de difusion en geometria XY

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez M, N. [CFE, Carretera Cardel-Nautla Km. 43.5, 91680 Veracruz (Mexico); Alonso V, G.; Valle G, E. del [IPN-ESFM, 07738 Mexico D.F. (Mexico)]. e-mail: nhmiranda@mexico.com

    2003-07-01

    In 1979, Hennart and collaborators applied several schemes of classic finite element in the numerical solution of the diffusion equations in X Y geometry and stationary state. Almost two decades then, in 1996, himself and other collaborators carried out a similar work but using nodal schemes type finite element. Continuing in this last direction, in this work a group it is described a set of several Hybrid Nodal schemes denominated (NH) as well as their application to solve the diffusion equations in multigroup in stationary state and X Y geometry. The term hybrid nodal it means that such schemes interpolate not only Legendre moments of face and of cell but also the values of the scalar flow of neutrons in the four corners of each cell or element of the spatial discretization of the domain of interest. All the schemes here considered are polynomials like they were it their predecessors. Particularly, its have developed and applied eight different hybrid nodal schemes that its are very nearby related with those developed by Hennart and collaborators in the past. It is treated of schemes in those that nevertheless that decreases the number of interpolation parameters it is conserved the accurate in relation to the bi-quadratic and bi-cubic schemes. Of these eight, three were described and applied in a previous work. It is the bi-lineal classic scheme as well as the hybrid nodal schemes, bi-quadratic and bi-cubic for that here only are described the other 5 hybrid nodal schemes although they are provided numerical results for several test problems with all them. (Author)

  13. A comparison of two nodal codes : Advanced nodal code (ANC) and analytic function expansion nodal (AFEN) code

    International Nuclear Information System (INIS)

    Chung, S.K.; Hah, C.J.; Lee, H.C.; Kim, Y.H.; Cho, N.Z.

    1996-01-01

    Modern nodal methods usually employs the transverse integration technique in order to reduce a multi-dimensional diffusion equation to one-dimensional diffusion equations. The use of the transverse integration technique requires two major approximations such as a transverse leakage approximation and a one-dimensional flux approximation. Both the transverse leakage and the one-dimensional flux are approximated by polynomials. ANC (Advanced Nodal Code) developed by Westinghouse employs a modern nodal expansion method for the flux calculation, the equivalence theory for the homogenization error reduction and a group theory for pin power recovery. Unlike the conventional modern nodal methods, AFEN (Analytic Function Expansion Nodal) method expands homogeneous flux distributions within a node into non-separable analytic basis functions, which eliminate two major approximations of the modern nodal methods. A comparison study of AFEN with ANC has been performed to see the applicability of AFEN to commercial PWR and different types of reactors such as MOX fueled reactor. The qualification comparison results demonstrate that AFEN methodology is accurate enough to apply for commercial PWR analysis. The results show that AFEN provides very accurate results (core multiplication factor and assembly power distribution) for cores that exhibit strong flux gradients as in a MOX loaded core. (author)

  14. Maternal Nodal inversely affects NODAL and STOX1 expression in the fetal placenta

    Directory of Open Access Journals (Sweden)

    Hari Krishna Thulluru

    2013-08-01

    Full Text Available Nodal, a secreted signaling protein from the TGFβ-super family plays a vital role during early embryonic development. Recently, it was found that maternal decidua-specific Nodal knockout mice show intrauterine growth restriction (IUGR and preterm birth. As the chromosomal location of NODAL is in the same linkage area as the susceptibility gene STOX1, associated with the familial form of early-onset, IUGR-complicated pre-eclampsia, their potential maternal-fetal interaction was investigated. Pre-eclamptic mothers with children who carried the STOX1 susceptibility allele themselves all carried the NODAL H165R SNP, which causes a 50% reduced activity. Surprisingly, in decidua Nodal knockout mice the fetal placenta showed up-regulation of STOX1 and NODAL expression. Conditioned media of human first trimester decidua and a human endometrial stromal cell line (T-HESC treated with siRNAs against NODAL or carrying the H165R SNP were also able to induce NODAL and STOX1 expression when added to SGHPL-5 first trimester extravillous trophoblast cells. Finally, a human TGFß-BMP-Signaling-Pathway PCR-Array on decidua and the T-HESC cell line with Nodal knockdown revealed upregulation of Activin-A, which was confirmed in conditioned media by ELISA. We show that maternal decidua Nodal knockdown gives upregulation of NODAL and STOX1 mRNA expression in fetal extravillous trophoblast cells, potentially via upregulation of Activin-A in the maternal decidua. As both Activin-A and Nodal have been implicated in pre-eclampsia, being increased in serum of pre-eclamptic women and upregulated in pre-eclamptic placentas respectively, this interaction at the maternal-fetal interface might play a substantial role in the development of pre-eclampsia.

  15. Quantum oscillations in nodal line systems

    Science.gov (United States)

    Yang, Hui; Moessner, Roderich; Lim, Lih-King

    2018-04-01

    We study signatures of magnetic quantum oscillations in three-dimensional nodal line semimetals at zero temperature. The extended nature of the degenerate bands can result in a Fermi surface geometry with topological genus one, as well as a Fermi surface of electron and hole pockets encapsulating the nodal line. Moreover, the underlying two-band model to describe a nodal line is not unique, in that there are two classes of Hamiltonian with distinct band topology giving rise to the same Fermi-surface geometry. After identifying the extremal cyclotron orbits in various magnetic field directions, we study their concomitant Landau levels and resulting quantum oscillation signatures. By Landau-fan-diagram analyses, we extract the nontrivial π Berry phase signature for extremal orbits linking the nodal line.

  16. Sensitivity of SBLOCA analysis to model nodalization

    International Nuclear Information System (INIS)

    Lee, C.; Ito, T.; Abramson, P.B.

    1983-01-01

    The recent Semiscale test S-UT-8 indicates the possibility for primary liquid to hang up in the steam generators during a SBLOCA, permitting core uncovery prior to loop-seal clearance. In analysis of Small Break Loss of Coolant Accidents with RELAP5, it is found that resultant transient behavior is quite sensitive to the selection of nodalization for the steam generators. Although global parameters such as integrated mass loss, primary inventory and primary pressure are relatively insensitive to the nodalization, it is found that the predicted distribution of inventory around the primary is significantly affected by nodalization. More detailed nodalization predicts that more of the inventory tends to remain in the steam generators, resulting in less inventory in the reactor vessel and therefore causing earlier and more severe core uncovery

  17. Twisted Vector Bundles on Pointed Nodal Curves

    Indian Academy of Sciences (India)

    Abstract. Motivated by the quest for a good compactification of the moduli space of -bundles on a nodal curve we establish a striking relationship between Abramovich's and Vistoli's twisted bundles and Gieseker vector bundles.

  18. VARIANT: VARIational anisotropic nodal transport for multidimensional Cartesian and hexadgonal geometry calculation

    International Nuclear Information System (INIS)

    Palmiotti, G.; Carrico, C.B.; Lewis, E.E.

    1995-10-01

    The theoretical basis, implementation information and numerical results are presented for VARIANT (VARIational Anisotropic Neutron Transport), a FORTRAN module of the DIF3D code system at Argonne National Laboratory. VARIANT employs the variational nodal method to solve multigroup steady-state neutron diffusion and transport problems. The variational nodal method is a hybrid finite element method that guarantees nodal balance and permits spatial refinement through the use of hierarchical complete polynomial trial functions. Angular variables are expanded with complete or simplified P 1 , P 3 or P 5 5 spherical harmonics approximations with full anisotropic scattering capability. Nodal response matrices are obtained, and the within-group equations are solved by red-black or four-color iteration, accelerated by a partitioned matrix algorithm. Fission source and upscatter iterations strategies follow those of DIF3D. Two- and three-dimensional Cartesian and hexagonal geometries are implemented. Forward and adjoint eigenvalue, fixed source, gamma heating, and criticality (concentration) search problems may be performed

  19. The NODAL system for the SPS

    International Nuclear Information System (INIS)

    Crowley-Milling, M.C.; Shering, G.C.

    1978-01-01

    A comprehensive description is given of the NODAL system used for computer control of the CERN Super-Proton Synchrotron. Details are given of NODAL, a high-level programming language based on FOCAL and SNOBOL4, designed for interactive use. It is shown how this interpretive language is used with a network of computers and how it can be extended by adding machine-code modules. The report updates and replaces an earlier one published in 1974. (Auth.)

  20. Nodal coupling by response matrix principles

    International Nuclear Information System (INIS)

    Ancona, A.; Becker, M.; Beg, M.D.; Harris, D.R.; Menezes, A.D.; VerPlanck, D.M.; Pilat, E.

    1977-01-01

    The response matrix approach has been used in viewing a reactor node in isolation and in characterizing the node by reflection and trans-emission factors. These are then used to generate invariant imbedding parameters, which in turn are used in a nodal reactor simulator code to compute core power distributions in two and three dimensions. Various nodal techniques are analyzed and converted into a single invariant imbedding formalism

  1. Magnonic triply-degenerate nodal points

    Science.gov (United States)

    Owerre, S. A.

    2017-12-01

    We generalize the concept of triply-degenerate nodal points to non-collinear antiferromagnets. Here, we introduce this concept to insulating quantum antiferromagnets on the decorated honeycomb lattice, with spin-1 bosonic quasiparticle excitations known as magnons. We demonstrate the existence of magnonic surface states with constant energy contours that form pairs of magnonic arcs connecting the surface projection of the magnonic triple nodal points. The quasiparticle excitations near the triple nodal points represent three-component bosons beyond that of magnonic Dirac, Weyl, and nodal-line cases. They can be regarded as a direct reflection of the intrinsic spin carried by magnons. Furthermore, we show that the magnonic triple nodal points can split into magnonic Weyl points, as the system transits from a non-collinear spin structure to a non-coplanar one with a non-zero scalar spin chirality. Our results not only apply to insulating antiferromagnets, but also provide a platform to seek for triple nodal points in metallic antiferromagnets.

  2. TISKTH-3: a couple neutronics/thermal-hydraulics code for the transient analysis of light water reactors

    International Nuclear Information System (INIS)

    Peng Muzhang; Zhang Quan; Wang Guoli; Zhang Yuman

    1988-01-01

    TISKTH-3 is a coupled neutronics/thermal-hydraulics code for the transient analysis. A 3-dimensional neutron kinetics equation solved by the Nodal Green's Function Method is used for the neutronics model of the code. A homogeneous equilibrium model with a complete boiling curve and two numerical solutions of the implicit and explicit scheme is used for the thermal-hydraulics model of the code. A 2-dimensional heat conduction equation with variable conductivity solved by the method of weighted residuals is used for the fuel rod heat transfer model of the code. TISKTH-3 is able to analyze the fast transient process and complicate accident situations in the core. The initative applications have shown that the stability and convergency in the calculations with the code are satisfactory

  3. TISKTH-3: a couple neutronics/thermal-hydraulics code for the transient analysis of light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Muzhang, Peng; Quan, Zhang; Guoli, Wang; Yuman, Zhang

    1988-03-01

    TISKTH-3 is a coupled neutronics/thermal-hydraulics code for the transient analysis. A 3-dimensional neutron kinetics equation solved by the Nodal Green's Function Method is used for the neutronics model of the code. A homogeneous equilibrium model with a complete boiling curve and two numerical solutions of the implicit and explicit scheme is used for the thermal-hydraulics model of the code. A 2-dimensional heat conduction equation with variable conductivity solved by the method of weighted residuals is used for the fuel rod heat transfer model of the code. TISKTH-3 is able to analyze the fast transient process and complicate accident situations in the core. The initative applications have shown that the stability and convergency in the calculations with the code are satisfactory.

  4. Discrete nodal integral transport-theory method for multidimensional reactor physics and shielding calculations

    International Nuclear Information System (INIS)

    Lawrence, R.D.; Dorning, J.J.

    1980-01-01

    A coarse-mesh discrete nodal integral transport theory method has been developed for the efficient numerical solution of multidimensional transport problems of interest in reactor physics and shielding applications. The method, which is the discrete transport theory analogue and logical extension of the nodal Green's function method previously developed for multidimensional neutron diffusion problems, utilizes the same transverse integration procedure to reduce the multidimensional equations to coupled one-dimensional equations. This is followed by the conversion of the differential equations to local, one-dimensional, in-node integral equations by integrating back along neutron flight paths. One-dimensional and two-dimensional transport theory test problems have been systematically studied to verify the superior computational efficiency of the new method

  5. Analysis of the applicability of acceleration methods for a triangular prism geometry nodal diffusion code

    International Nuclear Information System (INIS)

    Fujimura, Toichiro; Okumura, Keisuke

    2002-11-01

    A prototype version of a diffusion code has been developed to analyze the hexagonal core as reduced moderation reactor and the applicability of some acceleration methods have been investigated to accelerate the convergence of the iterative solution method. The hexagonal core is divided into regular triangular prisms in the three-dimensional code MOSRA-Prism and a polynomial expansion nodal method is applied to approximate the neutron flux distribution by a cubic polynomial. The multi-group diffusion equation is solved iteratively with ordinal inner and outer iterations and the effectiveness of acceleration methods is ascertained by applying an adaptive acceleration method and a neutron source extrapolation method, respectively. The formulation of the polynomial expansion nodal method is outlined in the report and the local and global effectiveness of the acceleration methods is discussed with various sample calculations. A new general expression of vacuum boundary condition, derived in the formulation is also described. (author)

  6. Two-energy group solution of the diffusion equation by the multidimensional nodal polynomial expansion method

    International Nuclear Information System (INIS)

    Ribeiro, R.D.M.; Vellozo, S.O.; Botelho, D.A.

    1983-01-01

    The EPON computer code based in a Nodal Polynomial Expansion Method, wrote in Fortran IV, for steady-state, square geometry, one-dimensional or two-dimensional geometry and for one or two-energy group is presented. The neutron and power flux distributions for nuclear power plants were calculated, comparing with codes that use similar or different methodologies. The availability, economy and speed of the methodology is demonstrated. (E.G.) [pt

  7. Moderator feedback effects in two-dimensional nodal methods for pressurized water reactor analysis

    International Nuclear Information System (INIS)

    Downar, T.J.

    1987-01-01

    A method was developed for incorporating moderator feedback effects in two-dimensional nodal codes used for pressurized water reactor (PWR) neutronic analysis. Equations for the assembly average quality and density are developed in terms of the assembly power calculated in two dimensions. The method is validated with a Westinghouse PWR using the Electric Power Research Institute code SIMULATE-E. Results show a several percent improvement is achieved in the two-dimensional power distribution prediction compared to methods without moderator feedback

  8. Neutron diffraction study of the formation kinetics of ordered antiphase domains in titanium carbohydride TiC{sub x}H{sub y}

    Energy Technology Data Exchange (ETDEWEB)

    Khidirov, I., E-mail: khidirov@inp.uz [Uzbekistan Academy of Sciences, Institute of Nuclear Physics (Uzbekistan)

    2015-09-15

    The kinetics of formation and growth of ordered antiphase domains (APDs) in titanium carbohydride TiC{sub 0.50}H{sub 0.21} has been investigated by neutron diffraction. A model of ordered APDs is proposed. It is established that the pronounced ordering of interstitial atoms and APDs begin at 450°C. It is shown that the period of ordered APDs (P ≈ 10–12) is independent of the exposure time at a constant temperature. It is found that the temperature of ordered APDs, T{sub OAPD}, increases nonlinearly with an increase in the carbon concentration in the range 0.50 ≤ C/Ti ≤ 0.70. The formation temperature of ordered APDs is found to correlate with the concentration dependence of the order–disorder transition temperature and be 0.60 of the order–disorder transition temperature: T{sub APD} = 0.60Τ{sub C}.

  9. Feasibility of creating a specialized reactimeter based on the inverse solution to kinetics equation with a current-mode neutron detector

    Energy Technology Data Exchange (ETDEWEB)

    Koshelev, A. S., E-mail: alexsander.coshelev@yandex.ru; Arapov, A. V.; Ovchinnikov, M. A. [Russian Federal Nuclear Center–All-Russian Research Institute of Experimental Physics (Russian Federation)

    2016-12-15

    The file-evaluation results of a reactimeter based on the inverse solution to the kinetics equation (ISKE) are presented, which were obtained using an operating hardware-measuring complex with a KNK-4 neutron detector working in the current mode. The processing of power-recording files of the BR-1M, BR-K1, and VIR-2M reactors of the Russian Federal Nuclear Center—All-Russian Research Institute of Experimental Physics, which was performed with the use of Excel simulation of the ISKE formalism, demonstrated the feasibility of implementation of the reactivity monitoring (during the operation of these reactors at stationary power) beginning from the level of ~5 × 10{sup –4}β{sub eff}.

  10. Recursive solutions for multi-group neutron kinetics diffusion equations in homogeneous three-dimensional rectangular domains with time dependent perturbations

    Energy Technology Data Exchange (ETDEWEB)

    Petersen, Claudio Z. [Universidade Federal de Pelotas, Capao do Leao (Brazil). Programa de Pos Graduacao em Modelagem Matematica; Bodmann, Bardo E.J.; Vilhena, Marco T. [Universidade Federal do Rio Grande do Sul, Porto Alegre, RS (Brazil). Programa de Pos-graduacao em Engenharia Mecanica; Barros, Ricardo C. [Universidade do Estado do Rio de Janeiro, Nova Friburgo, RJ (Brazil). Inst. Politecnico

    2014-12-15

    In the present work we solve in analytical representation the three dimensional neutron kinetic diffusion problem in rectangular Cartesian geometry for homogeneous and bounded domains for any number of energy groups and precursor concentrations. The solution in analytical representation is constructed using a hierarchical procedure, i.e. the original problem is reduced to a problem previously solved by the authors making use of a combination of the spectral method and a recursive decomposition approach. Time dependent absorption cross sections of the thermal energy group are considered with step, ramp and Chebyshev polynomial variations. For these three cases, we present numerical results and discuss convergence properties and compare our results to those available in the literature.

  11. Qualification of coupled 3D neutron kinetic/thermal hydraulic code systems by the calculation of a VVER-440 benchmark. Re-connection of an isolated loop

    Energy Technology Data Exchange (ETDEWEB)

    Kotsarev, Alexander; Lizorkin, Mikhail [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation); Bencik, Marek; Hadek, Jan [UJV Rez, a.s., Rez (Czech Republic); Kozmenkov, Yaroslav; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany)

    2016-09-15

    The 7th AER dynamic benchmark is a continuation of the efforts to validate the codes systematically for the estimation of the transient behavior of VVER type nuclear power plants. The main part of the benchmark is the simulation of the re-connection of an isolated circulation loop with low temperature in a VVER-440 plant. This benchmark was calculated by the National Research Centre ''Kurchatov Institute'' (with the code ATHLET/BIPR-VVER), UJV Rez (with the code RELAP5-3D {sup copyright}) and HZDR (with the code DYN3D/ATHLET). The paper gives an overview of the behavior of the main thermal hydraulic and neutron kinetic parameters in the provided solutions.

  12. An Adaptive Approach to Variational Nodal Diffusion Problems

    International Nuclear Information System (INIS)

    Zhang Hui; Lewis, E.E.

    2001-01-01

    An adaptive grid method is presented for the solution of neutron diffusion problems in two dimensions. The primal hybrid finite elements employed in the variational nodal method are used to reduce the diffusion equation to a coupled set of elemental response matrices. An a posteriori error estimator is developed to indicate the magnitude of local errors stemming from the low-order elemental interface approximations. An iterative procedure is implemented in which p refinement is applied locally by increasing the polynomial order of the interface approximations. The automated algorithm utilizes the a posteriori estimator to achieve local error reductions until an acceptable level of accuracy is reached throughout the problem domain. Application to a series of X-Y benchmark problems indicates the reduction of computational effort achievable by replacing uniform with adaptive refinement of the spatial approximations

  13. A nonlinear analytic function expansion nodal method for transient calculations

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Han Gyn; Park, Sang Yoon; Cho, Byung Oh; Zee, Sung Quun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    The nonlinear analytic function expansion nodal (AFEN) method is applied to the solution of the time-dependent neutron diffusion equation. Since the AFEN method requires both the particular solution and the homogeneous solution to the transient fixed source problem, the derivation of the solution method is focused on finding the particular solution efficiently. To avoid complicated particular solutions, the source distribution is approximated by quadratic polynomials and the transient source is constructed such that the error due to the quadratic approximation is minimized, In addition, this paper presents a new two-node solution scheme that is derived by imposing the constraint of current continuity at the interface corner points. The method is verified through a series of application to the NEACRP PWR rod ejection benchmark problems. 6 refs., 2 figs., 1 tab. (Author)

  14. A nonlinear analytic function expansion nodal method for transient calculations

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Han Gyn; Park, Sang Yoon; Cho, Byung Oh; Zee, Sung Quun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-12-31

    The nonlinear analytic function expansion nodal (AFEN) method is applied to the solution of the time-dependent neutron diffusion equation. Since the AFEN method requires both the particular solution and the homogeneous solution to the transient fixed source problem, the derivation of the solution method is focused on finding the particular solution efficiently. To avoid complicated particular solutions, the source distribution is approximated by quadratic polynomials and the transient source is constructed such that the error due to the quadratic approximation is minimized, In addition, this paper presents a new two-node solution scheme that is derived by imposing the constraint of current continuity at the interface corner points. The method is verified through a series of application to the NEACRP PWR rod ejection benchmark problems. 6 refs., 2 figs., 1 tab. (Author)

  15. ANDREA: Advanced nodal diffusion code for reactor analysis

    International Nuclear Information System (INIS)

    Belac, J.; Josek, R.; Klecka, L.; Stary, V.; Vocka, R.

    2005-01-01

    A new macro code is being developed at NRI which will allow coupling of the advanced thermal-hydraulics model with neutronics calculations as well as efficient use in core loading pattern optimization process. This paper describes the current stage of the macro code development. The core simulator is based on the nodal expansion method, Helios lattice code is used for few group libraries preparation. Standard features such as pin wise power reconstruction and feedback iterations on critical control rod position, boron concentration and reactor power are implemented. A special attention is paid to the system and code modularity in order to enable flexible and easy implementation of new features in future. Precision of the methods used in the macro code has been verified on available benchmarks. Testing against Temelin PWR operational data is under way (Authors)

  16. Error Estimation and Accuracy Improvements in Nodal Transport Methods

    International Nuclear Information System (INIS)

    Zamonsky, O.M.

    2000-01-01

    The accuracy of the solutions produced by the Discrete Ordinates neutron transport nodal methods is analyzed.The obtained new numerical methodologies increase the accuracy of the analyzed scheems and give a POSTERIORI error estimators. The accuracy improvement is obtained with new equations that make the numerical procedure free of truncation errors and proposing spatial reconstructions of the angular fluxes that are more accurate than those used until present. An a POSTERIORI error estimator is rigurously obtained for one dimensional systems that, in certain type of problems, allows to quantify the accuracy of the solutions. From comparisons with the one dimensional results, an a POSTERIORI error estimator is also obtained for multidimensional systems. LOCAL indicators, which quantify the spatial distribution of the errors, are obtained by the decomposition of the menctioned estimators. This makes the proposed methodology suitable to perform adaptive calculations. Some numerical examples are presented to validate the theoretical developements and to illustrate the ranges where the proposed approximations are valid

  17. CORTAP: a coupled neutron kinetics-heat transfer digital computer program for the dynamic simulation of the high temperature gas cooled reactor core

    International Nuclear Information System (INIS)

    Cleveland, J.C.

    1977-01-01

    CORTAP (Core Transient Analysis Program) was developed to predict the dynamic behavior of the High Temperature Gas Cooled Reactor (HTGR) core under normal operational transients and postulated accident conditions. CORTAP is used both as a stand-alone component simulation and as part of the HTGR nuclear steam supply (NSS) system simulation code ORTAP. The core thermal neutronic response is determined by solving the heat transfer equations for the fuel, moderator and coolant in an average powered region of the reactor core. The space independent neutron kinetics equations are coupled to the heat transfer equations through a rapidly converging iterative technique. The code has the capability to determine conservative fuel, moderator, and coolant temperatures in the ''hot'' fuel region. For transients involving a reactor trip, the core heat generation rate is determined from an expression for decay heat following a scram. Nonlinear effects introduced by temperature dependent fuel, moderator, and coolant properties are included in the model. CORTAP predictions will be compared with dynamic test results obtained from the Fort St. Vrain reactor owned by Public Service of Colorado, and, based on these comparisons, appropriate improvements will be made in CORTAP

  18. Coupling of unidimensional neutron kinetics to thermal hydraulics in parallel channels; Acoplamiento de cinetica neutronica unidimensional a canales termohidraulicos en paralelo

    Energy Technology Data Exchange (ETDEWEB)

    Cecenas F, M.; Campos G, R.M. [IIE, Av. Reforma 113, Col. Palmira, Cuernavaca, Morelos (Mexico)]. e-mail: mcf@iie.org.mx

    2003-07-01

    In this work the dynamic behavior of a consistent system in fifteen channels in parallel that represent the reactor core of a BWR type, coupled of a kinetic neutronic model in one dimension is studied by means of time series. The arrangement of channels is obtained collapsing the assemblies that it consists the core to an arrangement of channels prepared in straight lines, and it is coupled to the unidimensional solution of the neutron diffusion equation. This solution represents the radial power distribution, and initially the static solution is obtained to verify that the one modeling core is critic. The coupled set nuclear-thermal hydraulics it is solved numerically by means of a net of CPUs working in the outline teacher-slave by means of Parallel Virtual Machine (PVM), subject to the restriction that the pressure drop is equal for each channel, which is executed iterating on the refrigerant distribution. The channels are dimensioned according to the one Stability Benchmark of the Ringhals swedish plant, organized by the Nuclear Energy Agency in 1994. From the information of this benchmark it is obtained the axial power profile for each channel, which is assumed as invariant in the time. To obtain the time series, the system gets excited with white noise (sequence that statistically obeys to a normal distribution with zero media), so that the power generated in each channel it possesses the same ones characteristics of a typical signal obtained by means of the acquisition of those signals of neutron flux in a BWR reactor. (Author)

  19. The analytic nodal method in cylindrical geometry

    International Nuclear Information System (INIS)

    Prinsloo, Rian H.; Tomasevic, Djordje I.

    2008-01-01

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

  20. Encapsulation of nodal segments of lobelia chinensis

    Directory of Open Access Journals (Sweden)

    Weng Hing Thong

    2015-04-01

    Full Text Available Lobelia chinensis served as an important herb in traditional chinese medicine. It is rare in the field and infected by some pathogens. Therefore, encapsulation of axillary buds has been developed for in vitro propagation of L. chinensis. Nodal explants of L. chinensis were used as inclusion materials for encapsulation. Various combinations of calcium chloride and sodium alginate were tested. Encapsulation beads produced by mixing 50 mM calcium chloride and 3.5% sodium alginate supported the optimal in vitro conversion potential. The number of multiple shoots formed by encapsulated nodal segments was not significantly different from the average of shoots produced by non-encapsulated nodal segments. The encapsulated nodal segments regenerated in vitro on different medium. The optimal germination and regeneration medium was Murashige-Skoog medium. Plantlets regenerated from the encapsulated nodal segments were hardened, acclimatized and established well in the field, showing similar morphology with parent plants. This encapsulation technology would serve as an alternative in vitro regeneration system for L. chinensis.

  1. Neutron transport

    International Nuclear Information System (INIS)

    Berthoud, Georges; Ducros, Gerard; Feron, Damien; Guerin, Yannick; Latge, Christian; Limoge, Yves; Santarini, Gerard; Seiler, Jean-Marie; Vernaz, Etienne; Coste-Delclaux, Mireille; M'Backe Diop, Cheikh; Nicolas, Anne; Andrieux, Catherine; Archier, Pascal; Baudron, Anne-Marie; Bernard, David; Biaise, Patrick; Blanc-Tranchant, Patrick; Bonin, Bernard; Bouland, Olivier; Bourganel, Stephane; Calvin, Christophe; Chiron, Maurice; Damian, Frederic; Dumonteil, Eric; Fausser, Clement; Fougeras, Philippe; Gabriel, Franck; Gagnier, Emmanuel; Gallo, Daniele; Hudelot, Jean-Pascal; Hugot, Francois-Xavier; Dat Huynh, Tan; Jouanne, Cedric; Lautard, Jean-Jacques; Laye, Frederic; Lee, Yi-Kang; Lenain, Richard; Leray, Sylvie; Litaize, Olivier; Magnaud, Christine; Malvagi, Fausto; Mijuin, Dominique; Mounier, Claude; Naury, Sylvie; Nicolas, Anne; Noguere, Gilles; Palau, Jean-Marc; Le Pallec, Jean-Charles; Peneliau, Yannick; Petit, Odile; Poinot-Salanon, Christine; Raepsaet, Xavier; Reuss, Paul; Richebois, Edwige; Roque, Benedicte; Royer, Eric; Saint-Jean, Cyrille de; Santamarina, Alain; Serot, Olivier; Soldevila, Michel; Tommasi, Jean; Trama, Jean-Christophe; Tsilanizara, Aime; Behar, Christophe; Provitina, Olivier; Lecomte, Michael; Forestier, Alain; Bender, Alexandra; Parisot, Jean-Francois; Finot, Pierre

    2013-10-01

    This bibliographical note presents a reference book which addresses the study of neutron transport in matter, the study of conditions for a chain reaction and the study of modifications of matter composition due to nuclear reactions. This book presents the main nuclear data, their measurement, assessment and processing, and the spallation. It proposes an overview of methods applied for the study of neutron transport: basic equations and their derived forms, deterministic methods and Monte Carlo method of resolution of the Boltzmann equation, methods of resolution of generalized Bateman equations, methods of time resolution of space kinetics coupled equations. It presents the main calculation codes, discusses the qualification and experimental aspects, and gives an overview of neutron transport applications: neutron transport calculation of reactors, neutron transport coupled with other disciplines, physics of fuel cycle, criticality

  2. An adaptive mesh refinement approach for average current nodal expansion method in 2-D rectangular geometry

    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

  3. Sorption kinetics of superabsorbent polymers (SAPs) in fresh Portland cement-based pastes visualized and quantified by neutron radiography and correlated to the progress of cement hydration

    Energy Technology Data Exchange (ETDEWEB)

    Schroefl, Christof, E-mail: christof.schroefl@tu-dresden.de [Technische Universität Dresden, Fakultät Bauingenieurwesen, Institut für Baustoffe, DE-01062 Dresden (Germany); Mechtcherine, Viktor [Technische Universität Dresden, Fakultät Bauingenieurwesen, Institut für Baustoffe, DE-01062 Dresden (Germany); Vontobel, Peter; Hovind, Jan; Lehmann, Eberhard [Paul Scherrer Institut, Laboratory for Neutron Scattering and Imaging, CH-5232 Villigen/AG (Switzerland)

    2015-09-15

    Water sorption of two superabsorbent polymers in cement-based pastes has been characterized by neutron radiography. Cement pastes with W/C of 0.25 and 0.50 and one additionally containing silica fume (W/C = 0.42) were investigated. The SAPs differed in their inherent sorption kinetics in extracted cement pore solution (SAP 1: self-releasing; SAP 2: retentive). Desorption from SAP 1 started very early after paste preparation. Hence, its individual non-retentiveness governs its behavior only. SAP 2 released water into all matrices, but its kinetics were different. In the paste with the highest W/C, some moderate water release was recorded from the beginning. In the other two pastes, SAP 2 retained its stored liquid during the dormant period, i.e., up to the percolation threshold. Intense desorption then set in and continued throughout the acceleration period. These findings explain the pronouncedly higher efficiency of SAP 2 as internal curing admixture as compared to SAP 1.

  4. Sorption kinetics of superabsorbent polymers (SAPs) in fresh Portland cement-based pastes visualized and quantified by neutron radiography and correlated to the progress of cement hydration

    International Nuclear Information System (INIS)

    Schroefl, Christof; Mechtcherine, Viktor; Vontobel, Peter; Hovind, Jan; Lehmann, Eberhard

    2015-01-01

    Water sorption of two superabsorbent polymers in cement-based pastes has been characterized by neutron radiography. Cement pastes with W/C of 0.25 and 0.50 and one additionally containing silica fume (W/C = 0.42) were investigated. The SAPs differed in their inherent sorption kinetics in extracted cement pore solution (SAP 1: self-releasing; SAP 2: retentive). Desorption from SAP 1 started very early after paste preparation. Hence, its individual non-retentiveness governs its behavior only. SAP 2 released water into all matrices, but its kinetics were different. In the paste with the highest W/C, some moderate water release was recorded from the beginning. In the other two pastes, SAP 2 retained its stored liquid during the dormant period, i.e., up to the percolation threshold. Intense desorption then set in and continued throughout the acceleration period. These findings explain the pronouncedly higher efficiency of SAP 2 as internal curing admixture as compared to SAP 1

  5. Hybrid nodal methods in the solution of the diffusion equations in X Y geometry

    International Nuclear Information System (INIS)

    Hernandez M, N.; Alonso V, G.; Valle G, E. del

    2003-01-01

    In 1979, Hennart and collaborators applied several schemes of classic finite element in the numerical solution of the diffusion equations in X Y geometry and stationary state. Almost two decades then, in 1996, himself and other collaborators carried out a similar work but using nodal schemes type finite element. Continuing in this last direction, in this work a group it is described a set of several Hybrid Nodal schemes denominated (NH) as well as their application to solve the diffusion equations in multigroup in stationary state and X Y geometry. The term hybrid nodal it means that such schemes interpolate not only Legendre moments of face and of cell but also the values of the scalar flow of neutrons in the four corners of each cell or element of the spatial discretization of the domain of interest. All the schemes here considered are polynomials like they were it their predecessors. Particularly, its have developed and applied eight different hybrid nodal schemes that its are very nearby related with those developed by Hennart and collaborators in the past. It is treated of schemes in those that nevertheless that decreases the number of interpolation parameters it is conserved the accurate in relation to the bi-quadratic and bi-cubic schemes. Of these eight, three were described and applied in a previous work. It is the bi-lineal classic scheme as well as the hybrid nodal schemes, bi-quadratic and bi-cubic for that here only are described the other 5 hybrid nodal schemes although they are provided numerical results for several test problems with all them. (Author)

  6. Current trends in methods for neutron diffusion calculations

    International Nuclear Information System (INIS)

    Adams, C.H.

    1977-01-01

    Current work and trends in the application of neutron diffusion theory to reactor design and analysis are reviewed. Specific topics covered include finite-difference methods, synthesis methods, nodal calculations, finite-elements and perturbation theory

  7. Nodal Structure of the Electronic Wigner Function

    DEFF Research Database (Denmark)

    Schmider, Hartmut; Dahl, Jens Peder

    1996-01-01

    On the example of several atomic and small molecular systems, the regular behavior of nodal patterns in the electronic one-particle reduced Wigner function is demonstrated. An expression found earlier relates the nodal pattern solely to the dot-product of the position and the momentum vector......, if both arguments are large. An argument analogous to the ``bond-oscillatory principle'' for momentum densities links the nuclear framework in a molecule to an additional oscillatory term in momenta parallel to bonds. It is shown that these are visible in the Wigner function in terms of characteristic...

  8. A fast nodal neutron diffusion method for cartesian geometry

    International Nuclear Information System (INIS)

    Makai, M.; Maeder, C.

    1983-01-01

    A numerical method based on an analytical solution to the three-dimensional two-group diffusion equation has been derived assuming that the flux is a sum of the functions of one variable. In each mesh the incoming currents are used as boundary conditions. The final equations for the average flux and the outgoing currents are of the response matrix type. The method is presented in a form that can be extended to the general multigroup case. In the SEXI computer program developed on the basis of this method, the response matrix elements are recalculated in each outer iteration to minimize the data transfer between disk storage and central memory. The efficiency of the method is demonstrated for a light water reactor (LWR) benchmark problem. The SEXI program has been incorporated into the LWR simulator SILWER code as a possible option

  9. Reflected‑Point‑Reactor Kinetics Model for Neutron Coincidence Counting: Comments on the Equation for the Leakage Self‑Multiplication

    International Nuclear Information System (INIS)

    Croft, S.; McElroy, RD.; Favalli, A.; Hauck, D.; Henzlova, D.; Henzl, V.; Santi, PA.

    2015-01-01

    Passive neutron correlation counting is widely used, for example by international inspection agencies, for the non‑destructive assay of spontaneously fissile nuclear materials for nuclear safeguards. The mass of special nuclear material present in an item is usually estimated from the observed neutron counting rates by using equations based on mathematically describing the object as an isolated multiplying point‑like source. Calibration using representative physical standards can often adequately compensate for this theoretical oversimplification through the introduction and use of effective‑interpretational‑model‑parameters meaning that useful assay results are obtained. In this work we extend the point‑model treatment by including a simple reflector around the fissioning material. Specifically we show how the leakage self‑multiplication equation mathematically connects the traditional bare source and the reflected source cases. In doing so we explicitly demonstrate that although the presence of a simple reflector changes the leakage self‑multiplication the traditional bare‑item point model multiplicity equations retain the same mathematical form. Making and explaining this connection is important because it helps to explain and justify the practical success and use of the traditional point‑model equations even when the assumptions used to generate the key functional dependences are violated. We are not aware that this point has been recognized previously.

  10. A methodology for the coupling of RAMONA-3B neutron kinetics and TRAC-BF1 thermal-hydraulics

    International Nuclear Information System (INIS)

    Lopez, Arsenio Procopio; Morales Sandoval, Jaime B.

    2005-01-01

    The initial objective of this project was to directly couple the RAMONA and TRAC codes running on different PCs. The idea was to use the best part of each one and eliminate some of their limitations and widen the applicability of these codes to simulate different BWR and system components. It was required to try to minimize the amount of changes to present code subroutines and calculation procedures. If possible, just substitute values obtained in the parallel code. Preliminary results indicated that using a CHAN component of TRAC to model thermal-hydraulic phenomena for each neutronic channel modeled in RAMONA is rather difficult. Large amounts of CPU time consumption are obtained and lots of PCs would make this solution difficult, besides considerable large transients are introduced by the differences in thermal-hydraulic results of these codes. The substitution of the thermal-hydraulics of RAMONA, by the TRAC channel calculations, is possible but simulation of a null transient on both codes must be planed and a gradual change must be controlled by an additional supervisory subroutine. An indirect coupling of these codes, it is therefore proposed, in order to eliminate most of these limitations. In this indirect coupling, a thermal-hydraulic model of the average tube in a bundle and the global channel cooling fluid dynamics is programmed for each neutronic channel while the global reactor vessel and core is modeled by TRAC with just four channels and four rings. Results are more reliable, coupling is simpler and faster simulations are possible

  11. Isospectral graphs with identical nodal counts

    International Nuclear Information System (INIS)

    Oren, Idan; Band, Ram

    2012-01-01

    According to a recent conjecture, isospectral objects have different nodal count sequences (Gnutzmann et al 2005 J. Phys. A: Math. Gen. 38 8921–33). We study generalized Laplacians on discrete graphs, and use them to construct the first non-trivial counterexamples to this conjecture. In addition, these examples demonstrate a surprising connection between isospectral discrete and quantum graphs. (paper)

  12. Slip-band formation and dislocation kinetics in the stage I deformation of neutron-irradiated copper single crystals

    International Nuclear Information System (INIS)

    Kitajima, Sadakichi; Shinohara, Kazutoshi; Kutsuwada, Masanori

    1995-01-01

    The velocity of edge and screw dislocations moving in primary slip bands and the formation rate of primary slip bands were measured in stage I deformation of neutron-irradiated copper single crystals at different strain rates at room temperature using micro-cinematography and optical micrography. The average velocity of edge dislocations was larger at least by one order than that of screw ones, and that of screw dislocations did not depend so strongly on strain rate. The formation rate of primary slip bands was proportional to strain rate. From these results, it is concluded that (1) jogs produced on moving dislocations by cutting dislocation loops result in the difference in velocity between edge and screw dislocations and (2) the change in the density of mobile dislocations as well as velocity of dislocations is responsible for the change of plastic strain rate of a crystal. (author)

  13. A variable-order time-dependent neutron transport method for nuclear reactor kinetics using analytically-integrated space-time characteristics

    International Nuclear Information System (INIS)

    Hoffman, A. J.; Lee, J. C.

    2013-01-01

    A new time-dependent neutron transport method based on the method of characteristics (MOC) has been developed. Whereas most spatial kinetics methods treat time dependence through temporal discretization, this new method treats time dependence by defining the characteristics to span space and time. In this implementation regions are defined in space-time where the thickness of the region in time fulfills an analogous role to the time step in discretized methods. The time dependence of the local source is approximated using a truncated Taylor series expansion with high order derivatives approximated using backward differences, permitting the solution of the resulting space-time characteristic equation. To avoid a drastic increase in computational expense and memory requirements due to solving many discrete characteristics in the space-time planes, the temporal variation of the boundary source is similarly approximated. This allows the characteristics in the space-time plane to be represented analytically rather than discretely, resulting in an algorithm comparable in implementation and expense to one that arises from conventional time integration techniques. Furthermore, by defining the boundary flux time derivative in terms of the preceding local source time derivative and boundary flux time derivative, the need to store angularly-dependent data is avoided without approximating the angular dependence of the angular flux time derivative. The accuracy of this method is assessed through implementation in the neutron transport code DeCART. The method is employed with variable-order local source representation to model a TWIGL transient. The results demonstrate that this method is accurate and more efficient than the discretized method. (authors)

  14. Impact of cross-section generation procedures on the simulation of the VVER 1000 pump startup experiment in the OECD/DOE/CEA V1000CT benchmark by coupled 3-D thermal hydraulics/ neutron kinetics models

    International Nuclear Information System (INIS)

    Boyan D Ivanov; Kostadin N Ivanov; Sylvie Aniel; Eric Royer

    2005-01-01

    Full text of publication follows: In the framework of joint effort between the Nuclear Energy Agency (NEA) of OECD, the United States Department of Energy (US DOE), and the Commissariat a l'Energie Atomique (CEA), France a coupled 3-D thermal hydraulics/neutron kinetics benchmark was defined. The overall objective OECD/NEA V1000CT benchmark is to assess computer codes used in analysis of VVER-1000 reactivity transients where mixing phenomena (mass flow and temperature) in the reactor pressure vessel are complex. Original data from the Kozloduy-6 Nuclear Power Plant are available for the validation of computer codes: one experiment of pump start-up (V1000CT-1) and one experiment of steam generator isolation (V1000CT-2). Additional scenarios are defined for code-to-code comparison. As a 3D core model is necessary for a best-estimate computation of all the scenarios of the V1000CT benchmark, all participants were asked to develop their own core coupled 3-D thermal hydraulics/ neutron kinetics models based on the data available in the benchmark specifications. The first code to code comparisons based on the V1000CT-1 Exercise 2 specifications exhibited unacceptable discrepancies between 2 sets of results, one of them being close to experimental results. The present paper focuses first on the analysis of the observed discrepancies. The VVER 1000 3-D thermal hydraulics/neutron kinetics models are based on thermal-hydraulic and neutronic data homogenized at the assembly scale. The neutronic data, provided as part of the benchmark specifications, consist thus in a set of parametrized 2 group cross sections libraries representing the different assemblies and the reflectors. The origin of the high observed discrepancies was found to lie in the use of these neutronic libraries. The concern was then to find a way to provide neutronic data, compatible with all the benchmark participants neutronic models, that enable also comparisons with experimental results. An analysis of the

  15. Distribution of nuclear charge and angular momentum in chains 132-137, 99, and 102 of thermal neutron fission of 235U at various kinetic energies and charge states of the fragments

    International Nuclear Information System (INIS)

    Denschlag, H.O.; Braun, H.; Wolfsberg, K.

    1979-01-01

    The fission product yields of the members of the decay chains 132 to 137, 99, and 102 in 235 U(n/sub th/,f) were measured at various kinetic energies and ionic charge states of the fragments using the mass separator for unslowed fission products LOHENGRIN. The results are discussed with respect to four aspects: A preferential formation of neutron rich chain members found at high kinetic energy of the fragments is predominantly due to decreasing prompt neutron evaporation. A particularly large effect in chain 132 is attributed to the double shell closure in Sn-132. The persistence of an even-odd pairing effect in the yields throughout the range of kinetic energies studied leads to the conclusion that the high internal excitation energy of the fragments is tied up mainly in the form of collective energy (e.g., deformation energy) rather than single particle excitation. Generally, the yield distribution at constant kinetic energy is invariant with respect to the ionic charge state of the isotopes separated. Deviations from this behavior found in chains 99, 102, 133, and 136 are interpreted as being due to Auger events following a converted transition in the decay of ns-isomers taking place in the vacuum of the separator. A pronounced variation of the independent formation ratio of single isomeric states with the kinetic energy of the fragments is providing direct information on the controversial topic of the change of angular momentum of fission fragments as a function of deformation (scission distance). 34 references

  16. Implementation of the kinetics in the transport code AZTRAN

    International Nuclear Information System (INIS)

    Duran G, J. A.; Del Valle G, E.; Gomez T, A. M.

    2017-09-01

    This paper shows the implementation of the time dependence in the three-dimensional transport code AZTRAN (AZtlan TRANsport), which belongs to the AZTLAN platform, for the analysis of nuclear reactors (currently under development). The AZTRAN code with this implementation is able to numerically solve the time-dependent transport equation in XYZ geometry, for several energy groups, using the discrete ordinate method S n for the discretization of the angular variable, the nodal method RTN-0 for spatial discretization and method 0 for discretization in time. Initially, the code only solved the neutrons transport equation in steady state, so the implementation of the temporal part was made integrating the neutrons transport equation with respect to time and balance equations corresponding to the concentrations of delayed neutron precursors, for which method 0 was applied. After having directly implemented code kinetics, the improved quasi-static method was implemented, which is a tool for reducing computation time, where the angular flow is factored by the product of two functions called shape function and amplitude function, where the first is calculated for long time steps, called macro-steps and the second is resolved for small time steps called micro-steps. In the new version of AZTRAN several Benchmark problems that were taken from the literature were simulated, the problems used are of two and three dimensions which allowed corroborating the accuracy and stability of the code, showing in general in the reference tests a good behavior. (Author)

  17. The exponential function expansion of the intra-nodal cross sections for the spectral history gradient correction

    International Nuclear Information System (INIS)

    Cho, J. Y.; Noh, J. M.; Cheong, H. K.; Choo, H. K.

    1998-01-01

    In order to simplify the previous spectral history effect correction based on the polynomial expansion nodal method, a new spectral history effect correction is proposed. The new spectral history correction eliminates four microscopic depletion points out of total 13 depletion points in the previous correction by approximating the group cross sections with exponential function. The neutron flux to homogenize the group cross sections for the correction of the spectral history effect is calculated by the analytic function expansion nodal method in stead of the conventional polynomial expansion nodal method. This spectral history correction model is verified against the three MOX benchmark cores: a checkerboard type, a small core with 25 fuel assemblies, and a large core with 177 fuel assemblies. The benchmark results prove that this new spectral history correction model is superior to the previous one even with the reduced number of the local microscopic depletion points

  18. Uncertainty and Sensitivity of Neutron Kinetic Parameters in the Dynamic Response of a PWR Rod Ejection Accident Coupled Simulation

    Directory of Open Access Journals (Sweden)

    C. Mesado

    2012-01-01

    Full Text Available In nuclear safety analysis, it is very important to be able to simulate the different transients that can occur in a nuclear power plant with a very high accuracy. Although the best estimate codes can simulate the transients and provide realistic system responses, the use of nonexact models, together with assumptions and estimations, is a source of uncertainties which must be properly evaluated. This paper describes a Rod Ejection Accident (REA simulated using the coupled code RELAP5/PARCSv2.7 with a perturbation on the cross-sectional sets in order to determine the uncertainties in the macroscopic neutronic information. The procedure to perform the uncertainty and sensitivity (U&S analysis is a sampling-based method which is easy to implement and allows different procedures for the sensitivity analyses despite its high computational time. DAKOTA-Jaguar software package is the selected toolkit for the U&S analysis presented in this paper. The size of the sampling is determined by applying the Wilks’ formula for double tolerance limits with a 95% of uncertainty and with 95% of statistical confidence for the output variables. Each sample has a corresponding set of perturbations that will modify the cross-sectional sets used by PARCS. Finally, the intervals of tolerance of the output variables will be obtained by the use of nonparametric statistical methods.

  19. Statistical safety evaluation of BWR turbine trip scenario using coupled neutron kinetics and thermal hydraulics analysis code SKETCH-INS/TRACE5.0

    International Nuclear Information System (INIS)

    Ichikawa, Ryoko; Masuhara, Yasuhiro; Kasahara, Fumio

    2012-01-01

    The Best Estimate Plus Uncertainty (BEPU) method has been prepared for the regulatory cross-check analysis at Japan Nuclear Energy Safety Organization (JNES) on base of the three-dimensional neutron-kinetics/thermal-hydraulics coupled code SKETCH-INS/TRACE5.0. In the preparation, TRACE5.0 is verified against the large-scale thermal-hydraulic tests carried out with NUPEC facility. These tests were focused on the pressure drop of steam-liquid two phase flow and void fraction distribution. From the comparison of the experimental data with other codes (RELAP5/MOD3.3 and TRAC-BF1), TRACE5.0 was judged better than other codes. It was confirmed that TRACE5.0 has high reliability for thermal hydraulics behavior and are used as a best-estimate code for the statistical safety evaluation. Next, the coupled code SKETCH-INS/TRACE5.0 was applied to turbine trip tests performed at the Peach Bottom-2 BWR4 Plant. The turbine trip event shows the rapid power peak due to the voids collapse with the pressure increase. The analyzed peak value of core power is better simulated than the previous version SKETCH-INS/TRAC-BF1. And the statistical safety evaluation using SKETCH-INS/TRACE5.0 was applied to the loss of load transient for examining the influence of the choice of sampling method. (author)

  20. RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - user`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

    1998-03-01

    This document is the User`s Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code`s capabilities and limitations; Chapter 2 describes the code`s structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARC and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs.

  1. Coupling of the neutron-kinetic core model DYN3D with the thermal hydraulic code FLICA-4 within the NURESIM platform

    International Nuclear Information System (INIS)

    Gommlich, A.; Kliem, S.; Rohde, U.; Gomez, A.; Sanchez, V.

    2010-01-01

    Within the FP7 Collaborative Project NURISP (NUclear Reactor Integrated Simulation Project) new and significant steps will be done towards a European Reference Simulation Platform for applications relevant to present PWR and BWR and to future reactors. The first step towards this target has been made during the FP6 NURESIM Integrated Project, where the already common and well-proven NURESIM informatics platform has been developed. This platform is based on the open source software SALOME. The 3D neutron kinetic core model DYN3D developed at Forschungszentrum Dresden-Rossendorf is part of the NURESIM platform. Within the NURESIM project, a SALOME based pre-processor for creation of DYN3D input data sets via GUI has been developed. DYN3D has been implemented into SALOME as black box, which allowed an independent execution. A conversion of the DYN3D result file into SALOME format was developed which opened the possibility using SALOME tools to visualize DYN3D results. (orig.)

  2. Twin-domain size and bulk oxygen in-diffusion kinetics of YBa 2Cu 3O 6+x studied by neutron powder diffraction and gas volumetry

    Science.gov (United States)

    Poulsen, H. F.; Andersen, N. H.; Lebech, B.

    1991-02-01

    We report experimental results of twin-domain size and bulk oxygen in-diffusion kinetics of YBa 2Cu 3O 6+ x, which supplement a previous and simultaneous study of the structural phase diagram and oxygen equilibrium partial pressure. Analysis of neutron powder diffraction peak broadening show features which are identified to result from temperature independent twin-domain formation in to different orthorhombic phases with domain sizes and 250 and 350Å, respectively. The oxygen in-diffusion flow shows simple relaxation type behaviour J=J 0 exp( {-t}/{τ}) despite a rather broad particle size distribution. At higher temperatures, τ is activated with activation energies 0.55 and 0.25 eV in the tetragonal and orthorhombic phases, respectively. Comparison between twin-domain sizes and bulk oxygen in-diffusion time constants indicates that the twin-domain boundaries may contribute to the effective bulk oxygen in-diffusion. All our results may be interpreted in terms of the 2D ASYNNNI model description of the oxygen basal plane ordering, and they suggest that recent first principles interaction parameters should be modified.

  3. Studies on fission with ALADIN. Precise and simultaneous measurement of fission yields, total kinetic energy and total prompt neutron multiplicity at GSI

    International Nuclear Information System (INIS)

    Martin, Julie-Fiona; Taieb, Julien; Chatillon, Audrey; Belier, Gilbert; Boutoux, Guillaume; Ebran, Adeline; Gorbinet, Thomas; Grente, Lucie; Laurent, Benoit; Pellereau, Eric; Alvarez-Pol, Hector; Ayyad, Yassid; Benlliure, Jose; Cortina Gil, Dolores; Caamano, Manuel; Fernandez Dominguez, Beatriz; Paradela, Carlos; Ramos, Diego; Rodriguez-Sanchez, Jose-Luis; Vargas, Jossitt; Audouin, Laurent; Tassan-Got, Laurent; Aumann, Thomas; Casarejos, Enrique; Farget, Fanny; Rodriguez-Tajes, Carme; Heinz, Andreas; Jurado, Beatriz; Kelic-Heil, Aleksandra; Kurz, Nikolaus; Nociforo, Chiara; Pietri, Stephane; Rossi, Dominic; Schmidt, Karl-Heinz; Simon, Haik; Voss, Bernd; Weick, Helmut

    2015-01-01

    A novel technique for fission studies, based on the inverse kinematics approach, is presented. Following pioneering work in the nineties, the SOFIA Collaboration has designed and built an experimental set-up dedicated to the simultaneous measurement of isotopic yields, total kinetic energies and total prompt neutron multiplicities, by fully identifying both fission fragments in coincidence, for the very first time. This experiment, performed at GSI, permits to study the fission of a wide variety of fissioning systems, ranging from mercury to neptunium, possibly far from the valley of stability. A first experiment, performed in 2012, has provided a large array of unprecedented data regarding the nuclear fission process. An excerpt of the results is presented. With this solid starter, further improvements of the experimental set-up are considered, which are consistent with the expected developments at the GSI facility, in order to measure more fission observables in coincidence. The completeness reached in the SOFIA data, permits to scrutinize the correlations between the interesting features of fission, offering a very detailed insight in this still unraveled mechanism. (orig.)

  4. Studies on fission with ALADIN. Precise and simultaneous measurement of fission yields, total kinetic energy and total prompt neutron multiplicity at GSI

    Energy Technology Data Exchange (ETDEWEB)

    Martin, Julie-Fiona; Taieb, Julien; Chatillon, Audrey; Belier, Gilbert; Boutoux, Guillaume; Ebran, Adeline; Gorbinet, Thomas; Grente, Lucie; Laurent, Benoit; Pellereau, Eric [CEA DAM Bruyeres-le-Chatel, Arpajon (France); Alvarez-Pol, Hector; Ayyad, Yassid; Benlliure, Jose; Cortina Gil, Dolores; Caamano, Manuel; Fernandez Dominguez, Beatriz; Paradela, Carlos; Ramos, Diego; Rodriguez-Sanchez, Jose-Luis; Vargas, Jossitt [Universidad de Santiago de Compostela, Santiago de Compostela (Spain); Audouin, Laurent; Tassan-Got, Laurent [CNRS/IN2P3, IPNO, Orsay (France); Aumann, Thomas [Technische Universitaet Darmstadt, Darmstadt (Germany); Casarejos, Enrique [Universidad de Vigo, Vigo (Spain); Farget, Fanny; Rodriguez-Tajes, Carme [CNRS/IN2P3, GANIL, Caen (France); Heinz, Andreas [Chalmers University of Technology, Gothenburg (Sweden); Jurado, Beatriz [CNRS/IN2P3, CENBG, Gradignan (France); Kelic-Heil, Aleksandra; Kurz, Nikolaus; Nociforo, Chiara; Pietri, Stephane; Rossi, Dominic; Schmidt, Karl-Heinz; Simon, Haik; Voss, Bernd; Weick, Helmut [GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, Darmstadt (Germany)

    2015-12-15

    A novel technique for fission studies, based on the inverse kinematics approach, is presented. Following pioneering work in the nineties, the SOFIA Collaboration has designed and built an experimental set-up dedicated to the simultaneous measurement of isotopic yields, total kinetic energies and total prompt neutron multiplicities, by fully identifying both fission fragments in coincidence, for the very first time. This experiment, performed at GSI, permits to study the fission of a wide variety of fissioning systems, ranging from mercury to neptunium, possibly far from the valley of stability. A first experiment, performed in 2012, has provided a large array of unprecedented data regarding the nuclear fission process. An excerpt of the results is presented. With this solid starter, further improvements of the experimental set-up are considered, which are consistent with the expected developments at the GSI facility, in order to measure more fission observables in coincidence. The completeness reached in the SOFIA data, permits to scrutinize the correlations between the interesting features of fission, offering a very detailed insight in this still unraveled mechanism. (orig.)

  5. Pathology of nodal marginal zone lymphomas.

    Science.gov (United States)

    Pileri, Stefano; Ponzoni, Maurilio

    Nodal marginal zone B cell lymphomas (NMZLs) are a rare group of lymphoid disorders part of the spectrum of marginal zone B-cell lymphomas, which encompass splenic marginal one B-cell lymphoma (SMZL) and extra nodal marginal zone of B-cell lymphoma (EMZL), often of MALT-type. Two clinicopathological forms of NMZL are recognized: adult-type and pediatric-type, respectively. NMZLs show overlapping features with other types of MZ, but distinctive features as well. In this review, we will focus on the salient distinguishing features of NMZL mostly under morphological/immunophenotypical/molecular perspectives in views of the recent acquisitions and forthcoming updated 2016 WHO classification of lymphoid malignancies. Copyright © 2016 Elsevier Ltd. All rights reserved.

  6. Quantum anomalies in nodal line semimetals

    Science.gov (United States)

    Burkov, A. A.

    2018-04-01

    Topological semimetals are a new class of condensed matter systems with nontrivial electronic structure topology. Their unusual observable properties may often be understood in terms of quantum anomalies. In particular, Weyl and Dirac semimetals, which have point band-touching nodes, are characterized by the chiral anomaly, which leads to the Fermi arc surface states, anomalous Hall effect, negative longitudinal magnetoresistance, and planar Hall effect. In this paper, we explore analogous phenomena in nodal line semimetals. We demonstrate that such semimetals realize a three-dimensional analog of the parity anomaly, which is a known property of two-dimensional Dirac semimetals arising, for example, on the surface of a three-dimensional topological insulator. We relate one of the characteristic properties of nodal line semimetals, namely, the drumhead surface states, to this anomaly, and derive the field theory, which encodes the corresponding anomalous response.

  7. Temporal quadratic expansion nodal Green's function method

    International Nuclear Information System (INIS)

    Liu Cong; Jing Xingqing; Xu Xiaolin

    2000-01-01

    A new approach is presented to efficiently solve the three-dimensional space-time reactor dynamics equation which overcomes the disadvantages of current methods. In the Temporal Quadratic Expansion Nodal Green's Function Method (TQE/NGFM), the Quadratic Expansion Method (QEM) is used for the temporal solution with the Nodal Green's Function Method (NGFM) employed for the spatial solution. Test calculational results using TQE/NGFM show that its time step size can be 5-20 times larger than that of the Fully Implicit Method (FIM) for similar precision. Additionally, the spatial mesh size with NGFM can be nearly 20 times larger than that using the finite difference method. So, TQE/NGFM is proved to be an efficient reactor dynamics analysis method

  8. Acceleration of the FERM nodal program

    International Nuclear Information System (INIS)

    Nakata, H.

    1985-01-01

    It was tested three acceleration methods trying to reduce the number of outer iterations in the FERM nodal program. The results obtained indicated that the Chebychev polynomial acceleration method with variable degree results in a economy of 50% in the computer time. Otherwise, the acceleration method by source asymptotic extrapolation or by zonal rebalance did not result in economy of the global computer time, however some acceleration had been verified in outer iterations. (M.C.K.) [pt

  9. Acceleration of the nodal program FERM

    International Nuclear Information System (INIS)

    Nakata, H.

    1985-01-01

    Acceleration of the nodal FERM was tried by three acceleration schemes. Results of the calculations showed the best acceleration with the Tchebyshev method where the savings in the computing time were of the order of 50%. Acceleration with the Assymptotic Source Extrapoltation Method and with the Coarse-Mesh Rebalancing Method did not result in any improvement on the global computational time, although a reduction in the number of outer iterations was observed. (Author) [pt

  10. Nodal method for fast reactor analysis

    International Nuclear Information System (INIS)

    Shober, R.A.

    1979-01-01

    In this paper, a nodal method applicable to fast reactor diffusion theory analysis has been developed. This method has been shown to be accurate and efficient in comparison to highly optimized finite difference techniques. The use of an analytic solution to the diffusion equation as a means of determining accurate coupling relationships between nodes has been shown to be highly accurate and efficient in specific two-group applications, as well as in the current multigroup method

  11. Application of a nodal collocation approximation for the multidimensional PL equations to the 3D Takeda benchmark problems

    International Nuclear Information System (INIS)

    Capilla, M.; Talavera, C.F.; Ginestar, D.; Verdú, G.

    2012-01-01

    Highlights: ► The multidimensional P L approximation to the nuclear transport equation is reviewed. ► A nodal collocation method is developed for the spatial discretization of P L equations. ► Advantages of the method are lower dimension and good characterists of the associated algebraic eigenvalue problem. ► The P L nodal collocation method is implemented into the computer code SHNC. ► The SHNC code is verified with 2D and 3D benchmark eigenvalue problems from Takeda and Ikeda, giving satisfactory results. - Abstract: P L equations are classical approximations to the neutron transport equations, which are obtained expanding the angular neutron flux in terms of spherical harmonics. These approximations are useful to study the behavior of reactor cores with complex fuel assemblies, for the homogenization of nuclear cross-sections, etc., and most of these applications are in three-dimensional (3D) geometries. In this work, we review the multi-dimensional P L equations and describe a nodal collocation method for the spatial discretization of these equations for arbitrary odd order L, which is based on the expansion of the spatial dependence of the fields in terms of orthonormal Legendre polynomials. The performance of the nodal collocation method is studied by means of obtaining the k eff and the stationary power distribution of several 3D benchmark problems. The solutions are obtained are compared with a finite element method and a Monte Carlo method.

  12. New Multi-group Transport Neutronics (PHISICS) Capabilities for RELAP5-3D and its Application to Phase I of the OECD/NEA MHTGR-350 MW Benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Gerhard Strydom; Cristian Rabiti; Andrea Alfonsi

    2012-10-01

    PHISICS is a neutronics code system currently under development at the Idaho National Laboratory (INL). Its goal is to provide state of the art simulation capability to reactor designers. The different modules for PHISICS currently under development are a nodal and semi-structured transport core solver (INSTANT), a depletion module (MRTAU) and a cross section interpolation (MIXER) module. The INSTANT module is the most developed of the mentioned above. Basic functionalities are ready to use, but the code is still in continuous development to extend its capabilities. This paper reports on the effort of coupling the nodal kinetics code package PHISICS (INSTANT/MRTAU/MIXER) to the thermal hydraulics system code RELAP5-3D, to enable full core and system modeling. This will enable the possibility to model coupled (thermal-hydraulics and neutronics) problems with more options for 3D neutron kinetics, compared to the existing diffusion theory neutron kinetics module in RELAP5-3D (NESTLE). In the second part of the paper, an overview of the OECD/NEA MHTGR-350 MW benchmark is given. This benchmark has been approved by the OECD, and is based on the General Atomics 350 MW Modular High Temperature Gas Reactor (MHTGR) design. The benchmark includes coupled neutronics thermal hydraulics exercises that require more capabilities than RELAP5-3D with NESTLE offers. Therefore, the MHTGR benchmark makes extensive use of the new PHISICS/RELAP5-3D coupling capabilities. The paper presents the preliminary results of the three steady state exercises specified in Phase I of the benchmark using PHISICS/RELAP5-3D.

  13. Point-by-point model description of average prompt neutron data as a function of total kinetic energy of fission fragments

    International Nuclear Information System (INIS)

    Tudora, A.

    2013-01-01

    The experimental data of average prompt neutron multiplicity as a function of total kinetic energy of fragments <ν>(TKE) exhibit, especially in the case of 252 Cf(SF), different slopes dTKE/dν and different behaviours at low TKE values. The Point-by-Point (PbP) model can describe these different behaviours. The higher slope dTKE/dν and the flattening of <ν> at low TKE exhibited by a part of experimental data sets is very well reproduced when the PbP multi-parametric matrix ν(A,TKE) is averaged over a double distribution Y(A,TKE). The lower slope and the almost linear behaviour over the entire TKE range exhibited by other data sets is well described when the same matrix ν(A,TKE) is averaged over a single distribution Y(A). In the case of average prompt neutron energy in SCM as a function of TKE, different dTKE/dε slopes are also obtained by averaging the same PbP matrix ε(A,TKE) over Y(A,TKE) and over Y(A). The results are exemplified for 3 fissioning systems benefiting of experimental data as a function of TKE: 252 Cf(SF), 235 U(n th ,f) and 239 Pu(n th ,f). In the case of 234 U(n,f) for the first time it was possible to calculate <ν>(TKE) and <ε>(TKE) at many incident energies by averaging the PbP multi-parametric matrices over the experimental Y(A,TKE) distributions recently measured at IRMM for 14 incident energies in the range 0.3- 5 MeV. The results revealed that the slope dTKE/dν does not vary with the incident energy and the flattening of <ν> at low TKE values is more pronounced at low incident energies. The average model parameters dependences on TKE resulted from the PbP treatment allow the use of the most probable fragmentation approach, having the great advantage to provide results at many TKE values in a very short computing time compared to PbP and Monte Carlo treatments. (author)

  14. Whole core pin-by-pin coupled neutronic-thermal-hydraulic steady state and transient calculations using COBAYA3 code

    International Nuclear Information System (INIS)

    Jimenez, J.; Herrero, J. J.; Cuervo, D.; Aragones, J. M.

    2010-10-01

    Nowadays coupled 3-dimensional neutron kinetics and thermal-hydraulic core calculations are performed by applying a radial average channel approach using a meshing of one quarter of assembly in the best case. This approach does not take into account the subchannels effects due to the averaging of the physical fields and the loose of heterogeneity in the thermal-hydraulic model. Therefore the models do not have enough resolution to predict those subchannels effects which are important for the fuel design safety margins, because it is in the local scale, where we can search the hottest pellet or the maximum heat flux. The Polytechnic University of Madrid advanced multi-scale neutron-kinetics and thermal-hydraulics methodologies being implemented in COBAYA3 include domain decomposition by alternate core dissections for the local 3-dimensional fine-mesh scale problems (pin cells/subchannels) and an analytical nodal diffusion solver for the coarse mesh scale coupled with the thermal-hydraulic using a model of one channel per assembly or per quarter of assembly. In this work, we address the domain decomposition by the alternate core dissections methodology applied to solve coupled 3-dimensional neutronic-thermal-hydraulic problems at the fine-mesh scale. The neutronic-thermal-hydraulic coupling at the cell-subchannel scale allows the treatment of the effects of the detailed thermal-hydraulic feedbacks on cross-sections, thus resulting in better estimates of the local safety margins at the pin level. (Author)

  15. Impacts of Contingency Reserve on Nodal Price and Nodal Reliability Risk in Deregulated Power Systems

    DEFF Research Database (Denmark)

    Zhao, Qian; Wang, Peng; Goel, Lalit

    2013-01-01

    The deregulation of power systems allows customers to participate in power market operation. In deregulated power systems, nodal price and nodal reliability are adopted to represent locational operation cost and reliability performance. Since contingency reserve (CR) plays an important role...... in reliable operation, the CR commitment should be considered in operational reliability analysis. In this paper, a CR model based on customer reliability requirements has been formulated and integrated into power market settlement. A two-step market clearing process has been proposed to determine generation...

  16. Validation of the coupled neutron kinetic thermohydraulic code ATHLET/DYN3D with help of measured data of the OECD Turbine Trip Benchmarks. Final report

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.

    2003-12-01

    The project consisted in the validation of the coupled neutron kinetic/thermal hydraulic code system ATHLET/DYN3D for boiling water reactors by the participation at the OECD/NRC turbine trip benchmark. The benchmark defined by the OECD and the American NRC is based on an experiment with closure of the turbine stop valve which was carried out in 1977 in the nuclear power plant Peach Bottom 2 within the framework of a series of 3 experiments. In the experiment, the closure of the valve caused a pressure wave which propagated with attenuation into the reactor core. The condensation of steam in the reactor core caused by the increase of pressure lead to a positive reactivity insertion. The following rise of power was limited by the feedback and the insertion of the control rods. In the frame of the benchmark, the codes could be validated by comparisons with the measured results and the result of the other participants. The benchmark was divided into 3 phases or exercises. Phase I was used for checking the thermo-hydraulic model of the system using a given power release in the core. In phase II, three-dimensional core calculations were performed for given thermal-hydraulic boundary conditions. Coupled calculations were carried out for the selected experiment and four extreme scenarios in the phase III. In the frame of the project, FZR took part in phases II and III of the benchmark. The calculations for phase II were performed with DYN3D by using the assembly discontinuity factors (ADF) and 764 thermal-hydraulic channels (1 channel/assembly). The ATHLET input data set for the coolant system was obtained form the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS). It was slightly modified for the phase III calculations carried out with the parallel coupling of ATHLET and DYN3D. For spatially averaged parameters, a good agreement with the results of measurement and the results of other codes was achieved. The influence of the different models was investigated with the

  17. A nodal collocation method for the calculation of the lambda modes of the P L equations

    International Nuclear Information System (INIS)

    Capilla, M.; Talavera, C.F.; Ginestar, D.; Verdu, G.

    2005-01-01

    P L equations are classical approximations to the neutron transport equation admitting a diffusive form. Using this property, a nodal collocation method is developed for the P L approximations, which is based on the expansion of the flux in terms of orthonormal Legendre polynomials. This method approximates the differential lambda modes problem by an algebraic eigenvalue problem from which the fundamental and the subcritical modes of the system can be calculated. To test the performance of this method, two problems have been considered, a homogeneous slab, which admits an analytical solution, and a seven-region slab corresponding to a more realistic problem

  18. Development of a polynomial nodal model to the multigroup transport equation in one dimension

    International Nuclear Information System (INIS)

    Feiz, M.

    1986-01-01

    A polynomial nodal model that uses Legendre polynomial expansions was developed for the multigroup transport equation in one dimension. The development depends upon the least-squares minimization of the residuals using the approximate functions over the node. Analytical expressions were developed for the polynomial coefficients. The odd moments of the angular neutron flux over the half ranges were used at the internal interfaces, and the Marshak boundary condition was used at the external boundaries. Sample problems with fine-mesh finite-difference solutions of the diffusion and transport equations were used for comparison with the model

  19. Topological surface states in nodal superconductors.

    Science.gov (United States)

    Schnyder, Andreas P; Brydon, Philip M R

    2015-06-24

    Topological superconductors have become a subject of intense research due to their potential use for technical applications in device fabrication and quantum information. Besides fully gapped superconductors, unconventional superconductors with point or line nodes in their order parameter can also exhibit nontrivial topological characteristics. This article reviews recent progress in the theoretical understanding of nodal topological superconductors, with a focus on Weyl and noncentrosymmetric superconductors and their protected surface states. Using selected examples, we review the bulk topological properties of these systems, study different types of topological surface states, and examine their unusual properties. Furthermore, we survey some candidate materials for topological superconductivity and discuss different experimental signatures of topological surface states.

  20. Twisted vector bundles on pointed nodal curves

    Indian Academy of Sciences (India)

    R. Narasimhan (Krishtel eMaging) 1461 1996 Oct 15 13:05:22

    by identifying the points p1 and p2. If m ≥ 2, let R1,...,Rm−1 be m − 1 copies of the projective line P1 and let xi,yi be two distinct points in Ri. Let R be the nodal curve which arises from the union. R0 ⊔ R1 ⊔···⊔ Rm−1 ⊔ Rm by identifying p1 ∈ R0 and p2 ∈ Rm with x1 ∈ R1 and ym−1 ∈ Rm−1 respectively and by identifying ...

  1. Analytical representation for solution of the neutron point kinetics equation with time-dependent reactivity and free of the stiffness character; Representacao analitica da solucao da equacao de cinetica pontual para a reatividade variavel no tempo livre de rigidez

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Milena Wollmann da

    2013-08-01

    In this work, we report a genuine analytical representation for the solution of the neutron point kinetics equation free of the stiffness character, assuming that the reactivity is a continuous and sectionally continuous function of time. To this end, we initially cast the point kinetics equation in a first order linear differential equation. Next, we split the corresponding matrix as a sum of a diagonal matrix with a matrix, whose components contain the off-diagonal elements. Next, expanding the neutron density and the delayed neutron precursors concentrations in a truncated series, and replacing these expansions in the matrix equation, we come out with an equation, which allows to construct a recursive system, a first order matrix differential equation with source. The fundamental characteristic of this system relies on the fact that the corresponding matrix is diagonal, meanwhile the source term is written in terms of the matrix with the off-diagonal components. Further, the first equation of the recursive system has no source and satisfies the initial conditions. On the other hand, the remaining equations satisfy the null initial condition. Due to the diagonal feature of the matrix, we attain analytical solutions for these recursive equations. We also mention that we evaluate the results for any time value, without the analytical continuity because the purposed solution is free on the stiffness character. Finally, we present numerical simulations and comparisons against literature results, considering specific the applications for the following reactivity functions: constant, step, ramp, and sine. (author)

  2. The analytic nodal diffusion solver ANDES in multigroups for 3D rectangular geometry: Development and performance analysis

    International Nuclear Information System (INIS)

    Lozano, Juan-Andres; Garcia-Herranz, Nuria; Ahnert, Carol; Aragones, Jose-Maria

    2008-01-01

    In this work we address the development and implementation of the analytic coarse-mesh finite-difference (ACMFD) method in a nodal neutron diffusion solver called ANDES. The first version of the solver is implemented in any number of neutron energy groups, and in 3D Cartesian geometries; thus it mainly addresses PWR and BWR core simulations. The details about the generalization to multigroups and 3D, as well as the implementation of the method are given. The transverse integration procedure is the scheme chosen to extend the ACMFD formulation to multidimensional problems. The role of the transverse leakage treatment in the accuracy of the nodal solutions is analyzed in detail: the involved assumptions, the limitations of the method in terms of nodal width, the alternative approaches to implement the transverse leakage terms in nodal methods - implicit or explicit -, and the error assessment due to transverse integration. A new approach for solving the control rod 'cusping' problem, based on the direct application of the ACMFD method, is also developed and implemented in ANDES. The solver architecture turns ANDES into an user-friendly, modular and easily linkable tool, as required to be integrated into common software platforms for multi-scale and multi-physics simulations. ANDES can be used either as a stand-alone nodal code or as a solver to accelerate the convergence of whole core pin-by-pin code systems. The verification and performance of the solver are demonstrated using both proof-of-principle test cases and well-referenced international benchmarks

  3. Robust doubly charged nodal lines and nodal surfaces in centrosymmetric systems

    Science.gov (United States)

    Bzdušek, Tomáš; Sigrist, Manfred

    2017-10-01

    Weyl points in three spatial dimensions are characterized by a Z -valued charge—the Chern number—which makes them stable against a wide range of perturbations. A set of Weyl points can mutually annihilate only if their net charge vanishes, a property we refer to as robustness. While nodal loops are usually not robust in this sense, it has recently been shown using homotopy arguments that in the centrosymmetric extension of the AI symmetry class they nevertheless develop a Z2 charge analogous to the Chern number. Nodal loops carrying a nontrivial value of this Z2 charge are robust, i.e., they can be gapped out only by a pairwise annihilation and not on their own. As this is an additional charge independent of the Berry π -phase flowing along the band degeneracy, such nodal loops are, in fact, doubly charged. In this manuscript, we generalize the homotopy discussion to the centrosymmetric extensions of all Atland-Zirnbauer classes. We develop a tailored mathematical framework dubbed the AZ +I classification and show that in three spatial dimensions such robust and multiply charged nodes appear in four of such centrosymmetric extensions, namely, AZ +I classes CI and AI lead to doubly charged nodal lines, while D and BDI support doubly charged nodal surfaces. We remark that no further crystalline symmetries apart from the spatial inversion are necessary for their stability. We provide a description of the corresponding topological charges, and develop simple tight-binding models of various semimetallic and superconducting phases that exhibit these nodes. We also indicate how the concept of robust and multiply charged nodes generalizes to other spatial dimensions.

  4. Radiotherapy of adult nodal non Hodgkin's lymphoma

    International Nuclear Information System (INIS)

    Gamen, G.; Thirion, P.

    1999-01-01

    The role of radiotherapy in the treatment of nodal non-Hodgkin's lymphoma has been modified by the introduction of efficient chemotherapy and the development of different pathological classifications. The recommended treatment of early-stage aggressive lymphomas is primarily a combination chemotherapy. The interest of adjuvant radiotherapy remains unclear and has to be established through large prospective trials. If radiation therapy has to be delivered, the historical results of exclusive radiation therapy showed that involved-fields and a dose of 35-40 Gy (daily fraction of 1.8 Gy, 5 days a week) are the optimal schedule. The interest of radiotherapy in the treatment of advanced-stage aggressive lymphoma is yet to be proven. Further studies had to stratify localized stages according to the factors of the International Prognostic Index. For easy-stage low-grade lymphoma, radiotherapy remains the standard treatment. However, the appropriate technique to use is controversial. Involved-field irradiation at a dose of 35 Gy seems to be the optimal schedule, providing a 10 year disease-free survival rate of 50 % and no major toxicity. There is no standard indication of radiotherapy in the treatment advanced-stage low-grade lymphoma. For 'new' nodal lymphoma's types, the indication of radiotherapy cannot be established (mantle-zone lymphoma, marginal zone B-cell lymphoma) or must take into account the natural history (Burkitt's lymphoma, peripheral T-cell lymphoma) and the sensibility to others therapeutic methods. (authors)

  5. Error estimation for variational nodal calculations

    International Nuclear Information System (INIS)

    Zhang, H.; Lewis, E.E.

    1998-01-01

    Adaptive grid methods are widely employed in finite element solutions to both solid and fluid mechanics problems. Either the size of the element is reduced (h refinement) or the order of the trial function is increased (p refinement) locally to improve the accuracy of the solution without a commensurate increase in computational effort. Success of these methods requires effective local error estimates to determine those parts of the problem domain where the solution should be refined. Adaptive methods have recently been applied to the spatial variables of the discrete ordinates equations. As a first step in the development of adaptive methods that are compatible with the variational nodal method, the authors examine error estimates for use in conjunction with spatial variables. The variational nodal method lends itself well to p refinement because the space-angle trial functions are hierarchical. Here they examine an error estimator for use with spatial p refinement for the diffusion approximation. Eventually, angular refinement will also be considered using spherical harmonics approximations

  6. 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

  7. A nodal collocation approximation for the multi-dimensional PL equations - 2D applications

    International Nuclear Information System (INIS)

    Capilla, M.; Talavera, C.F.; Ginestar, D.; Verdu, G.

    2008-01-01

    A classical approach to solve the neutron transport equation is to apply the spherical harmonics method obtaining a finite approximation known as the P L equations. In this work, the derivation of the P L equations for multi-dimensional geometries is reviewed and a nodal collocation method is developed to discretize these equations on a rectangular mesh based on the expansion of the neutronic fluxes in terms of orthogonal Legendre polynomials. The performance of the method and the dominant transport Lambda Modes are obtained for a homogeneous 2D problem, a heterogeneous 2D anisotropic scattering problem, a heterogeneous 2D problem and a benchmark problem corresponding to a MOX fuel reactor core

  8. Two-dimensional analytical solution for nodal calculation of nuclear reactors

    International Nuclear Information System (INIS)

    Silva, Adilson C.; Pessoa, Paulo O.; Silva, Fernando C.; Martinez, Aquilino S.

    2017-01-01

    Highlights: • A proposal for a coarse mesh nodal method is presented. • The proposal uses the analytical solution of the two-dimensional neutrons diffusion equation. • The solution is performed homogeneous nodes with dimensions of the fuel assembly. • The solution uses four average fluxes on the node surfaces as boundary conditions. • The results show good accuracy and efficiency. - Abstract: In this paper, the two-dimensional (2D) neutron diffusion equation is analytically solved for two energy groups (2G). The spatial domain of reactor core is divided into a set of nodes with uniform nuclear parameters. To determine iteratively the multiplication factor and the neutron flux in the reactor we combine the analytical solution of the neutron diffusion equation with an iterative method known as power method. The analytical solution for different types of regions that compose the reactor is obtained, such as fuel and reflector regions. Four average fluxes in the node surfaces are used as boundary conditions for analytical solution. Discontinuity factors on the node surfaces derived from the homogenization process are applied to maintain averages reaction rates and the net current in the fuel assembly (FA). To validate the results obtained by the analytical solution a relative power density distribution in the FAs is determined from the neutron flux distribution and compared with the reference values. The results show good accuracy and efficiency.

  9. Torsionfree Sheaves over a Nodal Curve of Arithmetic Genus One

    Indian Academy of Sciences (India)

    We classify all isomorphism classes of stable torsionfree sheaves on an irreducible nodal curve of arithmetic genus one defined over C C . Let be a nodal curve of arithmetic genus one defined over R R , with exactly one node, such that does not have any real points apart from the node. We classify all isomorphism ...

  10. Wielandt method applied to the diffusion equations discretized by finite element nodal methods

    International Nuclear Information System (INIS)

    Mugica R, A.; Valle G, E. del

    2003-01-01

    Nowadays the numerical methods of solution to the diffusion equation by means of algorithms and computer programs result so extensive due to the great number of routines and calculations that should carry out, this rebounds directly in the execution times of this programs, being obtained results in relatively long times. This work shows the application of an acceleration method of the convergence of the classic method of those powers that it reduces notably the number of necessary iterations for to obtain reliable results, what means that the compute times they see reduced in great measure. This method is known in the literature like Wielandt method and it has incorporated to a computer program that is based on the discretization of the neutron diffusion equations in plate geometry and stationary state by polynomial nodal methods. In this work the neutron diffusion equations are described for several energy groups and their discretization by means of those called physical nodal methods, being illustrated in particular the quadratic case. It is described a model problem widely described in the literature which is solved for the physical nodal grade schemes 1, 2, 3 and 4 in three different ways: to) with the classic method of the powers, b) method of the powers with the Wielandt acceleration and c) method of the powers with the Wielandt modified acceleration. The results for the model problem as well as for two additional problems known as benchmark problems are reported. Such acceleration method can also be implemented to problems of different geometry to the proposal in this work, besides being possible to extend their application to problems in 2 or 3 dimensions. (Author)

  11. Elaboration of a nodal method to solve the steady state multigroup diffusion equation. Study and use of the multigroup diffusion code DAHRA

    International Nuclear Information System (INIS)

    Halilou, A.; Lounici, A.

    1981-01-01

    The subject is divided in two parts: In the first part a nodal method has been worked out to solve the steady state multigroup diffusion equation. This method belongs to the same set of nodal methods currently used to calculate the exact fission powers and neutron fluxes in a very short computing time. It has been tested on a two dimensional idealized reactors. The effective multiplication factor and the fission powers for each fuel element have been calculated. The second part consists in studying and mastering the multigroup diffusion code DAHRA - a reduced version of DIANE - a two dimensional code using finite difference method

  12. A nodal method of calculating power distributions for LWR-type reactors with square fuel lattices

    International Nuclear Information System (INIS)

    Hoeglund, Randolph.

    1980-06-01

    A nodal model is developed for calculating the power distribution in the core of a light water reactor with a square fuel lattice. The reactor core is divided into a number of more or less cubic nodes and a nodal coupling equation, which gives the thermal power density in one node as a function of the power densities in the neighbour nodes, is derived from the neutron diffusion equations for two energy groups. The three-dimensional power distribution can be computed iteratively using this coupling equation, for example following the point Jacobi, the Gauss-Seidel or the point successive overrelaxation scheme. The method has been included as the neutronic model in a reactor core simulation computer code BOREAS, where it is combined with a thermal-hydraulic model in order to make a simultaneous computation of the interdependent power and void distributions in a boiling water reactor possible. Also described in this report are a method for temporary one-dimensional iteration developed in order to accelerate the iterative solution of the problem and the Haling principle which is widely used in the planning of reloading operations for BWR reactors. (author)

  13. Neutron stars

    International Nuclear Information System (INIS)

    Irvine, J.M.

    1978-01-01

    The subject is covered in chapters entitled: introduction (resume of stellar evolution, gross characteristics of neutron stars); pulsars (pulsar characteristics, pulsars as neutron stars); neutron star temperatures (neutron star cooling, superfluidity and superconductivity in neutron stars); the exterior of neutron stars (the magnetosphere, the neutron star 'atmosphere', pulses); neutron star structure; neutron star equations of state. (U.K.)

  14. Topological surface states in nodal superconductors

    International Nuclear Information System (INIS)

    Schnyder, Andreas P; Brydon, Philip M R

    2015-01-01

    Topological superconductors have become a subject of intense research due to their potential use for technical applications in device fabrication and quantum information. Besides fully gapped superconductors, unconventional superconductors with point or line nodes in their order parameter can also exhibit nontrivial topological characteristics. This article reviews recent progress in the theoretical understanding of nodal topological superconductors, with a focus on Weyl and noncentrosymmetric superconductors and their protected surface states. Using selected examples, we review the bulk topological properties of these systems, study different types of topological surface states, and examine their unusual properties. Furthermore, we survey some candidate materials for topological superconductivity and discuss different experimental signatures of topological surface states. (topical review)

  15. Nodal aberration theory applied to freeform surfaces

    Science.gov (United States)

    Fuerschbach, Kyle; Rolland, Jannick P.; Thompson, Kevin P.

    2014-12-01

    When new three-dimensional packages are developed for imaging optical systems, the rotational symmetry of the optical system is often broken, changing its imaging behavior and making the optical performance worse. A method to restore the performance is to use freeform optical surfaces that compensate directly the aberrations introduced from tilting and decentering the optical surfaces. In order to effectively optimize the shape of a freeform surface to restore optical functionality, it is helpful to understand the aberration effect the surface may induce. Using nodal aberration theory the aberration fields induced by a freeform surface in an optical system are explored. These theoretical predications are experimentally validated with the design and implementation of an aberration generating telescope.

  16. 3-D neutron transport benchmarks

    International Nuclear Information System (INIS)

    Takeda, T.; Ikeda, H.

    1991-03-01

    A set of 3-D neutron transport benchmark problems proposed by the Osaka University to NEACRP in 1988 has been calculated by many participants and the corresponding results are summarized in this report. The results of K eff , control rod worth and region-averaged fluxes for the four proposed core models, calculated by using various 3-D transport codes are compared and discussed. The calculational methods used were: Monte Carlo, Discrete Ordinates (Sn), Spherical Harmonics (Pn), Nodal Transport and others. The solutions of the four core models are quite useful as benchmarks for checking the validity of 3-D neutron transport codes

  17. Measuring and calculational complex on the base of multi-detector system and the kinetics parametrs by the method of neutron pulsed source

    International Nuclear Information System (INIS)

    Veselova, G.P.; Grachev, A.V.; Ivanova, N.K.

    1985-01-01

    Description of hardware of a measuring and calculation complex (MCC) designed for measuring neutron-physical characteristics of a reactor by the neutron pulsed method simultaneously from 8 detectors and the MERA-60 computer software used by MCC for measuring dependences of neutron generation and life time on the reactivity change is presented. The complex serviceability is tested at one of the PEI physical stands. MCC operation during a year has demonstrated its high reliability and a possibility of being used both for pulsed and other methods for investigating physical stands without introduction of supplementary equipment

  18. Error Estimation and Accuracy Improvements in Nodal Transport Methods; Estimacion de Errores y Aumento de la Precision en Metodos Nodales de Transporte

    Energy Technology Data Exchange (ETDEWEB)

    Zamonsky, O M [Comision Nacional de Energia Atomica, Centro Atomico Bariloche (Argentina)

    2000-07-01

    The accuracy of the solutions produced by the Discrete Ordinates neutron transport nodal methods is analyzed.The obtained new numerical methodologies increase the accuracy of the analyzed scheems and give a POSTERIORI error estimators. The accuracy improvement is obtained with new equations that make the numerical procedure free of truncation errors and proposing spatial reconstructions of the angular fluxes that are more accurate than those used until present. An a POSTERIORI error estimator is rigurously obtained for one dimensional systems that, in certain type of problems, allows to quantify the accuracy of the solutions. From comparisons with the one dimensional results, an a POSTERIORI error estimator is also obtained for multidimensional systems. LOCAL indicators, which quantify the spatial distribution of the errors, are obtained by the decomposition of the menctioned estimators. This makes the proposed methodology suitable to perform adaptive calculations. Some numerical examples are presented to validate the theoretical developements and to illustrate the ranges where the proposed approximations are valid.

  19. A nodal expansion method using conformal mapping for hexagonal geometry

    International Nuclear Information System (INIS)

    Chao, Y.A.; Shatilla, Y.A.

    1993-01-01

    Hexagonal nodal methods adopting the same transverse integration process used for square nodal methods face the subtle theoretical problem that this process leads to highly singular nonphysical terms in the diffusion equation. Lawrence, in developing the DIF3D-N code, tried to approximate the singular terms with relatively simple polynomials. In the HEX-NOD code, Wagner ignored the singularities to simplify the diffusion equation and introduced compensating terms in the nodal equations to restore the nodal balance relation. More recently developed hexagonal nodal codes, such as HEXPE-DITE and the hexagonal version of PANTHER, used methods similar to Wagner's. It will be shown that for light water reactor applications, these two different approximations significantly degraded the accuracy of the respective method as compared to the established square nodal methods. Alternatively, the method of conformal mapping was suggested to map a hexagon to a rectangle, with the unique feature of leaving the diffusion operator invariant, thereby fundamentally resolving the problems associated with transverse integration. This method is now implemented in the Westinghouse hexagonal nodal code ANC-H. In this paper we report on the results of comparing the three methods for a variety of problems via benchmarking against the fine-mesh finite difference code

  20. Modifying nodal pricing method considering market participants optimality and reliability

    Directory of Open Access Journals (Sweden)

    A. R. Soofiabadi

    2015-06-01

    Full Text Available This paper develops a method for nodal pricing and market clearing mechanism considering reliability of the system. The effects of components reliability on electricity price, market participants’ profit and system social welfare is considered. This paper considers reliability both for evaluation of market participant’s optimality as well as for fair pricing and market clearing mechanism. To achieve fair pricing, nodal price has been obtained through a two stage optimization problem and to achieve fair market clearing mechanism, comprehensive criteria has been introduced for optimality evaluation of market participant. Social welfare of the system and system efficiency are increased under proposed modified nodal pricing method.

  1. PWR in-core nuclear fuel management optimization utilizing nodal (non-linear NEM) generalized perturbation theory

    International Nuclear Information System (INIS)

    Maldonado, G.I.; Turinsky, P.J.; Kropaczek, D.J.

    1993-01-01

    The computational capability of efficiently and accurately evaluate reactor core attributes (i.e., k eff and power distributions as a function of cycle burnup) utilizing a second-order accurate advanced nodal Generalized Perturbation Theory (GPT) model has been developed. The GPT model is derived from the forward non-linear iterative Nodal Expansion Method (NEM) strategy, thereby extending its inherent savings in memory storage and high computational efficiency to also encompass GPT via the preservation of the finite-difference matrix structure. The above development was easily implemented into the existing coarse-mesh finite-difference GPT-based in-core fuel management optimization code FORMOSA-P, thus combining the proven robustness of its adaptive Simulated Annealing (SA) multiple-objective optimization algorithm with a high-fidelity NEM GPT neutronics model to produce a powerful computational tool used to generate families of near-optimum loading patterns for PWRs. (orig.)

  2. Calculation of the power factor using the neutron diffusion hybrid equation

    International Nuclear Information System (INIS)

    Costa da Silva, Adilson; Carvalho da Silva, Fernando; Senra Martinez, Aquilino

    2013-01-01

    Highlights: ► A neutron diffusion hybrid equation with an external neutron source was used. ► Nodal expansion method to obtain the neutron flux was used. ► Nuclear power factors in each fuel element in the reactor core were calculated. ► The results obtained were very accurate. -- Abstract: In this paper, we used a neutron diffusion hybrid equation with an external neutron source to calculate nuclear power factors in each fuel element in the reactor core. We used the nodal expansion method to obtain the neutron flux for a given control rods bank position. The results were compared with results obtained for eigenvalue problem near criticality condition and fixed source problem during the start-up of the reactor, where external neutron sources are extremely important for the stabilization of external neutron detectors.

  3. The correction of pebble bed reactor nodal cross sections for the effects of leakage and depletion history

    Science.gov (United States)

    Hudson, Nathanael Harrison

    An accurate and computationally fast method to generate nodal cross sections for the Pebble Bed Reactor (PBR) was presented. In this method, named Spectral History Correction (SHC), a set of fine group microscopic cross section libraries, pre-computed at specified depletion and moderation states, was coupled with the nodal nuclide densities and group bucklings to compute the new fine group spectrum for each node. The relevant fine group cross-section library was then recollapsed to the local broad group cross-section structure with this new fine group spectrum. This library set was tracked in terms of fuel isotopic densities. Fine group modulation factors (to correct the homogeneous flux for heterogeneous effects) and fission spectra were also stored with the cross section library. As the PBR simulation converges to a steady state fuel cycle, the initial nodal cross section library becomes inaccurate due to the burnup of the fuel and the neutron leakage into and out of the node. Because of the recirculation of discharged fuel pebbles with fresh fuel pebbles, a node can consist of a collection of pebbles at various burnup stages. To account for the nodal burnup, the microscopic cross sections were combined with nodal averaged atom densities to approximate the fine group macroscopic cross-sections for that node. These constructed, homogeneous macroscopic cross sections within the node were used to calculate a numerical solution for the fine group spectrum with B1 theory. This new fine spectrum was used to collapse the pre-computed microscopic cross section library to the broad group structure employed by the fuel cycle code. This SHC technique was developed and practically implemented as a subroutine within the PBR fuel cycle code PEBBED. The SHC subroutine was called to recalculate the broad group cross sections during the code convergence. The result was a fast method that compared favorably to the benchmark scheme of cross section calculation with the lattice

  4. Progress in implementation of the neutronics model of HEXTRAN into APROS

    International Nuclear Information System (INIS)

    Rintala, J.

    2009-01-01

    A new three-dimensional nodal model for neutronics calculation is currently under implementation into APROS - Advanced PROcess Simulation environment - to conform the increasing accuracy requirements. The new model is based on an advanced nodal code HEXTRAN and its static version HEXBU-3D by VTT, Technical Research Centre of Finland. However, several improvements for the model are made and the whole model has been reprogrammed. They don't change the theory basement of the method, but rather makes the implementation more flexible. Currently the computational part of the program is ready and current work concentrates on testing and validation. User interface details and usability issues need also work in the future. In this paper, general information about the improvements of the theory is explained first. Then the latest validation results are given. Currently the dynamical characteristics are tested by calculating the AER's kinetic benchmarks for VVER-440 reactors. In this paper, the results for the first benchmark are shown for two version of the code. The first version is fully HEXTRAN-comparable code to test that the basic structure works as wanted. The second version is the actual improved model for APROS. (author)

  5. Progress in implementation of the neutronics model of HEXTRAN into APROS

    International Nuclear Information System (INIS)

    Jukka Rintala

    2009-01-01

    A new three-dimensional nodal model for neutronics calculation is currently under implementation into APROS - Advanced PROcess Simulation environment - to conform the increasing accuracy requirements. The new model is based on an advanced nodal code HEXTRAN and its static version HEXBU-3D by VTT, Technical Research Centre of Finland. However, several improvements for the model are made and the whole model has been reprogrammed. They don't change the theory basement of the method, but rather makes the implementation more flexible. Currently the computational part of the program is ready and current work concentrates on testing and validation. User interface details and usability issues need also work in the future. In this paper, general information about the improvements of the theory is explained first. Then the latest validation results are given. Currently the dynamical characteristics are tested by calculating the atomic energy research's kinetic benchmarks for WWER-440 reactors. In this paper, the results for the first benchmark are shown for two version of the code. The first version is fully HEXTRAN-comparable code to test that the basic structure works as wanted. The second version is the actual improved model for APROS. (Authors)

  6. Validation of new 3-D neutronics model in APROS for hexagonal geometry

    International Nuclear Information System (INIS)

    Rintala, J.

    2010-01-01

    APROS - Advanced PROcess Simulation environment-is a widely used simulation tool for nuclear power plant modelling. Earlier the three-dimensional neutronics calculation has been based on model using the difference method. The original three-dimensional core model is mainly used in power plant simulator applications, where it fits well because of its speed. For safety analysis purposes, however, a new model was considered to be an important improvement to meet the accuracy requirements. A sophisticated nodal model used already in HEXTRAN and TRAB-3D was decided to be implemented into APROS. The hexagonal part of the model has now been implemented and tested. For practical reasons, the model was programmed from scratch into APROS and also some small improvements were added and thus, an extensive validation program was necessary to prove the correct behaviour of the model. In this paper, the most important results from AER kinetic benchmarks 2 and 3 calculations are shown as well as the calculation results against data achieved LR-0 test reactor space-time kinetic experiments. Since the model is similar to the one in HEXTRAN, the results in benchmarks are compared to the results by it. In LR-0 calculations, results by both, original and new model are presented and compared to the measurements. The results shows that the implementation of the model has been successful and the new model improves the accuracy of three-dimensional neutronics calculation in APROS into the level required in safety analyses. (Author)

  7. Point kinetics modeling

    International Nuclear Information System (INIS)

    Kimpland, R.H.

    1996-01-01

    A normalized form of the point kinetics equations, a prompt jump approximation, and the Nordheim-Fuchs model are used to model nuclear systems. Reactivity feedback mechanisms considered include volumetric expansion, thermal neutron temperature effect, Doppler effect and void formation. A sample problem of an excursion occurring in a plutonium solution accidentally formed in a glovebox is presented

  8. CT simulation in nodal positive breast cancer

    International Nuclear Information System (INIS)

    Horst, E.; Schuck, A.; Moustakis, C.; Schaefer, U.; Micke, O.; Kronholz, H.L.; Willich, N.

    2001-01-01

    Background: A variety of solutions are used to match tangential fields and opposed lymph node fields in irradiation of nodal positive breast cancer. The choice is depending on the technical equipment which is available and the clinical situation. The CT simulation of a non-monoisocentric technique was evaluated in terms of accuracy and reproducibility. Patients, Material and Methods: The field match parameters were adjusted virtually at CT simulation and were compared with parameters derived mathematically. The coordinate transfer from the CT simulator to the conventional simulator was analyzed in 25 consecutive patients. Results: The angles adjusted virtually for a geometrically exact coplanar field match corresponded with the angles calculated for each set-up. The mean isocenter displacement was 5.7 mm and the total uncertainty of the coordinate transfer was 6.7 mm (1 SD). Limitations in the patient set-up became obvious because of the steep arm abduction necessary to fit the 70 cm CT gantry aperture. Required modifications of the arm position and coordinate transfer errors led to a significant shift of the marked matchline of >1.0 cm in eight of 25 patients (32%). Conclusion: The virtual CT simulation allows a precise and graphic definition of the field match parameters. However, modifications of the virtual set-up basically due to technical limitations were required in a total of 32% of cases, so that a hybrid technique was adapted at present that combines virtual adjustment of the ideal field alignment parameters with conventional simulation. (orig.) [de

  9. Present Status of GNF New Nodal Simulator

    International Nuclear Information System (INIS)

    Iwamoto, T.; Tamitani, M.; Moore, B.

    2001-01-01

    This paper presents core simulator consolidation work done at Global Nuclear Fuel (GNF). The unified simulator needs to supercede the capabilities of past simulator packages from the original GNF partners: GE, Hitachi, and Toshiba. At the same time, an effort is being made to produce a simulation package that will be a state-of-the-art analysis tool when released, in terms of the physics solution methodology and functionality. The core simulator will be capable and qualified for (a) high-energy cycles in the U.S. markets, (b) mixed-oxide (MOX) introduction in Japan, and (c) high-power density plants in Europe, etc. The unification of the lattice physics code is also in progress based on a transport model with collision probability methods. The AETNA core simulator is built upon the PANAC11 software base. The goal is to essentially replace the 1.5-energy-group model with a higher-order multigroup nonlinear nodal solution capable of the required modeling fidelity, while keeping highly automated library generation as well as functionality. All required interfaces to PANAC11 will be preserved, which minimizes the impact on users and process automation. Preliminary results show statistical accuracy improvement over the 1.5-group model

  10. BEACON: An application of nodal methods for operational support

    International Nuclear Information System (INIS)

    Boyd, W.A.; Nguyen, T.Q.

    1992-01-01

    A practical application of nodal methods is on-line plant operational support. However, to enable plant personnel to take full advantage of a nodal model to support plant operations, (a) a core nodal model must always be up to date with the current core history and conditions, (b) the nodal methods must be fast enough to allow numerous core calculations to be performed in minutes to support engineering decisions, and (c) the system must be easily accessible to engineering personnel at the reactor, their offices, or any other location considered appropriate. A core operational support package developed by Westinghouse called BEACON (best estimate analysis of core operations - nuclear) has been installed at several plants. Results from these plants and numerous in-core flux maps analyzed have demonstrated the accuracy of the model and the effectiveness of the methodology

  11. Aircraft Nodal Data Acquisition System (ANDAS), Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — Development of an Aircraft Nodal Data Acquisition System (ANDAS) based upon the short haul Zigbee networking standard is proposed. It employs a very thin (135 um)...

  12. Hybrid nodal loop metal: Unconventional magnetoresponse and material realization

    Science.gov (United States)

    Zhang, Xiaoming; Yu, Zhi-Ming; Lu, Yunhao; Sheng, Xian-Lei; Yang, Hui Ying; Yang, Shengyuan A.

    2018-03-01

    A nodal loop is formed by a band crossing along a one-dimensional closed manifold, with each point on the loop a linear nodal point in the transverse dimensions, and can be classified as type I or type II depending on the band dispersion. Here, we propose a class of nodal loops composed of both type-I and type-II points, which are hence termed as hybrid nodal loops. Based on first-principles calculations, we predict the realization of such loops in the existing electride material Ca2As . For a hybrid loop, the Fermi surface consists of coexisting electron and hole pockets that touch at isolated points for an extended range of Fermi energies, without the need for fine-tuning. This leads to unconventional magnetic responses, including the zero-field magnetic breakdown and the momentum-space Klein tunneling observable in the magnetic quantum oscillations, as well as the peculiar anisotropy in the cyclotron resonance.

  13. A nodal method based on matrix-response method

    International Nuclear Information System (INIS)

    Rocamora Junior, F.D.; Menezes, A.

    1982-01-01

    A nodal method based in the matrix-response method, is presented, and its application to spatial gradient problems, such as those that exist in fast reactors, near the core - blanket interface, is investigated. (E.G.) [pt

  14. Nodal prices determination with wind integration for radial ...

    African Journals Online (AJOL)

    With competitive electricity market operation, open access to the transmission and distribution network is essential ... The results have been obtained for IEEE 33 ...... The value of intermittent wind DG under nodal prices and amp – mile tariffs.

  15. Nodal aberration theory for wild-filed asymmetric optical systems

    Science.gov (United States)

    Chen, Yang; Cheng, Xuemin; Hao, Qun

    2016-10-01

    Nodal Aberration Theory (NAT) was used to calculate the zero field position in Full Field Display (FFD) for the given aberration term. Aiming at wide-filed non-rotational symmetric decentered optical systems, we have presented the nodal geography behavior of the family of third-order and fifth-order aberrations. Meanwhile, we have calculated the wavefront aberration expressions when one optical element in the system is tilted, which was not at the entrance pupil. By using a three-piece-cellphone lens example in optical design software CodeV, the nodal geography is testified under several situations; and the wavefront aberrations are calculated when the optical element is tilted. The properties of the nodal aberrations are analyzed by using Fringe Zernike coefficients, which are directly related with the wavefront aberration terms and usually obtained by real ray trace and wavefront surface fitting.

  16. STEP- A three-dimensional nodal diffusion code for LMR's

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeong Il; Kim, Taek Kyum [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-12-01

    STEP is a three-dimensional multigroup nodal diffusion code for the neutronics analysis of the LMR core. STEP employs DIF3D and HEXNOD nodal methods. In DIF3D, one-dimensional fluxes are approximated by polynomials while HEXNOD analytically solves transverse-integrated one-dimensional diffusion equations. The nodal equations are solved using a conventional fission source iteration procedure accelerated by coarse-mesh rebalancing and asymptotic extrapolation. At each fission source iteration, the interface currents for each group are computed by solving the response matrix equations with a known group source term. These partial currents are used to updata flux moments. This solution is accomplished by inner iteration, a series of sweeps through the spatial mesh. Inner iterations are performed by sweeping the axial mesh plane in a standard red-black checkerboard ordering, i.e. the odd-numbered planes are processed during the first pass, followed by the even-numbered planes on the second pass. On each plane, the nodes are swept in the four-color checkerboard ordering. STEP accepts microscopic cross section data from the CCCC standard interface file ISOTXS currently used for the neutronics analysis of LMR's at KAERI as well as macroscopic cross section data. Material cross sections are obtained by summing the product of atom densities and microscopic cross sections over all isotopes comprising the material. Energy is released from both fission ad capture. The thermal-hydraulics model calculates average fuel and coolant temperatures. STEP takes account of feedback effects from both fuel temperature and coolant temperature changes. The thermal-hydraulics model is a conservative, single channel model where there is no heat transfer between assemblies. Thus, STEP gives conservative results which, however, are of useful information for core design and can be useful tool for neutronics analysis of LMR core design and will be used for the base program of a future

  17. Three-dimensional static and dynamic reactor calculations by the nodal expansion method

    International Nuclear Information System (INIS)

    Christensen, B.

    1985-05-01

    This report reviews various method for the calculation of the neutron-flux- and power distribution in an nuclear reactor. The nodal expansion method (NEM) is especially described in much detail. The nodal expansion method solves the diffusion equation. In this method the reactor core is divided into nodes, typically 10 to 20 cm in each direction, and the average flux in each node is calculated. To obtain the coupling between the nodes the local flux inside each node is expressed by use of a polynomial expansion. The expansion is one-dimensional, so inside each node such three expansions occur. To calculate the expansion coefficients it is necessary that the polynomial expansion is a solution to the one-dimensional diffusion equation. When the one-dimensional diffusion equation is established a term with the transversal leakage occur, and this term is expanded after the same polynomials. The resulting equation system with the expansion coefficients as the unknowns is solved with weigthed residual technique. The nodal expansion method is built into a computer program (also called NEM), which is divided into two parts, one part for steady-state calculations and one part for dynamic calculations. It is possible to take advantage of symmetry properties of the reactor core. The program is very flexible with regard to the number of energy groups, the node size, the flux expansion order and the transverse leakage expansion order. The boundary of the core is described by albedos. The program and input to it are described. The program is tested on a number of examples extending from small theoretical one up to realistic reactor cores. Many calculations are done on the wellknown IAEA benchmark case. The calculations have tested the accuracy and the computing time for various node sizes and polynomial expansions. In the dynamic examples various strategies for variation of the time step-length have been tested. (author)

  18. Uniqueness Theorem for the Inverse Aftereffect Problem and Representation the Nodal Points Form

    Directory of Open Access Journals (Sweden)

    A. Neamaty

    2015-03-01

    Full Text Available In this paper, we consider a boundary value problem with aftereffect on a finite interval. Then, the asymptotic behavior of the solutions, eigenvalues, the nodal points and the associated nodal length are studied. We also calculate the numerical values of the nodal points and the nodal length. Finally, we prove the uniqueness theorem for the inverse aftereffect problem by applying any dense subset of the nodal points.

  19. Uniqueness Theorem for the Inverse Aftereffect Problem and Representation the Nodal Points Form

    OpenAIRE

    A. Neamaty; Sh. Akbarpoor; A. Dabbaghian

    2015-01-01

    In this paper, we consider a boundary value problem with aftereffect on a finite interval. Then, the asymptotic behavior of the solutions, eigenvalues, the nodal points and the associated nodal length are studied. We also calculate the numerical values of the nodal points and the nodal length. Finally, we prove the uniqueness theorem for the inverse aftereffect problem by applying any dense subset of the nodal points.

  20. Proliferation kinetics of cultured cells after irradiation with X-rays and 14 MeV neutrons studied by time-lapse cinematography

    Energy Technology Data Exchange (ETDEWEB)

    Kooi, M.W.; Stap, J.; Barendsen, G.W. (Amsterdam Univ. (Netherlands). Lab. for Radiobiology)

    1984-06-01

    Exponentially growing cells of an established line derived from a mouse osteosarcoma (MOS) have been studied by time-lapse cinematography after irradiation with 3 Gy of 200 kV X- rays or 1.5 Gy of 14 MeV neutrons. The results show that the radiation doses applied cause an equal elongation of the mean cell cycle time Tsub(c), which is largest in the irradiated cells but persists in the three subsequent generations. After 3 Gy of X-rays, mitotic delay is largest in cells irradiated in later stages of the cycle, but this difference is not observed after 1.5 Gy of 14 MeV neutrons. In subsequent generations the Tsub(c) values show larger variations among descendents of cells treated in the same stage of the cycle as compared to controls but this variation is equal for the doses of X-rays and neutrons applied. Division probability was significantly reduced in irradiated cells as well as in subsequent generations, whereas with neutrons as compared to X-rays the damage is expressed in earlier generations, with less variation as a function of the cell cycle.

  1. Proliferation kinetics of cultured cells after irradiation with X-rays and 14 MeV neutrons studied by time-lapse cinematography.

    Science.gov (United States)

    Kooi, M W; Stap, J; Barendsen, G W

    1984-06-01

    Exponentially growing cells of an established line derived from a mouse osteosarcoma (MOS) have been studied by time-lapse cinematography after irradiation with 3 Gy of 200 kV X-rays or 1.5 Gy of 14 MeV neutrons. Cell cycle times (Tc) of individual cells and their progeny in three subsequent generations as well as the occurrence of aberrant mitosis have been determined to evaluate the variation in expression of damage in relation to the stage in the intermitotic cycle and the radiation quality. The results show that the radiation doses applied cause an equal elongation of the mean Tc, which is largest in the irradiated cells but persists in the three subsequent generations. After 3 Gy of X-rays, mitotic delay is largest in cells irradiated in later stages of the cycle, but this difference is not observed after 1.5 Gy of 14 MeV neutrons. In subsequent generations the Tc values show larger variations among descendents of cells treated in the same stage of the cycle as compared to controls but this variation is equal for the doses of X-rays and neutrons applied. Division probability was significantly reduced in irradiated cells as well as in subsequent generations, whereby with neutrons as compared to X-rays the damage is expressed in earlier generations, with less variation as a function of the cell cycle.

  2. Proliferation kinetics of cultured cells after irradiation with X-rays and 14 MeV neutrons studied by time-lapse cinematography

    International Nuclear Information System (INIS)

    Kooi, M.W.; Stap, J.; Barendsen, G.W.

    1984-01-01

    Exponentially growing cells of an established line derived from a mouse osteosarcoma (MOS) have been studied by time-lapse cinematography after irradiation with 3 Gy of 200 kV X- rays or 1.5 Gy of 14 MeV neutrons. The results show that the radiation doses applied cause an equal elongation of the mean cell cycle time Tsub(c), which is largest in the irradiated cells but persists in the three subsequent generations. After 3 Gy of X-rays, mitotic delay is largest in cells irradiated in later stages of the cycle, but this difference is not observed after 1.5 Gy of 14 MeV neutrons. In subsequent generations the Tsub(c) values show larger variations among descendents of cells treated in the same stage of the cycle as compared to controls but this variation is equal for the doses of X-rays and neutrons applied. Division probability was significantly reduced in irradiated cells as well as in subsequent generations, whereas with neutrons as compared to X-rays the damage is expressed in earlier generations, with less variation as a function of the cell cycle. (author)

  3. Neutron radiotherapy

    International Nuclear Information System (INIS)

    Thomas, F.J.

    1987-01-01

    The rationale for the application of neutron radiation for the treatment of malignancies is well established based on radiobiological studies. These factors include the presence of tissue hypoxia, radiation response as a function of cell cycle kinetics, the repair capacity of the malignant cells and the regeneration of malignant cells during a fractionated course of radiation. Despite the constraints under which the clinical trials to date have been conducted, promising results have been obtained. Randomized trials have demonstrated that neutron therapy is the treatment of choice for inoperable salivary gland carcinomas. A randomized trial of the RTOG has demonstrated a probable advantage for neutron radiation in the treatment of advanced prostate carcinomas but is yet to be confirmed. An improvement in local control has also been observed for selected sarcomas. Equivocal or contradictory results have been obtained for squamous cell carcinomas of the head and neck, bronchogenic carcinomas, advanced rectal, transitional cell carcinomas of the bladder and cervical carcinomas. The practical consequences of these radiobiological and clinical observations on the current generation of clinical trials is discussed

  4. The ADO-nodal method for solving two-dimensional discrete ordinates transport problems

    International Nuclear Information System (INIS)

    Barichello, L.B.; Picoloto, C.B.; Cunha, R.D. da

    2017-01-01

    Highlights: • Two-dimensional discrete ordinates neutron transport. • Analytical Discrete Ordinates (ADO) nodal method. • Heterogeneous media fixed source problems. • Local solutions. - Abstract: In this work, recent results on the solution of fixed-source two-dimensional transport problems, in Cartesian geometry, are reported. Homogeneous and heterogeneous media problems are considered in order to incorporate the idea of arbitrary number of domain division into regions (nodes) when applying the ADO method, which is a method of analytical features, to those problems. The ADO-nodal formulation is developed, for each node, following previous work devoted to heterogeneous media problem. Here, however, the numerical procedure is extended to higher number of domain divisions. Such extension leads, in some cases, to the use of an iterative method for solving the general linear system which defines the arbitrary constants of the general solution. In addition to solve alternative heterogeneous media configurations than reported in previous works, the present approach allows comparisons with results provided by other metodologies generated with refined meshes. Numerical results indicate the ADO solution may achieve a prescribed accuracy using coarser meshes than other schemes.

  5. Application of nonlinear nodal diffusion generalized perturbation theory to nuclear fuel reload optimization

    International Nuclear Information System (INIS)

    Maldonado, G.I.; Turinsky, P.J.

    1995-01-01

    The determination of the family of optimum core loading patterns for pressurized water reactors (PWRs) involves the assessment of the core attributes for thousands of candidate loading patterns. For this reason, the computational capability to efficiently and accurately evaluate a reactor core's eigenvalue and power distribution versus burnup using a nodal diffusion generalized perturbation theory (GPT) model is developed. The GPT model is derived from the forward nonlinear iterative nodal expansion method (NEM) to explicitly enable the preservation of the finite difference matrix structure. This key feature considerably simplifies the mathematical formulation of NEM GPT and results in reduced memory storage and CPU time requirements versus the traditional response-matrix approach to NEM. In addition, a treatment within NEM GPT can account for localized nonlinear feedbacks, such as that due to fission product buildup and thermal-hydraulic effects. When compared with a standard nonlinear iterative NEM forward flux solve with feedbacks, the NEM GPT model can execute between 8 and 12 times faster. These developments are implemented within the PWR in-core nuclear fuel management optimization code FORMOSA-P, combining the robustness of its adaptive simulated annealing stochastic optimization algorithm with an NEM GPT neutronics model that efficiently and accurately evaluates core attributes associated with objective functions and constraints of candidate loading patterns

  6. 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

  7. NODAL3 Sensitivity Analysis for NEACRP 3D LWR Core Transient Benchmark (PWR

    Directory of Open Access Journals (Sweden)

    Surian Pinem

    2016-01-01

    Full Text Available This paper reports the results of sensitivity analysis of the multidimension, multigroup neutron diffusion NODAL3 code for the NEACRP 3D LWR core transient benchmarks (PWR. The code input parameters covered in the sensitivity analysis are the radial and axial node sizes (the number of radial node per fuel assembly and the number of axial layers, heat conduction node size in the fuel pellet and cladding, and the maximum time step. The output parameters considered in this analysis followed the above-mentioned core transient benchmarks, that is, power peak, time of power peak, power, averaged Doppler temperature, maximum fuel centerline temperature, and coolant outlet temperature at the end of simulation (5 s. The sensitivity analysis results showed that the radial node size and maximum time step give a significant effect on the transient parameters, especially the time of power peak, for the HZP and HFP conditions. The number of ring divisions for fuel pellet and cladding gives negligible effect on the transient solutions. For productive work of the PWR transient analysis, based on the present sensitivity analysis results, we recommend NODAL3 users to use 2×2 radial nodes per assembly, 1×18 axial layers per assembly, the maximum time step of 10 ms, and 9 and 1 ring divisions for fuel pellet and cladding, respectively.

  8. Some neutronics of innovative subcritical assembly with fast neutron spectrum

    International Nuclear Information System (INIS)

    Kiyavitskaya, H.; Fokov, Yu.; Rutkovskaya, Ch.; Sadovich, S.; Kasuk, D.; Gohar, Y.; Bolshinsky, I.

    2013-01-01

    Conclusion: • New assembly can be used to: • develop the experimental techniques and adapt the existing ones for monitoring the sub-criticality level, neutron spectra measurements, etc; • study the spatial kinetics of sub-critical and critical systems with fast neutron spectra; • measure the transmutation reaction rates of minor-actinides etc

  9. Fully kinetic simulations of megajoule-scale dense plasma focus

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, A.; Link, A.; Tang, V.; Halvorson, C.; May, M. [Lawrence Livermore National Laboratory, Livermore California 94550 (United States); Welch, D. [Voss Scientific, LLC, Albuquerque, New Mexico 87108 (United States); Meehan, B. T.; Hagen, E. C. [National Security Technologies, LLC, Las Vegas, Nevada 89030 (United States)

    2014-10-15

    Dense plasma focus (DPF) Z-pinch devices are sources of copious high energy electrons and ions, x-rays, and neutrons. Megajoule-scale DPFs can generate 10{sup 12} neutrons per pulse in deuterium gas through a combination of thermonuclear and beam-target fusion. However, the details of the neutron production are not fully understood and past optimization efforts of these devices have been largely empirical. Previously, we reported on the first fully kinetic simulations of a kilojoule-scale DPF and demonstrated that both kinetic ions and kinetic electrons are needed to reproduce experimentally observed features, such as charged-particle beam formation and anomalous resistivity. Here, we present the first fully kinetic simulation of a MegaJoule DPF, with predicted ion and neutron spectra, neutron anisotropy, neutron spot size, and time history of neutron production. The total yield predicted by the simulation is in agreement with measured values, validating the kinetic model in a second energy regime.

  10. Development of a 3D consistent 1D neutronics model for reactor core simulation

    International Nuclear Information System (INIS)

    Lee, Ki Bog; Joo, Han Gyu; Cho, Byung Oh; Zee, Sung Quun

    2001-02-01

    In this report a 3D consistent 1D model based on nonlinear analytic nodal method is developed to reproduce the 3D results. During the derivation, the current conservation factor (CCF) is introduced which guarantees the same axial neutron currents obtained from the 1D equation as the 3D reference values. Furthermore in order to properly use 1D group constants, a new 1D group constants representation scheme employing tables for the fuel temperature, moderator density and boron concentration is developed and functionalized for the control rod tip position. To test the 1D kinetics model with CCF, several steady state and transient calculations were performed and compared with 3D reference values. The errors of K-eff values were reduced about one tenth when using CCF without significant computational overhead. And the errors of power distribution were decreased to the range of one fifth or tenth at steady state calculation. The 1D kinetics model with CCF and the 1D group constant functionalization employing tables as a function of control rod tip position can provide preciser results at the steady state and transient calculation. Thus it is expected that the 1D kinetics model derived in this report can be used in the safety analysis, reactor real time simulation coupled with system analysis code, operator support system etc.

  11. Application of the HGPT methodology of reactor operation problems with a nodal mixed method

    International Nuclear Information System (INIS)

    Baudron, A.M.; Bruna, G.B.; Gandini, A.; Lautard, J.J.; Monti, S.; Pizzigati, G.

    1998-01-01

    The heuristically based generalized perturbation theory (HGPT), to first and higher order, applied to the neutron field of a reactor system, is discussed in relation to quasistatic problems. This methodology is of particular interest in reactor operation. In this application it may allow an on-line appraisal of the main physical responses of the reactor system when subject to alterations relevant to normal system exploitation, e.g. control rod movement, and/or soluble boron concentration changes to be introduced, for instance, for compensating power level variations following electrical network demands. In this paper, after describing the main features of the theory, its implementation into the diffusion, 3D mixed dual nodal code MINOS of the SAPHYR system is presented. The results from a small scale investigation performed on a simplified PWR system corroborate the validity of the methodology proposed

  12. Implementation of the kinetics in the transport code AZTRAN; Implementacion de la cinetica en el codigo de transporte AZTRAN

    Energy Technology Data Exchange (ETDEWEB)

    Duran G, J. A.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, San Pedro Zacatenco, 07738 Ciudad de Mexico (Mexico); Gomez T, A. M., E-mail: redfield1290@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2017-09-15

    This paper shows the implementation of the time dependence in the three-dimensional transport code AZTRAN (AZtlan TRANsport), which belongs to the AZTLAN platform, for the analysis of nuclear reactors (currently under development). The AZTRAN code with this implementation is able to numerically solve the time-dependent transport equation in XYZ geometry, for several energy groups, using the discrete ordinate method S{sub n} for the discretization of the angular variable, the nodal method RTN-0 for spatial discretization and method 0 for discretization in time. Initially, the code only solved the neutrons transport equation in steady state, so the implementation of the temporal part was made integrating the neutrons transport equation with respect to time and balance equations corresponding to the concentrations of delayed neutron precursors, for which method 0 was applied. After having directly implemented code kinetics, the improved quasi-static method was implemented, which is a tool for reducing computation time, where the angular flow is factored by the product of two functions called shape function and amplitude function, where the first is calculated for long time steps, called macro-steps and the second is resolved for small time steps called micro-steps. In the new version of AZTRAN several Benchmark problems that were taken from the literature were simulated, the problems used are of two and three dimensions which allowed corroborating the accuracy and stability of the code, showing in general in the reference tests a good behavior. (Author)

  13. Coupled Model of channels in parallel and neutron kinetics in two dimensions; Modelo acoplado de canales en paralelo y cinetica neutronica en dos dimensiones

    Energy Technology Data Exchange (ETDEWEB)

    Cecenas F, M.; Campos G, R.M. [Instituto de Investigaciones Electricas, Av. Reforma 113, Col. Palmira, 62490 Cuernavaca, Morelos (Mexico)]. E-mail: mcf@iie.org.mx; Valle G, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)

    2004-07-01

    In this work an arrangement of thermohydraulic channels is presented that represent those four quadrants of a nucleus of reactor type BWR. The channels are coupled to a model of neutronic in two dimensions that allow to generate the radial profile of power of the reactor. Nevertheless that the neutronic pattern is of two dimensions, it is supplemented with axial additional information when considering the axial profiles of power for each thermo hydraulic channel. The stationary state is obtained the one it imposes as frontier condition the same pressure drop for all the channels. This condition is satisfied to iterating on the flow of coolant in each channel to equal the pressure drop in all the channels. This stationary state is perturbed later on when modifying the values for the effective sections corresponding to an it assembles. The calculation in parallel of the neutronic and the thermo hydraulic is carried out with Vpm (Virtual parallel machine) by means of an outline teacher-slave in a local net of computers. (Author)

  14. Implications of inaccurate clinical nodal staging in pancreatic adenocarcinoma.

    Science.gov (United States)

    Swords, Douglas S; Firpo, Matthew A; Johnson, Kirsten M; Boucher, Kenneth M; Scaife, Courtney L; Mulvihill, Sean J

    2017-07-01

    Many patients with stage I-II pancreatic adenocarcinoma do not undergo resection. We hypothesized that (1) clinical staging underestimates nodal involvement, causing stage IIB to have a greater percent of resected patients and (2) this stage-shift causes discrepancies in observed survival. The Surveillance, Epidemiology, and End Results (SEER) research database was used to evaluate cause-specific survival in patients with pancreatic adenocarcinoma from 2004-2012. Survival was compared using the log-rank test. Single-center data on 105 patients who underwent resection of pancreatic adenocarcinoma without neoadjuvant treatment were used to compare clinical and pathologic nodal staging. In SEER data, medium-term survival in stage IIB was superior to IB and IIA, with median cause-specific survival of 14, 9, and 11 months, respectively (P < .001). Seventy-two percent of stage IIB patients underwent resection vs 28% in IB and 36% in IIA (P < .001). In our institutional data, 12.4% of patients had clinical evidence of nodal involvement vs 69.5% by pathologic staging (P < .001). Among clinical stage IA-IIA patients, 71.6% had nodal involvement by pathologic staging. Both SEER and institutional data support substantial underestimation of nodal involvement by clinical staging. This finding has implications in decisions regarding neoadjuvant therapy and analysis of outcomes in the absence of pathologic staging. Copyright © 2017 Elsevier Inc. All rights reserved.

  15. Development of an Analytic Nodal Diffusion Solver in Multi-groups for 3D Reactor Cores with Rectangular or Hexagonal Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Lozano, Juan Andres; Aragones, Jose Maria; Garcia-Herranz, Nuria [Universidad Politecnica de Madrid, 28006 Jose Gutierrez Abascal 2, Madrid (Spain)

    2008-07-01

    More accurate modelling of physical phenomena involved in present and future nuclear reactors requires a multi-scale and multi-physics approach. This challenge can be accomplished by the coupling of best-estimate core-physics, thermal-hydraulics and multi-physics solvers. In order to make viable that coupling, the current trends in reactor simulations are along the development of a new generation of tools based on user-friendly, modular, easily linkable, faster and more accurate codes to be integrated in common platforms. These premises are in the origin of the NURESIM Integrated Project within the 6. European Framework Program, which is envisaged to provide the initial step towards a Common European Standard Software Platform for nuclear reactors simulations. In the frame of this project and to reach the above-mentioned goals, a 3-D multigroup nodal solver for neutron diffusion calculations called ANDES (Analytic Nodal Diffusion Equation Solver) has been developed and tested in-depth in this Thesis. ANDES solves the steady-state and time-dependent neutron diffusion equation in three-dimensions and any number of energy groups, utilizing the Analytic Coarse-Mesh Finite-Difference (ACMFD) scheme to yield the nodal coupling equations. It can be applied to both Cartesian and triangular-Z geometries, so that simulations of LWR as well as VVER, HTR and fast reactors can be performed. The solver has been implemented in a fully encapsulated way, enabling it as a module to be readily integrated in other codes and platforms. In fact, it can be used either as a stand-alone nodal code or as a solver to accelerate the convergence of whole core pin-by-pin code systems. Verification of performance has shown that ANDES is a code with high order definition for whole core realistic nodal simulations. In this paper, the methodology developed and involved in ANDES is presented. (authors)

  16. Polarized neutrons

    International Nuclear Information System (INIS)

    Williams, W.G.

    1988-01-01

    The book on 'polarized neutrons' is intended to inform researchers in condensed matter physics and chemistry of the diversity of scientific problems that can be investigated using polarized neutron beams. The contents include chapters on:- neutron polarizers and instrumentation, polarized neutron scattering, neutron polarization analysis experiments and precessing neutron polarization. (U.K.)

  17. Solution of the transport equation in stationary state, in one and two dimensions, for BWR assemblies using nodal methods; Solucion de la ecuacion de transporte en estado estacionario, en 1 y 2 dimensiones, para ensambles tipo BWR usando metodos nodales

    Energy Technology Data Exchange (ETDEWEB)

    Xolocostli M, J V

    2002-07-01

    The main objective of this work is to solve the neutron transport equation in one and two dimensions (slab geometry and X Y geometry, respectively), with no time dependence, for BWR assemblies using nodal methods. In slab geometry, the nodal methods here used are the polynomial continuous (CMPk) and discontinuous (DMPk) families but only the Linear Continuous (also known as Diamond Difference), the Quadratic Continuous (QC), the Cubic Continuous (CC), the Step Discontinuous (also known as Backward Euler), the Linear Discontinuous (LD) and the Quadratic Discontinuous (QD) were considered. In all these schemes the unknown function, the angular neutron flux, is approximated as a sum of basis functions in terms of Legendre polynomials, associated to the values of the neutron flux in the edges (left, right, or both) and the Legendre moments in the cell, depending on the nodal scheme used. All these schemes were implemented in a computer program developed in previous thesis works and known with the name TNX. This program was modified for the purposes of this work. The program discreetizes the domain of concern in one dimension and determines numerically the angular neutron flux for each point of the discretization when the number of energy groups and regions are known starting from an initial approximation for the angular neutron flux being consistent with the boundary condition imposed for a given problem. Although only problems with two-energy groups were studied the computer program does not have limitations regarding the number of energy groups and the number of regions. The two problems analyzed with the program TNX have practically the same characteristics (fuel and water), with the difference that one of them has a control rod. In the part corresponding to two-dimensional problems, the implemented nodal methods were those designated as hybrids that consider not only the edge and cell Legendre moments, but also the values of the neutron flux in the corner points

  18. Solution of the transport equation in stationary state, in one and two dimensions, for BWR assemblies using nodal methods; Solucion de la ecuacion de transporte en estado estacionario, en 1 y 2 dimensiones, para ensambles tipo BWR usando metodos nodales

    Energy Technology Data Exchange (ETDEWEB)

    Xolocostli M, J.V

    2002-07-01

    The main objective of this work is to solve the neutron transport equation in one and two dimensions (slab geometry and X Y geometry, respectively), with no time dependence, for BWR assemblies using nodal methods. In slab geometry, the nodal methods here used are the polynomial continuous (CMPk) and discontinuous (DMPk) families but only the Linear Continuous (also known as Diamond Difference), the Quadratic Continuous (QC), the Cubic Continuous (CC), the Step Discontinuous (also known as Backward Euler), the Linear Discontinuous (LD) and the Quadratic Discontinuous (QD) were considered. In all these schemes the unknown function, the angular neutron flux, is approximated as a sum of basis functions in terms of Legendre polynomials, associated to the values of the neutron flux in the edges (left, right, or both) and the Legendre moments in the cell, depending on the nodal scheme used. All these schemes were implemented in a computer program developed in previous thesis works and known with the name TNX. This program was modified for the purposes of this work. The program discreetizes the domain of concern in one dimension and determines numerically the angular neutron flux for each point of the discretization when the number of energy groups and regions are known starting from an initial approximation for the angular neutron flux being consistent with the boundary condition imposed for a given problem. Although only problems with two-energy groups were studied the computer program does not have limitations regarding the number of energy groups and the number of regions. The two problems analyzed with the program TNX have practically the same characteristics (fuel and water), with the difference that one of them has a control rod. In the part corresponding to two-dimensional problems, the implemented nodal methods were those designated as hybrids that consider not only the edge and cell Legendre moments, but also the values of the neutron flux in the corner points

  19. Neutron--neutron logging

    International Nuclear Information System (INIS)

    Allen, L.S.

    1977-01-01

    A borehole logging tool includes a steady-state source of fast neutrons, two epithermal neutron detectors, and two thermal neutron detectors. A count rate meter is connected to each neutron detector. A first ratio detector provides an indication of the porosity of the formation surrounding the borehole by determining the ratio of the outputs of the two count rate meters connected to the two epithermal neutron detectors. A second ratio detector provides an indication of both porosity and macroscopic absorption cross section of the formation surrounding the borehole by determining the ratio of the outputs of the two count rate meters connected to the two thermal neutron detectors. By comparing the signals of the two ratio detectors, oil bearing zones and salt water bearing zones within the formation being logged can be distinguished and the amount of oil saturation can be determined. 6 claims, 2 figures

  20. On the Nodal Lines of Eisenstein Series on Schottky Surfaces

    Science.gov (United States)

    Jakobson, Dmitry; Naud, Frédéric

    2017-04-01

    On convex co-compact hyperbolic surfaces {X=Γ backslash H2}, we investigate the behavior of nodal curves of real valued Eisenstein series {F_λ(z,ξ)}, where {λ} is the spectral parameter, {ξ} the direction at infinity. Eisenstein series are (non-{L^2}) eigenfunctions of the Laplacian {Δ_X} satisfying {Δ_X F_λ=(1/4+λ^2)F_λ}. As {λ} goes to infinity (the high energy limit), we show that, for generic {ξ}, the number of intersections of nodal lines with any compact segment of geodesic grows like {λ}, up to multiplicative constants. Applications to the number of nodal domains inside the convex core of the surface are then derived.

  1. A computational study of nodal-based tetrahedral element behavior.

    Energy Technology Data Exchange (ETDEWEB)

    Gullerud, Arne S.

    2010-09-01

    This report explores the behavior of nodal-based tetrahedral elements on six sample problems, and compares their solution to that of a corresponding hexahedral mesh. The problems demonstrate that while certain aspects of the solution field for the nodal-based tetrahedrons provide good quality results, the pressure field tends to be of poor quality. Results appear to be strongly affected by the connectivity of the tetrahedral elements. Simulations that rely on the pressure field, such as those which use material models that are dependent on the pressure (e.g. equation-of-state models), can generate erroneous results. Remeshing can also be strongly affected by these issues. The nodal-based test elements as they currently stand need to be used with caution to ensure that their numerical deficiencies do not adversely affect critical values of interest.

  2. A theoretical study on a convergence problem of nodal methods

    Energy Technology Data Exchange (ETDEWEB)

    Shaohong, Z.; Ziyong, L. [Shanghai Jiao Tong Univ., 1954 Hua Shan Road, Shanghai, 200030 (China); Chao, Y. A. [Westinghouse Electric Company, P. O. Box 355, Pittsburgh, PA 15230-0355 (United States)

    2006-07-01

    The effectiveness of modern nodal methods is largely due to its use of the information from the analytical flux solution inside a homogeneous node. As a result, the nodal coupling coefficients depend explicitly or implicitly on the evolving Eigen-value of a problem during its solution iteration process. This poses an inherently non-linear matrix Eigen-value iteration problem. This paper points out analytically that, whenever the half wave length of an evolving node interior analytic solution becomes smaller than the size of that node, this non-linear iteration problem can become inherently unstable and theoretically can always be non-convergent or converge to higher order harmonics. This phenomenon is confirmed, demonstrated and analyzed via the simplest 1-D problem solved by the simplest analytic nodal method, the Analytic Coarse Mesh Finite Difference (ACMFD, [1]) method. (authors)

  3. Solution of the transport equation in stationary state, in one and two dimensions, for BWR assemblies using nodal methods

    International Nuclear Information System (INIS)

    Xolocostli M, J.V.

    2002-01-01

    The main objective of this work is to solve the neutron transport equation in one and two dimensions (slab geometry and X Y geometry, respectively), with no time dependence, for BWR assemblies using nodal methods. In slab geometry, the nodal methods here used are the polynomial continuous (CMPk) and discontinuous (DMPk) families but only the Linear Continuous (also known as Diamond Difference), the Quadratic Continuous (QC), the Cubic Continuous (CC), the Step Discontinuous (also known as Backward Euler), the Linear Discontinuous (LD) and the Quadratic Discontinuous (QD) were considered. In all these schemes the unknown function, the angular neutron flux, is approximated as a sum of basis functions in terms of Legendre polynomials, associated to the values of the neutron flux in the edges (left, right, or both) and the Legendre moments in the cell, depending on the nodal scheme used. All these schemes were implemented in a computer program developed in previous thesis works and known with the name TNX. This program was modified for the purposes of this work. The program discreetizes the domain of concern in one dimension and determines numerically the angular neutron flux for each point of the discretization when the number of energy groups and regions are known starting from an initial approximation for the angular neutron flux being consistent with the boundary condition imposed for a given problem. Although only problems with two-energy groups were studied the computer program does not have limitations regarding the number of energy groups and the number of regions. The two problems analyzed with the program TNX have practically the same characteristics (fuel and water), with the difference that one of them has a control rod. In the part corresponding to two-dimensional problems, the implemented nodal methods were those designated as hybrids that consider not only the edge and cell Legendre moments, but also the values of the neutron flux in the corner points

  4. Wielandt method applied to the diffusion equations discretized by finite element nodal methods; Metodo de Wielandt aplicado a las ecuaciones de difusion discretizadas por metodos nodales de elemento finito

    Energy Technology Data Exchange (ETDEWEB)

    Mugica R, A.; Valle G, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)]. e-mail: mugica@esfm.ipn.mx

    2003-07-01

    Nowadays the numerical methods of solution to the diffusion equation by means of algorithms and computer programs result so extensive due to the great number of routines and calculations that should carry out, this rebounds directly in the execution times of this programs, being obtained results in relatively long times. This work shows the application of an acceleration method of the convergence of the classic method of those powers that it reduces notably the number of necessary iterations for to obtain reliable results, what means that the compute times they see reduced in great measure. This method is known in the literature like Wielandt method and it has incorporated to a computer program that is based on the discretization of the neutron diffusion equations in plate geometry and stationary state by polynomial nodal methods. In this work the neutron diffusion equations are described for several energy groups and their discretization by means of those called physical nodal methods, being illustrated in particular the quadratic case. It is described a model problem widely described in the literature which is solved for the physical nodal grade schemes 1, 2, 3 and 4 in three different ways: to) with the classic method of the powers, b) method of the powers with the Wielandt acceleration and c) method of the powers with the Wielandt modified acceleration. The results for the model problem as well as for two additional problems known as benchmark problems are reported. Such acceleration method can also be implemented to problems of different geometry to the proposal in this work, besides being possible to extend their application to problems in 2 or 3 dimensions. (Author)

  5. Three Mile Island Unit 1 Main Steam Line Break Three-Dimensional Neutronics/Thermal-Hydraulics Analysis: Application of Different Coupled Codes

    International Nuclear Information System (INIS)

    D'Auria, Francesco; Moreno, Jose Luis Gago; Galassi, Giorgio Maria; Grgic, Davor; Spadoni, Antonino

    2003-01-01

    A comprehensive analysis of the double ended main steam line break (MSLB) accident assumed to occur in the Babcock and Wilcox Three Mile Island Unit 1 (TMI-1) has been carried out at the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione of the University of Pisa, Italy, in cooperation with the University of Zagreb, Croatia. The overall activity has been completed within the framework of the participation in the Organization for Economic Cooperation and Development-Committee on the Safety of Nuclear Installations-Nuclear Science Committee pressurized water reactor MSLB benchmark.Thermal-hydraulic system codes (various versions of Relap5), three-dimensional (3-D) neutronics codes (Parcs, Quabbox, and Nestle), and one subchannel code (Cobra) have been adopted for the analysis. Results from the following codes (or code versions) are assumed as reference:1. Relap5/mod3.2.2, beta version, coupled with the 3-D neutron kinetics Parcs code parallel virtual machine (PVM) coupling2. Relap5/mod3.2.2, gamma version, coupled with the 3-D neutron kinetics Quabbox code (direct coupling)3. Relap5/3D code coupled with the 3-D neutron kinetics Nestle code.The influence of PVM and of direct coupling is also discussed.Boundary and initial conditions of the system, including those relevant to the fuel status, have been supplied by Pennsylvania State University in cooperation with GPU Nuclear Corporation (the utility, owner of TMI) and the U.S. Nuclear Regulatory Commission. The comparison among the results obtained by adopting the same thermal-hydraulic nodalization and the coupled code version is discussed in this paper.The capability of the control rods to recover the accident has been demonstrated in all the cases as well as the capability of all the codes to predict the time evolution of the assigned transient. However, one stuck control rod caused some 'recriticality' or 'return to power' whose magnitude is largely affected by boundary and initial conditions

  6. Oddness of least energy nodal solutions on radial domains

    Directory of Open Access Journals (Sweden)

    Christopher Grumiau

    2010-07-01

    Full Text Available In this article, we consider the Lane-Emden problem $$displaylines{ Delta u(x + |{u(x}mathclose|^{p-2}u(x=0, quad hbox{for } xinOmega,cr u(x=0, quad hbox{for } xinpartialOmega, }$$ where $2 < p < 2^{*}$ and $Omega$ is a ball or an annulus in $mathbb{R}^{N}$, $Ngeq 2$. We show that, for p close to 2, least energy nodal solutions are odd with respect to an hyperplane -- which is their nodal surface. The proof ingredients are a constrained implicit function theorem and the fact that the second eigenvalue is simple up to rotations.

  7. Development and validation of a nodal code for core calculation

    International Nuclear Information System (INIS)

    Nowakowski, Pedro Mariano

    2004-01-01

    The code RHENO solves the multigroup three-dimensional diffusion equation using a nodal method of polynomial expansion.A comparative study has been made between this code and present internationals nodal diffusion codes, resulting that the RHENO is up to date.The RHENO has been integrated to a calculation line and has been extend to make burnup calculations.Two methods for pin power reconstruction were developed: modulation and imbedded. The modulation method has been implemented in a program, while the implementation of the imbedded method will be concluded shortly.The validation carried out (that includes experimental data of a MPR) show very good results and calculation efficiency

  8. Nodal algorithm derived from a new variational principle

    International Nuclear Information System (INIS)

    Watson, Fernando V.

    1995-01-01

    As a by-product of the research being carried on by the author on methods of recovering pin power distribution of PWR cores, a nodal algorithm based on a modified variational principle for the two group diffusion equations has been obtained. The main feature of the new algorithm is the low dimensionality achieved by the reduction of the original diffusion equations to a system of algebraic Eigen equations involving the average sources only, instead of sources and interface group currents used in conventional nodal methods. The advantage of this procedure is discussed and results generated by the new algorithm and by a finite difference code are compared. (author). 2 refs, 7 tabs

  9. Regional Nodal Irradiation in Early-Stage Breast Cancer.

    Science.gov (United States)

    Whelan, Timothy J; Olivotto, Ivo A; Parulekar, Wendy R; Ackerman, Ida; Chua, Boon H; Nabid, Abdenour; Vallis, Katherine A; White, Julia R; Rousseau, Pierre; Fortin, Andre; Pierce, Lori J; Manchul, Lee; Chafe, Susan; Nolan, Maureen C; Craighead, Peter; Bowen, Julie; McCready, David R; Pritchard, Kathleen I; Gelmon, Karen; Murray, Yvonne; Chapman, Judy-Anne W; Chen, Bingshu E; Levine, Mark N

    2015-07-23

    Most women with breast cancer who undergo breast-conserving surgery receive whole-breast irradiation. We examined whether the addition of regional nodal irradiation to whole-breast irradiation improved outcomes. We randomly assigned women with node-positive or high-risk node-negative breast cancer who were treated with breast-conserving surgery and adjuvant systemic therapy to undergo either whole-breast irradiation plus regional nodal irradiation (including internal mammary, supraclavicular, and axillary lymph nodes) (nodal-irradiation group) or whole-breast irradiation alone (control group). The primary outcome was overall survival. Secondary outcomes were disease-free survival, isolated locoregional disease-free survival, and distant disease-free survival. Between March 2000 and February 2007, a total of 1832 women were assigned to the nodal-irradiation group or the control group (916 women in each group). The median follow-up was 9.5 years. At the 10-year follow-up, there was no significant between-group difference in survival, with a rate of 82.8% in the nodal-irradiation group and 81.8% in the control group (hazard ratio, 0.91; 95% confidence interval [CI], 0.72 to 1.13; P=0.38). The rates of disease-free survival were 82.0% in the nodal-irradiation group and 77.0% in the control group (hazard ratio, 0.76; 95% CI, 0.61 to 0.94; P=0.01). Patients in the nodal-irradiation group had higher rates of grade 2 or greater acute pneumonitis (1.2% vs. 0.2%, P=0.01) and lymphedema (8.4% vs. 4.5%, P=0.001). Among women with node-positive or high-risk node-negative breast cancer, the addition of regional nodal irradiation to whole-breast irradiation did not improve overall survival but reduced the rate of breast-cancer recurrence. (Funded by the Canadian Cancer Society Research Institute and others; MA.20 ClinicalTrials.gov number, NCT00005957.).

  10. Prompt neutrons from {sup 236}U fission fragments

    Energy Technology Data Exchange (ETDEWEB)

    Boldeman, J W; Musgrove, A.R. de L.; Walsch, R L

    1971-03-01

    Measurements were made of prompt neutron emission in the thermal neutron fission of {sup 235}U. The mean neutron emission per fragment was obtained for particular values of the fragment mass and total kinetic energy. A direct neutron counting method was employed and a comparison made with data from previous experiments of this type. (author)

  11. Influence of the external neutron sources in the criticality prediction using 1/M curve

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, Valmir [COPPE/UFRJ, Programa de Engenharia Nuclear, Caixa Postal 68509, 21941-972 Rio de Janeiro (Brazil); Carvalho da Silva, Fernando [COPPE/UFRJ, Programa de Engenharia Nuclear, Caixa Postal 68509, 21941-972 Rio de Janeiro (Brazil); Martinez, Aquilino Senra [COPPE/UFRJ, Programa de Engenharia Nuclear, Caixa Postal 68509, 21941-972 Rio de Janeiro (Brazil)]. E-mail: aquilino@lmp.ufrj.br

    2005-11-15

    The influence of external neutron sources in the process to obtain the criticality condition is estimated. To reach this objective, the three-dimensional neutron diffusion equation in two groups of energy is solved, for a subcritical PWR reactor core with external neutron sources. The results are compared with the solution of the corresponding problem without external neutron sources, that is an eigenvalue problem. The method developed for this purposes it makes use of both the nodal method (for calculation of the neutron flux) and the finite differences method (for calculation of the adjoint flux). A coarse mesh finite difference method was developed for the adjoint flux calculation, which uses the output of the nodal expansion method. The results regarding the influence of the external neutron source presence for attaining criticality have shown that far from criticality it is necessary to calculate the reactivity values of the system.

  12. Influence of the external neutron sources in the criticality prediction using 1/M curve

    International Nuclear Information System (INIS)

    Pereira, Valmir; Carvalho da Silva, Fernando; Martinez, Aquilino Senra

    2005-01-01

    The influence of external neutron sources in the process to obtain the criticality condition is estimated. To reach this objective, the three-dimensional neutron diffusion equation in two groups of energy is solved, for a subcritical PWR reactor core with external neutron sources. The results are compared with the solution of the corresponding problem without external neutron sources, that is an eigenvalue problem. The method developed for this purposes it makes use of both the nodal method (for calculation of the neutron flux) and the finite differences method (for calculation of the adjoint flux). A coarse mesh finite difference method was developed for the adjoint flux calculation, which uses the output of the nodal expansion method. The results regarding the influence of the external neutron source presence for attaining criticality have shown that far from criticality it is necessary to calculate the reactivity values of the system

  13. ATLAS MDT neutron sensitivity measurement and modeling

    International Nuclear Information System (INIS)

    Ahlen, S.; Hu, G.; Osborne, D.; Schulz, A.; Shank, J.; Xu, Q.; Zhou, B.

    2003-01-01

    The sensitivity of the ATLAS precision muon detector element, the Monitored Drift Tube (MDT), to fast neutrons has been measured using a 5.5 MeV Van de Graaff accelerator. The major mechanism of neutron-induced signals in the drift tubes is the elastic collisions between the neutrons and the gas nuclei. The recoil nuclei lose kinetic energy in the gas and produce the signals. By measuring the ATLAS drift tube neutron-induced signal rate and the total neutron flux, the MDT neutron signal sensitivities were determined for different drift gas mixtures and for different neutron beam energies. We also developed a sophisticated simulation model to calculate the neutron-induced signal rate and signal spectrum for ATLAS MDT operation configurations. The calculations agree with the measurements very well. This model can be used to calculate the neutron sensitivities for different gaseous detectors and for neutron energies above those available to this experiment

  14. Kinetics of Ar isotopes during neutron irradiation: 39Ar loss from minerals as a source of error in 40Ar/39Ar dating

    International Nuclear Information System (INIS)

    Hess, J.C.; Lippolt, H.J.

    1986-01-01

    The loss of 39 Ar from minerals in the course of neutron activation for 40 Ar/ 39 Ar dating is studied by directly measuring the loss rates in vacuum-sealed ampoules. Biotite shows 39 Ar losses between 0.1% and 16%. These losses are predominantly due to diffusion processes from K-poor alteration-phase intergrowths in the biotites at the elevated temperatures during the irradiation. Estimates for the irradiation temperatures range from 150 0 to 180 0 C. Direct 39 Ar recoil loss from biotite seems to be minor compared to difussion loss of recoil-implanted 39 Ar. Precise 40 Ar/ 39 Ar dating of biotites therefore requires the measurement of the 39 Ar losses during irradiation. Glauconite loses not only neutron-induced Ar isotopes ( 39 Ar: 20-22%, 37 Ar: 17-19%) but also radiogenic 40 Ar(∼9%). Slight 39 Ar losses are also observed for light micas (0.2% and 0.35%), hornblendes (0.1%) and sanidines (200 and 700 ppm). 25 refs.; 4 figs.; 6 tabs

  15. A Nodal and Finite Difference Hybrid Method for Pin-by-Pin Heterogeneous Three-Dimensional Light Water Reactor Diffusion Calculations

    International Nuclear Information System (INIS)

    Lee, Deokjung; Downar, Thomas J.; Kim, Yonghee

    2004-01-01

    An innovative hybrid spatial discretization method is proposed to improve the computational efficiency of pin-wise heterogeneous three-dimensional light water reactor (LWR) core neutronics analysis. The newly developed method employs the standard finite difference method in the x and y directions and the well-known nodal methods [nodal expansion method (NEM) and analytic nodal method (ANM) as needed] in the z direction. Four variants of the hybrid method are investigated depending on the axial nodal methodologies: HYBRID A, NEM with the conventional quadratic transverse leakage; HYBRID B, the conventional NEM method except that the transverse-leakage shapes are obtained from a fine-mesh local problem (FMLP) around the control rod tip; HYBRID C, the same as HYBRID B except that ANM with a high-order transverse leakage obtained from the FMLP is used in the vicinity of the control rod tip; and HYBRID D, the same as HYBRID C except that the transverse leakage is determined using the buckling approximation instead of the FMLP around the control rod tip. Benchmark calculations demonstrate that all the hybrid algorithms are consistent and stable and that the HYBRID C method provides the best numerical performance in the case of rodded LWR problems with pin-wise homogenized cross sections

  16. The merits of DNA content and cell kinetic parameters for the assessment of intrinsic cellular radiosensitivity to photon and high-LET neutron irradiation

    International Nuclear Information System (INIS)

    Theron, C.S.; Serafin, A.; Bohm, L.; Slabbert, J.P.

    1997-01-01

    Differences of the intrinsic cellular radiosensitivity between tumours make the selection of patients for specific radiation schedules very difficult. The reasons for these variations are still unclear, but are thought to be due to genomic and cellular characteristics. Radiosensitivities vary between cell cycle stages, with S-phase cells being most radioresistant and G2/M phase cells most radiosensitive. It is also well established that most tumour cells have an abnormal ploidy. DNA content and cellular proliferation kinetics therefore could influence the intrinsic radiosensitivity. This prompted us to assess the merits of these parameters as predictors of radiation response. (authors)

  17. Magnetohydrodynamics of neutron star interiors

    International Nuclear Information System (INIS)

    Easson, I.; Pethick, C.J.

    1979-01-01

    Magnetohydrodynamic equations for the charged particles in the fluid interior of a neutron star are derived from the Landau-Boltzmann kinetic equations. It is assumed that the protons are normal and the neutrons are superfluid. The dissipative processes associated with the weak interactions are shown to be negligible except in very hot neutron stars; we neglect them here. Among the topics discussed are: the influence of the neutron-proton nuclear force (Fermi liquid corrections) on the magnetohydrodynamics; the effects of the magnetic field on the pressure, viscosity, and heat conductivity tensors; the plasma equation of state; and the form of the generalized Ohm's law

  18. Extension of the linear nodal method to large concrete building calculations

    International Nuclear Information System (INIS)

    Childs, R.L.; Rhoades, W.A.

    1985-01-01

    The implementation of the linear nodal method in the TORT code is described, and the results of a mesh refinement study to test the effectiveness of the linear nodal and weighted diamond difference methods available in TORT are presented

  19. Experimental determination of neutron lifetimes through macroscopic neutron noise in the IPEN/MB-01 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gonnelli, Eduardo; Diniz, Ricardo [Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP Travessa R-400, 05508-900, Cidade Universitaria, Sao Paulo (Brazil)

    2013-05-06

    The neutron lifetimes of the core, reflector, and global were experimentally obtained through macroscopic neutron noise in the IPEN/MB-01 reactor for five levels of subcriticality. The theoretical Auto Power Spectral Densities were derived by point kinetic equations taking the reflector effect into account, and one of the approaches consider an additional group of delayed neutrons.

  20. The Nodal Location of Metastases in Melanoma Sentinel Lymph Nodes

    DEFF Research Database (Denmark)

    Riber-Hansen, Rikke; Nyengaard, Jens; Hamilton-Dutoit, Stephen

    2009-01-01

    BACKGROUND: The design of melanoma sentinel lymph node (SLN) histologic protocols is based on the premise that most metastases are found in the central parts of the nodes, but the evidence for this belief has never been thoroughly tested. METHODS: The nodal location of melanoma metastases in 149...

  1. Real-time control of power systems using nodal prices

    NARCIS (Netherlands)

    Jokic, A.; Lazar, M.; Bosch, van den P.P.J.

    2009-01-01

    This article presents a novel control scheme for achieving optimal power balancing and congestion management in electrical power systems via nodal prices. We develop a dynamic controller that guarantees economically optimal steady-state operation while respecting all line flow constraints in

  2. Topological transport in Dirac nodal-line semimetals

    Science.gov (United States)

    Rui, W. B.; Zhao, Y. X.; Schnyder, Andreas P.

    2018-04-01

    Topological nodal-line semimetals are characterized by one-dimensional Dirac nodal rings that are protected by the combined symmetry of inversion P and time-reversal T . The stability of these Dirac rings is guaranteed by a quantized ±π Berry phase and their low-energy physics is described by a one-parameter family of (2+1)-dimensional quantum field theories exhibiting the parity anomaly. Here we study the Berry-phase supported topological transport of P T -invariant nodal-line semimetals. We find that small inversion breaking allows for an electric-field-induced anomalous transverse current, whose universal component originates from the parity anomaly. Due to this Hall-like current, carriers at opposite sides of the Dirac nodal ring flow to opposite surfaces when an electric field is applied. To detect the topological currents, we propose a dumbbell device, which uses surface states to filter charges based on their momenta. Suggestions for experiments and device applications are discussed.

  3. Hyoid bone chondrosarcoma with cervical nodal metastasis: A case ...

    African Journals Online (AJOL)

    Background: Hyoid bone chondrosarcoma is a very rare condition. This study presents a case report of low-grade chondrosarcoma of hyoid bone with cervical nodal metastasis. The study also presents preoperative radiological investigations, pathological examination and the follow-up of the case. Case presentation: A 42 ...

  4. Note on the nodal line of the p-Laplacian

    Directory of Open Access Journals (Sweden)

    Abdel R. El Amrouss

    2006-09-01

    Full Text Available In this paper, we prove that the length of the nodal line of the eigenfunctions associated to the second eigenvalue of the problem $$ -Delta_p u = lambda ho (x |u|^{p-2}u quad hbox{in } Omega $$ with the Dirichlet conditions is not bounded uniformly with respect to the weight.

  5. A nodal method based on the response-matrix method

    International Nuclear Information System (INIS)

    Cunha Menezes Filho, A. da; Rocamora Junior, F.D.

    1983-02-01

    A nodal approach based on the Response-Matrix method is presented with the purpose of investigating the possibility of mixing two different allocations in the same problem. It is found that the use of allocation of albedo combined with allocation of direct reflection produces good results for homogeneous fast reactor configurations. (Author) [pt

  6. CRY 1AB trangenic cowpea obtained by nodal electroporation ...

    African Journals Online (AJOL)

    Electroporation-mediated genetic transformation was used to introduce Cry 1 Ab insecticidal gene into cowpea. Nodal buds were electroporated in planta with a plasmid carrying the Cry 1Ab and antibiotic resistance npt II genes driven by a 35S CaMV promoter. T1 seeds derived from electroporated branches were selected ...

  7. Ultracold neutrons

    International Nuclear Information System (INIS)

    Steenstrup, S.

    Briefly surveys recent developments in research work with ultracold neutrons (neutrons of very low velocity, up to 10 m/s at up to 10 -7 eV and 10 -3 K). Slow neutrons can be detected in an ionisation chamber filled with B 10 F 3 . Very slow neutrons can be used for investigations into the dipole moment of neutrons. Neutrons of large wave length have properties similar to those of light. The limit angle for total reflection is governed by the wave length and by the material. Total reflection can be used to filter ultracold neutrons out of the moderator material of a reactor. Total reflection can also be used to store ultracold neutrons but certain problems with storage have not yet been clarified. Slow neutrons can be made to lose speed in a neutron turbine, and come out as ultracold neutrons. A beam of ultracold neutrons could be used in a neutron microscope. (J.S.)

  8. Modification of the ANC Nodal Code for analysis of PWR assembly bow

    International Nuclear Information System (INIS)

    Franceschini, Fausto; Fetterman, Robert J.; Little, David C.

    2008-01-01

    Refueling operations at certain PWR cores have revealed fuel assemblies with assembly bow that was higher than expected. As the fuel assemblies bow, the gaps between assemblies change from the uniform nominal configuration. This causes a change in the water volume which affects neutron moderation and thereby power distribution, fuel depletion history, rod internal pressure, etc., with non-trivial impacts on the safety analysis. Westinghouse has developed a new methodology for incorporation of assembly bow in its reload safety analysis package. As part of the new process, the standard Westinghouse reactor physics tool for core analysis, the Advanced Nodal Code ANC, has been modified. The modified ANC, ANCGAP, enables explicit treatment of three-dimensional gap distributions in its neutronic calculations; its accuracy is similar to that of the standard ANC, as demonstrated through an extensive benchmark campaign conducted over a variety of fuel compositions and challenging gap configurations. These features make ANCGAP a crucial tool in the Westinghouse assembly bow package. (authors)

  9. Modification of the ANC Nodal Code for analysis of PWR assembly bow

    Energy Technology Data Exchange (ETDEWEB)

    Franceschini, Fausto; Fetterman, Robert J.; Little, David C. [Westinghouse Electric Company LLC, Pittsburgh PA (United States)

    2008-07-01

    Refueling operations at certain PWR cores have revealed fuel assemblies with assembly bow that was higher than expected. As the fuel assemblies bow, the gaps between assemblies change from the uniform nominal configuration. This causes a change in the water volume which affects neutron moderation and thereby power distribution, fuel depletion history, rod internal pressure, etc., with non-trivial impacts on the safety analysis. Westinghouse has developed a new methodology for incorporation of assembly bow in its reload safety analysis package. As part of the new process, the standard Westinghouse reactor physics tool for core analysis, the Advanced Nodal Code ANC, has been modified. The modified ANC, ANCGAP, enables explicit treatment of three-dimensional gap distributions in its neutronic calculations; its accuracy is similar to that of the standard ANC, as demonstrated through an extensive benchmark campaign conducted over a variety of fuel compositions and challenging gap configurations. These features make ANCGAP a crucial tool in the Westinghouse assembly bow package. (authors)

  10. 47 CFR 101.503 - Digital Electronic Message Service Nodal Stations.

    Science.gov (United States)

    2010-10-01

    ... Service § 101.503 Digital Electronic Message Service Nodal Stations. 10.6 GHz DEMS Nodal Stations may be... 47 Telecommunication 5 2010-10-01 2010-10-01 false Digital Electronic Message Service Nodal Stations. 101.503 Section 101.503 Telecommunication FEDERAL COMMUNICATIONS COMMISSION (CONTINUED) SAFETY...

  11. A two-dimensional, semi-analytic expansion method for nodal calculations

    International Nuclear Information System (INIS)

    Palmtag, S.P.

    1995-08-01

    Most modern nodal methods used today are based upon the transverse integration procedure in which the multi-dimensional flux shape is integrated over the transverse directions in order to produce a set of coupled one-dimensional flux shapes. The one-dimensional flux shapes are then solved either analytically or by representing the flux shape by a finite polynomial expansion. While these methods have been verified for most light-water reactor applications, they have been found to have difficulty predicting the large thermal flux gradients near the interfaces of highly-enriched MOX fuel assemblies. A new method is presented here in which the neutron flux is represented by a non-seperable, two-dimensional, semi-analytic flux expansion. The main features of this method are (1) the leakage terms from the node are modeled explicitly and therefore, the transverse integration procedure is not used, (2) the corner point flux values for each node are directly edited from the solution method, and a corner-point interpolation is not needed in the flux reconstruction, (3) the thermal flux expansion contains hyperbolic terms representing analytic solutions to the thermal flux diffusion equation, and (4) the thermal flux expansion contains a thermal to fast flux ratio term which reduces the number of polynomial expansion functions needed to represent the thermal flux. This new nodal method has been incorporated into the computer code COLOR2G and has been used to solve a two-dimensional, two-group colorset problem containing uranium and highly-enriched MOX fuel assemblies. The results from this calculation are compared to the results found using a code based on the traditional transverse integration procedure

  12. Analysis of 2D reactor core using linear perturbation theory and nodal finite element methods

    International Nuclear Information System (INIS)

    Adrian Mugica; Edmundo del Valle

    2005-01-01

    In this work the multigroup steady state neutron diffusion equations are solved using the nodal finite element method (NFEM) and the Linear Perturbation Theory (LPT) for XY geometry. The NFEM used corresponds to the Raviart-Thomas schemes RT0 and RT1, interpolating 5 and 12 parameters respectively in each node of the space discretization. The accuracy of these methods is related with the dimension of the space approximation and the mesh size. Therefore, using fine meshes and the RT0 or RT1 nodal methods leads to a large an interesting eigenvalue problem. The finite element method used to discretize the weak formulation of the diffusion equations is the Galerkin one. The algebraic structure of the discrete eigenvalue problem is obtained and solved using the Wielandt technique and the BGSTAB iterative method using the SPARSKIT package developed by Yousef Saad. The results obtained with LPT show good agreement with the results obtained directly for the perturbed problem. In fact, the cpu time to solve a single problem, the unperturbed and the perturbed one, is practically the same but when one is focused in shuffling many times two different assemblies in the core then the LPT technique becomes quite useful to get good approximations in a short time. This particular problem was solved for one quarter-core with NFEM. Thus, the computer program based on LPT can be used to perform like an analysis tool in the fuel reload optimization or combinatory analysis to get reload patterns in nuclear power plants once that it had been incorporated with the thermohydraulic aspects needed to simulate accurately a real problem. The maximum differences between the NFEM and LPT for the three LWR reactor cores are about 250 pcm. This quantity is considered an acceptable value for this kind of analysis. (authors)

  13. Non-linear triangle-based polynomial expansion nodal method for hexagonal core analysis

    International Nuclear Information System (INIS)

    Cho, Jin Young; Cho, Byung Oh; Joo, Han Gyu; Zee, Sung Qunn; Park, Sang Yong

    2000-09-01

    This report is for the implementation of triangle-based polynomial expansion nodal (TPEN) method to MASTER code in conjunction with the coarse mesh finite difference(CMFD) framework for hexagonal core design and analysis. The TPEN method is a variation of the higher order polynomial expansion nodal (HOPEN) method that solves the multi-group neutron diffusion equation in the hexagonal-z geometry. In contrast with the HOPEN method, only two-dimensional intranodal expansion is considered in the TPEN method for a triangular domain. The axial dependence of the intranodal flux is incorporated separately here and it is determined by the nodal expansion method (NEM) for a hexagonal node. For the consistency of node geometry of the MASTER code which is based on hexagon, TPEN solver is coded to solve one hexagonal node which is composed of 6 triangular nodes directly with Gauss elimination scheme. To solve the CMFD linear system efficiently, stabilized bi-conjugate gradient(BiCG) algorithm and Wielandt eigenvalue shift method are adopted. And for the construction of the efficient preconditioner of BiCG algorithm, the incomplete LU(ILU) factorization scheme which has been widely used in two-dimensional problems is used. To apply the ILU factorization scheme to three-dimensional problem, a symmetric Gauss-Seidel Factorization scheme is used. In order to examine the accuracy of the TPEN solution, several eigenvalue benchmark problems and two transient problems, i.e., a realistic VVER1000 and VVER440 rod ejection benchmark problems, were solved and compared with respective references. The results of eigenvalue benchmark problems indicate that non-linear TPEN method is very accurate showing less than 15 pcm of eigenvalue errors and 1% of maximum power errors, and fast enough to solve the three-dimensional VVER-440 problem within 5 seconds on 733MHz PENTIUM-III. In the case of the transient problems, the non-linear TPEN method also shows good results within a few minute of

  14. Intercomparison of the finite difference and nodal discrete ordinates and surface flux transport methods for a LWR pool-reactor benchmark problem in X-Y geometry

    International Nuclear Information System (INIS)

    O'Dell, R.D.; Stepanek, J.; Wagner, M.R.

    1983-01-01

    The aim of the present work is to compare and discuss the three of the most advanced two dimensional transport methods, the finite difference and nodal discrete ordinates and surface flux method, incorporated into the transport codes TWODANT, TWOTRAN-NODAL, MULTIMEDIUM and SURCU. For intercomparison the eigenvalue and the neutron flux distribution are calculated using these codes in the LWR pool reactor benchmark problem. Additionally the results are compared with some results obtained by French collision probability transport codes MARSYAS and TRIDENT. Because the transport solution of this benchmark problem is close to its diffusion solution some results obtained by the finite element diffusion code FINELM and the finite difference diffusion code DIFF-2D are included

  15. The variational nodal method: history and recent accomplishments

    International Nuclear Information System (INIS)

    Lewis, E.E.

    2004-01-01

    The variational nodal method combines spherical harmonics expansions in angle with hybrid finite element techniques is space to obtain multigroup transport response matrix algorithms applicable to both deep penetration and reactor core physics problems. This survey briefly recounts the method's history and reviews its capabilities. The variational basis for the approach is presented and two methods for obtaining discretized equations in the form of response matrices are detailed. The first is that contained the widely used VARIANT code, while the second incorporates newly developed integral transport techniques into the variational nodal framework. The two approaches are combined with a finite sub element formulation to treat heterogeneous nodes. Applications are presented for both a deep penetration problem and to an OECD benchmark consisting of LWR MOX fuel assemblies. Ongoing work is discussed. (Author)

  16. Radiological signs of extra nodal abdominal involvements in lymphoma

    International Nuclear Information System (INIS)

    Carro, A.I.; Alegre, N.; Cervera, J.L.; Montero, A.I.

    1998-01-01

    To assess abdominal CT images in lymphoma patients for the study of extra nodal abdominal involvement. Ninety-two patients diagnosed as having lymphoma were studied retrospectively. All the patients underwent abdominopelvic CT with oral and intravenous contrast (except in one patient who was allergic). In every case, the diagnosis was confirmed by biopsy or radiological follow-up after treatment had been completed. Fifty-two patients (56.5%) presented infiltration of extra nodal organs. The organs most frequently involved were liver and spleen, followed by the gastrointestinal tract, the musculoskeletal system and the genitourinary tract. The findings in this study coincide with those reported elsewhere with the exception of the splenic involvement the incidence of which was lower in the present series. (Author) 17 refs

  17. On the nodal structure of atomic and molecular Wigner functions

    International Nuclear Information System (INIS)

    Dahl, J.P.; Schmider, H.

    1996-01-01

    In previous work on the phase-space representation of quantum mechanics, we have presented detailed pictures of the electronic one-particle reduced Wigner function for atoms and small molecules. In this communication, we focus upon the nodal structure of the function. On the basis of the simplest systems, we present an expression which relates the oscillatory decay of the Wigner function solely to the dot product of the position and momentum vector, if both arguments are large. We then demonstrate the regular behavior of nodal patterns for the larger systems. For the molecular systems, an argument analogous to the open-quotes bond-oscillatory principleclose quotes for momentum densities links the nuclear framework to an additional oscillatory term in momenta parallel to bonds. It is shown that these are visible in the Wigner function in terms of characteristic nodes

  18. Topological and trivial magnetic oscillations in nodal loop semimetals

    Science.gov (United States)

    Oroszlány, László; Dóra, Balázs; Cserti, József; Cortijo, Alberto

    2018-05-01

    Nodal loop semimetals are close descendants of Weyl semimetals and possess a topologically dressed band structure. We argue by combining the conventional theory of magnetic oscillation with topological arguments that nodal loop semimetals host coexisting topological and trivial magnetic oscillations. These originate from mapping the topological properties of the extremal Fermi surface cross sections onto the physics of two dimensional semi-Dirac systems, stemming from merging two massless Dirac cones. By tuning the chemical potential and the direction of magnetic field, a sharp transition is identified from purely trivial oscillations, arising from the Landau levels of a normal two dimensional (2D) electron gas, to a phase where oscillations of topological and trivial origin coexist, originating from 2D massless Dirac and semi-Dirac points, respectively. These could in principle be directly identified in current experiments.

  19. Ischemic stroke associated with radio frequency ablation for nodal reentry

    International Nuclear Information System (INIS)

    Diaz M, Juan C; Duran R, Carlos E; Perafan B, Pablo; Pava M, Luis F

    2010-01-01

    Atrioventricular nodal reentry tachycardia is the most common type of paroxysmal supraventricular tachycardia. In those patients in whom drug therapy is not effective or not desired, radio frequency ablation is an excellent therapeutic method. Although overall these procedures are fast and safe, several complications among which ischemic stroke stands out, have been reported. We present the case of a 41 year old female patient with repetitive episodes of tachycardia due to nodal reentry who was treated with radiofrequency ablation. Immediately after the procedure she presented focal neurologic deficit consistent with ischemic stroke in the right medial cerebral artery territory. Angiography with angioplastia and abxicimab was performed and then tissue plasminogen activator (rtPA) was locally infused, with appropriate clinical and angiographic outcome.

  20. RELAP 4/MOD 6 boiling water nodalization study

    International Nuclear Information System (INIS)

    Sonneck, G.; Pfau, H.

    1985-09-01

    The risk of nuclear steam supply systems is dominated by the core melt accidents. The first step to a realistic assessment of these sequences is the successful prediction of a loss of coolant event in a test loop. One of the codes for that is RELAP 4/MOD 6 and one of the important options in this code is the nodalization. The base of this work is the test LOCA No. 1 FIX II in Studsvik (Sweden) which also served as the OECD International Standard Problem 15. This report discusses the influence of different nodalizations, of different distributions of pressure, water and structural heat as well as of different bubble rise options, break flow coefficients, and heat transfer time steps. The most important result is that a simple RELAP 4/MOD6 model with less than 10 volumes is able to predict an experiment as LOCA No. 1 in FIX II successfully using only a fraction of the usual computing time. (Author)

  1. The variational nodal method: some history and recent activity

    International Nuclear Information System (INIS)

    Lewis, E.E.; Smith, M.A.; Palmiotti, G.

    2005-01-01

    The variational nodal method combines spherical harmonics expansions in angle with hybrid finite element techniques in space to obtain multigroup transport response matrix algorithms applicable to a wide variety of reactor physics problems. This survey briefly recounts the method's history and reviews its capabilities. Two methods for obtaining discretized equations in the form of response matrices are compared. The first is that contained the widely used VARIANT code, while the second incorporates more recently developed integral transport techniques into the variational nodal framework. The two approaches are combined with a finite sub-element formulation to treat heterogeneous nodes. Results are presented for application to a deep penetration problem and to an OECD benchmark consisting of LWR Mox fuel assemblies. Ongoing work is discussed. (authors)

  2. Hybrid microscopic depletion model in nodal code DYN3D

    International Nuclear Information System (INIS)

    Bilodid, Y.; Kotlyar, D.; Shwageraus, E.; Fridman, E.; Kliem, S.

    2016-01-01

    Highlights: • A new hybrid method of accounting for spectral history effects is proposed. • Local concentrations of over 1000 nuclides are calculated using micro depletion. • The new method is implemented in nodal code DYN3D and verified. - Abstract: The paper presents a general hybrid method that combines the micro-depletion technique with correction of micro- and macro-diffusion parameters to account for the spectral history effects. The fuel in a core is subjected to time- and space-dependent operational conditions (e.g. coolant density), which cannot be predicted in advance. However, lattice codes assume some average conditions to generate cross sections (XS) for nodal diffusion codes such as DYN3D. Deviation of local operational history from average conditions leads to accumulation of errors in XS, which is referred as spectral history effects. Various methods to account for the spectral history effects, such as spectral index, burnup-averaged operational parameters and micro-depletion, were implemented in some nodal codes. Recently, an alternative method, which characterizes fuel depletion state by burnup and 239 Pu concentration (denoted as Pu-correction) was proposed, implemented in nodal code DYN3D and verified for a wide range of history effects. The method is computationally efficient, however, it has applicability limitations. The current study seeks to improve the accuracy and applicability range of Pu-correction method. The proposed hybrid method combines the micro-depletion method with a XS characterization technique similar to the Pu-correction method. The method was implemented in DYN3D and verified on multiple test cases. The results obtained with DYN3D were compared to those obtained with Monte Carlo code Serpent, which was also used to generate the XS. The observed differences are within the statistical uncertainties.

  3. SPANDOM - source projection analytic nodal discrete ordinates method

    International Nuclear Information System (INIS)

    Kim, Tae Hyeong; Cho, Nam Zin

    1994-01-01

    We describe a new discrete ordinates nodal method for the two-dimensional transport equation. We solve the discrete ordinates equation analytically after the source term is projected and represented in polynomials. The method is applied to two fast reactor benchmark problems and compared with the TWOHEX code. The results indicate that the present method accurately predicts not only multiplication factor but also flux distribution

  4. HEXAN - a hexagonal nodal code for solving the diffusion equation

    International Nuclear Information System (INIS)

    Makai, M.

    1982-07-01

    This report describes the theory of and provides a user's manual for the HEXAN program, which is a nodal program for the solution of the few-group diffusion equation in hexagonal geometry. Based upon symmetry considerations, the theory provides an analytical solution in a homogeneous node. WWER and HTGR test problem solutions are presented. The equivalence of the finite-difference scheme and the response matrix method is proven. The properties of a symmetric node's response matrix are investigated. (author)

  5. NOMAD: a nodal microscopic analysis method for nuclear fuel depletion

    International Nuclear Information System (INIS)

    Rajic, H.L.; Ougouag, A.M.

    1987-01-01

    Recently developed assembly homogenization techniques made possible very efficient global burnup calculations based on modern nodal methods. There are two possible ways of modeling the global depletion process: macroscopic and microscopic depletion models. Using a microscopic global depletion approach NOMAD (NOdal Microscopic Analysis Method for Nuclear Fuel Depletion), a multigroup, two- and three-dimensional, multicycle depletion code was devised. The code uses the ILLICO nodal diffusion model. The formalism of the ILLICO methodology is extended to treat changes in the macroscopic cross sections during a depletion cycle without recomputing the coupling coefficients. This results in a computationally very efficient method. The code was tested against a well-known depletion benchmark problem. In this problem a two-dimensional pressurized water reactor is depleted through two cycles. Both cycles were run with 1 x 1 and 2 x 2 nodes per assembly. It is obvious that the one node per assembly solution gives unacceptable results while the 2 x 2 solution gives relative power errors consistently below 2%

  6. Applications of a systematic homogenization theory for nodal diffusion methods

    International Nuclear Information System (INIS)

    Zhang, Hong-bin; Dorning, J.J.

    1992-01-01

    The authors recently have developed a self-consistent and systematic lattice cell and fuel bundle homogenization theory based on a multiple spatial scales asymptotic expansion of the transport equation in the ratio of the mean free path to the reactor characteristics dimension for use with nodal diffusion methods. The mathematical development leads naturally to self-consistent analytical expressions for homogenized diffusion coefficients and cross sections and flux discontinuity factors to be used in nodal diffusion calculations. The expressions for the homogenized nuclear parameters that follow from the systematic homogenization theory (SHT) are different from those for the traditional flux and volume-weighted (FVW) parameters. The calculations summarized here show that the systematic homogenization theory developed recently for nodal diffusion methods yields accurate values for k eff and assembly powers even when compared with the results of a fine mesh transport calculation. Thus, it provides a practical alternative to equivalence theory and GET (Ref. 3) and to simplified equivalence theory, which requires auxiliary fine-mesh calculations for assemblies embedded in a typical environment to determine the discontinuity factors and the equivalent diffusion coefficient for a homogenized assembly

  7. Flow-based market coupling. Stepping stone towards nodal pricing?

    International Nuclear Information System (INIS)

    Van der Welle, A.J.

    2012-07-01

    For achieving one internal energy market for electricity by 2014, market coupling is deployed to integrate national markets into regional markets and ultimately one European electricity market. The extent to which markets can be coupled depends on the available transmission capacities between countries. Since interconnections are congested from time to time, congestion management methods are deployed to divide the scarce available transmission capacities over market participants. For further optimization of the use of available transmission capacities while maintaining current security of supply levels, flow-based market coupling (FBMC) will be implemented in the CWE region by 2013. Although this is an important step forward, important hurdles for efficient congestion management remain. Hence, flow based market coupling is compared to nodal pricing, which is often considered as the most optimal solution from theoretical perspective. In the context of decarbonised power systems it is concluded that advantages of nodal pricing are likely to exceed its disadvantages, warranting further development of FBMC in the direction of nodal pricing.

  8. Neutron reflectometry

    International Nuclear Information System (INIS)

    Van Well, A.A.

    1999-01-01

    Neutron research where reflection, refraction, and interference play an essential role is generally referred to as 'neutron optics'. The neutron wavelength, the scattering length density and the magnetic properties of the material determine the critical angle for total reflection. The theoretical background of neutron reflection, experimental methods and the interpretation of reflection data are presented. (K.A.)

  9. Kinetic Typography

    DEFF Research Database (Denmark)

    van Leeuwen, Theo; Djonov, Emilia

    2014-01-01

    After discussing broad cultural drivers behind the development of kinetic typography, the chapter outlines an approach to analysing kinetic typography which is based on Halliday's theory of transitivity, as applied by Kress and Van Leeuwen to visual images.......After discussing broad cultural drivers behind the development of kinetic typography, the chapter outlines an approach to analysing kinetic typography which is based on Halliday's theory of transitivity, as applied by Kress and Van Leeuwen to visual images....

  10. Patterns of practice of regional nodal irradiation in breast cancer: results of the European Organization for Research and Treatment of Cancer (EORTC) NOdal Radiotherapy (NORA) survey

    NARCIS (Netherlands)

    Belkacemi, Y.; Kaidar-Person, O.; Poortmans, P.; Ozsahin, M.; Valli, M.-C.; Russell, N.; Kunkler, I.; Hermans, J.; Kuten, A.; van Tienhoven, G.; Westenberg, H.

    2015-01-01

    Predicting outcome of breast cancer (BC) patients based on sentinel lymph node (SLN) status without axillary lymph node dissection (ALND) is an area of uncertainty. It influences the decision-making for regional nodal irradiation (RNI). The aim of the NORA (NOdal RAdiotherapy) survey was to examine

  11. Neutron and thermal dynamics of a gaseous core fission reactor

    International Nuclear Information System (INIS)

    van Dam, H.; Kuijper, J.C.; Stekelenburg, A.J.C.; Hoogenboom, J.E.; Boersma-Klein, W.; Kistemaker, J.

    1989-01-01

    In this paper neutron kinetics and thermal dynamics of a Gaseous Core Fission Reactor with magnetical pumping are shown to have many unconventional aspects. Attention is focused on the properties of the fuel gas, the non-linear neutron kinetics and the energy balance in thermodynamical cycles

  12. Neutron radiography with ultracold neutrons

    International Nuclear Information System (INIS)

    Bates, J.C.

    1981-01-01

    The neutron transmission factor of very thin films may be low if the neutron energy is comparable to the pseudo-potential of the film material. Surprisingly, perhaps, it is relatively easy to obtain neutrons with such low energies in sufficient numbers to produce neutron radiographs. (orig.)

  13. 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)

  14. Explicit formulation of a nodal transport method for discrete ordinates calculations in two-dimensional fixed-source problems

    Energy Technology Data Exchange (ETDEWEB)

    Tres, Anderson [Universidade Federal do Rio Grande do Sul, Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Matematica Aplicada; Becker Picoloto, Camila [Universidade Federal do Rio Grande do Sul, Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica; Prolo Filho, Joao Francisco [Universidade Federal do Rio Grande do Sul, Porto Alegre, RS (Brazil). Inst de Matematica, Estatistica e Fisica; Dias da Cunha, Rudnei; Basso Barichello, Liliane [Universidade Federal do Rio Grande do Sul, Porto Alegre, RS (Brazil). Inst de Matematica

    2014-04-15

    In this work a study of two-dimensional fixed-source neutron transport problems, in Cartesian geometry, is reported. The approach reduces the complexity of the multidimensional problem using a combination of nodal schemes and the Analytical Discrete Ordinates Method (ADO). The unknown leakage terms on the boundaries that appear from the use of the derivation of the nodal scheme are incorporated to the problem source term, such as to couple the one-dimensional integrated solutions, made explicit in terms of the x and y spatial variables. The formulation leads to a considerable reduction of the order of the associated eigenvalue problems when combined with the usual symmetric quadratures, thereby providing solutions that have a higher degree of computational efficiency. Reflective-type boundary conditions are introduced to represent the domain on a simpler form than that previously considered in connection with the ADO method. Numerical results obtained with the technique are provided and compared to those present in the literature. (orig.)

  15. Improvement of spatial discretization error on the semi-analytic nodal method using the scattered source subtraction method

    International Nuclear Information System (INIS)

    Yamamoto, Akio; Tatsumi, Masahiro

    2006-01-01

    In this paper, the scattered source subtraction (SSS) method is newly proposed to improve the spatial discretization error of the semi-analytic nodal method with the flat-source approximation. In the SSS method, the scattered source is subtracted from both side of the diffusion or the transport equation to make spatial variation of the source term to be small. The same neutron balance equation is still used in the SSS method. Since the SSS method just modifies coefficients of node coupling equations (those used in evaluation for the response of partial currents), its implementation is easy. Validity of the present method is verified through test calculations that are carried out in PWR multi-assemblies configurations. The calculation results show that the SSS method can significantly improve the spatial discretization error. Since the SSS method does not have any negative impact on execution time, convergence behavior and memory requirement, it will be useful to reduce the spatial discretization error of the semi-analytic nodal method with the flat-source approximation. (author)

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

    International Nuclear Information System (INIS)

    Duerigen, S.; Fridman, E.

    2011-01-01

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

  17. Kinetics of Pressurized Water Reactors with Hot or Cold Moderators

    Energy Technology Data Exchange (ETDEWEB)

    Norinder, O

    1960-11-15

    The set of neutron kinetic equations developed in this report permits the use of long integration steps during stepwise integration. Thermal relations which describe the transfer of heat from fuel to coolant are derived. The influence upon the kinetic behavior of the reactor of a number of parameters is studied. A comparison of the kinetic properties of the hot and cold moderators is given.

  18. Neutron Skins and Neutron Stars

    OpenAIRE

    Piekarewicz, J.

    2013-01-01

    The neutron-skin thickness of heavy nuclei provides a fundamental link to the equation of state of neutron-rich matter, and hence to the properties of neutron stars. The Lead Radius Experiment ("PREX") at Jefferson Laboratory has recently provided the first model-independence evidence on the existence of a neutron-rich skin in 208Pb. In this contribution we examine how the increased accuracy in the determination of neutron skins expected from the commissioning of intense polarized electron be...

  19. Solution of the transport equation in stationary state and X Y geometry, using continuous and discontinuous hybrid nodal schemes; Solucion de la ecuacion de transporte en estado estacionario y geometria X Y, usando esquemas nodales hibridos continuos y discontinuos

    Energy Technology Data Exchange (ETDEWEB)

    Xolocostli M, V.; Valle G, E. del [IPN-ESFM, 07738 Mexico D.F. (Mexico); Alonso V, G. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: xvicente@hotmail.com

    2003-07-01

    In this work it is described the development and the application of the NH-FEM schemes, Hybrid Nodal schemes using the Finite Element method in the solution of the neutron transport equation in stationary state and X Y geometry, of which two families of schemes were developed, one of which corresponds to the continuous and the other to the discontinuous ones, inside those first its are had the Bi-Quadratic Bi Q, and to the Bi-cubic BiC, while for the seconds the Discontinuous Bi-lineal DBiL and the Discontinuous Bi-quadratic DBiQ. These schemes were implemented in a program to which was denominated TNHXY, Transport of neutrons with Hybrid Nodal schemes in X Y geometry. One of the immediate applications of the schemes NH-FEM it will be in the analysis of assemblies of nuclear fuel, particularly of the BWR type. The validation of the TNHXY program was made with two test problems or benchmark, already solved by other authors with numerical techniques and to compare results. The first of them consists in an it BWR fuel assemble in an arrangement 7x7 without rod and with control rod providing numerical results. The second is a fuel assemble of mixed oxides (MOX) in an arrangement 10x10. This last problem it is known as the Benchmark problem WPPR of the NEA Data Bank and the results are compared with those of other commercial codes as HELIOS, MCNP-4B and CPM-3. (Author)

  20. Analytical applications for delayed neutrons

    International Nuclear Information System (INIS)

    Eccleston, G.W.

    1983-01-01

    Analytical formulations that describe the time dependence of neutron populations in nuclear materials contain delayed-neutron dependent terms. These terms are important because the delayed neutrons, even though their yields in fission are small, permit control of the fission chain reaction process. Analytical applications that use delayed neutrons range from simple problems that can be solved with the point reactor kinetics equations to complex problems that can only be solved with large codes that couple fluid calculations with the neutron dynamics. Reactor safety codes, such as SIMMER, model transients of the entire reactor core using coupled space-time neutronics and comprehensive thermal-fluid dynamics. Nondestructive delayed-neutron assay instruments are designed and modeled using a three-dimensional continuous-energy Monte Carlo code. Calculations on high-burnup spent fuels and other materials that contain a mix of uranium and plutonium isotopes require accurate and complete information on the delayed-neutron periods, yields, and energy spectra. A continuing need exists for delayed-neutron parameters for all the fissioning isotopes

  1. Neutron Dosimetry

    International Nuclear Information System (INIS)

    Vanhavere, F.

    2001-01-01

    The objective of SCK-CEN's R and D programme on neutron dosimetry is to improve the determination of neutron doses by studying neutron spectra, neutron dosemeters and shielding adaptations. In 2000, R and D focused on the contiued investigation of the bubble detectors type BD-PND and BDT, in particular their sensitivity and temperature dependence; the updating of SCK-CEN's criticality dosemeter, the investigation of the characteristics of new thermoluminescent materials and their use in neutron dosemetry; and the investigation of neutron shielding

  2. Neutron Dosimetry

    Energy Technology Data Exchange (ETDEWEB)

    Vanhavere, F

    2001-04-01

    The objective of SCK-CEN's R and D programme on neutron dosimetry is to improve the determination of neutron doses by studying neutron spectra, neutron dosemeters and shielding adaptations. In 2000, R and D focused on the contiued investigation of the bubble detectors type BD-PND and BDT, in particular their sensitivity and temperature dependence; the updating of SCK-CEN's criticality dosemeter, the investigation of the characteristics of new thermoluminescent materials and their use in neutron dosemetry; and the investigation of neutron shielding.

  3. Radiotherapy studies and extra-nodal non-Hodgkin lymphomas, progress and challenges

    DEFF Research Database (Denmark)

    Specht, L

    2012-01-01

    Extra-nodal lymphomas may arise in any organ, and different histological subtypes occur in distinct patterns. Prognosis and treatment depend not only on the histological subtype and disease extent, but also on the particular involved extra-nodal organ. The clinical course and response to treatment...... for the more common extra-nodal organs, e.g. stomach, Waldeyer's ring, skin and brain, are fairly well known and show significant variation. A few randomised trials have been carried out testing the role of radiotherapy in these lymphomas. However, for most extra-nodal lymphomas, randomised trials have...... not been carried out, and treatment decisions are made on small patient series and extrapolations from nodal lymphomas. Hopefully, wide international collaboration will make controlled clinical trials possible in the less common extra-nodal lymphomas. Modern highly conformal radiotherapy allows better...

  4. A Nodal-independent and tissue-intrinsic mechanism controls heart-looping chirality

    Science.gov (United States)

    Noël, Emily S.; Verhoeven, Manon; Lagendijk, Anne Karine; Tessadori, Federico; Smith, Kelly; Choorapoikayil, Suma; den Hertog, Jeroen; Bakkers, Jeroen

    2013-11-01

    Breaking left-right symmetry in bilateria is a major event during embryo development that is required for asymmetric organ position, directional organ looping and lateralized organ function in the adult. Asymmetric expression of Nodal-related genes is hypothesized to be the driving force behind regulation of organ laterality. Here we identify a Nodal-independent mechanism that drives asymmetric heart looping in zebrafish embryos. In a unique mutant defective for the Nodal-related southpaw gene, preferential dextral looping in the heart is maintained, whereas gut and brain asymmetries are randomized. As genetic and pharmacological inhibition of Nodal signalling does not abolish heart asymmetry, a yet undiscovered mechanism controls heart chirality. This mechanism is tissue intrinsic, as explanted hearts maintain ex vivo retain chiral looping behaviour and require actin polymerization and myosin II activity. We find that Nodal signalling regulates actin gene expression, supporting a model in which Nodal signalling amplifies this tissue-intrinsic mechanism of heart looping.

  5. Elsevier Trophoblast Research Award lecture: The multifaceted role of Nodal signaling during mammalian reproduction.

    Science.gov (United States)

    Park, C B; Dufort, D

    2011-03-01

    Nodal, a secreted signaling protein in the transforming growth factor-beta (TGF-β) superfamily, has established roles in vertebrate development. However, components of the Nodal signaling pathway are also expressed at the maternal-fetal interface and have been implicated in many processes of mammalian reproduction. Emerging evidence indicates that Nodal and its extracellular inhibitor Lefty are expressed in the uterus and complex interactions between the two proteins mediate menstruation, decidualization and embryo implantation. Furthermore, several studies have shown that Nodal from both fetal and maternal sources may regulate trophoblast cell fate and facilitate placentation as both embryonic and uterine-specific Nodal knockout mouse strains exhibit disrupted placenta morphology. Here we review the established and prospective roles of Nodal signaling in facilitating successful pregnancy, including recent evidence supporting a potential link to parturition and preterm birth. Copyright © 2011 Elsevier Ltd. All rights reserved.

  6. Reactivity Coefficient Calculation for AP1000 Reactor Using the NODAL3 Code

    Science.gov (United States)

    Pinem, Surian; Malem Sembiring, Tagor; Tukiran; Deswandri; Sunaryo, Geni Rina

    2018-02-01

    The reactivity coefficient is a very important parameter for inherent safety and stability of nuclear reactors operation. To provide the safety analysis of the reactor, the calculation of changes in reactivity caused by temperature is necessary because it is related to the reactor operation. In this paper, the temperature reactivity coefficients of fuel and moderator of the AP1000 core are calculated, as well as the moderator density and boron concentration. All of these coefficients are calculated at the hot full power condition (HFP). All neutron diffusion constant as a function of temperature, water density and boron concentration were generated by the SRAC2006 code. The core calculations for determination of the reactivity coefficient parameter are done by using NODAL3 code. The calculation results show that the fuel temperature, moderator temperature and boron reactivity coefficients are in the range between -2.613 pcm/°C to -4.657pcm/°C, -1.00518 pcm/°C to 1.00649 pcm/°C and -9.11361 pcm/ppm to -8.0751 pcm/ppm, respectively. For the water density reactivity coefficients, the positive reactivity occurs at the water temperature less than 190 °C. The calculation results show that the reactivity coefficients are accurate because the results have a very good agreement with the design value.

  7. Benchmarking of the PHOENIX-P/ANC [Advanced Nodal Code] advanced nuclear design system

    International Nuclear Information System (INIS)

    Nguyen, T.Q.; Liu, Y.S.; Durston, C.; Casadei, A.L.

    1988-01-01

    At Westinghouse, an advanced neutronic methods program was designed to improve the quality of the predictions, enhance flexibility in designing advanced fuel and related products, and improve design lead time. Extensive benchmarking data is presented to demonstrate the accuracy of the Advanced Nodal Code (ANC) and the PHOENIX-P advanced lattice code. Qualification data to demonstrate the accuracy of ANC include comparison of key physics parameters against a fine-mesh diffusion theory code, TORTISE. Benchmarking data to demonstrate the validity of the PHOENIX-P methodologies include comparison of physics predictions against critical experiments, isotopics measurements and measured power distributions from spatial criticals. The accuracy of the PHOENIX-P/ANC Advanced Design System is demonstrated by comparing predictions of hot zero power physics parameters and hot full power core follow against measured data from operating reactors. The excellent performance of this system for a broad range of comparisons establishes the basis for implementation of these tools for core design, licensing and operational follow of PWR [pressurized water reactor] cores at Westinghouse

  8. MOSRA-Light; high speed three-dimensional nodal diffusion code for vector computers

    Energy Technology Data Exchange (ETDEWEB)

    Okumura, Keisuke [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-10-01

    MOSRA-Light is a three-dimensional neutron diffusion calculation code for X-Y-Z geometry. It is based on the 4th order polynomial nodal expansion method (NEM). As the 4th order NEM is not sensitive to mesh sizes, accurate calculation is possible by the use of coarse meshes of about 20 cm. The drastic decrease of number of unknowns in a 3-dimensional problem results in very fast computation. Furthermore, it employs newly developed computation algorithm `boundary separated checkerboard sweep method` appropriate to vector computers. This method is very efficient because the speedup factor by vectorization increases, as a scale of problem becomes larger. Speed-up factor compared to the scalar calculation is from 20 to 40 in the case of PWR core calculation. Considering the both effects by the vectorization and the coarse mesh method, total speedup factor is more than 1000 as compared with conventional scalar code with the finite difference method. MOSRA-Light can be available on most of vector or scalar computers with the UNIX or it`s similar operating systems (e.g. freeware like Linux). Users can easily install it by the help of the conversation style installer. This report contains the general theory of NEM, the fast computation algorithm, benchmark calculation results and detailed information for usage of this code including input data instructions and sample input data. (author)

  9. MOSRA-Light; high speed three-dimensional nodal diffusion code for vector computers

    International Nuclear Information System (INIS)

    Okumura, Keisuke

    1998-10-01

    MOSRA-Light is a three-dimensional neutron diffusion calculation code for X-Y-Z geometry. It is based on the 4th order polynomial nodal expansion method (NEM). As the 4th order NEM is not sensitive to mesh sizes, accurate calculation is possible by the use of coarse meshes of about 20 cm. The drastic decrease of number of unknowns in a 3-dimensional problem results in very fast computation. Furthermore, it employs newly developed computation algorithm 'boundary separated checkerboard sweep method' appropriate to vector computers. This method is very efficient because the speedup factor by vectorization increases, as a scale of problem becomes larger. Speed-up factor compared to the scalar calculation is from 20 to 40 in the case of PWR core calculation. Considering the both effects by the vectorization and the coarse mesh method, total speedup factor is more than 1000 as compared with conventional scalar code with the finite difference method. MOSRA-Light can be available on most of vector or scalar computers with the UNIX or it's similar operating systems (e.g. freeware like Linux). Users can easily install it by the help of the conversation style installer. This report contains the general theory of NEM, the fast computation algorithm, benchmark calculation results and detailed information for usage of this code including input data instructions and sample input data. (author)

  10. Unstructured characteristic method embedded with variational nodal method using domain decomposition techniques

    Energy Technology Data Exchange (ETDEWEB)

    Girardi, E.; Ruggieri, J.M. [CEA Cadarache (DER/SPRC/LEPH), 13 - Saint-Paul-lez-Durance (France). Dept. d' Etudes des Reacteurs; Santandrea, S. [CEA Saclay, Dept. Modelisation de Systemes et Structures DM2S/SERMA/LENR, 91 - Gif sur Yvette (France)

    2005-07-01

    This paper describes a recently-developed extension of our 'Multi-methods,multi-domains' (MM-MD) method for the solution of the multigroup transport equation. Based on a domain decomposition technique, our approach allows us to treat the one-group equation by cooperatively employing several numerical methods together. In this work, we describe the coupling between the Method of Characteristics (integro-differential equation, unstructured meshes) with the Variational Nodal Method (even parity equation, cartesian meshes). Then, the coupling method is applied to the benchmark model of the Phebus experimental facility (Cea Cadarache). Our domain decomposition method give us the capability to employ a very fine mesh in describing a particular fuel bundle with an appropriate numerical method (MOC), while using a much large mesh size in the rest of the core, in conjunction with a coarse-mesh method (VNM). This application shows the benefits of our MM-MD approach, in terms of accuracy and computing time: the domain decomposition method allows us to reduce the Cpu time, while preserving a good accuracy of the neutronic indicators: reactivity, core-to-bundle power coupling coefficient and flux error. (authors)

  11. Unstructured characteristic method embedded with variational nodal method using domain decomposition techniques

    International Nuclear Information System (INIS)

    Girardi, E.; Ruggieri, J.M.

    2005-01-01

    This paper describes a recently-developed extension of our 'Multi-methods,multi-domains' (MM-MD) method for the solution of the multigroup transport equation. Based on a domain decomposition technique, our approach allows us to treat the one-group equation by cooperatively employing several numerical methods together. In this work, we describe the coupling between the Method of Characteristics (integro-differential equation, unstructured meshes) with the Variational Nodal Method (even parity equation, cartesian meshes). Then, the coupling method is applied to the benchmark model of the Phebus experimental facility (Cea Cadarache). Our domain decomposition method give us the capability to employ a very fine mesh in describing a particular fuel bundle with an appropriate numerical method (MOC), while using a much large mesh size in the rest of the core, in conjunction with a coarse-mesh method (VNM). This application shows the benefits of our MM-MD approach, in terms of accuracy and computing time: the domain decomposition method allows us to reduce the Cpu time, while preserving a good accuracy of the neutronic indicators: reactivity, core-to-bundle power coupling coefficient and flux error. (authors)

  12. Solution of the transport equation in stationary state and X Y geometry, using continuous and discontinuous hybrid nodal schemes

    International Nuclear Information System (INIS)

    Xolocostli M, V.; Valle G, E. del; Alonso V, G.

    2003-01-01

    In this work it is described the development and the application of the NH-FEM schemes, Hybrid Nodal schemes using the Finite Element method in the solution of the neutron transport equation in stationary state and X Y geometry, of which two families of schemes were developed, one of which corresponds to the continuous and the other to the discontinuous ones, inside those first its are had the Bi-Quadratic Bi Q, and to the Bi-cubic BiC, while for the seconds the Discontinuous Bi-lineal DBiL and the Discontinuous Bi-quadratic DBiQ. These schemes were implemented in a program to which was denominated TNHXY, Transport of neutrons with Hybrid Nodal schemes in X Y geometry. One of the immediate applications of the schemes NH-FEM it will be in the analysis of assemblies of nuclear fuel, particularly of the BWR type. The validation of the TNHXY program was made with two test problems or benchmark, already solved by other authors with numerical techniques and to compare results. The first of them consists in an it BWR fuel assemble in an arrangement 7x7 without rod and with control rod providing numerical results. The second is a fuel assemble of mixed oxides (MOX) in an arrangement 10x10. This last problem it is known as the Benchmark problem WPPR of the NEA Data Bank and the results are compared with those of other commercial codes as HELIOS, MCNP-4B and CPM-3. (Author)

  13. Nodal wear model: corrosion in carbon blast furnace hearths

    International Nuclear Information System (INIS)

    Verdeja, L. F.; Gonzalez, R.; Alfonso, A.; Barbes, M. F.

    2003-01-01

    Criteria developed for the Nodal Wear Model (NWM) were applied to estimate the shape of the corrosion profiles that a blast furnace hearth may acquire during its campaign. Taking into account design of the hearth, the boundary conditions, the characteristics of the refractory materials used and the operation conditions of the blast furnace, simulation of wear profiles with central well, mushroom and elephant foot shape were accomplished. The foundations of the NWM are constructed considering that the corrosion of the refractory is a function of the temperature present at each point (node) of the liquid metal-refractory interface and the corresponding physical and chemical characteristics of the corrosive fluid. (Author) 31 refs

  14. Mechanism of polyuria and natriuresis in atrioventricular nodal tachycardia.

    Science.gov (United States)

    Canepa-Anson, R; Williams, M; Marshall, J; Mitsuoka, T; Lightman, S; Sutton, R

    1984-01-01

    A woman with tachycardia associated with polyuria was investigated. Electrophysiological analysis showed that the tachycardia was an atrioventricular nodal re-entrant tachycardia. Programmed stimulation was then used to provoke and sustain the tachycardia for 40 minutes. Polyuria, with an appreciable increase in free water clearance, was observed. This was associated with reduction in plasma and urinary arginine vasopressin concentrations. Appreciable natriuresis also developed. These results support the hypothesis that the polyuria with increased free water clearance and the natriuresis occurring during sustained tachycardia in man are due to inhibition of secretion of vasopressin and the release of natriuretic factor. PMID:6434116

  15. Approximate Schur complement preconditioning of the lowest order nodal discretizations

    Energy Technology Data Exchange (ETDEWEB)

    Moulton, J.D.; Ascher, U.M. [Univ. of British Columbia, Vancouver, British Columbia (Canada); Morel, J.E. [Los Alamos National Lab., NM (United States)

    1996-12-31

    Particular classes of nodal methods and mixed hybrid finite element methods lead to equivalent, robust and accurate discretizations of 2nd order elliptic PDEs. However, widespread popularity of these discretizations has been hindered by the awkward linear systems which result. The present work exploits this awkwardness, which provides a natural partitioning of the linear system, by defining two optimal preconditioners based on approximate Schur complements. Central to the optimal performance of these preconditioners is their sparsity structure which is compatible with Dendy`s black box multigrid code.

  16. A survey of available margin in a PWR RIA with statistical methods and 3D kinetics

    International Nuclear Information System (INIS)

    Riverola Gurruchaga, J.; Nunez Rodriguez, T.

    2010-01-01

    This paper investigates the recovery of margin in a PWR RIA simulation with 3D kinetics, due to statistical techniques. The chosen reference core is a typical 12 feet, 17*17 PWR, with very low leakage loading pattern strategy and gadolinium oxide as burnable poison. The PARCS calculated average nuclear power and nodal power are transferred to a hot spot model for a sequential calculation of fuel temperature and enthalpy responses allowing for independent hypothesis in both calculations. The hot spot analysis is done with a pellet type model with RELAP. The analysis is done at HZP and EOC, since this state is the most limiting one respect to the enthalpy rise criterion, compared to other burn-up condition or initial power cases. In this work, the enthalpy increase is estimated with several statistical methods of propagation of uncertainties: order statistics, parametric statistics, surface response and sensitivities. A discussion on the advantages and disadvantages of each method is also presented. This statistical analysis is also useful to confirm a previous classification of parameters and assumptions according to their importance for the simulation, and found to be consistent with the state of the art in the published literature. These parameters include ejected rod worth and ejection time, delayed neutron fraction and yields, nuclear power peaking factor, and Doppler. (authors)

  17. Nodal line optimization and its application to violin top plate design

    Science.gov (United States)

    Yu, Yonggyun; Jang, In Gwun; Kim, In Kyum; Kwak, Byung Man

    2010-10-01

    In the literature, most problems of structural vibration have been formulated to adjust a specific natural frequency: for example, to maximize the first natural frequency. In musical instruments like a violin; however, mode shapes are equally important because they are related to sound quality in the way that natural frequencies are related to the octave. The shapes of nodal lines, which represent the natural mode shapes, are generally known to have a unique feature for good violins. Among the few studies on mode shape optimization, one typical study addresses the optimization of nodal point location for reducing vibration in a one-dimensional beam structure. However, nodal line optimization, which is required in violin plate design, has not yet been considered. In this paper, the central idea of controlling the shape of the nodal lines is proposed and then applied to violin top plate design. Finite element model for a violin top plate was constructed using shell elements. Then, optimization was performed to minimize the square sum of the displacement of selected nodes located along the target nodal lines by varying the thicknesses of the top plate. We conducted nodal line optimization for the second and the fifth modes together at the same time, and the results showed that the nodal lines obtained match well with the target nodal lines. The information on plate thickness distribution from nodal line optimization would be valuable for tailored trimming of a violin top plate for the given performances.

  18. Time-dependent patterning of the mesoderm and endoderm by Nodal signals in zebrafish

    Directory of Open Access Journals (Sweden)

    Dougan Scott T

    2007-03-01

    Full Text Available Abstract Background The vertebrate body plan is generated during gastrulation with the formation of the three germ layers. Members of the Nodal-related subclass of the TGF-β superfamily induce and pattern the mesoderm and endoderm in all vertebrates. In zebrafish, two nodal-related genes, called squint and cyclops, are required in a dosage-dependent manner for the formation of all derivatives of the mesoderm and endoderm. These genes are expressed dynamically during the blastula stages and may have different roles at different times. This question has been difficult to address because conditions that alter the timing of nodal-related gene expression also change Nodal levels. We utilized a pharmacological approach to conditionally inactivate the ALK 4, 5 and 7 receptors during the blastula stages without disturbing earlier signaling activity. This permitted us to directly examine when Nodal signals specify cell types independently of dosage effects. Results We show that two drugs, SB-431542 and SB-505124, completely block the response to Nodal signals when added to embryos after the mid-blastula transition. By blocking Nodal receptor activity at later stages, we demonstrate that Nodal signaling is required from the mid-to-late blastula period to specify sequentially, the somites, notochord, blood, Kupffer's vesicle, hatching gland, heart, and endoderm. Blocking Nodal signaling at late times prevents specification of cell types derived from the embryo margin, but not those from more animal regions. This suggests a linkage between cell fate and length of exposure to Nodal signals. Confirming this, cells exposed to a uniform Nodal dose adopt progressively more marginal fates with increasing lengths of exposure. Finally, cell fate specification is delayed in squint mutants and accelerated when Nodal levels are elevated. Conclusion We conclude that (1 Nodal signals are most active during the mid-to-late blastula stages, when nodal-related gene

  19. Neutron radiography

    International Nuclear Information System (INIS)

    Hrdlicka, Z.

    1977-01-01

    Neutron radiography is a radiographic method using a neutron beam of a defined geometry. The neutron source usually consists of a research reactor, a specialized neutron radiography reactor or the 252 Cf radioisotope source. There are two types of the neutron radiography display system, viz., a system producing neutron radiography images by a photographic process or a system allowing a visual display, eg., using a television monitor. The method can be used wherever X-ray radiography is used except applications in the radiography of humans. The neutron radiography unit at UJV uses the WWR-S reactor as the neutron source and both types of the above mentioned display system. (J.P.)

  20. Fast neutron dosimetry

    International Nuclear Information System (INIS)

    DeLuca, P.M. Jr.; Pearson, D.W.

    1991-01-01

    During 1988--1990 the magnetic resonance dosimetry project was completed, as were the 250 MeV proton shielding measurements. The first cellular experiment using human cells in vitro at the 1 GeV electron storage ring was also accomplished. More detail may be found in DOE Report number-sign DOE/EV/60417-002 and the open literature cited in the individual progress subsections. We report Kinetic Energy Released in Matter (KERMA), factor measurements in several elements of critical importance to neutron radiation therapy and radiation protection for space habitation and exploration for neutron energies below 30 MeV. The results of this effort provide the only direct measurements of the oxygen and magnesium kerma factors above 20 MeV neutron energy, and the only measurements of the iron kerma factor above 15 MeV. They provide data of immediate relevance to neutron radiotherapy and impose strict criteria for normalizing and testing nuclear models used to calculate kerma factors at higher neutron energies

  1. The neutron

    International Nuclear Information System (INIS)

    Kredov, B.M.

    1979-01-01

    The history of the neutron is displayed on the basis of contributions by scientists who produced outstanding results in neutron research (part 1), of summarizing discoveries and theories which led to the discovery of the neutron and the resulting development of nuclear physics (part 2), and of fundamental papers written by Rutherford, Chadwick, Iwanenko, and others (appendix). Of interest to physicists, historians, and students

  2. Neutron techniques

    International Nuclear Information System (INIS)

    Charlton, J.S.

    1986-01-01

    The way in which neutrons interact with matter such as slowing-down, diffusion, neutron absorption and moderation are described. The use of neutron techniques in industry, in moisture gages, level and interface measurements, the detection of blockages, boron analysis in ore feedstock and industrial radiography are discussed. (author)

  3. Neutron Transport Methods for Accelerator-Driven Systems

    International Nuclear Information System (INIS)

    Nicholas Tsoulfanidis; Elmer Lewis

    2005-01-01

    The objective of this project has been to develop computational methods that will enable more effective analysis of Accelerator Driven Systems (ADS). The work is centered at the University of Missouri at Rolla, with a subcontract at Northwestern University, and close cooperation with the Nuclear Engineering Division at Argonne National Laboratory. The work has fallen into three categories. First, the treatment of the source for neutrons originating from the spallation target which drives the neutronics calculations of the ADS. Second, the generalization of the nodal variational method to treat the R-Z geometry configurations frequently needed for scoping calculations in Accelerator Driven Systems. Third, the treatment of void regions within variational nodal methods as needed to treat the accelerator beam tube

  4. Reactor kinetics methods development. Final report

    International Nuclear Information System (INIS)

    Hansen, K.F.; Henry, A.F.

    1978-01-01

    This report is a qualitative summary of research conducted at MIT from 1967 to 1977 in the area of reactor kinetics methods. The objectives of the research were to find methods of integration of various mathematical models of nuclear reactor transients. From the beginning the work was aimed at numerical integration methods. Specific areas of research, discussed in more detail following, included: integration of multigroup diffusion theory models by finite difference and finite element methods; response matrix and nodal methods; coarse-mesh homogenization; and special treatment of boundary conditions

  5. Exact boundary controllability of nodal profile for quasilinear hyperbolic systems

    CERN Document Server

    Li, Tatsien; Gu, Qilong

    2016-01-01

    This book provides a comprehensive overview of the exact boundary controllability of nodal profile, a new kind of exact boundary controllability stimulated by some practical applications. This kind of controllability is useful in practice as it does not require any precisely given final state to be attained at a suitable time t=T by means of boundary controls, instead it requires the state to exactly fit any given demand (profile) on one or more nodes after a suitable time t=T by means of boundary controls. In this book we present a general discussion of this kind of controllability for general 1-D first order quasilinear hyperbolic systems and for general 1-D quasilinear wave equations on an interval as well as on a tree-like network using a modular-structure construtive method, suggested in LI Tatsien's monograph "Controllability and Observability for Quasilinear Hyperbolic Systems"(2010), and we establish a complete theory on the local exact boundary controllability of nodal profile for 1-D quasilinear hyp...

  6. Nodal Diffusion Burnable Poison Treatment for Prismatic Reactor Cores

    International Nuclear Information System (INIS)

    Ougouag, A.M.; Ferrer, R.M.

    2010-01-01

    The prismatic block version of the High Temperature Reactor (HTR) considered as a candidate Very High Temperature Reactor (VHTR)design may use burnable poison pins in locations at some corners of the fuel blocks (i.e., assembly equivalent structures). The presence of any highly absorbing materials, such as these burnable poisons, within fuel blocks for hexagonal geometry, graphite-moderated High Temperature Reactors (HTRs) causes a local inter-block flux depression that most nodal diffusion-based method have failed to properly model or otherwise represent. The location of these burnable poisons near vertices results in an asymmetry in the morphology of the assemblies (or blocks). Hence the resulting inadequacy of traditional homogenization methods, as these 'spread' the actually local effect of the burnable poisons throughout the assembly. Furthermore, the actual effect of the burnable poison is primarily local with influence in its immediate vicinity, which happens to include a small region within the same assembly as well as similar regions in the adjacent assemblies. Traditional homogenization methods miss this artifact entirely. This paper presents a novel method for treating the local effect of the burnable poison explicitly in the context of a modern nodal method.

  7. Hereditary bone dysplasia with pathological fractures and nodal osteoarthropathy

    International Nuclear Information System (INIS)

    Arendse, Regan; Brink, Paul; Beighton, Peter

    2009-01-01

    A father and daughter both had multiple pathological fractures and nodal osteoarthropathy. The father, aged 50 years, had at least 20 healed fractures of the axial and appendicular skeleton, sustained by minor trauma over his 50-year lifespan, many of which had been surgically fixed prior to his first presentation to us. Fractures of the clavicles, thoracic cage and long bones of the arms and legs, had healed with malalignment and deformity. Healed fractures were complicated by ankylosis of the cervical vertebrae and both elbows. He also had osteoarthritis of the hands, with exuberant osteophytosis, and profound perceptive deafness. His general health was good, his intellect and facies were normal, and his sclerae were white. The daughter, aged 27 years, had sustained at least seven fractures of the axial and appendicular skeleton following trivial injuries, in distribution similar to those of the father. She had also experienced painful swelling of the fingers, which preceded progressive development of nodal osteoarthropathy. Her hearing was normal. In both individuals, biochemical and immunological investigations yielded normal results. It was not possible for molecular studies to be undertaken. Pedigree data were consistent with autosomal dominant transmission, and this disorder appeared to be a previously undocumented heritable skeletal dysplasia. (orig.)

  8. Hereditary bone dysplasia with pathological fractures and nodal osteoarthropathy

    Energy Technology Data Exchange (ETDEWEB)

    Arendse, Regan [University of Stellenbosch, Department of Medicine, Tygerberg Hospital, Stellenbosch (South Africa); University of Cape Town, Division of Rheumatology, Groote Schuur Hospital, Cape Town (South Africa); Brink, Paul [University of Stellenbosch, Department of Medicine, Tygerberg Hospital, Stellenbosch (South Africa); Beighton, Peter [University of Cape Town, Division of Human Genetics, Faculty of Health Sciences, Cape Town (South Africa)

    2009-12-15

    A father and daughter both had multiple pathological fractures and nodal osteoarthropathy. The father, aged 50 years, had at least 20 healed fractures of the axial and appendicular skeleton, sustained by minor trauma over his 50-year lifespan, many of which had been surgically fixed prior to his first presentation to us. Fractures of the clavicles, thoracic cage and long bones of the arms and legs, had healed with malalignment and deformity. Healed fractures were complicated by ankylosis of the cervical vertebrae and both elbows. He also had osteoarthritis of the hands, with exuberant osteophytosis, and profound perceptive deafness. His general health was good, his intellect and facies were normal, and his sclerae were white. The daughter, aged 27 years, had sustained at least seven fractures of the axial and appendicular skeleton following trivial injuries, in distribution similar to those of the father. She had also experienced painful swelling of the fingers, which preceded progressive development of nodal osteoarthropathy. Her hearing was normal. In both individuals, biochemical and immunological investigations yielded normal results. It was not possible for molecular studies to be undertaken. Pedigree data were consistent with autosomal dominant transmission, and this disorder appeared to be a previously undocumented heritable skeletal dysplasia. (orig.)

  9. Dirac Magnon Nodal Loops in Quasi-2D Quantum Magnets.

    Science.gov (United States)

    Owerre, S A

    2017-07-31

    In this report, we propose a new concept of one-dimensional (1D) closed lines of Dirac magnon nodes in two-dimensional (2D) momentum space of quasi-2D quantum magnetic systems. They are termed "2D Dirac magnon nodal-line loops". We utilize the bilayer honeycomb ferromagnets with intralayer coupling J and interlayer coupling J L , which is realizable in the honeycomb chromium compounds CrX 3 (X ≡ Br, Cl, and I). However, our results can also exist in other layered quasi-2D quantum magnetic systems. Here, we show that the magnon bands of the bilayer honeycomb ferromagnets overlap for J L  ≠ 0 and form 1D closed lines of Dirac magnon nodes in 2D momentum space. The 2D Dirac magnon nodal-line loops are topologically protected by inversion and time-reversal symmetry. Furthermore, we show that they are robust against weak Dzyaloshinskii-Moriya interaction Δ DM  magnon edge modes.

  10. Physical kinetics

    International Nuclear Information System (INIS)

    Lifschitz, E.M.; Pitajewski, L.P.

    1983-01-01

    The textbook covers the subject under the following headings: kinetic gas theory, diffusion approximation, collisionless plasma, collisions within the plasma, plasma in the magnetic field, theory of instabilities, dielectrics, quantum fluids, metals, diagram technique for nonequilibrium systems, superconductors, and kinetics of phase transformations

  11. Heparin kinetics

    International Nuclear Information System (INIS)

    Swart, C.A.M. de.

    1983-01-01

    The author has studied the kinetics of heparin and heparin fractions after intravenous administration in humans and in this thesis the results of this study are reported. Basic knowledge about the physico-chemical properties of heparin and its interactions with proteins resulting in anticoagulant and lipolytic effects are discussed in a review (chapter II), which also comprises some clinical aspects of heparin therapy. In chapter III the kinetics of the anticoagulant effect are described after intravenous administration of five commercial heparin preparations. A mathematical model is presented that fits best to these kinetics. The kinetics of the anticoagulant and lipolytic effects after intravenous injection of various 35 S-radiolabelled heparin fractions and their relationship with the disappearance of the radiolabel are described in chapter IV. Chapter V gives a description of the kinetics of two radiolabels after injection of in vitro formed complexes consisting of purified, 125 I-radiolabelled antithrombin III and various 35 S-radiolabelled heparin fractions. (Auth.)

  12. Assessment of Effect on LBLOCA PCT for Change in Upper Head Nodalization

    International Nuclear Information System (INIS)

    Kang, Dong Gu; Huh, Byung Gil; Yoo, Seung Hun; Bang, Youngseok; Seul, Kwangwon; Cho, Daehyung

    2014-01-01

    In this study, the best estimate plus uncertainty (BEPU) analysis of LBLOCA for original and modified nodalizations was performed, and the effect on LBLOCA PCT for change in upper head nodalization was assessed. In this study, the best estimate plus uncertainty (BEPU) analysis of LBLOCA for original and modified nodalizations was performed, and the effect on LBLOCA PCT for change in upper head nodalization was assessed. It is confirmed that modification of upper head nodalization influences PCT behavior, especially in the reflood phase. In conclusions, the modification of nodalization to reflect design characteristic of upper head temperature should be done to predict PCT behavior accurately in LBLOCA analysis. In the best estimate (BE) method with the uncertainty evaluation, the system nodalization is determined by the comparative studies of the experimental data. Up to now, it was assumed that the temperature of the upper dome in OPR-1000 was close to that of the cold leg. However, it was found that the temperature of the upper head/dome might be a little lower than or similar to that of the hot leg through the evaluation of the detailed design data. Since the higher upper head temperature affects blowdown quenching and peak cladding temperature in the reflood phase, the nodalization for upper head should be modified

  13. LOLA SYSTEM: A code block for nodal PWR simulation. Part. I - Simula-3 Code

    Energy Technology Data Exchange (ETDEWEB)

    Aragones, J M; Ahnert, C; Gomez Santamaria, J; Rodriguez Olabarria, I

    1985-07-01

    Description of the theory and users manual of the SIMULA-3 code, which is part of the core calculation system by nodal theory in one group, called LOLA SYSTEM. SIMULA-3 is the main module of the system, it uses a modified nodal theory, with interface leakages equivalent to the diffusion theory. (Author) 4 refs.

  14. Rules for Phase Shifts of Quantum Oscillations in Topological Nodal-Line Semimetals

    Science.gov (United States)

    Li, Cequn; Wang, C. M.; Wan, Bo; Wan, Xiangang; Lu, Hai-Zhou; Xie, X. C.

    2018-04-01

    Nodal-line semimetals are topological semimetals in which band touchings form nodal lines or rings. Around a loop that encloses a nodal line, an electron can accumulate a nontrivial π Berry phase, so the phase shift in the Shubnikov-de Haas (SdH) oscillation may give a transport signature for the nodal-line semimetals. However, different experiments have reported contradictory phase shifts, in particular, in the WHM nodal-line semimetals (W =Zr /Hf , H =Si /Ge , M =S /Se /Te ). For a generic model of nodal-line semimetals, we present a systematic calculation for the SdH oscillation of resistivity under a magnetic field normal to the nodal-line plane. From the analytical result of the resistivity, we extract general rules to determine the phase shifts for arbitrary cases and apply them to ZrSiS and Cu3 PdN systems. Depending on the magnetic field directions, carrier types, and cross sections of the Fermi surface, the phase shift shows rich results, quite different from those for normal electrons and Weyl fermions. Our results may help explore transport signatures of topological nodal-line semimetals and can be generalized to other topological phases of matter.

  15. LOLA SYSTEM: A code block for nodal PWR simulation. Part. I - Simula-3 Code

    International Nuclear Information System (INIS)

    Aragones, J. M.; Ahnert, C.; Gomez Santamaria, J.; Rodriguez Olabarria, I.

    1985-01-01

    Description of the theory and users manual of the SIMULA-3 code, which is part of the core calculation system by nodal theory in one group, called LOLA SYSTEM. SIMULA-3 is the main module of the system, it uses a modified nodal theory, with interface leakages equivalent to the diffusion theory. (Author) 4 refs

  16. Application of the RT-0 nodal methods and NRMPO matrix-response to the cycles 1 and 2 of the LVC

    International Nuclear Information System (INIS)

    Delfin L, A.; Hernandez L, H.; Alonso V, G.

    2005-01-01

    The nodal methods the same as that of matrix-response are used to develop numeric calculations, so much in static as dynamics of reactors, in one, two and three dimensions. The topic of this work is to apply the equations modeled in the RPM0 program, obtained when using the nodal scheme RT-0 (Raviart-Thomas index zero) in the neutron diffusion equation in stationary state X Y geometry, applying finite differences centered in mesh and lineal reactivity; also, to use those equations captured in the NRMPO program developed by E. Malambu that uses the matrix-response method in X Y geometry. The numeric results of the radial distribution of power by fuel assembly of the unit 1, in the cycles 1 and 2 of the CLV obtained by both methods, they are compared with the calculations obtained with the CM-PRESTO code that is a neutronic-thermo hydraulic simulator in three dimensions. The comparison of the radial distribution of power in the cycles 1 and 2 of the CLV with the CM-PRESTO code, it presents for RPM0 maximum errors of 8.2% and 12.4% and for NRMPO 31.2% and 61.3% respectively. The results show that it can be feasible to use the program RPM0 like a quick and efficient tool in the multicycle analysis in the fuel management. (Author)

  17. Neutron radiography

    International Nuclear Information System (INIS)

    Hiraoka, Eiichi

    1988-01-01

    The thermal neutron absorption coefficient is essentially different from the X-ray absorption coefficient. Each substance has a characteristic absorption coefficient regardless of its density. Neutron deams have the following features: (1) neutrons are not transmitted efficiently by low molecular weight substances, (2) they are transmitted efficiently by heavy metals, and (3) the transmittance differs among isotopes. Thus, neutron beams are suitable for cheking for foreign matters in heavy metals and testing of composites consisting of both heavy and light materials. A neutron source generates fast neutrons, which should be converted into thermal neutrons by reducing their energy. Major neutron souces include nuclear reactors, radioisotopes and particle accelerators. Photographic films and television systems are mainly used to observe neutron transmission images. Computers are employed for image processing, computerized tomography and three-dimensional analysis. The major applications of neutron radiography include inspection of neclear fuel; evaluation of material for airplane; observation of fuel in the engine and oil in the hydraulic systems in airplanes; testing of composite materials; etc. (Nogami, K.)

  18. Opposing nodal and BMP signals regulate left-right asymmetry in the sea urchin larva.

    Directory of Open Access Journals (Sweden)

    Yi-Jyun Luo

    Full Text Available Nodal and BMP signals are important for establishing left-right (LR asymmetry in vertebrates. In sea urchins, Nodal signaling prevents the formation of the rudiment on the right side. However, the opposing pathway to Nodal signaling during LR axis establishment is not clear. Here, we revealed that BMP signaling is activated in the left coelomic pouch, specifically in the veg2 lineage, but not in the small micromeres. By perturbing BMP activities, we demonstrated that BMP signaling is required for activating the expression of the left-sided genes and the formation of the left-sided structures. On the other hand, Nodal signals on the right side inhibit BMP signaling and control LR asymmetric separation and apoptosis of the small micromeres. Our findings show that BMP signaling is the positive signal for left-sided development in sea urchins, suggesting that the opposing roles of Nodal and BMP signals in establishing LR asymmetry are conserved in deuterostomes.

  19. A spectral nodal method for discrete ordinates problems in x,y geometry

    International Nuclear Information System (INIS)

    Barros, R.C. de; Larsen, E.W.

    1991-06-01

    A new nodal method is proposed for the solution of S N problems in x- y-geometry. This method uses the Spectral Green's Function (SGF) scheme for solving the one-dimensional transverse-integrated nodal transport equations with no spatial truncation error. Thus, the only approximations in the x, y-geometry nodal method occur in the transverse leakage terms, as in diffusion theory. We approximate these leakage terms using a flat or constant approximation, and we refer to the resulting method as the SGF-Constant Nodal (SGF-CN) method. We show in numerical calculations that the SGF-CN method is much more accurate than other well-known transport nodal methods for coarse-mesh deep-penetration S N problems, even though the transverse leakage terms are approximated rather simply. (author)

  20. Dual Atrioventricular Nodal Pathways Physiology: A Review of Relevant Anatomy, Electrophysiology, and Electrocardiographic Manifestations

    Directory of Open Access Journals (Sweden)

    Bhalaghuru Chokkalingam Mani, MD

    2014-01-01

    Full Text Available More than half a century has passed since the concept of dual atrioventricular (AV nodal pathways physiology was conceived. Dual AV nodal pathways have been shown to be responsible for many clinical arrhythmia syndromes, most notably AV nodal reentrant tachycardia. Although there has been a considerable amount of research on this topic, the subject of dual AV nodal pathways physiology remains heavily debated and discussed. Despite advances in understanding arrhythmia mechanisms and the widespread use of invasive electrophysiologic studies, there is still disagreement on the anatomy and physiology of the AV node that is the basis of discontinuous antegrade AV conduction. The purpose of this paper is to review the concept of dual AV nodal pathways physiology and its varied electrocardiographic manifestations.

  1. Neutron-nucleus interactions and fission. Chapter 1

    International Nuclear Information System (INIS)

    1998-01-01

    The central problem in nuclear-reactor kinetics is to predict the evolution in time of the neutron population in a multiplying medium. Point kinetics allows study of the global behaviour of the neutron population from the average properties of the medium. Before tackling, in the following chapters, the equations governing the time variation of the reactor power (proportional to the total neutron population), the properties of a neutron-multiplying medium shall be discussed briefly. After recalling a number of definitions, a qualitative description shall be given of the principal nuclear reactions at play in a self-sustaining chain reaction, with emphasis on the source of fission neutrons. Since delayed neutrons play a crucial role in reactor kinetics, their production in a reactor shall be described in greater detail. (author)

  2. An integral nodal variational method for multigroup criticality calculations

    International Nuclear Information System (INIS)

    Lewis, E.E.; Tsoulfanidis, N.

    2003-01-01

    An integral formulation of the variational nodal method is presented and applied to a series of benchmark critically problems. The method combines an integral transport treatment of the even-parity flux within the spatial node with an odd-parity spherical harmonics expansion of the Lagrange multipliers at the node interfaces. The response matrices that result from this formulation are compatible with those in the VARIANT code at Argonne National Laboratory. Either homogeneous or heterogeneous nodes may be employed. In general, for calculations requiring higher-order angular approximations, the integral method yields solutions with comparable accuracy while requiring substantially less CPU time and memory than the standard spherical harmonics expansion using the same spatial approximations. (author)

  3. On-line application of the PANTHER advanced nodal code

    International Nuclear Information System (INIS)

    Hutt, P.K.; Knight, M.P.

    1992-01-01

    Over the last few years, Nuclear Electric has developed an integrated core performance code package for both light water reactors (LWRs) and advanced gas-cooled reactors (AGRs) that can perform a comprehensive range of calculations for fuel cycle design, safety analysis, and on-line operational support for such plants. The package consists of the following codes: WIMS for lattice physics, PANTHER whole reactor nodal flux and AGR thermal hydraulics, VIPRE for LWR thermal hydraulics, and ENIGMA for fuel performance. These codes are integrated within a UNIX-based interactive system called the Reactor Physics Workbench (RPW), which provides an interactive graphic user interface and quality assurance records/data management. The RPW can also control calculational sequences and data flows. The package has been designed to run both off-line and on-line accessing plant data through the RPW

  4. Fourier-Accelerated Nodal Solvers (FANS) for homogenization problems

    Science.gov (United States)

    Leuschner, Matthias; Fritzen, Felix

    2017-11-01

    Fourier-based homogenization schemes are useful to analyze heterogeneous microstructures represented by 2D or 3D image data. These iterative schemes involve discrete periodic convolutions with global ansatz functions (mostly fundamental solutions). The convolutions are efficiently computed using the fast Fourier transform. FANS operates on nodal variables on regular grids and converges to finite element solutions. Compared to established Fourier-based methods, the number of convolutions is reduced by FANS. Additionally, fast iterations are possible by assembling the stiffness matrix. Due to the related memory requirement, the method is best suited for medium-sized problems. A comparative study involving established Fourier-based homogenization schemes is conducted for a thermal benchmark problem with a closed-form solution. Detailed technical and algorithmic descriptions are given for all methods considered in the comparison. Furthermore, many numerical examples focusing on convergence properties for both thermal and mechanical problems, including also plasticity, are presented.

  5. Nodal domains on isospectral quantum graphs: the resolution of isospectrality?

    International Nuclear Information System (INIS)

    Band, Ram; Shapira, Talia; Smilansky, Uzy

    2006-01-01

    We present and discuss isospectral quantum graphs which are not isometric. These graphs are the analogues of the isospectral domains in R 2 which were introduced recently in Gordon et al (1992 Bull. Am. Math. Soc. 27 134-8), Chapman (1995 Am. Math. Mon. 102 124), Buser et al (1994 Int. Math. Res. Not. 9 391-400), Okada and Shudo (2001 J. Phys. A: Math. Gen. 34 5911-22), Jakobson et al (2006 J. Comput. Appl. Math. 194 141-55) and Levitin et al (2006 J. Phys. A: Math. Gen. 39 2073-82)) all based on Sunada's construction of isospectral domains (Sunada T 1985 Ann. Math. 121 196-86). After presenting some of the properties of these graphs, we discuss a few examples which support the conjecture that by counting the nodal domains of the corresponding eigenfunctions one can resolve the isospectral ambiguity

  6. INTERMITTENT ANTIARYTHMIC THERAPY OF ARIOVENTICULAR NODAL REENTRY TACHYCARDIA IN CHILDREN

    Directory of Open Access Journals (Sweden)

    Boris Djindjic

    2008-04-01

    Full Text Available Until recent advances in pharmacology and clinical cardiology regarding farmacodynamics of antiarrhythmic drugs and their efficiency in patients with refractory paroxysmal supraventricular tachycardia, chronic prophylactic therapy was the only treatment option for patients refusing catheter ablation. Another treatment option, also known by eponym “pill in pocket” have been shown to be equally useful and efficacious.The aim of our study was prospective examination of children with refractory atrioventricular nodal reentry tachycardia (AVNRT who were withdrawn from chronic antiarrhythmic prophylactic therapy and started with intermittent oral beta blocker treatment (propranolol at dosage 1 mg/kg - max 80 mg.Twelve children (8 boys and 4 girls with AVNRT were included in the study. Four children did not have arrhythmia during first six months after withdrawal and 7 were successfully treated without complication.Intermittent antiarrhythmic therapy in children with AVNRT could be very efficacious and useful treatment option which significantly improves their quality of life.

  7. CAISO flicks switch on nodal scheme and lights stay on

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2009-06-15

    In 2000-01, two years after introducing a competitive wholesale power auction in California - with a separate day-ahead zonal market operated by the California Power Exchange and a zonal market for ancillary services and balancing energy operated by the California Independent System Operator (CAISO) - the California market collapsed from exorbitant prices, flagrant gaming, and abuse of market power. Nine years later, CAISO introduced a nodal pricing auction for the wholesale market in April, replacing the zonal scheme, which was among many causes of the original market's demise. With nearly 3,000 nodes on the network, high prices in one region do not affect prices everywhere on the system. After investing some $200 million to upgrade the software, countless delays, and 18 months of market simulation and testing, the new auction was introduced and nothing unusual happened.

  8. Ultrasound beam characteristics of a symmetric nodal origami based array

    Science.gov (United States)

    Bilgunde, Prathamesh N.; Bond, Leonard J.

    2018-04-01

    Origami-the ancient art of paper folding-is being explored in acoustics for effective focusing of sound. In this short communication, we present a numerical investigation of beam characteristics for an origami based ultrasound array. A spatial re-configuration of array elements is performed based upon the symmetric nodal origami. The effect of fold angle on the ultrasound beam is evaluated using frequency domain and transient finite element analysis. It was found that increase in the fold angle reduces near field length by 58% and also doubles the beam intensity as compared to the linear array. Transient analysis also indicated 80% reduction in the -6dB beam width, which can improve the lateral resolution of phased array. Such a spatially re-configurable array could potentially be used in the future to reduce the cost of electronics in the phased array instrumentation.

  9. Space-angle approximations in the variational nodal method

    International Nuclear Information System (INIS)

    Lewis, E. E.; Palmiotti, G.; Taiwo, T.

    1999-01-01

    The variational nodal method is formulated such that the angular and spatial approximations maybe examined separately. Spherical harmonic, simplified spherical harmonic, and discrete ordinate approximations are coupled to the primal hybrid finite element treatment of the spatial variables. Within this framework, two classes of spatial trial functions are presented: (1) orthogonal polynomials for the treatment of homogeneous nodes and (2) bilinear finite subelement trial functions for the treatment of fuel assembly sized nodes in which fuel-pin cell cross sections are represented explicitly. Polynomial and subelement trial functions are applied to benchmark water-reactor problems containing MOX fuel using spherical harmonic and simplified spherical harmonic approximations. The resulting accuracy and computing costs are compared

  10. Comparison of PANTHER nodal solutions in hexagonal-z geometry

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  11. Static benchmarking of the NESTLE advanced nodal code

    International Nuclear Information System (INIS)

    Mosteller, R.D.

    1997-01-01

    Results from the NESTLE advanced nodal code are presented for multidimensional numerical benchmarks representing four different types of reactors, and predictions from NESTLE are compared with measured data from pressurized water reactors (PWRs). The numerical benchmarks include cases representative of PWRs, boiling water reactors (BWRs), CANDU heavy water reactors (HWRs), and high-temperature gas-cooled reactors (HTGRs). The measured PWR data include critical soluble boron concentrations and isothermal temperature coefficients of reactivity. The results demonstrate that NESTLE correctly solves the multigroup diffusion equations for both Cartesian and hexagonal geometries, that it reliably calculates k eff and reactivity coefficients for PWRs, and that--subsequent to the incorporation of additional thermal-hydraulic models--it will be able to perform accurate calculations for the corresponding parameters in BWRs, HWRs, and HTGRs as well

  12. Contemporary Management of Recurrent Nodal Disease in Differentiated Thyroid Carcinoma

    Science.gov (United States)

    Na’ara, Shorook; Amit, Moran; Fridman, Eran; Gil, Ziv

    2016-01-01

    Differentiated thyroid carcinoma (DTC) comprises over 90% of thyroid tumors and includes papillary and follicular carcinomas. Patients with DTC have an excellent prognosis, with a 10-year survival rate of over 90%. However, the risk of recurrent tumor ranges between 5% and 30% within 10 years of the initial diagnosis. Cervical lymph node disease accounts for the majority of recurrences and in most cases is detected during follow-up by ultrasound or elevated levels of serum thyroglobulin. Recurrent disease is accompanied by increased morbidity. The mainstay of treatment of nodal recurrence is surgical management. We provide an overview of the literature addressing surgical management of recurrent or persistent lymph node disease in patients with DTC. PMID:26886954

  13. Kinetic Interface

    DEFF Research Database (Denmark)

    2009-01-01

    A kinetic interface for orientation detection in a video training system is disclosed. The interface includes a balance platform instrumented with inertial motion sensors. The interface engages a participant's sense of balance in training exercises.......A kinetic interface for orientation detection in a video training system is disclosed. The interface includes a balance platform instrumented with inertial motion sensors. The interface engages a participant's sense of balance in training exercises....

  14. Kinetics of Surfactant Desorption at an Air–Solution Interface

    KAUST Repository

    Morgan, C. E.; Breward, C. J. W.; Griffiths, I. M.; Howell, P. D.; Penfold, J.; Thomas, R. K.; Tucker, I.; Petkov, J. T.; Webster, J. R. P.

    2012-01-01

    The kinetics of re-equilibration of the anionic surfactant sodium dodecylbenzene sulfonate at the air-solution interface have been studied using neutron reflectivity. The experimental arrangement incorporates a novel flow cell in which the subphase

  15. Neutron detector

    Science.gov (United States)

    Stephan, Andrew C [Knoxville, TN; Jardret,; Vincent, D [Powell, TN

    2011-04-05

    A neutron detector has a volume of neutron moderating material and a plurality of individual neutron sensing elements dispersed at selected locations throughout the moderator, and particularly arranged so that some of the detecting elements are closer to the surface of the moderator assembly and others are more deeply embedded. The arrangement captures some thermalized neutrons that might otherwise be scattered away from a single, centrally located detector element. Different geometrical arrangements may be used while preserving its fundamental characteristics. Different types of neutron sensing elements may be used, which may operate on any of a number of physical principles to perform the function of sensing a neutron, either by a capture or a scattering reaction, and converting that reaction to a detectable signal. High detection efficiency, an ability to acquire spectral information, and directional sensitivity may be obtained.

  16. Fusion neutronics

    CERN Document Server

    Wu, Yican

    2017-01-01

    This book provides a systematic and comprehensive introduction to fusion neutronics, covering all key topics from the fundamental theories and methodologies, as well as a wide range of fusion system designs and experiments. It is the first-ever book focusing on the subject of fusion neutronics research. Compared with other nuclear devices such as fission reactors and accelerators, fusion systems are normally characterized by their complex geometry and nuclear physics, which entail new challenges for neutronics such as complicated modeling, deep penetration, low simulation efficiency, multi-physics coupling, etc. The book focuses on the neutronics characteristics of fusion systems and introduces a series of theories and methodologies that were developed to address the challenges of fusion neutronics, and which have since been widely applied all over the world. Further, it introduces readers to neutronics design’s unique principles and procedures, experimental methodologies and technologies for fusion systems...

  17. Neutron spectometers

    International Nuclear Information System (INIS)

    Poortmans, F.

    1977-01-01

    Experimental work in the field of low-energy neutron physics can be subdivided into two classes: 1)Study of the decay process of the compound-nucleus state as for example the study of the capture gamma rays and of the neutron induced fission process; 2)Study of the reaction mechanism, mainly by measuring the reaction cross-sections and resonance parameters. These neutron cross-sections and resonance parameters are also important data required for many technological applications especially for reactor development programmes. In general, the second class of experiments impose other requirements on the neutron spectrometer than the first class. In most cases, a better neutron energy resolution and a broader neutron energy range are required for the study of the reaction mechanism than for the study of various aspects of the decay process. (author)

  18. Nodal wear model: corrosion in carbon blast furnace hearths

    Directory of Open Access Journals (Sweden)

    Verdeja, L. F.

    2003-06-01

    Full Text Available Criterions developed for the Nodal Wear Model (NWM were applied to estimate the shape of the corrosion profiles that a blast furnace hearth may acquire during its campaign. Taking into account design of the hearth, the boundary conditions, the characteristics of the refractory materials used and the operation conditions of the blast furnace, simulation of wear profiles with central well, mushroom and elephant foot shape were accomplished. The foundations of the NWM are constructed considering that the corrosion of the refractory is a function of the temperature present at each point (node of the liquid metal-refractory interface and the corresponding physical and chemical characteristics of the corrosive fluid.

    Se aplican los criterios del Modelo de Desgaste Nodal (MDN para la estimación de los perfiles de corrosión que podría ir adquiriendo el crisol de un homo alto durante su campaña. Atendiendo al propio diseño del crisol, a las condiciones límites de contorno, a las características del material refractario utilizado y a las condiciones de operación del horno, se consiguen simular perfiles de desgaste con "pozo central", con "forma de seta" ó de "pie de elefante". Los fundamentos del MDN se apoyan en la idea de considerar que la corrosión del refractario es función de la temperatura que el sistema pueda presentar en cada punto (nodo de la intercara refractario-fundido y de las correspondientes características físico-químicas del fluido corrosivo.

  19. Calculation of accurate albedo boundary conditions for three-dimensional nodal diffusion codes by the method of characteristics

    International Nuclear Information System (INIS)

    Petkov, Petko T.

    2000-01-01

    Most of the few-group three-dimensional nodal diffusion codes used for neutronics calculations of the WWER reactors use albedo type boundary conditions on the core-reflector boundary. The conventional albedo are group-to-group reflection probabilities, defined on each outer node face. The method of characteristics is used to calculate accurate albedo by the following procedure. A many-group two-dimensional heterogeneous core-reflector problem, including a sufficient part of the core and detailed description of the adjacent reflector, is solved first. From this solution the angular flux on the core-reflector boundary is calculated in all groups for all traced neutron directions. Accurate boundary conditions can be calculated for the radial, top and bottom reflectors as well as for the absorber part of the WWER-440 reactor control assemblies. The algorithm can be used to estimate also albedo, coupling outer node faces on the radial reflector in the axial direction. Numerical results for the WWER-440 reactor are presented. (Authors)

  20. Neutron exposure

    International Nuclear Information System (INIS)

    Prillinger, G.; Konynenburg, R.A. van

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 6, LWR-PV neutron transport calculations and dosimetry methods and how they are combined to evaluate the neutron exposure of the steel of pressure vessels are discussed. An effort to correlate neutron exposure parameters with damage is made