A multi-level surface rebalancing approach for efficient convergence acceleration of 3D full core multi-group fine grid nodal diffusion iterations

International Nuclear Information System (INIS)

Geemert, René van

2014-01-01

Highlights: • New type of multi-level rebalancing approach for nodal transport. • Generally improved and more mesh-independent convergence behavior. • Importance for intended regime of 3D pin-by-pin core computations. - Abstract: A new multi-level surface rebalancing (MLSR) approach has been developed, aimed at enabling an improved non-linear acceleration of nodal flux iteration convergence in 3D steady-state and transient reactor simulation. This development is meant specifically for anticipating computational needs for solving envisaged multi-group diffusion-like SP N calculations with enhanced mesh resolution (i.e. 3D multi-box up to 3D pin-by-pin grid). For the latter grid refinement regime, the previously available multi-level coarse mesh rebalancing (MLCMR) strategy has been observed to become increasingly inefficient with increasing 3D mesh resolution. Furthermore, for very fine 3D grids that feature a very fine axial mesh as well, non-convergence phenomena have been observed to emerge. In the verifications pursued up to now, these problems have been resolved by the new approach. The novelty arises from taking the interface current balance equations defined over all Cartesian box edges, instead of the nodal volume-integrated process-rate balance equation, as an appropriate restriction basis for setting up multi-level acceleration of fine grid interface current iterations. The new restriction strategy calls for the use of a newly derived set of adjoint spectral equations that are needed for computing a limited set of spectral response vectors per node. This enables a straightforward determination of group-condensed interface current spectral coupling operators that are of crucial relevance in the new rebalancing setup. Another novelty in the approach is a new variational method for computing the neutronic eigenvalue. Within this context, the latter is treated as a control parameter for driving another, newly defined and numerically more fundamental

Present status of reactor physics in the United States and Japan-II. 6. 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 supersede the capabilities of past simulator packages from the original GNF partners: GE (Ref. 1), Hitachi (Ref. 2), and Toshiba (Ref. 3). 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. The status of the GNF new nodal simulator is presented. It is built on a highly automated software base by combining the best technologies of GE, Hitachi, and Toshiba and will provide a BWR core analysis tool with high functionality and fidelity. (authors)

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)

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

Procedure to Generate the MPACT Multigroup Library

Energy Technology Data Exchange (ETDEWEB)

Kim, Kang Seog [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-12-17

The CASL neutronics simulator MPACT is under development for the neutronics and T-H coupled simulation for the light water reactor. The objective of this document is focused on reviewing the current procedure to generate the MPACT multigroup library. Detailed methodologies and procedures are included in this document for further discussion to improve the MPACT multigroup library.

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

A closed-form solution for the two-dimensional transport equation by the LTS{sub N} nodal method in the energy range of Compton effect

Energy Technology Data Exchange (ETDEWEB)

Rodriguez, B.D.A., E-mail: barbararodriguez@furg.b [Universidade Federal do Rio Grande, Instituto de Matematica, Estatistica e Fisica, Rio Grande, RS (Brazil); Vilhena, M.T., E-mail: vilhena@mat.ufrgs.b [Universidade Federal do Rio Grande do Sul, Departamento de Matematica Pura e Aplicada, Porto Alegre, RS (Brazil); Hoff, G., E-mail: hoff@pucrs.b [Pontificia Universidade Catolica do Rio Grande do Sul, Faculdade de Fisica, Porto Alegre, RS (Brazil); Bodmann, B.E.J., E-mail: bardo.bodmann@ufrgs.b [Universidade Federal do Rio Grande do Sul, Departamento de Matematica Pura e Aplicada, Porto Alegre, RS (Brazil)

2011-01-15

In the present work we report on a closed-form solution for the two-dimensional Compton transport equation by the LTS{sub N} nodal method in the energy range of Compton effect. The solution is determined using the LTS{sub N} nodal approach for homogeneous and heterogeneous rectangular domains, assuming the Klein-Nishina scattering kernel and a multi-group model. The solution is obtained by two one-dimensional S{sub N} equation systems resulting from integrating out one of the orthogonal variables of the S{sub N} equations in the rectangular domain. The leakage angular fluxes are approximated by exponential forms, which allows to determine a closed-form solution for the photons transport equation. The angular flux and the parameters of the medium are used for the calculation of the absorbed energy in rectangular domains with different dimensions and compositions. In this study, only the absorbed energy by Compton effect is considered. We present numerical simulations and comparisons with results obtained by using the simulation platform GEANT4 (version 9.1) with its low energy libraries.

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)

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

Numerical analysis for multi-group neutron-diffusion equation using Radial Point Interpolation Method (RPIM)

International Nuclear Information System (INIS)

Kim, Kyung-O; Jeong, Hae Sun; Jo, Daeseong

2017-01-01

Highlights: • Employing the Radial Point Interpolation Method (RPIM) in numerical analysis of multi-group neutron-diffusion equation. • Establishing mathematical formation of modified multi-group neutron-diffusion equation by RPIM. • Performing the numerical analysis for 2D critical problem. - Abstract: A mesh-free method is introduced to overcome the drawbacks (e.g., mesh generation and connectivity definition between the meshes) of mesh-based (nodal) methods such as the finite-element method and finite-difference method. In particular, the Point Interpolation Method (PIM) using a radial basis function is employed in the numerical analysis for the multi-group neutron-diffusion equation. The benchmark calculations are performed for the 2D homogeneous and heterogeneous problems, and the Multiquadrics (MQ) and Gaussian (EXP) functions are employed to analyze the effect of the radial basis function on the numerical solution. Additionally, the effect of the dimensionless shape parameter in those functions on the calculation accuracy is evaluated. According to the results, the radial PIM (RPIM) can provide a highly accurate solution for the multiplication eigenvalue and the neutron flux distribution, and the numerical solution with the MQ radial basis function exhibits the stable accuracy with respect to the reference solutions compared with the other solution. The dimensionless shape parameter directly affects the calculation accuracy and computing time. Values between 1.87 and 3.0 for the benchmark problems considered in this study lead to the most accurate solution. The difference between the analytical and numerical results for the neutron flux is significantly increased in the edge of the problem geometry, even though the maximum difference is lower than 4%. This phenomenon seems to arise from the derivative boundary condition at (x,0) and (0,y) positions, and it may be necessary to introduce additional strategy (e.g., the method using fictitious points and

Numerical method for multigroup one-dimensional SN eigenvalue problems with no spatial truncation error

International Nuclear Information System (INIS)

Abreu, M.P.; Filho, H.A.; Barros, R.C.

1993-01-01

The authors describe a new nodal method for multigroup slab-geometry discrete ordinates S N eigenvalue problems that is completely free from all spatial truncation errors. The unknowns in the method are the node-edge angular fluxes, the node-average angular fluxes, and the effective multiplication factor k eff . The numerical values obtained for these quantities are exactly those of the dominant analytic solution of the S N eigenvalue problem apart from finite arithmetic considerations. This method is based on the use of the standard balance equation and two nonstandard auxiliary equations. In the nonmultiplying regions, e.g., the reflector, we use the multigroup spectral Green's function (SGF) auxiliary equations. In the fuel regions, we use the multigroup spectral diamond (SD) auxiliary equations. The SD auxiliary equation is an extension of the conventional auxiliary equation used in the diamond difference (DD) method. This hybrid characteristic of the SD-SGF method improves both the numerical stability and the convergence rate

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

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)

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)

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

Development and verification of a high performance multi-group SP3 transport capability in the ARTEMIS core simulator

International Nuclear Information System (INIS)

Van Geemert, Rene

2008-01-01

For satisfaction of future global customer needs, dedicated efforts are being coordinated internationally and pursued continuously at AREVA NP. The currently ongoing CONVERGENCE project is committed to the development of the ARCADIA R next generation core simulation software package. ARCADIA R will be put to global use by all AREVA NP business regions, for the entire spectrum of core design processes, licensing computations and safety studies. As part of the currently ongoing trend towards more sophisticated neutronics methodologies, an SP 3 nodal transport concept has been developed for ARTEMIS which is the steady-state and transient core simulation part of ARCADIA R . For enabling a high computational performance, the SP N calculations are accelerated by applying multi-level coarse mesh re-balancing. In the current implementation, SP 3 is about 1.4 times as expensive computationally as SP 1 (diffusion). The developed SP 3 solution concept is foreseen as the future computational workhorse for many-group 3D pin-by-pin full core computations by ARCADIA R . With the entire numerical workload being highly parallelizable through domain decomposition techniques, associated CPU-time requirements that adhere to the efficiency needs in the nuclear industry can be expected to become feasible in the near future. The accuracy enhancement obtainable by using SP 3 instead of SP 1 has been verified by a detailed comparison of ARTEMIS 16-group pin-by-pin SP N results with KAERI's DeCart reference results for the 2D pin-by-pin Purdue UO 2 /MOX benchmark. This article presents the accuracy enhancement verification and quantifies the achieved ARTEMIS-SP 3 computational performance for a number of 2D and 3D multi-group and multi-box (up to pin-by-pin) core computations. (authors)

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

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)

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%

A closed-form solution for the two-dimensional transport equation by the LTSN nodal method in the range of Compton Effect

International Nuclear Information System (INIS)

Rodriguez, Barbara D.A.; Tullio de Vilhena, Marco; Hoff, Gabriela

2008-01-01

In this paper we report a two-dimensional LTS N nodal solution for homogeneous and heterogeneous rectangular domains, assuming the Klein-Nishina scattering kernel and multigroup model. The main idea relies on the solution of the two one-dimensional S N equations resulting from transverse integration of the S N equations in the rectangular domain by the LTS N nodal method, considering the leakage angular fluxes approximated by exponential, which allow us to determine a closed-form solution for the photons transport equation. The angular flux and the parameters of the medium are used for the calculation of the absorbed energy in rectangular domains with different dimensions and compositions. The incoming photons will be tracked until their whole energy is deposited and/or they leave the domain of interest. In this study, the absorbed energy by Compton Effect will be considered. The remaining effects will not be taken into account. We present numerical simulations and comparisons with results obtained by using Geant4 (version 9.1) program which applies the Monte Carlo's technique to low energy libraries for a two-dimensional problem assuming the Klein-Nishina scattering kernel. (authors)

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)

Interface discontinuity factors in the modal Eigenspace of the multigroup diffusion matrix

International Nuclear Information System (INIS)

Garcia-Herranz, N.; Herrero, J.J.; Cuervo, D.; Ahnert, C.

2011-01-01

Interface discontinuity factors based on the Generalized Equivalence Theory are commonly used in nodal homogenized diffusion calculations so that diffusion average values approximate heterogeneous higher order solutions. In this paper, an additional form of interface correction factors is presented in the frame of the Analytic Coarse Mesh Finite Difference Method (ACMFD), based on a correction of the modal fluxes instead of the physical fluxes. In the ACMFD formulation, implemented in COBAYA3 code, the coupled multigroup diffusion equations inside a homogenized region are reduced to a set of uncoupled modal equations through diagonalization of the multigroup diffusion matrix. Then, physical fluxes are transformed into modal fluxes in the Eigenspace of the diffusion matrix. It is possible to introduce interface flux discontinuity jumps as the difference of heterogeneous and homogeneous modal fluxes instead of introducing interface discontinuity factors as the ratio of heterogeneous and homogeneous physical fluxes. The formulation in the modal space has been implemented in COBAYA3 code and assessed by comparison with solutions using classical interface discontinuity factors in the physical space. (author)

The General-Use Nodal Network Solver (GUNNS) Modeling Package for Space Vehicle Flow System Simulation

Science.gov (United States)

Harvey, Jason; Moore, Michael

2013-01-01

The General-Use Nodal Network Solver (GUNNS) is a modeling software package that combines nodal analysis and the hydraulic-electric analogy to simulate fluid, electrical, and thermal flow systems. GUNNS is developed by L-3 Communications under the TS21 (Training Systems for the 21st Century) project for NASA Johnson Space Center (JSC), primarily for use in space vehicle training simulators at JSC. It has sufficient compactness and fidelity to model the fluid, electrical, and thermal aspects of space vehicles in real-time simulations running on commodity workstations, for vehicle crew and flight controller training. It has a reusable and flexible component and system design, and a Graphical User Interface (GUI), providing capability for rapid GUI-based simulator development, ease of maintenance, and associated cost savings. GUNNS is optimized for NASA's Trick simulation environment, but can be run independently of Trick.

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)

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

Multi-Group Covariance Data Generation from Continuous-Energy Monte Carlo Transport Calculations

International Nuclear Information System (INIS)

Lee, Dong Hyuk; Shim, Hyung Jin

2015-01-01

The sensitivity and uncertainty (S/U) methodology in deterministic tools has been utilized for quantifying uncertainties of nuclear design parameters induced by those of nuclear data. The S/U analyses which are based on multi-group cross sections can be conducted by an simple error propagation formula with the sensitivities of nuclear design parameters to multi-group cross sections and the covariance of multi-group cross section. The multi-group covariance data required for S/U analysis have been produced by nuclear data processing codes such as ERRORJ or PUFF from the covariance data in evaluated nuclear data files. However in the existing nuclear data processing codes, an asymptotic neutron flux energy spectrum, not the exact one, has been applied to the multi-group covariance generation since the flux spectrum is unknown before the neutron transport calculation. It can cause an inconsistency between the sensitivity profiles and the covariance data of multi-group cross section especially in resolved resonance energy region, because the sensitivities we usually use are resonance self-shielded while the multi-group cross sections produced from an asymptotic flux spectrum are infinitely-diluted. In order to calculate the multi-group covariance estimation in the ongoing MC simulation, mathematical derivations for converting the double integration equation into a single one by utilizing sampling method have been introduced along with the procedure of multi-group covariance tally

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)

ESELEM 4: a code for calculating fine neutron spectrum and multi-group cross sections in plate lattice

International Nuclear Information System (INIS)

Nakagawa, Masayuki; Katsuragi, Satoru; Narita, Hideo.

1976-07-01

The multi-group treatment has been used in the design study of fast reactors and analysis of experiments at fast critical assemblies. The accuracy of the multi-group cross sections therefore affects strongly the results of these analyses. The ESELEM 4 code has been developed to produce multi-group cross sections with an advanced method from the nuclear data libraries used in the JAERI Fast set. ESELEM 4 solves integral transport equation by the collision probability method in plate lattice geometry to obtain the fine neutron spectrum. A typical fine group mesh width is 0.008 in lethargy unit. The multi-group cross sections are calculated by weighting the point data with the fine structure neutron flux. Some devices are applied to reduce computation time and computer core storage required for the calculation. The slowing down sources are calculated with the use of a recurrence formula derived for elastic and inelastic scattering. The broad group treatment is adopted above 2 MeV for dealing with both light any heavy elements. Also the resonance cross sections of heavy elements are represented in a broad group structure, for which we use the values of the JAERI Fast set. The library data are prepared by the PRESM code from ENDF/A type nuclear data files. The cross section data can be compactly stored in the fast computer core memory for saving the core storage and data processing time. The programme uses the variable dimensions to increase its flexibility. The users' guide for ESELEM 4 and PRESM is also presented in this report. (auth.)

A transient, Hex-Z nodal code corrected by discontinuity factors

International Nuclear Information System (INIS)

Shatilla, Y.A.M.; Henry, A.F.

1993-01-01

This document constitutes Volume 1 of the Final Report of a three-year study supported by the special Research Grant Program for Nuclear Energy Research set up by the US Department of Energy. The original motivation for the work was to provide a fast and accurate computer program for the analysis of transients in heavy water or graphite-moderated reactors being considered as candidates for the New Production Reactor. Thus, part of the funding was by way of pass-through money from the Savannah River Laboratory. With this intent in mind, a three-dimensional (Hex-Z), general-energy-group transient, nodal code was created, programmed, and tested. In order to improve accuracy, correction terms, called open-quotes discontinuity factors,close quotes were incorporated into the nodal equations. Ideal values of these factors force the nodal equations to provide node-integrated reaction rates and leakage rates across nodal surfaces that match exactly those edited from a more exact reference calculation. Since the exact reference solution is needed to compute the ideal discontinuity factors, the fact that they result in exact nodal equations would be of little practical interest were it not that approximate discontinuity factors, found at a greatly reduced cost, often yield very accurate results. For example, for light-water reactors, discontinuity factors found from two-dimensional, fine-mesh, multigroup transport solutions for two-dimensional cuts of a fuel assembly provide very accurate predictions of three-dimensional, full-core power distributions. The present document (volume 1) deals primarily with the specification, programming and testing of the three-dimensional, Hex-Z computer program. The program solves both the static (eigenvalue) and transient, general-energy-group, nodal equations corrected by user-supplied discontinuity factors

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.

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)

Multigroup calculations of low-energy neutral transport in tokamak plasmas

International Nuclear Information System (INIS)

Gilligan, J.G.; Gralnick, S.L.; Price, W.G. Jr.; Kammash, T.

1978-01-01

Multigroup discrete ordinates methods avoid many of the approximations that have been used in previous neutral transport analyses. Of particular interest are the neutral profiles generated as an integral part of larger plasma system simulation codes. To determine the appropriateness of utilizing a particular multigroup code, ANISN, for this purpose, results are compared with the neutral transport module of the Duechs code. For a typical TFTR plasma, predicted neutral densities differ by a maximum factor of three on axis and outfluxes at the plasma boundary by approximately 40%. This is found to be significant for a neutral transport module. Possible sources of the observed discrepancies are indicated from an analysis of the approximations used in the Duechs model. Recommendations are made concerning the future application of the multigroup method. (author)

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

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.

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

The orphan receptor ALK7 and the Activin receptor ALK4 mediate signaling by Nodal proteins during vertebrate development

Science.gov (United States)

Reissmann, Eva; Jörnvall, Henrik; Blokzijl, Andries; Andersson, Olov; Chang, Chenbei; Minchiotti, Gabriella; Persico, M. Graziella; Ibáñez, Carlos F.; Brivanlou, Ali H.

2001-01-01

Nodal proteins have crucial roles in mesendoderm formation and left–right patterning during vertebrate development. The molecular mechanisms of signal transduction by Nodal and related ligands, however, are not fully understood. In this paper, we present biochemical and functional evidence that the orphan type I serine/threonine kinase receptor ALK7 acts as a receptor for mouse Nodal and Xenopus Nodal-related 1 (Xnr1). Receptor reconstitution experiments indicate that ALK7 collaborates with ActRIIB to confer responsiveness to Xnr1 and Nodal. Both receptors can independently bind Xnr1. In addition, Cripto, an extracellular protein genetically implicated in Nodal signaling, can independently interact with both Xnr1 and ALK7, and its expression greatly enhances the ability of ALK7 and ActRIIB to respond to Nodal ligands. The Activin receptor ALK4 is also able to mediate Nodal signaling but only in the presence of Cripto, with which it can also interact directly. A constitutively activated form of ALK7 mimics the mesendoderm-inducing activity of Xnr1 in Xenopus embryos, whereas a dominant-negative ALK7 specifically blocks the activities of Nodal and Xnr1 but has little effect on other related ligands. In contrast, a dominant-negative ALK4 blocks all mesoderm-inducing ligands tested, including Nodal, Xnr1, Xnr2, Xnr4, and Activin. In agreement with a role in Nodal signaling, ALK7 mRNA is localized to the ectodermal and organizer regions of Xenopus gastrula embryos and is expressed during early stages of mouse embryonic development. Therefore, our results indicate that both ALK4 and ALK7 can mediate signal transduction by Nodal proteins, although ALK7 appears to be a receptor more specifically dedicated to Nodal signaling. PMID:11485994

Nuclear data processing and multigroup cross section generation

International Nuclear Information System (INIS)

Kim, Jeong Do; Kil, Chung Sub

1996-01-01

The multigroup constants for WIMS/CASMO were updated with ENDF/B-VI and were tested. The continuous energy MCNP library developed last year was validated against the LWR-simulated critical experiments. The MCNP library will be used to design and analyze nuclear and shielding facilities. The system for generation of MATXS multigroup library and TRANSX code, which is able to prepare the data for the discrete ordinates and diffusion codes from the MATXS library, was established. The MATXS libraries for analyses of thermal and fast critical experiments were generated and tested. The MATXS/TRANSX system for the discrete ordinates and diffusion codes will be useful for nuclear analyses. 10 tabs., 5 figs., 17 refs. (Author)

MORET: Version 4.B. A multigroup Monte Carlo criticality code

International Nuclear Information System (INIS)

Jacquet, Olivier; Miss, Joachim; Courtois, Gerard

2003-01-01

MORET 4 is a three dimensional multigroup Monte Carlo code which calculates the effective multiplication factor (keff) of any configurations more or less complex as well as reaction rates in the different volumes of the geometry and the leakage out of the system. MORET 4 is the Monte Carlo code of the APOLLO2-MORET 4 standard route of CRISTAL, the French criticality package. It is the most commonly used Monte Carlo code for French criticality calculations. During the last four years, the MORET 4 team has developed or improved the following major points: modernization of the geometry, implementation of perturbation algorithms, source distribution convergence, statistical detection of stationarity, unbiased variance estimation and creation of pre-processing and post-processing tools. The purpose of this paper is not only to present the new features of MORET but also to detail clearly the physical models and the mathematical methods used in the code. (author)

Cyclotron radiation by a multi-group method

International Nuclear Information System (INIS)

Chu, T.C.

1980-01-01

A multi-energy group technique is developed to study conditions under which cyclotron radiation emission can shift a Maxwellian electron distribution into a non-Maxwellian; and if the electron distribution is non-Maxwellian, to study the rate of cyclotron radiation emission as compared to that emitted by a Maxwellian having the same mean electron density and energy. The assumptions in this study are: the electrons should be in an isotropic medium and the magnetic field should be uniform. The multi-group technique is coupled into a multi-group Fokker-Planck computer code to study electron behavior under the influence of cyclotron radiation emission in a self-consistent fashion. Several non-Maxwellian distributions were simulated to compare their cyclotron emissions with the corresponding energy and number density equivalent Maxwellian distribtions

FENDL multigroup libraries

International Nuclear Information System (INIS)

Ganesan, S.; Muir, D.W.

1992-01-01

Selected neutron reaction nuclear data libraries and photon-atomic interaction cross section libraries for elements of interest to the IAEA's program on Fusion Evaluated Nuclear Data Library (FENDL) have been processed into MATXSR format using the NJOY system on the VAX4000 computer of the IAEA. This document lists the resulting multigroup data libraries. All the multigroup data generated are available cost-free upon request from the IAEA Nuclear Data Section. (author). 9 refs

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

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)

THEMIS-4: a coherent punctual and multigroup cross section library for Monte Carlo and SN codes from ENDF/B4

International Nuclear Information System (INIS)

Dejonghe, G.; Gonnord, J.; Monnier, A.; Nimal, J.C.

1983-05-01

The THEMIS cross section processing system has been developped to produce punctual data for MONTE CARLO and coherent multigroup data for SN codes from ENDF/B. The THEMIS-4 data base has been generated from ENDF/B4 using the system and can be accessed by the 3-D Monte Carlo system TRIPOLI-2 and by the SN codes ANISN and DOT. An interpretation of ORNL fusion shielding benchmark is presented

An improved one-and-a-half group BWR core simulator for a new-generation core management system

International Nuclear Information System (INIS)

Iwamoto, Tatsuya; Yamamoto, Munenari

2000-01-01

An improved one-and-a-half group core simulator method for a next-generation BWR core management system is presented. In the improved method, intranodal spectral index (thermal to fast flux ratio) is expanded with analytic solutions to the diffusion equation, and the nodal power density and the interface net current are calculated, taking the intranodal flux shape into consideration. A unique method was developed for assembly heterogeneity correction. Thus eliminating the insufficiencies of the conventional one-and-a-half group method, we can have accurate power distributions as well as local peaking factors for cores having large spectral mismatch between fuel assemblies. The historical effects of spectral mismatch are also considered in both nodal power and local peaking calculations. Although reflectors are not solved explicitly, there is essentially no need for core dependent adjustable parameters, since boundary conditions are derived in the same manner as in the interior nodes. Calculation time for nodal solutions is comparable to that for the conventional method, and is less than 1/10 of a few-group nodal simulator. Verifications of the present method were made by comparing the results with those obtained by heterogeneous fine-mesh multi-group core depletion calculations, and the accuracy was shown to be fairly good. (author)

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.

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)

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)

Sensitivity analysis of MIDAS tests using SPACE code. Effect of nodalization

International Nuclear Information System (INIS)

Eom, Shin; Oh, Seung-Jong; Diab, Aya

2018-01-01

The nodalization sensitivity analysis for the ECCS (Emergency Core Cooling System) bypass phe�nomena was performed using the SPACE (Safety and Performance Analysis CodE) thermal hydraulic analysis computer code. The results of MIDAS (Multi-�dimensional Investigation in Downcomer Annulus Simulation) test were used. The MIDAS test was conducted by the KAERI (Korea Atomic Energy Research Institute) for the performance evaluation of the ECC (Emergency Core Cooling) bypass phenomenon in the DVI (Direct Vessel Injection) system. The main aim of this study is to examine the sensitivity of the SPACE code results to the number of thermal hydraulic channels used to model the annulus region in the MIDAS experiment. The numerical model involves three nodalization cases (4, 6, and 12 channels) and the result show that the effect of nodalization on the bypass fraction for the high steam flow rate MIDAS tests is minimal. For computational efficiency, a 4 channel representation is recommended for the SPACE code nodalization. For the low steam flow rate tests, the SPACE code over-�predicts the bypass fraction irrespective of the nodalization finesse. The over-�prediction at low steam flow may be attributed to the difficulty to accurately represent the flow regime in the vicinity of the broken cold leg.

Sensitivity analysis of MIDAS tests using SPACE code. Effect of nodalization

Energy Technology Data Exchange (ETDEWEB)

Eom, Shin; Oh, Seung-Jong; Diab, Aya [KEPCO International Nuclear Graduate School (KINGS), Ulsan (Korea, Republic of). Dept. of NPP Engineering

2018-02-15

The nodalization sensitivity analysis for the ECCS (Emergency Core Cooling System) bypass phe�nomena was performed using the SPACE (Safety and Performance Analysis CodE) thermal hydraulic analysis computer code. The results of MIDAS (Multi-�dimensional Investigation in Downcomer Annulus Simulation) test were used. The MIDAS test was conducted by the KAERI (Korea Atomic Energy Research Institute) for the performance evaluation of the ECC (Emergency Core Cooling) bypass phenomenon in the DVI (Direct Vessel Injection) system. The main aim of this study is to examine the sensitivity of the SPACE code results to the number of thermal hydraulic channels used to model the annulus region in the MIDAS experiment. The numerical model involves three nodalization cases (4, 6, and 12 channels) and the result show that the effect of nodalization on the bypass fraction for the high steam flow rate MIDAS tests is minimal. For computational efficiency, a 4 channel representation is recommended for the SPACE code nodalization. For the low steam flow rate tests, the SPACE code over-�predicts the bypass fraction irrespective of the nodalization finesse. The over-�prediction at low steam flow may be attributed to the difficulty to accurately represent the flow regime in the vicinity of the broken cold leg.

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

Generating and verification of ACE-multigroup library for MCNP

International Nuclear Information System (INIS)

Chen Chaobin; Hu Zehua; Chen Yixue; Wu Jun; Yang Shouhai

2012-01-01

The Monte Carlo code MCNP can handle multigroup calculations and a sample multigroup set based on ENDF/B-V, MGXSNP, is available for MCNP for coupled neutron-photon transport. However, this library is not suit- able for all problems, and there is a need for users to be able to generate multigroup libraries tailored to their specific applications. For these purposes CSPT (cross section processing tool) is created to generate multigroup library for MCNP from deterministic multigroup cross sections (GENDF or ANISN format at present). Several ACE-multigroup libraries based on ENDF/B-VII.0 converted and verified in this work, we drawn the conclusion that the CSPT code works correctly and the libraries produced are credible. (authors)

Nodal methods with non linear feedback for the three dimensional resolution of the diffusion's multigroup equations

International Nuclear Information System (INIS)

Ferri, A.A.

1986-01-01

Nodal methods applied in order to calculate the power distribution in a nuclear reactor core are presented. These methods have received special attention, because they yield accurate results in short computing times. Present nodal schemes contain several unknowns per node and per group. In the methods presented here, non linear feedback of the coupling coefficients has been applied to reduce this number to only one unknown per node and per group. The resulting algorithm is a 7- points formula, and the iterative process has proved stable in the response matrix scheme. The intranodal flux shape is determined by partial integration of the diffusion equations over two of the coordinates, leading to a set of three coupled one-dimensional equations. These can be solved by using a polynomial approximation or by integration (analytic solution). The tranverse net leakage is responsible for the coupling between the spatial directions, and two alternative methods are presented to evaluate its shape: direct parabolic approximation and local model expansion. Numerical results, which include the IAEA two-dimensional benchmark problem illustrate the efficiency of the developed methods. (M.E.L.) [es

Implantation of multigroup diffusion code 2DB in the IEAv CDC CYBER 170/750 system, and its preliminary evaluation

International Nuclear Information System (INIS)

Prati, A.; Anaf, J.

1988-09-01

The IBM version of the multigroup diffusion code 2DB was implemented in the IEAv CDC CYBER 170/750 system. It was optimized relative to the use of the central memory, limited to 132 K-words, through the memory manager CMM and its partition into three source codes: rectangular and cylindrical geometries, triangular geometry and hexagonal geometry. The reactangular, triangular and hexagonal geometry nodal options were revised and optimized. A fast reactor and a PWR type thermal reactor sample cases were studied. The results are presented and analized. An updated 2DB code user's manual was written in Portugueses and published separately. (author) [pt

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.

Multigroup and coupled forward-adjoint Monte Carlo calculation efficiencies for secondary neutron doses from proton beams

International Nuclear Information System (INIS)

Kelsey IV, Charles T.; Prinja, Anil K.

2011-01-01

We evaluate the Monte Carlo calculation efficiency for multigroup transport relative to continuous energy transport using the MCNPX code system to evaluate secondary neutron doses from a proton beam. We consider both fully forward simulation and application of a midway forward adjoint coupling method to the problem. Previously we developed tools for building coupled multigroup proton/neutron cross section libraries and showed consistent results for continuous energy and multigroup proton/neutron transport calculations. We observed that forward multigroup transport could be more efficient than continuous energy. Here we quantify solution efficiency differences for a secondary radiation dose problem characteristic of proton beam therapy problems. We begin by comparing figures of merit for forward multigroup and continuous energy MCNPX transport and find that multigroup is 30 times more efficient. Next we evaluate efficiency gains for coupling out-of-beam adjoint solutions with forward in-beam solutions. We use a variation of a midway forward-adjoint coupling method developed by others for neutral particle transport. Our implementation makes use of the surface source feature in MCNPX and we use spherical harmonic expansions for coupling in angle rather than solid angle binning. The adjoint out-of-beam transport for organs of concern in a phantom or patient can be coupled with numerous forward, continuous energy or multigroup, in-beam perturbations of a therapy beam line configuration. Out-of-beam dose solutions are provided without repeating out-of-beam transport. (author)

Multi-group transport methods for high-resolution neutron activation analysis

International Nuclear Information System (INIS)

Burns, K. A.; Smith, L. E.; Gesh, C. J.; Shaver, M. W.

2009-01-01

The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples. In these applications, high-resolution gamma-ray spectrometers are used to preserve as much information as possible about the emitted photon flux, which consists of both continuum and characteristic gamma rays with discrete energies. Monte Carlo transport is the most commonly used modeling tool for this type of problem, but computational times for many problems can be prohibitive. This work explores the use of multi-group deterministic methods for the simulation of neutron activation problems. Central to this work is the development of a method for generating multi-group neutron-photon cross-sections in a way that separates the discrete and continuum photon emissions so that the key signatures in neutron activation analysis (i.e., the characteristic line energies) are preserved. The mechanics of the cross-section preparation method are described and contrasted with standard neutron-gamma cross-section sets. These custom cross-sections are then applied to several benchmark problems. Multi-group results for neutron and photon flux are compared to MCNP results. Finally, calculated responses of high-resolution spectrometers are compared. Preliminary findings show promising results when compared to MCNP. A detailed discussion of the potential benefits and shortcomings of the multi-group-based approach, in terms of accuracy, and computational efficiency, is provided. (authors)

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)

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

The BWR core simulator COSIMA with 2 group nodal flux expansion and control rod history

International Nuclear Information System (INIS)

Hoejerup, C.F.

1989-08-01

The boiling water simulator NOTAM has been modified and improved in several aspects: - The ''1 1/2'' energy group TRILUX nodal flux solution method has been exchanged with a 2 group modal expansion method. - Control rod ''history'' has been introduced. - Precalculated instrument factors have been introduced. The paper describes these improvements, which were considered sufficiently large to justify a new name to the programme: COSIMA. (author)

Offensive Strategy in the 2D Soccer Simulation League Using Multi-Group Ant Colony Optimization

Directory of Open Access Journals (Sweden)

Shengbing Chen

2016-02-01

Full Text Available The 2D soccer simulation league is one of the best test beds for the research of artificial intelligence (AI. It has achieved great successes in the domain of multi-agent cooperation and machine learning. However, the problem of integral offensive strategy has not been solved because of the dynamic and unpredictable nature of the environment. In this paper, we present a novel offensive strategy based on multi-group ant colony optimization (MACO-OS. The strategy uses the pheromone evaporation mechanism to count the preference value of each attack action in different environments, and saves the values of success rate and preference in an attack information tree in the background. The decision module of the attacker then selects the best attack action according to the preference value. The MACO-OS approach has been successfully implemented in our 2D soccer simulation team in RoboCup competitions. The experimental results have indicated that the agents developed with this strategy, along with related techniques, delivered outstanding performances.

On the convergence of multigroup discrete-ordinates approximations

International Nuclear Information System (INIS)

Victory, H.D. Jr.; Allen, E.J.; Ganguly, K.

1987-01-01

Our analysis is divided into two distinct parts which we label for convenience as Part A and Part B. In Part A, we demonstrate that the multigroup discrete-ordinates approximations are well-defined and converge to the exact transport solution in any subcritical setting. For the most part, we focus on transport in two-dimensional Cartesian geometry. A Nystroem technique is used to extend the discrete ordinates multigroup approximates to all values of the angular and energy variables. Such an extension enables us to employ collectively compact operator theory to deduce stability and convergence of the approximates. In Part B, we perform a thorough convergence analysis for the multigroup discrete-ordinates method for an anisotropically-scattering subcritical medium in slab geometry. The diamond-difference and step-characteristic spatial approximation methods are each studied. The multigroup neutron fluxes are shown to converge in a Banach space setting under realistic smoothness conditions on the solution. This is the first thorough convergence analysis for the fully-discretized multigroup neutron transport equations

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)

Introduction of corrections taking into account interdependence of multigroup constants to the results of multigroup perturbation theory calculations

International Nuclear Information System (INIS)

Raskach, K. F.

2012-01-01

In multigroup calculations of reactivity and sensitivity coefficients, methodical errors can appear if the interdependence of multigroup constants is not taken into account. For this effect to be taken into account, so-called implicit components of the aforementioned values are introduced. A simple technique for computing these values is proposed. It is based on the use of subgroup parameters.

Burnup simulations of an inert matrix fuel using a two region, multigroup reactor physics model

Energy Technology Data Exchange (ETDEWEB)

Schneider, E. [Dept. of Mechanical Engineering, Univ. of Texas at Austin, 1 Univ. Place C2200, Austin, TX 78712 (United States); Deinert, M.; Bingham Cady, K. [Dept. of Theoretical and Applied Mechanics, Cornell Univ., Ithaca, NY 14853 (United States)

2006-07-01

Determining the time dependent concentration of isotopes in a nuclear reactor core is of fundamental importance to analysis of nuclear fuel cycles and the impact of spent fuels on long term storage facilities. We present a fast, conceptually simple tool for performing burnup calculations applicable to obtaining isotopic balances as a function of fuel burnup. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to determine the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. The model has been tested against benchmarked results for LWRs burning UOX and MOX, as well as MONTEBURNS simulations of zirconium oxide based IMF, all with strong fidelity. As an illustrative example, VBUDS burnup calculation results for an IMF fuel are presented in this paper. (authors)

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)

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

AMPX: a modular code system for generating coupled multigroup neutron-gamma libraries from ENDF/B

Energy Technology Data Exchange (ETDEWEB)

Greene, N.M.; Lucius, J.L.; Petrie, L.M.; Ford, W.E. III; White, J.E.; Wright, R.Q.

1976-03-01

AMPX is a modular system for producing coupled multigroup neutron-gamma cross section sets. Basic neutron and gamma cross-section data for AMPX are obtained from ENDF/B libraries. Most commonly used operations required to generate and collapse multigroup cross-section sets are provided in the system. AMPX is flexibly dimensioned; neutron group structures, and gamma group structures, and expansion orders to represent anisotropic processes are all arbitrary and limited only by available computer core and budget. The basic processes provided will (1) generate multigroup neutron cross sections; (2) generate multigroup gamma cross sections; (3) generate gamma yields for gamma-producing neutron interactions; (4) combine neutron cross sections, gamma cross sections, and gamma yields into final ''coupled sets''; (5) perform one-dimensional discrete ordinates transport or diffusion theory calculations for neutrons and gammas and, on option, collapse the cross sections to a broad-group structure, using the one-dimensional results as weighting functions; (6) plot cross sections, on option, to facilitate the ''evaluation'' of a particular multigroup set of data; (7) update and maintain multigroup cross section libraries in such a manner as to make it not only easy to combine new data with previously processed data but also to do it in a single pass on the computer; and (8) output multigroup cross sections in convenient formats for other codes. (auth)

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)

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)

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)

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.

Multigroup Moderation Test in Generalized Structured Component Analysis

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Angga Dwi Mulyanto

2016-05-01

Full Text Available Generalized Structured Component Analysis (GSCA is an alternative method in structural modeling using alternating least squares. GSCA can be used for the complex analysis including multigroup. GSCA can be run with a free software called GeSCA, but in GeSCA there is no multigroup moderation test to compare the effect between groups. In this research we propose to use the T test in PLS for testing moderation Multigroup on GSCA. T test only requires sample size, estimate path coefficient, and standard error of each group that are already available on the output of GeSCA and the formula is simple so the user does not need a long time for analysis.

CHARTB multigroup transport package

International Nuclear Information System (INIS)

Baker, L.

1979-03-01

The physics and numerical implementation of the radiation transport routine used in the CHARTB MHD code are discussed. It is a one-dimensional (Cartesian, cylindrical, and spherical symmetry), multigroup,, diffusion approximation. Tests and applications will be discussed as well

Calculation of multigroup reaction rates for the Ghana Research ...

African Journals Online (AJOL)

The discrete ordinate spatial model, which pro-vides solution to the differential form of the transport equation by the Carlson-SN (N=4) approach was adopted to solve the Ludwig-Boltzmann multigroup neutron transport equation for this analysis. The results show that for any fissile resonance absorber, the reaction rates ...

Range calculations using multigroup transport methods

International Nuclear Information System (INIS)

Hoffman, T.J.; Robinson, M.T.; Dodds, H.L. Jr.

1979-01-01

Several aspects of radiation damage effects in fusion reactor neutron and ion irradiation environments are amenable to treatment by transport theory methods. In this paper, multigroup transport techniques are developed for the calculation of particle range distributions. These techniques are illustrated by analysis of Au-196 atoms recoiling from (n,2n) reactions with gold. The results of these calculations agree very well with range calculations performed with the atomistic code MARLOWE. Although some detail of the atomistic model is lost in the multigroup transport calculations, the improved computational speed should prove useful in the solution of fusion material design problems

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)

Multigroup cross section collapsing optimization of a He-3 detector assembly model using deterministic transport techniques

International Nuclear Information System (INIS)

Huang, Mi; Yi, Ce; Manalo, Kevin L.; Sjoden, Glenn E.

2011-01-01

Multigroup optimization is performed on a neutron detector assembly to examine the validity of transport response in forward and adjoint modes. For SN transport simulations, we discuss the multigroup collapse of an 80 group library to 40, 30, and 16 groups, constructed from using the 3-D parallel PENTRAN and macroscopic cross section collapsing with YGROUP contribution weighting. The difference in using P_1 and P_3 Legendre order in scattering cross sections is investigated; also, associated forward and adjoint transport responses are calculated. We conclude that for the block analyzed, a 30 group cross section optimizes both computation time and accuracy relative to the 80 group transport calculations. (author)

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.).

Analytic function expansion nodal method for nuclear reactor core design

International Nuclear Information System (INIS)

Noh, Hae Man

1995-02-01

than the analytic function. The second variation of the AFEN method we developed is the AFEN/PEN hybrid method. This method is designed especially for the multigroup reactor analysis. This hybrid method solves the diffusion equations for the fast energy groups by the PEN method, and those for the thermal energy groups by the AFEN method. This method is based on the observation that the fast group neutron flux distributions are generally so smooth that they can be approximated by a high-order polynomial and that, on the other hand, the thermal fluxes require the analytic function expansion for the representation of their strong gradients near the interface between assemblies having different neutronic properties. The results of benchmark problems on which this method was tested indicate that performance of the hybrid method is much better than that of the PEN method and is nearly the same to that of the AFEN method. In order for the AFEN method and its variations to be used in analyzing the neutron behavior in an actual reactor core, we also developed a new burnup correction model to reduce the errors in nodal flux distributions induced by the intranodal burnup gradients. It is essential for the nodal methods to maintain their accuracy in fuel depletion analysis. The burnup correction model developed in this study homogenizes equivalently the node with the burnup-induced cross section variations into the homogeneous node with the equivalent parameters such as the flux-volume-weighted constant cross sections and the discontinuity factors. The results of a benchmark problem show that this model eliminates almost all the errors in the nodal unknowns which are induced by the intranodal burnup gradients

Nodalization effects on RELAP5 results related to MTR research reactor transient scenarios

Directory of Open Access Journals (Sweden)

Khedr Ahmed

2005-01-01

Full Text Available The present work deals with the anal y sis of RELAP5 results obtained from the evaluation study of the total loss of flow transient with the deficiency of the heat removal system in a research reactor using two different nodalizations. It focuses on the effect of nodalization on the thermal-hydraulic evaluation of the re search reactor. The analysis of RELAP5 results has shown that nodalization has a big effect on the predicted scenario of the postulated transient. There fore, great care should be taken during the nodalization of the reactor, especially when the avail able experimental or measured data are insufficient for making a complete qualification of the nodalization. Our analysis also shows that the research reactor pool simulation has a great effect on the evaluation of natural circulation flow and on other thermal-hydraulic parameters during the loss of flow transient. For example, the on set time of core boiling changes from less than 2000 s to 15000 s, starting from the beginning of the transient. This occurs if the pool is simulated by two vertical volumes in stead of one vertical volume.

Multigroup or multipoint thermal neutron data preparation. Programme SIGMA

International Nuclear Information System (INIS)

Matausek, M.V.; Kunc, M.

1974-01-01

When calculating the space energy distribution of thermal neutrons in reactor lattices, in either the multigroup or the multipoint approximation, it is convenient to divide the problem into two independent parts. Firstly, for all material regions of the given reactor lattice cell, the group or the point values of cross sections, scattering kernel and the outer source of thermal neutrons are calculated by a data preparation programme. These quantities are then used as input, by the programme which solves multigroup or multipoint transport equations, to generate the space energy neutron spectra in the cell considered and to determine the related integral quantities, namely the different reaction rates. The present report deals with the first part of the problem. An algorithm for constructing a set of thermal neutron input data, to be used with the multigroup or multipoint version of the code MULTI /1,2,3/, is presented and the new version of the programme SIGMA /4/, written in FORTRAN IV for the CDC-3600 computer, is described. For a given reactor cell material, composed of a number of different isotopes, this programme calculates the group or the point values of the scattering macroscopic absorption cross section, macroscopic scattering cross section, kernel and the outer source of thermal neutrons. Numerous options are foreseen in the programme, concerning the energy variation of cross sections and a scattering kernel, concerning the weighting spectrum in multigroup scheme or the procedure for constructing the scattering matrix in the multipoint scheme and, finally, concerning the organization of output. The details of the calculational algorithm are presented in Section 2 of the paper. Section 3 contains the description of the programme and the instructions for its use (author)

Verification and validation of multi-group library MUSE1.0 created from ENDF/B-VII.0

International Nuclear Information System (INIS)

Chen Yixue; Wu Jun; Yang Shouhai; Zhang Bin; Lu Daogang; Chen Chaobin

2010-01-01

A multi-group library set named MUSE1.0 with 172-neutron group and 42-photon group is produced based on ENDF/B-VII.0 using NJOY code. Weight function of the multi-group library set is taken from the Vitanim-e library and the max legendre order of scattering matrix is six. All the nuclides have thermal scattering data created using free-gas scattering law and 10 Bondarenko background cross sections se lected to generate the self-shielded multi-group cross sections. The final libraries have GENDF-format, MATXS-format and ACE-multi-group sub-libraries and each sub-library generated under 4 temperatures(293 K,600 K,800 K and 900 K). This paper provides a summary of the procedure to produce the library set and a detail description of the validation of the multi-group library set by several critical benchmark devices and shielding benchmark devices using MCNP code. The ability to handle the thermal neutron transport and resonance self-shielding problems are investigated specially. In the end, we draw the conclusion that the multi-group libraries produced is credible and can be used in the R and D process of Supercritical Water Reactor Design. (authors)

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.

User manual for version 4.3 of the Tripoli-4 Monte-Carlo method particle transport computer code

International Nuclear Information System (INIS)

Both, J.P.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B.

2003-01-01

This manual relates to Version 4.3 TRIPOLI-4 code. TRIPOLI-4 is a computer code simulating the transport of neutrons, photons, electrons and positrons. It can be used for radiation shielding calculations (long-distance propagation with flux attenuation in non-multiplying media) and neutronic calculations (fissile medium, criticality or sub-criticality basis). This makes it possible to calculate k eff (for criticality), flux, currents, reaction rates and multi-group cross-sections. TRIPOLI-4 is a three-dimensional code that uses the Monte-Carlo method. It allows for point-wise description in terms of energy of cross-sections and multi-group homogenized cross-sections and features two modes of geometrical representation: surface and combinatorial. The code uses cross-section libraries in ENDF/B format (such as JEF2-2, ENDF/B-VI and JENDL) for point-wise description cross-sections in APOTRIM format (from the APOLLO2 code) or a format specific to TRIPOLI-4 for multi-group description. (authors)

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.)

User manual for version 4.3 of the Tripoli-4 Monte-Carlo method particle transport computer code; Notice d'utilisation du code Tripoli-4, version 4.3: code de transport de particules par la methode de Monte Carlo

Energy Technology Data Exchange (ETDEWEB)

Both, J.P.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B

2003-07-01

This manual relates to Version 4.3 TRIPOLI-4 code. TRIPOLI-4 is a computer code simulating the transport of neutrons, photons, electrons and positrons. It can be used for radiation shielding calculations (long-distance propagation with flux attenuation in non-multiplying media) and neutronic calculations (fissile medium, criticality or sub-criticality basis). This makes it possible to calculate k{sub eff} (for criticality), flux, currents, reaction rates and multi-group cross-sections. TRIPOLI-4 is a three-dimensional code that uses the Monte-Carlo method. It allows for point-wise description in terms of energy of cross-sections and multi-group homogenized cross-sections and features two modes of geometrical representation: surface and combinatorial. The code uses cross-section libraries in ENDF/B format (such as JEF2-2, ENDF/B-VI and JENDL) for point-wise description cross-sections in APOTRIM format (from the APOLLO2 code) or a format specific to TRIPOLI-4 for multi-group description. (authors)

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

Review of multigroup nuclear cross-section processing

Energy Technology Data Exchange (ETDEWEB)

Trubey, D.K.; Hendrickson, H.R. (comps.)

1978-10-01

These proceedings consist of 18 papers given at a seminar--workshop on ''Multigroup Nuclear Cross-Section Processing'' held at Oak Ridge, Tennessee, March 14--16, 1978. The papers describe various computer code systems and computing algorithms for producing multigroup neutron and gamma-ray cross sections from evaluated data, and experience with several reference data libraries. Separate abstracts were prepared for 13 of the papers. The remaining five have already been cited in ERA, and may be located by referring to the entry CONF-780334-- in the Report Number Index. (RWR)

Self-shielding phenomenon modelling in multigroup transport code Apollo-2; Modelisation du phenomene d'autoprotection dans le code de transport multigroupe Apollo 2

Energy Technology Data Exchange (ETDEWEB)

Coste-Delclaux, M

2006-03-15

This document describes the improvements carried out for modelling the self-shielding phenomenon in the multigroup transport code APOLLO2. They concern the space and energy treatment of the slowing-down equation, the setting up of quadrature formulas to calculate reaction rates, the setting-up of a method that treats directly a resonant mixture and the development of a sub-group method. We validate these improvements either in an elementary or in a global way. Now, we obtain, more accurate multigroup reaction rates and we are able to carry out a reference self-shielding calculation on a very fine multigroup mesh. To end, we draw a conclusion and give some prospects on the remaining work. (author)

NDS multigroup cross section libraries

International Nuclear Information System (INIS)

DayDay, N.

1981-12-01

A summary description and documentation of the multigroup cross section libraries which exist at the IAEA Nuclear Data Section are given in this report. The libraries listed are available either on tape or in printed form. (author)

Performance of a fine-grained parallel model for multi-group nodal-transport calculations in three-dimensional pin-by-pin reactor geometry

International Nuclear Information System (INIS)

Masahiro, Tatsumi; Akio, Yamamoto

2003-01-01

A production code SCOPE2 was developed based on the fine-grained parallel algorithm by the red/black iterative method targeting parallel computing environments such as a PC-cluster. It can perform a depletion calculation in a few hours using a PC-cluster with the model based on a 9-group nodal-SP3 transport method in 3-dimensional pin-by-pin geometry for in-core fuel management of commercial PWRs. The present algorithm guarantees the identical convergence process as that in serial execution, which is very important from the viewpoint of quality management. The fine-mesh geometry is constructed by hierarchical decomposition with introduction of intermediate management layer as a block that is a quarter piece of a fuel assembly in radial direction. A combination of a mesh division scheme forcing even meshes on each edge and a latency-hidden communication algorithm provided simplicity and efficiency to message passing to enhance parallel performance. Inter-processor communication and parallel I/O access were realized using the MPI functions. Parallel performance was measured for depletion calculations by the 9-group nodal-SP3 transport method in 3-dimensional pin-by-pin geometry with 340 x 340 x 26 meshes for full core geometry and 170 x 170 x 26 for quarter core geometry. A PC cluster that consists of 24 Pentium-4 processors connected by the Fast Ethernet was used for the performance measurement. Calculations in full core geometry gave better speedups compared to those in quarter core geometry because of larger granularity. Fine-mesh sweep and feedback calculation parts gave almost perfect scalability since granularity is large enough, while 1-group coarse-mesh diffusion acceleration gave only around 80%. The speedup and parallel efficiency for total computation time were 22.6 and 94%, respectively, for the calculation in full core geometry with 24 processors. (authors)

Status report on multigroup cross section generation code development for high-fidelity deterministic neutronics simulation system

International Nuclear Information System (INIS)

Yang, W.S.; Lee, C.H.

2008-01-01

Under the fast reactor simulation program launched in April 2007, development of an advanced multigroup cross section generation code was initiated in July 2007, in conjunction with the development of the high-fidelity deterministic neutron transport code UNIC. The general objectives are to simplify the existing multi-step schemes and to improve the resolved and unresolved resonance treatments. Based on the review results of current methods and the fact that they have been applied successfully to fast critical experiment analyses and fast reactor designs for last three decades, the methodologies of the ETOE-2/MC 2 -2/SDX code system were selected as the starting set of methodologies for multigroup cross section generation for fast reactor analysis. As the first step for coupling with the UNIC code and use in a parallel computing environment, the MC 2 -2 code was updated by modernizing the memory structure and replacing old data management package subroutines and functions with FORTRAN 90 based routines. Various modifications were also made in the ETOE-2 and MC 2 -2 codes to process the ENDF/B-VII.0 data properly. Using the updated ETOE-2/MC 2 -2 code system, the ENDF/B-VII.0 data was successfully processed for major heavy and intermediate nuclides employed in sodium-cooled fast reactors. Initial verification tests of the MC 2 -2 libraries generated from ENDF/B-VII.0 data were performed by inter-comparison of twenty-one group infinite dilute total cross sections obtained from MC 2 -2, VIM, and NJOY. For almost all nuclides considered, MC 2 -2 cross sections agreed very well with those from VIM and NJOY. Preliminary validation tests of the ENDF/B-VII.0 libraries of MC 2 -2 were also performed using a set of sixteen fast critical benchmark problems. The deterministic results based on MC 2 -2/TWODANT calculations were in good agreement with MCNP solutions within ∼0.25% Δρ, except a few small LANL fast assemblies. Relative to the MCNP solution, the MC 2 -2/TWODANT

Status report on multigroup cross section generation code development for high-fidelity deterministic neutronics simulation system.

Energy Technology Data Exchange (ETDEWEB)

Yang, W. S.; Lee, C. H. (Nuclear Engineering Division)

2008-05-16

Under the fast reactor simulation program launched in April 2007, development of an advanced multigroup cross section generation code was initiated in July 2007, in conjunction with the development of the high-fidelity deterministic neutron transport code UNIC. The general objectives are to simplify the existing multi-step schemes and to improve the resolved and unresolved resonance treatments. Based on the review results of current methods and the fact that they have been applied successfully to fast critical experiment analyses and fast reactor designs for last three decades, the methodologies of the ETOE-2/MC{sup 2}-2/SDX code system were selected as the starting set of methodologies for multigroup cross section generation for fast reactor analysis. As the first step for coupling with the UNIC code and use in a parallel computing environment, the MC{sup 2}-2 code was updated by modernizing the memory structure and replacing old data management package subroutines and functions with FORTRAN 90 based routines. Various modifications were also made in the ETOE-2 and MC{sup 2}-2 codes to process the ENDF/B-VII.0 data properly. Using the updated ETOE-2/MC{sup 2}-2 code system, the ENDF/B-VII.0 data was successfully processed for major heavy and intermediate nuclides employed in sodium-cooled fast reactors. Initial verification tests of the MC{sup 2}-2 libraries generated from ENDF/B-VII.0 data were performed by inter-comparison of twenty-one group infinite dilute total cross sections obtained from MC{sup 2}-2, VIM, and NJOY. For almost all nuclides considered, MC{sup 2}-2 cross sections agreed very well with those from VIM and NJOY. Preliminary validation tests of the ENDF/B-VII.0 libraries of MC{sup 2}-2 were also performed using a set of sixteen fast critical benchmark problems. The deterministic results based on MC{sup 2}-2/TWODANT calculations were in good agreement with MCNP solutions within {approx}0.25% {Delta}{rho}, except a few small LANL fast assemblies

Multiarea Transmission Cost Allocation in Large Power Systems Using the Nodal Pricing Control Approach

Directory of Open Access Journals (Sweden)

M. Ghayeni

2010-12-01

Full Text Available This paper proposes an algorithm for transmission cost allocation (TCA in a large power system based on nodal pricing approach using the multi-area scheme. The nodal pricing approach is introduced to allocate the transmission costs by the control of nodal prices in a single area network. As the number of equations is dependent on the number of buses and generators, this method will be very time consuming for large power systems. To solve this problem, the present paper proposes a new algorithm based on multi-area approach for regulating the nodal prices, so that the simulation time is greatly reduced and therefore the TCA problem with nodal pricing approach will be applicable for large power systems. In addition, in this method the transmission costs are allocated to users more equitable. Since the higher transmission costs in an area having a higher reliability are paid only by users of that area in contrast with the single area method, in which these costs are allocated to all users regardless of their locations. The proposed method is implemented on the IEEE 118 bus test system which comprises three areas. Results show that with application of multi-area approach, the simulation time is greatly reduced and the transmission costs are also allocated to users with less variation in new nodal prices with respect to the single area approach.

Self-shielding phenomenon modelling in multigroup transport code Apollo-2; Modelisation du phenomene d'autoprotection dans le code de transport multigroupe Apollo 2

Energy Technology Data Exchange (ETDEWEB)

Coste-Delclaux, M

2006-03-15

This document describes the improvements carried out for modelling the self-shielding phenomenon in the multigroup transport code APOLLO2. They concern the space and energy treatment of the slowing-down equation, the setting up of quadrature formulas to calculate reaction rates, the setting-up of a method that treats directly a resonant mixture and the development of a sub-group method. We validate these improvements either in an elementary or in a global way. Now, we obtain, more accurate multigroup reaction rates and we are able to carry out a reference self-shielding calculation on a very fine multigroup mesh. To end, we draw a conclusion and give some prospects on the remaining work. (author)

Generalization of Spectral Green's Function nodal method for slab-geometry fixed-source adjoint transport problems in SN formulation

International Nuclear Information System (INIS)

Curbelo, Jesus P.; Silva, Odair P. da; Barros, Ricardo C.

2017-01-01

Presented here is the application of the adjoint technique for solving source{detector discrete ordinates (S N ) transport problems by using a spectral nodal method. For slab-geometry adjoint S-N model, the adjoint spectral Green's function method (SGF † ) is extended to multigroup problems considering arbitrary L'th-order of scattering anisotropy, and the possibility of non{zero prescribed boundary conditions for the forward S N transport problems. The SGF † method converges numerical solutions that are completely free from spatial truncation errors. In order to generate numerical solutions of the SGF † equations, we use the partial adjoint one{node block inversion (NBI) iterative scheme. Partial adjoint NBI scheme uses the most recent estimates for the node-edge adjoint angular Fluxes in the outgoing directions of a given discretization node, to solve the resulting adjoint SN problem in that node for all the adjoint angular fluxes in the incoming directions, which constitute the outgoing adjoint angular fluxes for the adjacent node in the sweeping directions. Numerical results are given to illustrate the present spectral nodal method features and some advantages of using the adjoint technique in source-detector problems. author)

Nodal price volatility reduction and reliability enhancement of restructured power systems considering demand-price elasticity

International Nuclear Information System (INIS)

Goel, L.; Wu, Qiuwei; Wang, Peng

2008-01-01

With the development of restructured power systems, the conventional 'same for all customers' electricity price is getting replaced by nodal prices. Electricity prices will fluctuate with time and nodes. In restructured power systems, electricity demands will interact mutually with prices. Customers may shift some of their electricity consumption from time slots of high electricity prices to those of low electricity prices if there is a commensurate price incentive. The demand side load shift will influence nodal prices in return. This interaction between demand and price can be depicted using demand-price elasticity. This paper proposes an evaluation technique incorporating the impact of the demand-price elasticity on nodal prices, system reliability and nodal reliabilities of restructured power systems. In this technique, demand and price correlations are represented using the demand-price elasticity matrix which consists of self/cross-elasticity coefficients. Nodal prices are determined using optimal power flow (OPF). The OPF and customer damage functions (CDFs) are combined in the proposed reliability evaluation technique to assess the reliability enhancement of restructured power systems considering demand-price elasticity. The IEEE reliability test system (RTS) is simulated to illustrate the developed techniques. The simulation results show that demand-price elasticity reduces the nodal price volatility and improves both the system reliability and nodal reliabilities of restructured power systems. Demand-price elasticity can therefore be utilized as a possible efficient tool to reduce price volatility and to enhance the reliability of restructured power systems. (author)

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

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

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

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

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.

Barrier tunneling of the loop-nodal semimetal in the hyperhoneycomb lattice

Science.gov (United States)

Guan, Ji-Huan; Zhang, Yan-Yang; Lu, Wei-Er; Xia, Yang; Li, Shu-Shen

2018-05-01

We theoretically investigate the barrier tunneling in the 3D model of the hyperhoneycomb lattice, which is a nodal-line semimetal with a Dirac loop at zero energy. In the presence of a rectangular potential, the scattering amplitudes for different injecting states around the nodal loop are calculated, by using analytical treatments of the effective model, as well as numerical simulations of the tight binding model. In the low energy regime, states with remarkable transmissions are only concentrated in a small range around the loop plane. When the momentum of the injecting electron is coplanar with the nodal loop, nearly perfect transmissions can occur for a large range of injecting azimuthal angles if the potential is not high. For higher potential energies, the transmission shows a resonant oscillation with the potential, but still with peaks being perfect transmissions that do not decay with the potential width. These strikingly robust transports of the loop-nodal semimetal can be approximately explained by a momentum dependent Dirac Hamiltonian.

MUXS: a code to generate multigroup cross sections for sputtering calculations

International Nuclear Information System (INIS)

Hoffman, T.J.; Robinson, M.T.; Dodds, H.L. Jr.

1982-10-01

This report documents MUXS, a computer code to generate multigroup cross sections for charged particle transport problems. Cross sections generated by MUXS can be used in many multigroup transport codes, with minor modifications to these codes, to calculate sputtering yields, reflection coefficients, penetration distances, etc

The value of nodal information in predicting lung cancer relapse using 4DPET/4DCT

Energy Technology Data Exchange (ETDEWEB)

Li, Heyse, E-mail: heyse.li@mail.utoronto.ca [Department of Mechanical and Industrial Engineering, University of Toronto, 5 King’s College Road, Toronto, Ontario M5S 3G8 (Canada); Becker, Nathan; Raman, Srinivas [Radiation Oncology, UHN Princess Margaret Cancer Centre, 610 University of Avenue, Toronto, Ontario M5T 2M9 (Canada); Chan, Timothy C. Y. [Department of Mechanical and Industrial Engineering, University of Toronto, 5 King’s College Road, Toronto, Ontario M5S 3G8, Canada and Techna Institute for the Advancement of Technology for Health, 124 - 100 College Street, Toronto, Ontario M5G 1P5 (Canada); Bissonnette, Jean-Pierre [Radiation Oncology, UHN Princess Margaret Cancer Centre, 610 University of Avenue, Toronto, Ontario M5T 2M9, Canada and Techna Institute for the Advancement of Technology for Health, 124 - 100 College Street, Toronto, Ontario M5G 1P5 (Canada)

2015-08-15

Purpose: There is evidence that computed tomography (CT) and positron emission tomography (PET) imaging metrics are prognostic and predictive in nonsmall cell lung cancer (NSCLC) treatment outcomes. However, few studies have explored the use of standardized uptake value (SUV)-based image features of nodal regions as predictive features. The authors investigated and compared the use of tumor and node image features extracted from the radiotherapy target volumes to predict relapse in a cohort of NSCLC patients undergoing chemoradiation treatment. Methods: A prospective cohort of 25 patients with locally advanced NSCLC underwent 4DPET/4DCT imaging for radiation planning. Thirty-seven image features were derived from the CT-defined volumes and SUVs of the PET image from both the tumor and nodal target regions. The machine learning methods of logistic regression and repeated stratified five-fold cross-validation (CV) were used to predict local and overall relapses in 2 yr. The authors used well-known feature selection methods (Spearman’s rank correlation, recursive feature elimination) within each fold of CV. Classifiers were ranked on their Matthew’s correlation coefficient (MCC) after CV. Area under the curve, sensitivity, and specificity values are also presented. Results: For predicting local relapse, the best classifier found had a mean MCC of 0.07 and was composed of eight tumor features. For predicting overall relapse, the best classifier found had a mean MCC of 0.29 and was composed of a single feature: the volume greater than 0.5 times the maximum SUV (N). Conclusions: The best classifier for predicting local relapse had only tumor features. In contrast, the best classifier for predicting overall relapse included a node feature. Overall, the methods showed that nodes add value in predicting overall relapse but not local relapse.

Kalpakkam multigroup cross section set for fast reactor applications - status and performance

International Nuclear Information System (INIS)

Ramanadhan, M.M.; Gopalakrishnan, M.M.

1986-01-01

This report documents the status of the presently created set of multigroup constants at Kalpakkam. The list of nuclides processed and the details of multigroup structure are given. Also included are the particulars of dilutions and temperatures for each nuclide in the multigroup cross section set for which self shielding factors have been calculated. Using this new multigroup cross section set, measured integral quantities such as K-eff, central reaction rate ratios, central reactivity worths etc. were calculated for a few fast critical benchmark assemblies and the calculated values of neutronic parameters obtained were compared with those obtained using the available Cadarache cross section library and those published in literature for ENDF/B-IV based set and Japanese evaluated nuclear data library (JENDL). The details of analyses are documented along with the conclusions. (author). 17 refs., 12 tabs

The Suppression of Energy Discretization Errors in Multigroup Transport Calculations

International Nuclear Information System (INIS)

Larsen, Edward

2013-01-01

The Objective of this project is to develop, implement, and test new deterministric methods to solve, as efficiently as possible, multigroup neutron transport problems having an extremely large number of groups. Our approach was to (i) use the standard CMFD method to 'coarsen' the space-angle grid, yielding a multigroup diffusion equation, and (ii) use a new multigrid-in-space-and-energy technique to efficiently solve the multigroup diffusion problem. The overall strategy of (i) how to coarsen the spatial an energy grids, and (ii) how to navigate through the various grids, has the goal of minimizing the overall computational effort. This approach yields not only the fine-grid solution, but also coarse-group flux-weighted cross sections that can be used for other related problems.

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.

HAMMER, 1-D Multigroup Neutron Transport Infinite System Cell Calculation for Few-Group Diffusion Calculation

International Nuclear Information System (INIS)

Honeck, H.C.

1984-01-01

1 - Description of problem or function: HAMMER performs infinite lattice, one-dimensional cell multigroup calculations, followed (optionally) by one-dimensional, few-group, multi-region reactor calculations with neutron balance edits. 2 - Method of solution: Infinite lattice parameters are calculated by means of multigroup transport theory, composite reactor parameters by few-group diffusion theory. 3 - Restrictions on the complexity of the problem: - Cell calculations - maxima of: 30 thermal groups; 54 epithermal groups; 20 space points; 20 regions; 18 isotopes; 10 mixtures; 3 thermal up-scattering mixtures; 200 resonances per group; no overlap or interference; single level only. - Reactor calculations - maxima of : 40 regions; 40 mixtures; 250 space points; 4 groups

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)

A numerical method for multigroup slab-geometry discrete ordinates problems with no spatial truncation error

International Nuclear Information System (INIS)

Barros, R.C. de; Larsen, E.W.

1991-01-01

A generalization of the one-group Spectral Green's Function (SGF) method is developed for multigroup, slab-geometry discrete ordinates (S N ) problems. The multigroup SGF method is free from spatial truncation errors; it generated numerical values for the cell-edge and cell-average angular fluxes that agree with the analytic solution of the multigroup S N equations. Numerical results are given to illustrate the method's accuracy

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)

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)

CASTRO: A NEW COMPRESSIBLE ASTROPHYSICAL SOLVER. III. MULTIGROUP RADIATION HYDRODYNAMICS

International Nuclear Information System (INIS)

Zhang, W.; Almgren, A.; Bell, J.; Howell, L.; Burrows, A.; Dolence, J.

2013-01-01

We present a formulation for multigroup radiation hydrodynamics that is correct to order O(v/c) using the comoving-frame approach and the flux-limited diffusion approximation. We describe a numerical algorithm for solving the system, implemented in the compressible astrophysics code, CASTRO. CASTRO uses a Eulerian grid with block-structured adaptive mesh refinement based on a nested hierarchy of logically rectangular variable-sized grids with simultaneous refinement in both space and time. In our multigroup radiation solver, the system is split into three parts: one part that couples the radiation and fluid in a hyperbolic subsystem, another part that advects the radiation in frequency space, and a parabolic part that evolves radiation diffusion and source-sink terms. The hyperbolic subsystem and the frequency space advection are solved explicitly with high-order Godunov schemes, whereas the parabolic part is solved implicitly with a first-order backward Euler method. Our multigroup radiation solver works for both neutrino and photon radiation.

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

SERKON program for compiling a multigroup library to be used in BETTY calculation

International Nuclear Information System (INIS)

Nguyen Phuoc Lan.

1982-11-01

A SERKON-type program was written to compile data sets generated by FEDGROUP-3 into a multigroup library for BETTY calculation. A multigroup library was generated from the ENDF/B-IV data file and tested against the TRX-1 and TRX-2 lattices with good results. (author)

MCFT: a program for calculating fast and thermal neutron multigroup constants

International Nuclear Information System (INIS)

Yang Shunhai; Sang Xinzeng

1993-01-01

MCFT is a program for calculating the fast and thermal neutron multigroup constants, which is redesigned from some codes for generation of thermal neutron multigroup constants and for fast neutron multigroup constants adapted on CYBER 825 computer. It uses indifferently as basic input with the evaluated nuclear data contained in the ENDF/B (US), KEDAK (Germany) and UK (United Kingdom) libraries. The code includes a section devoted to the generation of resonant Doppler broadened cross section in the framework of single-or multi-level Breit-Wigner formalism. The program can compute the thermal neutron scattering law S (α, β, T) as the input data in tabular, free gas or diffusion motion form. It can treat up to 200 energy groups and Legendre moments up to P 5 . The output consists of various reaction multigroup constants in all neutron energy range desired in the nuclear reactor design and calculation. Three options in input file can be used by the user. The output format is arbitrary and defined by user with a minimum of program modification. The program includes about 15,000 cards and 184 subroutines. FORTRAN 5 computer language is used. The operation system is under NOS 2 on computer CYBER 825

Generalization of Spectral Green's Function nodal method for slab-geometry fixed-source adjoint transport problems in S{sub N} formulation

Energy Technology Data Exchange (ETDEWEB)

Curbelo, Jesus P.; Silva, Odair P. da; Barros, Ricardo C. [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Instituto Politecnico. Programa de Pos-graduacao em Modelagem Computacional; Garcia, Carlos R., E-mail: cgh@instec.cu [Departamento de Ingenieria Nuclear, Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba)

2017-07-01

Presented here is the application of the adjoint technique for solving source-detector discrete ordinates (S{sub N}) transport problems by using a spectral nodal method. For slab-geometry adjoint S-N model, the adjoint spectral Green's function method (SGF{sup †}) is extended to multigroup problems considering arbitrary L'th-order of scattering anisotropy, and the possibility of non-zero prescribed boundary conditions for the forward S{sub N} transport problems. The SGF{sup †} method converges numerical solutions that are completely free from spatial truncation errors. In order to generate numerical solutions of the SGF{sup †} equations, we use the partial adjoint one-node block inversion (NBI) iterative scheme. Partial adjoint NBI scheme uses the most recent estimates for the node-edge adjoint angular Fluxes in the outgoing directions of a given discretization node, to solve the resulting adjoint SN problem in that node for all the adjoint angular fluxes in the incoming directions, which constitute the outgoing adjoint angular fluxes for the adjacent node in the sweeping directions. Numerical results are given to illustrate the present spectral nodal method features and some advantages of using the adjoint technique in source-detector problems. author)

The Multigroup Neutron Diffusion Equations/1 Space Dimension

Energy Technology Data Exchange (ETDEWEB)

Linde, Sven

1960-06-15

A description is given of a program for the Ferranti Mercury computer which solves the one-dimensional multigroup diffusion equations in plane, cylindrical or spherical geometry, and also approximates automatically a two-dimensional solution by separating the space variables. In section A the method of calculation is outlined and the preparation of data for two group problems is described. The spatial separation of two-dimensional equations is considered in section B. Section C covers the multigroup equations. These parts are self contained and include all information required for the use of the program. Details of the numerical methods are given in section D. Three sample problems are solved in section E. Punching and operating instructions are given in an appendix.

The Multigroup Neutron Diffusion Equations/1 Space Dimension

International Nuclear Information System (INIS)

Linde, Sven

1960-06-01

A description is given of a program for the Ferranti Mercury computer which solves the one-dimensional multigroup diffusion equations in plane, cylindrical or spherical geometry, and also approximates automatically a two-dimensional solution by separating the space variables. In section A the method of calculation is outlined and the preparation of data for two group problems is described. The spatial separation of two-dimensional equations is considered in section B. Section C covers the multigroup equations. These parts are self contained and include all information required for the use of the program. Details of the numerical methods are given in section D. Three sample problems are solved in section E. Punching and operating instructions are given in an appendix

Research of the application of multi-group libraries based on ENDF/B-VII library in the reactor design

International Nuclear Information System (INIS)

Mi Aijun; Li Junjie

2010-01-01

In this paper the multi-group libraries were constructed by processing ENDF/B-VII neutron incident files into multi-group structure, and the application of the multi-group libraries in the pressurized-water reactor(PWR) design was studied. The construction of the multi-group library is realized by using the NJOY nuclear data processing system. The code can process the neutron cross section files form ENDF format to MATXS format which was required in SN code. Two dimension transport theory code of discrete ordinates DORT was used to verify the multi-group libraries and the method of the construction by comparing calculations for some representative benchmarks. We made the PWR shielding calculation by using the multi-group libraries and studied the influence of the parameters involved during the construction of the libraries such as group structure, temperatures and weight functions on the shielding design of the PWR. This work is the preparation for the construction of the multi-group library which will be used in PWR shielding design in engineering. (authors)

Proposal to extend CSEWG neutron and photon multigroup structures for wider applications

International Nuclear Information System (INIS)

LaBauve, R.J.; Wilson, W.B.

1976-02-01

The 239-group neutron multigroup structure recommended by the Codes and Formats Subcommittee of the cross section evaluation working group (CSEWG) for use in LMFBR design is not well suited for application in certain other areas, particularly thermal reactor design. This report describes a proposal for a neutron group structure consisting of 347 groups, which is an extension of the CSEWG group structure into the thermal range, and also includes more detail in other energy ranges important in LWR, HTGR, GCFR, and CTR design. Similarly, a proposed extension of the CSEWG 94-group photon multigroup structure to 103 groups is described. A subset of the neutron multigroup structure, consisting of 154 groups and for use in power reactor studies, is also presented

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.

Application of direct discrete method (DDM) to multigroup neutron transport problems

International Nuclear Information System (INIS)

Vosoughi, Naser; Salehi, Ali Akbar; Shahriari, Majid

2003-01-01

The Direct Discrete Method (DDM), which produced excellent results for one-group neutron transport problems, has been developed for multigroup energy. A multigroup neutron transport discrete equation has been produced for a cylindrical shape fuel element with and without associated coolant regions with two boundary conditions. The calculations are illustrated for two-group energy by graphs showing the fast and thermal fluxes. The validity of the results are tested against the results obtained by the ANISN code. (author)

Time-dependent patterning of the mesoderm and endoderm by Nodal signals in zebrafish

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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

Radiotherapy for esthesioneuroblastoma: is elective nodal irradiation warranted in the multimodality treatment approach?

Science.gov (United States)

Noh, O Kyu; Lee, Sang-wook; Yoon, Sang Min; Kim, Sung Bae; Kim, Sang Yoon; Kim, Chang Jin; Jo, Kyung Ja; Choi, Eun Kyung; Song, Si Yeol; Kim, Jong Hoon; Ahn, Seung Do

2011-02-01

The role of elective nodal irradiation (ENI) in radiotherapy for esthesioneuroblastoma (ENB) has not been clearly defined. We analyzed treatment outcomes of patients with ENB and the frequency of cervical nodal failure in the absence of ENI. Between August 1996 and December 2007, we consulted with 19 patients with ENB regarding radiotherapy. Initial treatment consisted of surgery alone in 2 patients; surgery and postoperative radiotherapy in 4; surgery and adjuvant chemotherapy in 1; surgery, postoperative radiotherapy, and chemotherapy in 3; and chemotherapy followed by radiotherapy or concurrent chemoradiotherapy in 5. Five patients did not receive planned radiotherapy because of disease progression. Including 2 patients who received salvage radiotherapy, 14 patients were treated with radiotherapy. Elective nodal irradiation was performed in 4 patients with high-risk factors, including 3 with cervical lymph node metastasis at presentation. Fourteen patients were analyzable, with a median follow-up of 27 months (range, 7-64 months). The overall 3-year survival rate was 73.4%. Local failure occurred in 3 patients (21.4%), regional cervical failure in 3 (21.4%), and distant failure in 2 (14.3%). No cervical nodal failure occurred in patients treated with combined systemic chemotherapy regardless of ENI. Three cervical failures occurred in the 4 patients treated with ENI or neck dissection (75%), none of whom received systemic chemotherapy. ENI during radiotherapy for ENB seems to play a limited role in preventing cervical nodal failure. Omitting ENI may be an option if patients are treated with a combination of radiotherapy and chemotherapy. Copyright © 2011 Elsevier Inc. All rights reserved.

Radiotherapy for Esthesioneuroblastoma: Is Elective Nodal Irradiation Warranted in the Multimodality Treatment Approach?

International Nuclear Information System (INIS)

Noh, O Kyu; Lee, Sang-wook; Yoon, Sang Min; Kim, Sung Bae; Kim, Sang Yoon; Kim, Chang Jin; Jo, Kyung Ja; Choi, Eun Kyung; Song, Si Yeol; Kim, Jong Hoon; Ahn, Seung Do

2011-01-01

Purpose: The role of elective nodal irradiation (ENI) in radiotherapy for esthesioneuroblastoma (ENB) has not been clearly defined. We analyzed treatment outcomes of patients with ENB and the frequency of cervical nodal failure in the absence of ENI. Methods and Materials: Between August 1996 and December 2007, we consulted with 19 patients with ENB regarding radiotherapy. Initial treatment consisted of surgery alone in 2 patients; surgery and postoperative radiotherapy in 4; surgery and adjuvant chemotherapy in 1; surgery, postoperative radiotherapy, and chemotherapy in 3; and chemotherapy followed by radiotherapy or concurrent chemoradiotherapy in 5. Five patients did not receive planned radiotherapy because of disease progression. Including 2 patients who received salvage radiotherapy, 14 patients were treated with radiotherapy. Elective nodal irradiation was performed in 4 patients with high-risk factors, including 3 with cervical lymph node metastasis at presentation. Results: Fourteen patients were analyzable, with a median follow-up of 27 months (range, 7-64 months). The overall 3-year survival rate was 73.4%. Local failure occurred in 3 patients (21.4%), regional cervical failure in 3 (21.4%), and distant failure in 2 (14.3%). No cervical nodal failure occurred in patients treated with combined systemic chemotherapy regardless of ENI. Three cervical failures occurred in the 4 patients treated with ENI or neck dissection (75%), none of whom received systemic chemotherapy. Conclusions: ENI during radiotherapy for ENB seems to play a limited role in preventing cervical nodal failure. Omitting ENI may be an option if patients are treated with a combination of radiotherapy and chemotherapy.

A new coupling kernel for the three-dimensional simulation of a boiling water reactor core by the nodal coupling method

International Nuclear Information System (INIS)

Gupta, N.K.

1981-01-01

A new coupling kernel is developed for the three-dimensional (3-D) simulation of Boiling Water Reactors (BWR's) by the nodal coupling method. The new kernel depends not only on the properties of the node under consideration but also on the properties of its neighbouring nodes. This makes the kernel more useful in particular for fuel bundles lying in a surrounding of different nuclear characteristics, e.g. for a controlled bundle in the surrounding of uncontrolled bundles or vice-versa. The main parameter in the new kernel is a space-dependent factor obtained from the ratio of thermal-to-fast flux. The average value of the above ratio for each node is evaluated analytically. The kernel is incorporated in a 3-D BWR core simulation program MOGS. As an experimental verification of the model, the cycle-6 operations of the two units of the Tarapur Atomic Power Station (TAPS) are simulated and the result of the simulation are compared with Travelling Incore Probe (TIP) data. (orig.)

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.

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.

BWR modeling capability and Scale/Triton lattice-to-core integration of the Nestle nodal simulator - 331

International Nuclear Information System (INIS)

Galloway, J.; Hernandez, H.; Maldonado, G.I.; Jessee, M.; Popov, E.; Clarno, K.

2010-01-01

This article reports the status of recent and substantial enhancements made to the NESTLE nodal core simulator, a code originally developed at North Carolina State University (NCSU) of which version 5.2.1 has been available for several years through the Oak Ridge National Laboratory (ORNL) Radiation Safety Information Computational Center (RSICC) software repository. In its released and available form, NESTLE is a seasoned, well-developed and extensively tested code system particularly useful to model PWRs. In collaboration with NCSU, University of Tennessee (UT) and ORNL researchers have recently developed new enhancements for the NESTLE code, including the implementation of a two-phase drift-flux thermal hydraulic and flow redistribution model to facilitate modeling of Boiling Water Reactors (BWRs) as well as the development of an integrated coupling of SCALE/TRITON lattice physics to NESTLE so to produce an end-to-end capability for reactor simulations. These latest advancements implemented into NESTLE as well as an update of other ongoing efforts of this project are herein reported. (authors)

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

Final report [on solving the multigroup diffusion equations

International Nuclear Information System (INIS)

Birkhoff, G.

1975-01-01

Progress achieved in the development of variational methods for solving the multigroup neutron diffusion equations is described. An appraisal is made of the extent to which improved variational methods could advantageously replace difference methods currently used

Multi-Group Library Generation with Explicit Resonance Interference Using Continuous Energy Monte Carlo Calculation

Energy Technology Data Exchange (ETDEWEB)

Park, Ho Jin; Cho, Jin Young [KAERI, Daejeon (Korea, Republic of); Kim, Kang Seog [Oak Ridge National Laboratory, Oak Ridge (United States); Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

2016-05-15

In this study, multi-group cross section libraries for the DeCART code were generated using a new procedure. The new procedure includes generating the RI tables based on the MC calculations, correcting the effective fission product yield calculations, and considering most of the fission products as resonant nuclides. KAERI (Korea Atomic Energy Research Institute) has developed the transport lattice code KARMA (Kernel Analyzer by Ray-tracing Method for fuel Assembly) and DeCART (Deterministic Core Analysis based on Ray Tracing) for a multi-group neutron transport analysis of light water reactors (LWRs). These codes adopt the method of characteristics (MOC) to solve the multi-group transport equation and resonance fixed source problem, the subgroup and the direct iteration method with resonance integral tables for resonance treatment. With the development of the DeCART and KARMA code, KAERI has established its own library generation system for a multi-group transport calculation. In the KAERI library generation system, the multi-group average cross section and resonance integral (RI) table are generated and edited using PENDF (point-wise ENDF) and GENDF (group-wise ENDF) produced by the NJOY code. The new method does not need additional processing because the MC method can handle any geometry information and material composition. In this study, the new method is applied to the dominant resonance nuclide such as U{sup 235} and U{sup 238} and the conventional method is applied to the minor resonance nuclides. To examine the newly generated multi-group cross section libraries, various benchmark calculations such as pin-cell, FA, and core depletion problem are performed and the results are compared with the reference solutions. Overall, the results by the new method agree well with the reference solution. The new procedure based on the MC method were verified and provided the multi-group library that can be used in the SMR nuclear design analysis.

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

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)

Proposal to extend CSEWG neutron and photon multigroup structures for wider applications. [Tables

Energy Technology Data Exchange (ETDEWEB)

LaBauve, R.J.; Wilson, W.B.

1976-02-01

The 239-group neutron multigroup structure recommended by the Codes and Formats Subcommittee of the cross section evaluation working group (CSEWG) for use in LMFBR design is not well suited for application in certain other areas, particularly thermal reactor design. This report describes a proposal for a neutron group structure consisting of 347 groups, which is an extension of the CSEWG group structure into the thermal range, and also includes more detail in other energy ranges important in LWR, HTGR, GCFR, and CTR design. Similarly, a proposed extension of the CSEWG 94-group photon multigroup structure to 103 groups is described. A subset of the neutron multigroup structure, consisting of 154 groups and for use in power reactor studies, is also presented.

Atlas-Based Segmentation Improves Consistency and Decreases Time Required for Contouring Postoperative Endometrial Cancer Nodal Volumes

International Nuclear Information System (INIS)

Young, Amy V.; Wortham, Angela; Wernick, Iddo; Evans, Andrew; Ennis, Ronald D.

2011-01-01

Purpose: Accurate target delineation of the nodal volumes is essential for three-dimensional conformal and intensity-modulated radiotherapy planning for endometrial cancer adjuvant therapy. We hypothesized that atlas-based segmentation ('autocontouring') would lead to time savings and more consistent contours among physicians. Methods and Materials: A reference anatomy atlas was constructed using the data from 15 postoperative endometrial cancer patients by contouring the pelvic nodal clinical target volume on the simulation computed tomography scan according to the Radiation Therapy Oncology Group 0418 trial using commercially available software. On the simulation computed tomography scans from 10 additional endometrial cancer patients, the nodal clinical target volume autocontours were generated. Three radiation oncologists corrected the autocontours and delineated the manual nodal contours under timed conditions while unaware of the other contours. The time difference was determined, and the overlap of the contours was calculated using Dice's coefficient. Results: For all physicians, manual contouring of the pelvic nodal target volumes and editing the autocontours required a mean ± standard deviation of 32 ± 9 vs. 23 ± 7 minutes, respectively (p = .000001), a 26% time savings. For each physician, the time required to delineate the manual contours vs. correcting the autocontours was 30 ± 3 vs. 21 ± 5 min (p = .003), 39 ± 12 vs. 30 ± 5 min (p = .055), and 29 ± 5 vs. 20 ± 5 min (p = .0002). The mean overlap increased from manual contouring (0.77) to correcting the autocontours (0.79; p = .038). Conclusion: The results of our study have shown that autocontouring leads to increased consistency and time savings when contouring the nodal target volumes for adjuvant treatment of endometrial cancer, although the autocontours still required careful editing to ensure that the lymph nodes at risk of recurrence are properly included in the target volume.

Development of a 3D-Multigroup program to simulate anomalous diffusion phenomena in the nuclear reactors

International Nuclear Information System (INIS)

Maleki Moghaddam, Nader; Afarideh, Hossein; Espinosa-Paredes, Gilberto

2015-01-01

Highlights: • The new version of neutron diffusion equation for simulating anomalous diffusion is presented. • Application of fractional calculus in the nuclear reactor is revealed. • A 3D-Multigroup program is developed based on the fractional operators. • The super-diffusion and sub-diffusion phenomena are modeled in the nuclear reactors core. - Abstract: The diffusion process is categorized in three parts, normal diffusion, super-diffusion and sub-diffusion. The classical neutron diffusion equation is used to model normal diffusion. A new scheme of derivatives is required to model anomalous diffusion phenomena. The fractional space derivatives are employed to model anomalous diffusion processes where a plume of particles spreads at an inconsistent rate with the classical Brownian motion model. In the fractional diffusion equation, the fractional Laplacians are used; therefore the statistical jump length of neutrons is unrestricted. It is clear that the fractional Laplacians are capable to model the anomalous phenomena in nuclear reactors. We have developed a NFDE-3D (neutron fractional diffusion equation) as a core calculation code to model normal and anomalous diffusion phenomena. The NFDE-3D is validated against the LMW-LWR reactor. The results demonstrate that reactors exhibit complex behavior versus order of the fractional derivatives which depends on the competition between neutron absorption and super-diffusion phenomenon

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.

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

Tumor microvessel density–associated mast cells in canine nodal lymphoma

Science.gov (United States)

Mann, Elizabeth; Whittington, Lisa

2014-01-01

Objective: Mast cells are associated in angiogenesis in various human and animal neoplasms. However, association of mast cells with tumor microvessel density in canine lymphoma was not previously documented. The objective of the study is to determine if mast cells are increased in canine nodal lymphomas and to evaluate their correlation with tumor microvessel density and grading of lymphomas. Methods: Nodal lymphomas from 33 dogs were studied and compared with nonneoplastic lymph nodes from 6 dogs as control. Mast cell count was made on Toluidine blue stained sections. Immunohistochemistry using antibody against Factor VIII was employed to visualize and determine microvessel density. Results: The mast cell count in lymphoma (2.95 ± 2.4) was significantly higher (p < 0.05) than that in the control (0.83 ± 0.3) and was positively correlated with tumor microvessel density (r = 0.44, p = 0.009). Significant difference was not observed in mast cell count and tumor microvessel density among different gradings of lymphomas. Conclusions: Mast cells are associated with tumor microvessel density in canine nodal lymphoma with no significant difference among gradings of lymphomas. Mast cells may play an important role in development of canine nodal lymphomas. Further detailed investigation on the role of mast cells as important part of tumor microenvironment in canine nodal lymphomas is recommended. PMID:26770752

Tumor microvessel density–associated mast cells in canine nodal lymphoma

Directory of Open Access Journals (Sweden)

Moges Woldemeskel

2014-11-01

Full Text Available Objective: Mast cells are associated in angiogenesis in various human and animal neoplasms. However, association of mast cells with tumor microvessel density in canine lymphoma was not previously documented. The objective of the study is to determine if mast cells are increased in canine nodal lymphomas and to evaluate their correlation with tumor microvessel density and grading of lymphomas. Methods: Nodal lymphomas from 33 dogs were studied and compared with nonneoplastic lymph nodes from 6 dogs as control. Mast cell count was made on Toluidine blue stained sections. Immunohistochemistry using antibody against Factor VIII was employed to visualize and determine microvessel density. Results: The mast cell count in lymphoma (2.95 ± 2.4 was significantly higher (p < 0.05 than that in the control (0.83 ± 0.3 and was positively correlated with tumor microvessel density (r = 0.44, p = 0.009. Significant difference was not observed in mast cell count and tumor microvessel density among different gradings of lymphomas. Conclusions: Mast cells are associated with tumor microvessel density in canine nodal lymphoma with no significant difference among gradings of lymphomas. Mast cells may play an important role in development of canine nodal lymphomas. Further detailed investigation on the role of mast cells as important part of tumor microenvironment in canine nodal lymphomas is recommended.

Continuous energy Monte Carlo method based homogenization multi-group constants calculation

International Nuclear Information System (INIS)

Li Mancang; Wang Kan; Yao Dong

2012-01-01

The efficiency of the standard two-step reactor physics calculation relies on the accuracy of multi-group constants from the assembly-level homogenization process. In contrast to the traditional deterministic methods, generating the homogenization cross sections via Monte Carlo method overcomes the difficulties in geometry and treats energy in continuum, thus provides more accuracy parameters. Besides, the same code and data bank can be used for a wide range of applications, resulting in the versatility using Monte Carlo codes for homogenization. As the first stage to realize Monte Carlo based lattice homogenization, the track length scheme is used as the foundation of cross section generation, which is straight forward. The scattering matrix and Legendre components, however, require special techniques. The Scattering Event method was proposed to solve the problem. There are no continuous energy counterparts in the Monte Carlo calculation for neutron diffusion coefficients. P 1 cross sections were used to calculate the diffusion coefficients for diffusion reactor simulator codes. B N theory is applied to take the leakage effect into account when the infinite lattice of identical symmetric motives is assumed. The MCMC code was developed and the code was applied in four assembly configurations to assess the accuracy and the applicability. At core-level, A PWR prototype core is examined. The results show that the Monte Carlo based multi-group constants behave well in average. The method could be applied to complicated configuration nuclear reactor core to gain higher accuracy. (authors)

A survey of TRIPOLI-4

International Nuclear Information System (INIS)

Both, J.P.; Derriennic, H.; Morillon, B.; Nimal, J.C.

1994-01-01

A survey of the new version of the TRIPOLI code used in shielding calculations, is presented. The main new features introduced in this version are: combinatorial geometry, a multigroup automatic weighting scheme, and complete treatment of nuclear evaluation files; a simulation implementation including strategy for parallelization of the code is presented. Some benchmark calculations are also presented. 7 figs., 5 tabs., 4 refs

Nuclear data and multigroup methods in fast reactor calculations

International Nuclear Information System (INIS)

Gur, Y.

1975-03-01

The work deals with fast reactor multigroup calculations, and the efficient treatment of basic nuclear data, which serves as raw material for the calculations. Its purpose is twofold: to build a computer code system that handles a large, detailed library of basic neutron cross section data, (such as ENDF/B-III) and yields a compact set of multigroup cross sections for reactor calculations; to use the code system for comparative analysis of different libraries, in order to discover basic uncertainties that still exist in the measurement of neutron cross sections, and to determine their influence upon uncertainties in nuclear calculations. A program named NANICK which was written in two versions is presented. The first handles the American basic data library, ENDF/B-III, while the second handles the German basic data library, KEDAK. The mathematical algorithm is identical in both versions, and only the file management is different. This program calculates infinitely diluted multigroup cross sections and scattering matrices. It is complemented by the program NASIF that calculates shielding factors from resonance parameters. Different versions of NASIF were written to handle ENDF/B-III or KEDAK. New methods for evaluating in reactor calculations the long term behavior of the neutron flux as well as its fine structure are described and an efficient calculation of the shielding factors from resonance parameters is offered. (B.G.)

Patterns of failure after the reduced volume approach for elective nodal irradiation in nasopharyngeal carcinoma.

Science.gov (United States)

Seol, Ki Ho; Lee, Jeong Eun

2016-03-01

To evaluate the patterns of nodal failure after radiotherapy (RT) with the reduced volume approach for elective neck nodal irradiation (ENI) in nasopharyngeal carcinoma (NPC). Fifty-six NPC patients who underwent definitive chemoradiotherapy with the reduced volume approach for ENI were reviewed. The ENI included retropharyngeal and level II lymph nodes, and only encompassed the echelon inferior to the involved level to eliminate the entire neck irradiation. Patients received either moderate hypofractionated intensity-modulated RT for a total of 72.6 Gy (49.5 Gy to elective nodal areas) or a conventional fractionated three-dimensional conformal RT for a total of 68.4-72 Gy (39.6-45 Gy to elective nodal areas). Patterns of failure, locoregional control, and survival were analyzed. The median follow-up was 38 months (range, 3 to 80 months). The out-of-field nodal failure when omitting ENI was none. Three patients developed neck recurrences (one in-field recurrence in the 72.6 Gy irradiated nodal area and two in the elective irradiated region of 39.6 Gy). Overall disease failure at any site developed in 11 patients (19.6%). Among these, there were six local failures (10.7%), three regional failures (5.4%), and five distant metastases (8.9%). The 3-year locoregional control rate was 87.1%, and the distant failure-free rate was 90.4%; disease-free survival and overall survival at 3 years was 80% and 86.8%, respectively. No patient developed nodal failure in the omitted ENI site. Our investigation has demonstrated that the reduced volume approach for ENI appears to be a safe treatment approach in NPC.

Patterns of failure after the reduced volume approach for elective nodal irradiation in nasopharyngeal carcinoma

Energy Technology Data Exchange (ETDEWEB)

Seol, Ki Ho; Lee, Jeong Eun [Dept. of Radiation Oncology, Kyungpook National University School of Medicine, Daegu (Korea, Republic of)

2016-03-15

To evaluate the patterns of nodal failure after radiotherapy (RT) with the reduced volume approach for elective neck nodal irradiation (ENI) in nasopharyngeal carcinoma (NPC). Fifty-six NPC patients who underwent definitive chemoradiotherapy with the reduced volume approach for ENI were reviewed. The ENI included retropharyngeal and level II lymph nodes, and only encompassed the echelon inferior to the involved level to eliminate the entire neck irradiation. Patients received either moderate hypofractionated intensity-modulated RT for a total of 72.6 Gy (49.5 Gy to elective nodal areas) or a conventional fractionated three-dimensional conformal RT for a total of 68.4-72 Gy (39.6-45 Gy to elective nodal areas). Patterns of failure, locoregional control, and survival were analyzed. The median follow-up was 38 months (range, 3 to 80 months). The out-of-field nodal failure when omitting ENI was none. Three patients developed neck recurrences (one in-field recurrence in the 72.6 Gy irradiated nodal area and two in the elective irradiated region of 39.6 Gy). Overall disease failure at any site developed in 11 patients (19.6%). Among these, there were six local failures (10.7%), three regional failures (5.4%), and five distant metastases (8.9%). The 3-year locoregional control rate was 87.1%, and the distant failure-free rate was 90.4%; disease-free survival and overall survival at 3 years was 80% and 86.8%, respectively. No patient developed nodal failure in the omitted ENI site. Our investigation has demonstrated that the reduced volume approach for ENI appears to be a safe treatment approach in NPC.

Patterns of failure after the reduced volume approach for elective nodal irradiation in nasopharyngeal carcinoma

International Nuclear Information System (INIS)

Seol, Ki Ho; Lee, Jeong Eun

2016-01-01

To evaluate the patterns of nodal failure after radiotherapy (RT) with the reduced volume approach for elective neck nodal irradiation (ENI) in nasopharyngeal carcinoma (NPC). Fifty-six NPC patients who underwent definitive chemoradiotherapy with the reduced volume approach for ENI were reviewed. The ENI included retropharyngeal and level II lymph nodes, and only encompassed the echelon inferior to the involved level to eliminate the entire neck irradiation. Patients received either moderate hypofractionated intensity-modulated RT for a total of 72.6 Gy (49.5 Gy to elective nodal areas) or a conventional fractionated three-dimensional conformal RT for a total of 68.4-72 Gy (39.6-45 Gy to elective nodal areas). Patterns of failure, locoregional control, and survival were analyzed. The median follow-up was 38 months (range, 3 to 80 months). The out-of-field nodal failure when omitting ENI was none. Three patients developed neck recurrences (one in-field recurrence in the 72.6 Gy irradiated nodal area and two in the elective irradiated region of 39.6 Gy). Overall disease failure at any site developed in 11 patients (19.6%). Among these, there were six local failures (10.7%), three regional failures (5.4%), and five distant metastases (8.9%). The 3-year locoregional control rate was 87.1%, and the distant failure-free rate was 90.4%; disease-free survival and overall survival at 3 years was 80% and 86.8%, respectively. No patient developed nodal failure in the omitted ENI site. Our investigation has demonstrated that the reduced volume approach for ENI appears to be a safe treatment approach in NPC

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

Three- and four-noded planar elements using absolute nodal coordinate formulation

International Nuclear Information System (INIS)

Olshevskiy, Alexander; Dmitrochenko, Oleg; Kim, Changwan

2013-01-01

This paper investigates two new types of planar finite elements containing three and four nodes. These elements are the reduced forms of the spatial plate elements employing the absolute nodal coordinate approach. Elements of the first type use translations of nodes and global slopes as nodal coordinates and have 18 and 24 degrees of freedom. The slopes facilitate the prevention of the shear locking effect in bending problems. Furthermore, the slopes accurately describe the deformed shape of the elements. Triangular and quadrilateral elements of the second type use translational degrees of freedom only and, therefore, can be utilized successfully in problems without bending. These simple elements with 6 and 8 degrees of freedom are identical to the elements used in conventional formulation of the finite element method from the kinematical point of view. Similarly to the famous problem called “flying spaghetti” which is used often as a benchmark for beam elements, a kind of “flying lasagna” is simulated for the planar elements. Numerical results of simulations are presented.

An energy recondensation method using the discrete generalized multigroup energy expansion theory

International Nuclear Information System (INIS)

Zhu Lei; Forget, Benoit

2011-01-01

Highlights: → Discrete-generalized multigroup method was implemented as a recondensation scheme. → Coarse group cross-sections were recondensed from core-level solution. → Neighboring effect of reflector and MOX bundle was improved. → Methodology was shown to be fully consistent when a flat angular flux approximation is used. - Abstract: In this paper, the discrete generalized multigroup (DGM) method was used to recondense the coarse group cross-sections using the core level solution, thus providing a correction for neighboring effect found at the core level. This approach was tested using a discrete ordinates implementation in both 1-D and 2-D. Results indicate that 2 or 3 iterations can substantially improve the flux and fission density errors associated with strong interfacial spectral changes as found in the presence of strong absorbers, reflector of mixed-oxide fuel. The methodology is also proven to be fully consistent with the multigroup methodology as long as a flat-flux approximation is used spatially.

Global dynamics of a novel multi-group model for computer worms

International Nuclear Information System (INIS)

Gong Yong-Wang; Song Yu-Rong; Jiang Guo-Ping

2013-01-01

In this paper, we study worm dynamics in computer networks composed of many autonomous systems. A novel multi-group SIQR (susceptible-infected-quarantined-removed) model is proposed for computer worms by explicitly considering anti-virus measures and the network infrastructure. Then, the basic reproduction number of worm R 0 is derived and the global dynamics of the model are established. It is shown that if R 0 is less than or equal to 1, the disease-free equilibrium is globally asymptotically stable and the worm dies out eventually, whereas, if R 0 is greater than 1, one unique endemic equilibrium exists and it is globally asymptotically stable, thus the worm persists in the network. Finally, numerical simulations are given to illustrate the theoretical results. (general)

Investigations of safety-related parameters applying a new multi-group diffusion code for HTR transients

International Nuclear Information System (INIS)

Kasselmann, S.; Druska, C.; Lauer, A.

2010-01-01

The energy spectra of fast and thermal neutrons from fission reactions in the FZJ code TINTE are modelled by two broad energy groups. Present demands for increased numerical accuracy led to the question of how precise the 2-group approximation is compared to a multi-group model. Therefore a new simulation program called MGT (Multi Group TINTE) has recently been developed which is able to handle up to 43 energy groups. Furthermore, an internal spectrum calculation for the determination of cross-sections can be performed for each time step and location within the reactor. In this study the multi-group energy models are compared to former calculations with only two energy groups. Different scenarios (normal operation and design-basis accidents) have been defined for a high temperature pebble bed reactor design with annular core. The effect of an increasing number of energy groups on safety-related parameters like the fuel and coolant temperature, the nuclear heat source or the xenon concentration is studied. It has been found that for the studied scenarios the use of up to 8 energy groups is a good trade-off between precision and a tolerable amount of computing time. (orig.)

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

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

Effects of Out-of-Plane Disorder on the Nodal Quasiparticle and Superconducting Gap in Single-Layer Bi_2Sr_1.6Ln_0.4CuO_6 delta (Ln = La, Nd, Gd)

Energy Technology Data Exchange (ETDEWEB)

Hashimoto, M.

2011-01-04

How out-of-plane disorder affects the electronic structure has been investigated for the single-layer cuprates Bi{sub 2}Sr{sub 1.6}Ln{sub 0.4}CuO{sub 6+{delta}} (Ln = La, Nd, Gd) by angle-resolved photoemission spectroscopy. We have observed that, with increasing disorder, while the Fermi surface shape and band dispersions are not affected, the quasi-particle width increases, the anti-nodal gap is enhanced and the superconducting gap in the nodal region is depressed. The results indicate that the superconductivity is significantly depressed by out-of-plane disorder through the enhancement of the anti-nodal gap and the depression of the superconducting gap in the nodal region.

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

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...

Multigroup neutron transport equation in the diffusion and P{sub 1} approximation

Energy Technology Data Exchange (ETDEWEB)

Obradovic, D [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

1970-07-01

Investigations of the properties of the multigroup transport operator, width and without delayed neutrons in the diffusion and P{sub 1} approximation, is performed using Keldis's theory of operator families as well as a technique . recently used for investigations into the properties of the general linearized Boltzmann operator. It is shown that in the case without delayed neutrons, multigroup transport operator in the diffusion and P{sub 1} approximation possesses a complete set of generalized eigenvectors. A formal solution to the initial value problem is also given. (author)

Complex of two-dimensional multigroup programs for neutron-physical computations of nuclear reactor

International Nuclear Information System (INIS)

Karpov, V.A.; Protsenko, A.N.

1975-01-01

Briefly stated mathematical aspects of the two-dimensional multigroup method of neutron-physical computation of nuclear reactor. Problems of algorithmization and BESM-6 computer realisation of multigroup diffuse approximations in hexagonal and rectangular calculated lattices are analysed. The results of computation of fast critical assembly having complicated composition of the core are given. The estimation of computation accuracy of criticality, neutron fields distribution and efficiency of absorbing rods by means of computer programs developed is done. (author)

Extension of the analytic nodal diffusion solver ANDES to triangular-Z geometry and coupling with COBRA-IIIc for hexagonal core analysis

International Nuclear Information System (INIS)

Lozano, Juan-Andres; Jimenez, Javier; Garcia-Herranz, Nuria; Aragones, Jose-Maria

2010-01-01

In this paper the extension of the multigroup nodal diffusion code ANDES, based on the Analytic Coarse Mesh Finite Difference (ACMFD) method, from Cartesian to hexagonal geometry is presented, as well as its coupling with the thermal-hydraulic (TH) code COBRA-IIIc for hexagonal core analysis. In extending the ACMFD method to hexagonal assemblies, triangular-Z nodes are used. In the radial plane, a direct transverse integration procedure is applied along the three directions that are orthogonal to the triangle interfaces. The triangular nodalization avoids the singularities, that appear when applying transverse integration to hexagonal nodes, and allows the advantage of the mesh subdivision capabilities implicit within that geometry. As for the thermal-hydraulics, the extension of the coupling scheme to hexagonal geometry has been performed with the capability to model the core using either assembly-wise channels (hexagonal mesh) or a higher refinement with six channels per fuel assembly (triangular mesh). Achieving this level of TH mesh refinement with COBRA-IIIc code provides a better estimation of the in-core 3D flow distribution, improving the TH core modelling. The neutronics and thermal-hydraulics coupled code, ANDES/COBRA-IIIc, previously verified in Cartesian geometry core analysis, can also be applied now to full three-dimensional VVER core problems, as well as to other thermal and fast hexagonal core designs. Verification results are provided, corresponding to the different cases of the OECD/NEA-NSC VVER-1000 Coolant Transient Benchmarks.

Study on the Control Strategy of Shifting Time Involving Multigroup Clutches

Directory of Open Access Journals (Sweden)

Zhen Zhu

2016-01-01

Full Text Available This paper focuses on the control strategy of shifting time involving multigroup clutches for a hydromechanical continuously variable transmission (HMCVT. The dynamic analyses of mathematical models are presented in this paper, and the simulation models are used to study the control strategy of HMCVT. Simulations are performed in Simulation X platform to investigate the shifting time of clutches under different operating conditions. On this basis, simulation analysis and test verification of two typical conditions, which play the decisive roles for the shifting quality, are carried out. The results show that there are differences in the shifting time of the two typical conditions. In the shifting process from the negative transmission of hydromechanical ranges to the positive transmission of hydromechanical ranges, the control strategy based on the shifting time is switching the clutches of shifting mechanism firstly and then disengaging a group of clutches of planetary gear mechanism and engaging another group of the clutches of planetary gear mechanism lastly. In the shifting process from the hydraulic range to the hydromechanical range, the control strategy based on the shifting time is switching the clutches of hydraulic shifting mechanism and planetary gear mechanism at first and then engaging the clutch of shifting mechanism.

A Multigroup diffusion solver using pseudo transient continuation for a radiation-hydrodynamic code with patch-based AMR

Energy Technology Data Exchange (ETDEWEB)

Shestakov, A I; Offner, S R

2006-09-21

We present a scheme to solve the nonlinear multigroup radiation diffusion (MGD) equations. The method is incorporated into a massively parallel, multidimensional, Eulerian radiation-hydrodynamic code with adaptive mesh refinement (AMR). The patch-based AMR algorithm refines in both space and time creating a hierarchy of levels, coarsest to finest. The physics modules are time-advanced using operator splitting. On each level, separate 'level-solve' packages advance the modules. Our multigroup level-solve adapts an implicit procedure which leads to a two-step iterative scheme that alternates between elliptic solves for each group with intra-cell group coupling. For robustness, we introduce pseudo transient continuation ({Psi}tc). We analyze the magnitude of the {Psi}tc parameter to ensure positivity of the resulting linear system, diagonal dominance and convergence of the two-step scheme. For AMR, a level defines a subdomain for refinement. For diffusive processes such as MGD, the refined level uses Dirichet boundary data at the coarse-fine interface and the data is derived from the coarse level solution. After advancing on the fine level, an additional procedure, the sync-solve (SS), is required in order to enforce conservation. The MGD SS reduces to an elliptic solve on a combined grid for a system of G equations, where G is the number of groups. We adapt the 'partial temperature' scheme for the SS; hence, we reuse the infrastructure developed for scalar equations. Results are presented. We consider a multigroup test problem with a known analytic solution. We demonstrate utility of {Psi}tc by running with increasingly larger timesteps. Lastly, we simulate the sudden release of energy Y inside an Al sphere (r = 15 cm) suspended in air at STP. For Y = 11 kT, we find that gray radiation diffusion and MGD produce similar results. However, if Y = 1 MT, the two packages yield different results. Our large Y simulation contradicts a long-standing theory

A Multigroup diffusion Solver Using Pseudo Transient Continuation for a Radiaiton-Hydrodynamic Code with Patch-Based AMR

Energy Technology Data Exchange (ETDEWEB)

Shestakov, A I; Offner, S R

2007-03-02

We present a scheme to solve the nonlinear multigroup radiation diffusion (MGD) equations. The method is incorporated into a massively parallel, multidimensional, Eulerian radiation-hydrodynamic code with adaptive mesh refinement (AMR). The patch-based AMR algorithm refines in both space and time creating a hierarchy of levels, coarsest to finest. The physics modules are time-advanced using operator splitting. On each level, separate 'level-solve' packages advance the modules. Our multigroup level-solve adapts an implicit procedure which leads to a two-step iterative scheme that alternates between elliptic solves for each group with intra-cell group coupling. For robustness, we introduce pseudo transient continuation ({Psi}tc). We analyze the magnitude of the {Psi}tc parameter to ensure positivity of the resulting linear system, diagonal dominance and convergence of the two-step scheme. For AMR, a level defines a subdomain for refinement. For diffusive processes such as MGD, the refined level uses Dirichet boundary data at the coarse-fine interface and the data is derived from the coarse level solution. After advancing on the fine level, an additional procedure, the sync-solve (SS), is required in order to enforce conservation. The MGD SS reduces to an elliptic solve on a combined grid for a system of G equations, where G is the number of groups. We adapt the 'partial temperature' scheme for the SS; hence, we reuse the infrastructure developed for scalar equations. Results are presented. We consider a multigroup test problem with a known analytic solution. We demonstrate utility of {Psi}tc by running with increasingly larger timesteps. Lastly, we simulate the sudden release of energy Y inside an Al sphere (r = 15 cm) suspended in air at STP. For Y = 11 kT, we find that gray radiation diffusion and MGD produce similar results. However, if Y = 1 MT, the two packages yield different results. Our large Y simulation contradicts a long-standing theory

Development and testing of multigroup library with correction of self-shielding effects in fusion-fission hybrid reactor

International Nuclear Information System (INIS)

Zou Jun; He Zhaozhong; Zeng Qin; Qiu Yuefeng; Wang Minghuang

2010-01-01

A multigroup library HENDL2.1/SS (Hybrid Evaluated Nuclear Data Library/Self-Shielding) based on ENDF/B-VII.0 evaluate data has been generated using Bondarenko and flux calculator method for the correction of self-shielding effect of neutronics analyses. To validate the reliability of the multigroup library HENDL2.1/SS, transport calculations for fusion-fission hybrid system FDS-I were performed in this paper. It was verified that the calculations with the HENDL2.1/SS gave almost the same results with MCNP calculations and were better than calculations with the HENDL2.0/MG which is another multigroup library without self-shielding correction. The test results also showed that neglecting resonance self-shielding caused underestimation of the K eff , neutron fluxes and waste transmutation ratios in the multigroup calculations of FDS-I.

Tumour thickness as a predictor of nodal metastases in oral cancer: comparison between tongue and floor of mouth subsites.

Science.gov (United States)

Balasubramanian, Deepak; Ebrahimi, Ardalan; Gupta, Ruta; Gao, Kan; Elliott, Michael; Palme, Carsten E; Clark, Jonathan R

2014-12-01

To identify whether tumour thickness as a predictor of nodal metastases in oral squamous cell carcinoma differs between tongue and floor of mouth (FOM) subsites. Retrospective review of 343 patients treated between 1987 and 2012. The neck was considered positive in the presence of pathologically proven nodal metastases on neck dissection or during follow-up. There were 222 oral tongue and 121 FOM tumours. In patients with FOM tumours 2.1-4mm thick, the rate of nodal metastases was 41.7%. In contrast, for tongue cancers of a similar thickness the rate was only 11.2%. This increased to 38.5% in patients with tongue cancers that were 4.1-6mm thick. Comparing these two subsites, FOM cancers cross the critical 20% threshold of probability for nodal metastases between 1 and 2mm whereas tongue cancers cross the 20% threshold just under 4mm thickness. On logistic regression adjusting for relevant covariates, there was a significant difference in the propensity for nodal metastases based on tumour thickness according to subsite (p=0.028). Thin FOM tumours (2.1-4mm) have a high rate of nodal metastases. Elective neck dissection is appropriate in FOM tumours ⩾2mm thick and in tongue tumours ⩾4mm thick. Copyright © 2014 Elsevier Ltd. All rights reserved.

Ultrasound-guided core biopsy: an effective method of detecting axillary nodal metastases.

LENUS (Irish Health Repository)

Solon, Jacqueline G

2012-02-01

BACKGROUND: Axillary nodal status is an important prognostic predictor in patients with breast cancer. This study evaluated the sensitivity and specificity of ultrasound-guided core biopsy (Ax US-CB) at detecting axillary nodal metastases in patients with primary breast cancer, thereby determining how often sentinel lymph node biopsy could be avoided in node positive patients. STUDY DESIGN: Records of patients presenting to a breast unit between January 2007 and June 2010 were reviewed retrospectively. Patients who underwent axillary ultrasonography with or without preoperative core biopsy were identified. Sensitivity, specificity, positive predictive value, and negative predictive value for ultrasonography and percutaneous biopsy were evaluated. RESULTS: Records of 718 patients were reviewed, with 445 fulfilling inclusion criteria. Forty-seven percent (n = 210\\/445) had nodal metastases, with 110 detected by Ax US-CB (sensitivity 52.4%, specificity 100%, positive predictive value 100%, negative predictive value 70.1%). Axillary ultrasonography without biopsy had sensitivity and specificity of 54.3% and 97%, respectively. Lymphovascular invasion was an independent predictor of nodal metastases (sensitivity 60.8%, specificity 80%). Ultrasound-guided core biopsy detected more than half of all nodal metastases, sparing more than one-quarter of all breast cancer patients an unnecessary sentinel lymph node biopsy. CONCLUSIONS: Axillary ultrasonography, when combined with core biopsy, is a valuable component of the management of patients with primary breast cancer. Its ability to definitively identify nodal metastases before surgical intervention can greatly facilitate a patient\\'s preoperative integrated treatment plan. In this regard, we believe our study adds considerably to the increasing data, which indicate the benefit of Ax US-CB in the preoperative detection of nodal metastases.

The Use of System Codes in Scaling Studies: Relevant Techniques for Qualifying NPP Nodalizations for Particular Scenarios

Directory of Open Access Journals (Sweden)

V. Martinez-Quiroga

2014-01-01

Full Text Available System codes along with necessary nodalizations are valuable tools for thermal hydraulic safety analysis. Qualifying both codes and nodalizations is an essential step prior to their use in any significant study involving code calculations. Since most existing experimental data come from tests performed on the small scale, any qualification process must therefore address scale considerations. This paper describes the methodology developed at the Technical University of Catalonia in order to contribute to the qualification of Nuclear Power Plant nodalizations by means of scale disquisitions. The techniques that are presented include the so-called Kv-scaled calculation approach as well as the use of “hybrid nodalizations” and “scaled-up nodalizations.” These methods have revealed themselves to be very helpful in producing the required qualification and in promoting further improvements in nodalization. The paper explains both the concepts and the general guidelines of the method, while an accompanying paper will complete the presentation of the methodology as well as showing the results of the analysis of scaling discrepancies that appeared during the posttest simulations of PKL-LSTF counterpart tests performed on the PKL-III and ROSA-2 OECD/NEA Projects. Both articles together produce the complete description of the methodology that has been developed in the framework of the use of NPP nodalizations in the support to plant operation and control.

Reference calculations on critical assemblies with Apollo2 code working with a fine multigroup mesh; Calculs de reference avec un maillage multigroupe fin sur des assemblages critiques par Apollo2

Energy Technology Data Exchange (ETDEWEB)

Aggery, A

1999-12-01

The objective of this thesis is to add to the multigroup transport code APOLLO2 the capability to perform deterministic reference calculations, for any type of reactor, using a very fine energy mesh of several thousand groups. This new reference tool allows us to validate the self-shielding model used in industrial applications, to perform depletion calculations, differential effects calculations, critical buckling calculations or to evaluate precisely data required by the self shielding model. At its origin, APOLLO2 was designed to perform routine calculations with energy meshes around one hundred groups. That is why, in the current format of cross sections libraries, almost each value of the multigroup energy transfer matrix is stored. As this format is not convenient for a high number of groups (concerning memory size), we had to search out a new format for removal matrices and consequently to modify the code. In the new format we found, only some values of removal matrices are kept (these values depend on a reconstruction precision choice), the other ones being reconstructed by a linear interpolation, what reduces the size of these matrices. Then we had to show that APOLLO2 working with a fine multigroup mesh had the capability to perform reference calculations on any assembly geometry. For that, we successfully carried out the validation with several calculations for which we compared APOLLO2 results (obtained with the universal mesh of 11276 groups) to results obtained with Monte Carlo codes (MCNP, TRIPOLI4). Physical analysis led with this new tool have been very fruitful and show a great potential for such an R and D tool. (author)

Sensitivity analysis to a RELAP5 nodalization developed for a typical TRIGA research reactor

International Nuclear Information System (INIS)

Reis, Patrícia A.L.; Costa, Antonella L.; Pereira, Claubia; Silva, Clarysson A.M.; Veloso, Maria Auxiliadora F.

2012-01-01

Highlights: ► We investigated how much the code results are affected by the code user. ► Two essential modifications were made on a previously validated nodalization. ► We used the RELAP5 code to predict the results. ► Results highlight the necessity of sensitivity analysis to have the ideal modeling. - Abstract: The main aim of this work is to identify how much the code results are affected by the code user in the choice of, for example, the number of thermal hydraulic channels in a nuclear reactor nodalization. To perform this, two essential modifications were made on a previously validated nodalization for analysis of steady-state and forced recirculation off transient in the IPR-R1 TRIGA research reactor. Experimental data were taken as reference to compare the behavior of the reactor for two different types of modeling. The results highlight the necessity of sensitivity analysis to obtain the ideal modeling to simulate a specific system.

CTD: a computer program to solve the three dimensional multi-group diffusion equation in X, Y, Z, and triangular Z geometries

Energy Technology Data Exchange (ETDEWEB)

Fletcher, J K

1973-05-01

CTD is a computer program written in Fortran 4 to solve the multi-group diffusion theory equations in X, Y, Z and triangular Z geometries. A power print- out neutron balance and breeding gain are also produced. 4 references. (auth)

Nodalization Preparation for the Transient Simulation of Cooling System for One Line Mode of RSG-GAS

International Nuclear Information System (INIS)

Sukmanto Dibyo; Susyadi; Tagor MS; Darwis Isnaeni

2004-01-01

Cooling system is important component in RSG-GAS. To carry out the transient simulation of one line-cooling mode, the model of RSG-GAS has been prepared. To illustrate the transient condition, the RELAP5.MOD3 computer code the existing input files were used. This Input consist of kinetic, thermal, hydraulic and geometries data. Modification and decrement of number of nodalization has been done to simplification as well as running time. The reasonable result of model is arranged to determine the initial condition of input data therefore steady state condition have agreement to the analysis result of one line cooling mode of RSG-GAS. Parameter investigated are transient temperatures of cooling system after decreasing of secondary cooling system occur as function of time. These parameters can be requested using input of Minor Edit Request Simulation is conducted at the reactor power of 15 MW steady-state for one-line cooling mode in which the primary and secondary cooling of 430 kg/sec and 550 kg/sec respectively. Decreasing of secondary cooling flow is caused by pump trip. As a consequence, the control rod drop due to reactor protection system. The negative reactivity of control rod causes decreasing of reactor power. Change of pattern for the primary and secondary cooling system can be known. After that simulation depicts that increasing of temperatures occur at the certain moment since initiation temperature conditions, due to reactor shut down, curve inclined move going down. (author)

Pattern of Progression after Stereotactic Body Radiotherapy for Oligometastatic Prostate Cancer Nodal Recurrences.

Science.gov (United States)

Ost, P; Jereczek-Fossa, B A; Van As, N; Zilli, T; Tree, A; Henderson, D; Orecchia, R; Casamassima, F; Surgo, A; Miralbell, R; De Meerleer, G

2016-09-01

To report the relapse pattern of stereotactic body radiotherapy (SBRT) for oligorecurrent nodal prostate cancer (PCa). PCa patients with ≤3 lymph nodes (N1/M1a) at the time of recurrence were treated with SBRT. SBRT was defined as a radiotherapy dose of at least 5 Gy per fraction to a biological effective dose of at least 80 Gy to all metastatic sites. Distant progression-free survival was defined as the time interval between the first day of SBRT and appearance of new metastatic lesions, outside the high-dose region. Relapses after SBRT were recorded and compared with the initially treated site. Secondary end points were local control, time to palliative androgen deprivation therapy and toxicity scored using the Common Terminology Criteria for Adverse Events v4.0. Overall, 89 metastases were treated in 72 patients. The median distant progression-free survival was 21 months (95% confidence interval 16-25 months) with 88% of patients having ≤3 metastases at the time of progression. The median time from first SBRT to the start of palliative androgen deprivation therapy was 44 months (95% confidence interval 17-70 months). Most relapses (68%) occurred in nodal regions. Relapses after pelvic nodal SBRT (n = 36) were located in the pelvis (n = 14), retroperitoneum (n = 1), pelvis and retroperitoneum (n = 8) or in non-nodal regions (n = 13). Relapses after SBRT for extrapelvic nodes (n = 5) were located in the pelvis (n = 1) or the pelvis and retroperitoneum (n = 4). Late grade 1 and 2 toxicity was observed in 17% (n = 12) and 4% of patients (n = 3). SBRT for oligometastatic PCa nodal recurrences is safe. Most subsequent relapses are again nodal and oligometastatic. Copyright © 2016 The Royal College of Radiologists. Published by Elsevier Ltd. All rights reserved.

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

Multi-level nonlinear diffusion acceleration method for multigroup transport k-Eigenvalue problems

International Nuclear Information System (INIS)

Anistratov, Dmitriy Y.

2011-01-01

The nonlinear diffusion acceleration (NDA) method is an efficient and flexible transport iterative scheme for solving reactor-physics problems. This paper presents a fast iterative algorithm for solving multigroup neutron transport eigenvalue problems in 1D slab geometry. The proposed method is defined by a multi-level system of equations that includes multigroup and effective one-group low-order NDA equations. The Eigenvalue is evaluated in the exact projected solution space of smallest dimensionality, namely, by solving the effective one- group eigenvalue transport problem. Numerical results that illustrate performance of the new algorithm are demonstrated. (author)

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.

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...

Study of flow over object problems by a nodal discontinuous Galerkin-lattice Boltzmann method

Science.gov (United States)

Wu, Jie; Shen, Meng; Liu, Chen

2018-04-01

The flow over object problems are studied by a nodal discontinuous Galerkin-lattice Boltzmann method (NDG-LBM) in this work. Different from the standard lattice Boltzmann method, the current method applies the nodal discontinuous Galerkin method into the streaming process in LBM to solve the resultant pure convection equation, in which the spatial discretization is completed on unstructured grids and the low-storage explicit Runge-Kutta scheme is used for time marching. The present method then overcomes the disadvantage of standard LBM for depending on the uniform meshes. Moreover, the collision process in the LBM is completed by using the multiple-relaxation-time scheme. After the validation of the NDG-LBM by simulating the lid-driven cavity flow, the simulations of flows over a fixed circular cylinder, a stationary airfoil and rotating-stationary cylinders are performed. Good agreement of present results with previous results is achieved, which indicates that the current NDG-LBM is accurate and effective for flow over object problems.

Development and Validation of NODAL-LAMBDA Program for the Calculation of the Sub-criticality of LAMDA MODES By Nodal Methods in BWR reactors

International Nuclear Information System (INIS)

Munoz-Cobo, J. L.; Merino, R.; Escriva, A.; Melara, J.; Concejal, A.

2014-01-01

We have developed a 3D code with two energy groups and diffusion theory that is capable of calculating eigenvalues lambda of a BWR reactor using nodal methods and boundary conditions that calculates ALBEDO NODAL-LAMBDA from the properties of the reflector code itself. The code calculates the sub-criticality of the first harmonic, which is involved in the stability against oscillations reactor out of phase, and which is needed for calculating the decay rate for data out of phase oscillations. The code is very fast and in a few seconds is able to make a calculation of the first eigenvalues and eigenvectors, discretized solving the problem with different matrix elements zero. The code uses the LAPACK and ARPACK libraries. It was necessary to modify the LAPACK library to perform various operations with five non-diagonal matrices simultaneously in order to reduce the number of calls to bookstores and simplify the procedure for calculating the matrices in compressed format CSR. The code is validated by comparing it with the results for SIMULATE different cases and making 3D BENCHMAR of the IAEA. (Author)

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

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...

FINELM: a multigroup finite element diffusion code. Part II

International Nuclear Information System (INIS)

Davierwalla, D.M.

1981-05-01

The author presents the axisymmetric case in cylindrical coordinates for the finite element multigroup neutron diffusion code, FINELM. The numerical acceleration schemes incorporated viz. the Lebedev extrapolations and the coarse mesh rebalancing, space collapsing, are discussed. A few benchmark computations are presented as validation of the code. (Auth.)

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

Cassandre : a two-dimensional multigroup diffusion code for reactor transient analysis

International Nuclear Information System (INIS)

Arien, B.; Daniels, J.

1986-12-01

CASSANDRE is a two-dimensional (x-y or r-z) finite element neutronics code with thermohydraulics feedback for reactor dynamics prior to the disassembly phase. It uses the multigroup neutron diffusion theory. Its main characteristics are the use of a generalized quasistatic model, the use of a flexible multigroup point-kinetics algorithm allowing for spectral matching and the use of a finite element description. The code was conceived in order to be coupled with any thermohydraulics module, although thermohydraulics feedback is only considered in r-z geometry. In steady state criticality search is possible either by control rod insertion or by homogeneous poisoning of the coolant. This report describes the main characterstics of the code structure and provides all the information needed to use the code. (Author)

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

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

Generation, Testing, and Validation of a WIMS-D/4 Multigroup Cross-Section Library Based on the JENDL-3.2 Nuclear Data

International Nuclear Information System (INIS)

Rahman, Mafizur; Takano, Hideki

2001-01-01

A new 69-group library of multigroup constants for the lattice code WIMS-D/4 has been generated with an improved resonance treatment, processing nuclear data from JENDL-3.2 by NJOY91.108. A parallel ENDF/B-VI based library has also been constructed for intercomparison of results. Benchmark calculations for a number of thermal reactor critical assemblies of both uranium and plutonium fuels have been performed with the code WIMS-D/4.1 with its three different libraries: the original WIMS library (NEA-0329/10) and the new ENDF/B-VI and JENDL-3.2 based libraries. The results calculated with both ENDF and JENDL based libraries show a similar tendency and are found in better agreement with the experimental values. Benchmark parameters are further calculated with the comprehensive lattice code SRAC95. The results from SRAC95 and WIMS-D/4.1 (both using JENDL-3.2 based libraries) agree well with each other. The new library is also verified for its applicability to mixed-oxide cores of varying plutonium contents

RZ calculations for self shielded multigroup cross sections

Energy Technology Data Exchange (ETDEWEB)

Li, M.; Sanchez, R.; Zmijarevic, I.; Stankovski, Z. [Commissariat a l' Energie Atomique CEA, Direction de l' Energie Nucleaire, DEN/DM2S/SERMA/LENR, 91191 Gif-sur-Yvette Cedex (France)

2006-07-01

A collision probability method has been implemented for RZ geometries. The method accounts for white albedo, specular and translation boundary condition on the top and bottom surfaces of the geometry and for a white albedo condition on the outer radial surface. We have applied the RZ CP method to the calculation of multigroup self shielded cross sections for Gadolinia absorbers in BWRs. (authors)

RZ calculations for self shielded multigroup cross sections

International Nuclear Information System (INIS)

Li, M.; Sanchez, R.; Zmijarevic, I.; Stankovski, Z.

2006-01-01

A collision probability method has been implemented for RZ geometries. The method accounts for white albedo, specular and translation boundary condition on the top and bottom surfaces of the geometry and for a white albedo condition on the outer radial surface. We have applied the RZ CP method to the calculation of multigroup self shielded cross sections for Gadolinia absorbers in BWRs. (authors)

Fluorine-18-Fluorodeoxyglucose PET in the mediastinal nodal staging of bronchogenic carcinoma.

Energy Technology Data Exchange (ETDEWEB)

Berlangieri, S.U.; Scott, A.M.; Knight, S.; Pointon, O.; Thomas, D.L.; O``Keefe, G.; Chan, J.G.; Egen, G.F.; Tochon-Danguy, H.J.; Clarke, C.P.; McKay, W.J. [Austin Hospital, Melbourne, VIC (Australia). Centre for Positron Emission Tomography and the Departments of Nuclear Medicine and Thoracic Surgery

1998-03-01

Full text: Non-invasive methods of pre-operative staging of non-small cell bronchogenic carcinoma are inaccurate. To determine the clinical role of positron emission tomography (PET) in the mediastinal staging of lung carcinoma, {sup 18}F-fluorodeoxyglucose (FDG) studies were performed in 25 patients with suspected non-small cell bronchogenic carcinoma and correlated with pathology. The patients comprised 20 men and 5 women (mean age 63; range 43-78 y). All patients had proven non-small cell lung carcinoma, except two, one patient with benign inflammatory disease and the other with small cell carcinoma. The FDG PET studies were acquired on a Siemens 951131R body tomography over 2-3 bed positions to include the thorax and mediastinum. The PET images were interpreted for tumour involvement of mediastinal nodes according to the American Thoracic Society classification and scored for confidence of tumour presence on a 5 point scale. The intensity of glucose metabolism was compared to mediastinal blood pool activity and graded on a 4 point scale. FDG PET correctly excluded ipsilateral mediastinal nodal (N2) disease in 16 of 16 patients. Six of nine patients with N2 disease were correctly identified by FDG PET. Of the three patients with N2 nodal involvement not detected by PET, each had single station nodal disease, and in two patients the primary lesions abutted the involved nodal group. A total of 104 nodal stations were sampled or examined at surgery. FDG PET correctly excluded disease in 83/83 (100% specificity) negative nodal stations. FDG PET is a promising non-invasive functional imaging modality for the mediastinal staging of bronchogenic carcinoma.

Non-analogue Monte Carlo method, application to neutron simulation; Methode de Monte Carlo non analogue, application a la simulation des neutrons

Energy Technology Data Exchange (ETDEWEB)

Morillon, B.

1996-12-31

With most of the traditional and contemporary techniques, it is still impossible to solve the transport equation if one takes into account a fully detailed geometry and if one studies precisely the interactions between particles and matters. Only the Monte Carlo method offers such a possibility. However with significant attenuation, the natural simulation remains inefficient: it becomes necessary to use biasing techniques where the solution of the adjoint transport equation is essential. The Monte Carlo code Tripoli has been using such techniques successfully for a long time with different approximate adjoint solutions: these methods require from the user to find out some parameters. If this parameters are not optimal or nearly optimal, the biases simulations may bring about small figures of merit. This paper presents a description of the most important biasing techniques of the Monte Carlo code Tripoli ; then we show how to calculate the importance function for general geometry with multigroup cases. We present a completely automatic biasing technique where the parameters of the biased simulation are deduced from the solution of the adjoint transport equation calculated by collision probabilities. In this study we shall estimate the importance function through collision probabilities method and we shall evaluate its possibilities thanks to a Monte Carlo calculation. We compare different biased simulations with the importance function calculated by collision probabilities for one-group and multigroup problems. We have run simulations with new biasing method for one-group transport problems with isotropic shocks and for multigroup problems with anisotropic shocks. The results show that for the one-group and homogeneous geometry transport problems the method is quite optimal without splitting and russian roulette technique but for the multigroup and heterogeneous X-Y geometry ones the figures of merit are higher if we add splitting and russian roulette technique.

Multigroup computation of the temperature-dependent Resonance Scattering Model (RSM) and its implementation

Energy Technology Data Exchange (ETDEWEB)

Ghrayeb, S. Z. [Dept. of Mechanical and Nuclear Engineering, Pennsylvania State Univ., 230 Reber Building, Univ. Park, PA 16802 (United States); Ouisloumen, M. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States); Ougouag, A. M. [Idaho National Laboratory, MS-3860, PO Box 1625, Idaho Falls, ID 83415 (United States); Ivanov, K. N.

2012-07-01

A multi-group formulation for the exact neutron elastic scattering kernel is developed. This formulation is intended for implementation into a lattice physics code. The correct accounting for the crystal lattice effects influences the estimated values for the probability of neutron absorption and scattering, which in turn affect the estimation of core reactivity and burnup characteristics. A computer program has been written to test the formulation for various nuclides. Results of the multi-group code have been verified against the correct analytic scattering kernel. In both cases neutrons were started at various energies and temperatures and the corresponding scattering kernels were tallied. (authors)

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...

Validation of RETRAN-03 by simulating a peach bottom turbine trip and boiloff at the full integral simulation test facility

International Nuclear Information System (INIS)

Westacott, J.L.; Peterson, C.E.

1992-01-01

This paper reports that the RETRAN-03 computer code is validated by simulating two tests that were performed at the Full Integral Simulation Test (FIST) facility. The RETRAN-03 results of a turbine trip (test 4PTT1) and failure to maintain water level at decay power (test T1QUV) are compared with the FIST test data. The RETRAN-03 analysis of test 4PTT1 is compared with a previous TRAC-BWR analysis of the test. Sensitivity to various model nodalizations and RETRAN-03 slip options are studied by comparing results of test T1QUV. The predicted thermal-hydraulic responses of both tests agree well with the test data. The pressure response of test 4PTT1 and the boiloff rate for test T1QUV are accurately predicted. Core uncovery time is found to be sensitive to the upper downcomer and upper plenum nodalization. The RETRAN-03 algebraic and dynamic slip options produce similar results for test T1QUV

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

Incidental Prophylactic Nodal Irradiation and Patterns of Nodal Relapse in Inoperable Early Stage NSCLC Patients Treated With SBRT: A Case-Matched Analysis

Energy Technology Data Exchange (ETDEWEB)

Lao, Louis [Department of Radiation Oncology, Princess Margaret Cancer Centre, University of Toronto, Toronto, Ontario (Canada); Department of Radiation Oncology, Auckland City Hospital, Auckland (New Zealand); Hope, Andrew J. [Department of Radiation Oncology, Princess Margaret Cancer Centre, University of Toronto, Toronto, Ontario (Canada); Maganti, Manjula [Department of Biostatistics, Princess Margaret Cancer Centre, University of Toronto, Toronto, Ontario (Canada); Brade, Anthony; Bezjak, Andrea; Saibishkumar, Elantholi P.; Giuliani, Meredith; Sun, Alexander [Department of Radiation Oncology, Princess Margaret Cancer Centre, University of Toronto, Toronto, Ontario (Canada); Cho, B. C. John, E-mail: john.cho@rmp.uhn.on.ca [Department of Radiation Oncology, Princess Margaret Cancer Centre, University of Toronto, Toronto, Ontario (Canada)

2014-09-01

Purpose: Reported rates of non-small cell lung cancer (NSCLC) nodal failure following stereotactic body radiation therapy (SBRT) are lower than those reported in the surgical series when matched for stage. We hypothesized that this effect was due to incidental prophylactic nodal irradiation. Methods and Materials: A prospectively collected group of medically inoperable early stage NSCLC patients from 2004 to 2010 was used to identify cases with nodal relapses. Controls were matched to cases, 2:1, controlling for tumor volume (ie, same or greater) and tumor location (ie, same lobe). Reference (normalized to equivalent dose for 2-Gy fractions [EQD2]) point doses at the ipsilateral hilum and carina, demographic data, and clinical outcomes were extracted from the medical records. Univariate conditional logistical regression analyses were performed with variables of interest. Results: Cases and controls were well matched except for size. The controls, as expected, had larger gross tumor volumes (P=.02). The mean ipsilateral hilar doses were 9.6 Gy and 22.4 Gy for cases and controls, respectively (P=.014). The mean carinal doses were 7.0 Gy and 9.2 Gy, respectively (P=.13). Mediastinal nodal relapses, with and without ipsilateral hilar relapse, were associated with mean ipsilateral hilar doses of 3.6 Gy and 19.8 Gy, respectively (P=.01). The conditional density plot appears to demonstrate an inverse dose-effect relationship between ipsilateral hilar normalized total dose and risk of ipsilateral hilar relapse. Conclusions: Incidental hilar dose greater than 20 Gy is significantly associated with fewer ipsilateral hilar relapses in inoperable early stage NSCLC patients treated with SBRT.

Incidental Prophylactic Nodal Irradiation and Patterns of Nodal Relapse in Inoperable Early Stage NSCLC Patients Treated With SBRT: A Case-Matched Analysis

International Nuclear Information System (INIS)

Lao, Louis; Hope, Andrew J.; Maganti, Manjula; Brade, Anthony; Bezjak, Andrea; Saibishkumar, Elantholi P.; Giuliani, Meredith; Sun, Alexander; Cho, B. C. John

2014-01-01

Purpose: Reported rates of non-small cell lung cancer (NSCLC) nodal failure following stereotactic body radiation therapy (SBRT) are lower than those reported in the surgical series when matched for stage. We hypothesized that this effect was due to incidental prophylactic nodal irradiation. Methods and Materials: A prospectively collected group of medically inoperable early stage NSCLC patients from 2004 to 2010 was used to identify cases with nodal relapses. Controls were matched to cases, 2:1, controlling for tumor volume (ie, same or greater) and tumor location (ie, same lobe). Reference (normalized to equivalent dose for 2-Gy fractions [EQD2]) point doses at the ipsilateral hilum and carina, demographic data, and clinical outcomes were extracted from the medical records. Univariate conditional logistical regression analyses were performed with variables of interest. Results: Cases and controls were well matched except for size. The controls, as expected, had larger gross tumor volumes (P=.02). The mean ipsilateral hilar doses were 9.6 Gy and 22.4 Gy for cases and controls, respectively (P=.014). The mean carinal doses were 7.0 Gy and 9.2 Gy, respectively (P=.13). Mediastinal nodal relapses, with and without ipsilateral hilar relapse, were associated with mean ipsilateral hilar doses of 3.6 Gy and 19.8 Gy, respectively (P=.01). The conditional density plot appears to demonstrate an inverse dose-effect relationship between ipsilateral hilar normalized total dose and risk of ipsilateral hilar relapse. Conclusions: Incidental hilar dose greater than 20 Gy is significantly associated with fewer ipsilateral hilar relapses in inoperable early stage NSCLC patients treated with SBRT

Phase I Trial of Pelvic Nodal Dose Escalation With Hypofractionated IMRT for High-Risk Prostate Cancer

Energy Technology Data Exchange (ETDEWEB)

Adkison, Jarrod B.; McHaffie, Derek R.; Bentzen, Soren M.; Patel, Rakesh R.; Khuntia, Deepak [Department of Human Oncology, University of Wisconsin Carbone Cancer Center, School of Medicine and Public Health, Madison, WI (United States); Petereit, Daniel G. [Department of Radiation Oncology, John T. Vucurevich Regional Cancer Care Institute, Rapid City Regional Hospital, Rapid City, SD (United States); Hong, Theodore S.; Tome, Wolfgang [Department of Human Oncology, University of Wisconsin Carbone Cancer Center, School of Medicine and Public Health, Madison, WI (United States); Ritter, Mark A., E-mail: ritter@humonc.wisc.edu [Department of Human Oncology, University of Wisconsin Carbone Cancer Center, School of Medicine and Public Health, Madison, WI (United States)

2012-01-01

Purpose: Toxicity concerns have limited pelvic nodal prescriptions to doses that may be suboptimal for controlling microscopic disease. In a prospective trial, we tested whether image-guided intensity-modulated radiation therapy (IMRT) can safely deliver escalated nodal doses while treating the prostate with hypofractionated radiotherapy in 5 Vulgar-Fraction-One-Half weeks. Methods and Materials: Pelvic nodal and prostatic image-guided IMRT was delivered to 53 National Comprehensive Cancer Network (NCCN) high-risk patients to a nodal dose of 56 Gy in 2-Gy fractions with concomitant treatment of the prostate to 70 Gy in 28 fractions of 2.5 Gy, and 50 of 53 patients received androgen deprivation for a median duration of 12 months. Results: The median follow-up time was 25.4 months (range, 4.2-57.2). No early Grade 3 Radiation Therapy Oncology Group or Common Terminology Criteria for Adverse Events v.3.0 genitourinary (GU) or gastrointestinal (GI) toxicities were seen. The cumulative actuarial incidence of Grade 2 early GU toxicity (primarily alpha blocker initiation) was 38%. The rate was 32% for Grade 2 early GI toxicity. None of the dose-volume descriptors correlated with GU toxicity, and only the volume of bowel receiving {>=}30 Gy correlated with early GI toxicity (p = 0.029). Maximum late Grades 1, 2, and 3 GU toxicities were seen in 30%, 25%, and 2% of patients, respectively. Maximum late Grades 1 and 2 GI toxicities were seen in 30% and 8% (rectal bleeding requiring cautery) of patients, respectively. The estimated 3-year biochemical control (nadir + 2) was 81.2 {+-} 6.6%. No patient manifested pelvic nodal failure, whereas 2 experienced paraaortic nodal failure outside the field. The six other clinical failures were distant only. Conclusions: Pelvic IMRT nodal dose escalation to 56 Gy was delivered concurrently with 70 Gy of hypofractionated prostate radiotherapy in a convenient, resource-efficient, and well-tolerated 28-fraction schedule. Pelvic nodal dose

MC²-3: Multigroup Cross Section Generation Code for Fast Reactor Analysis

Energy Technology Data Exchange (ETDEWEB)

Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Yang, W. S. [Argonne National Lab. (ANL), Argonne, IL (United States)

2013-11-08

The MC²-3 code is a Multigroup Cross section generation Code for fast reactor analysis, developed by improving the resonance self-shielding and spectrum calculation methods of MC²-2 and integrating the one-dimensional cell calculation capabilities of SDX. The code solves the consistent P1 multigroup transport equation using basic neutron data from ENDF/B data files to determine the fundamental mode spectra for use in generating multigroup neutron cross sections. A homogeneous medium or a heterogeneous slab or cylindrical unit cell problem is solved in ultrafine (~2000) or hyperfine (~400,000) group levels. In the resolved resonance range, pointwise cross sections are reconstructed with Doppler broadening at specified isotopic temperatures. The pointwise cross sections are directly used in the hyperfine group calculation whereas for the ultrafine group calculation, self-shielded cross sections are prepared by numerical integration of the pointwise cross sections based upon the narrow resonance approximation. For both the hyperfine and ultrafine group calculations, unresolved resonances are self-shielded using the analytic resonance integral method. The ultrafine group calculation can also be performed for two-dimensional whole-core problems to generate region-dependent broad-group cross sections. Multigroup cross sections are written in the ISOTXS format for a user-specified group structure. The code is executable on UNIX, Linux, and PC Windows systems, and its library includes all isotopes of the ENDF/BVII. 0 data.

MVP/GMVP version 3. General purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods

International Nuclear Information System (INIS)

Nagaya, Yasunobu; Okumura, Keisuke; Sakurai, Takeshi; Mori, Takamasa

2017-03-01

In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two Monte Carlo codes MVP (continuous-energy method) and GMVP (multigroup method) have been developed at Japan Atomic Energy Agency. The codes have adopted a vectorized algorithm and have been developed for vector-type supercomputers. They also support parallel processing with a standard parallelization library MPI and thus a speed-up of Monte Carlo calculations can be achieved on general computing platforms. The first and second versions of the codes were released in 1994 and 2005, respectively. They have been extensively improved and new capabilities have been implemented. The major improvements and new capabilities are as follows: (1) perturbation calculation for effective multiplication factor, (2) exact resonant elastic scattering model, (3) calculation of reactor kinetics parameters, (4) photo-nuclear model, (5) simulation of delayed neutrons, (6) generation of group constants. This report describes the physical model, geometry description method used in the codes, new capabilities and input instructions. (author)

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.

RADHEAT-V4: a code system to generate multigroup constants and analyze radiation transport for shielding safety evaluation

International Nuclear Information System (INIS)

Yamano, Naoki; Minami, Kazuyoshi; Koyama, Kinji; Naito, Yoshitaka.

1989-03-01

A modular code system RADHEAT-V4 has been developed for performing precisely neutron and photon transport analyses, and shielding safety evaluations. The system consists of the functional modules for producing coupled multi-group neutron and photon cross section sets, for analyzing the neutron and photon transport, and for calculating the atom displacement and the energy deposition due to radiations in nuclear reactor or shielding material. A precise method named Direct Angular Representation (DAR) has been developed for eliminating an error associated with the method of the finite Legendre expansion in evaluating angular distributions of cross sections and radiation fluxes. The DAR method implemented in the code system has been described in detail. To evaluate the accuracy and applicability of the code system, some test calculations on strong anisotropy problems have been performed. From the results, it has been concluded that RADHEAT-V4 is successfully applicable to evaluating shielding problems accurately for fission and fusion reactors and radiation sources. The method employed in the code system is very effective in eliminating negative values and oscillations of angular fluxes in a medium having an anisotropic source or strong streaming. Definitions of the input data required in various options of the code system and the sample problems are also presented. (author)

Recent validation experience with multigroup cross-section libraries and scale

International Nuclear Information System (INIS)

Bowman, S.M.; Wright, R.Q.; DeHart, M.D.; Parks, C.V.; Petrie, L.M.

1995-01-01

This paper will discuss the results obtained and lessons learned from an extensive validation of new ENDF/B-V and ENDF/B-VI multigroup cross-section libraries using analyses of critical experiments. The KENO V. a Monte Carlo code in version 4.3 of the SCALE computer code system was used to perform the critical benchmark calculations via the automated SCALE sequence CSAS25. The cross-section data were processed by the SCALE automated problem-dependent resonance-processing procedure included in this sequence. Prior to calling KENO V.a, CSAS25 accesses BONAMI to perform resonance self-shielding for nuclides with Bondarenko factors and NITAWL-II to process nuclides with resonance parameter data via the Nordheim Integral Treatment

Multi-level methods for solving multigroup transport eigenvalue problems in 1D slab geometry

International Nuclear Information System (INIS)

Anistratov, D. Y.; Gol'din, V. Y.

2009-01-01

A methodology for solving eigenvalue problems for the multigroup neutron transport equation in 1D slab geometry is presented. In this paper we formulate and compare different variants of nonlinear multi-level iteration methods. They are defined by means of multigroup and effective one-group low-order quasi diffusion (LOQD) equations. We analyze the effects of utilization of the effective one-group LOQD problem for estimating the eigenvalue. We present numerical results to demonstrate the performance of the iteration algorithms in different types of reactor-physics problems. (authors)

Multi-group neutron transport theory

International Nuclear Information System (INIS)

Zelazny, R.; Kuszell, A.

1962-01-01

Multi-group neutron transport theory. In the paper the general theory of the application of the K. M. Case method to N-group neutron transport theory in plane geometry is given. The eigenfunctions (distributions) for the system of Boltzmann equations have been derived and the completeness theorem has been proved. By means of general solution two examples important for reactor and shielding calculations are given: the solution of a critical and albedo problem for a slab. In both cases the system of singular integral equations for expansion coefficients into a full set of eigenfunction distributions has been reduced to the system of Fredholm-type integral equations. Some results can be applied also to some spherical problems. (author) [fr

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)

Regeneration of three sweet potato (Ipomea batatas (L.)) accessions via meristem, Nodal and callus induction

International Nuclear Information System (INIS)

Addae-Frimpomaah, F.

2012-11-01

In vitro regeneration of three sweet potato accessions UE007, UK-BNARI and SA-BNARI using meristem, nodal cuttings or callus induction was studied. Meristematic explants cultured on Murashige and Skoog (1962) basal medium supplemented with low concentration of benzylaminopurine (BAP) or kinetin resulted in callus with or without shoot development which delayed shoot emergence. The degree of callus development increased as the concentration of the cytokinin in the culture medium increased. Although, callus development was comparatively lower on kinetin amended medium than BAP amended medium, Murashige and Skoog medium supplemented with 0.25mg/1BAP had the highest shoot induction (80%). For further differentiation of callus or shoots into distinct stem and leaves, the culture were transferred into fresh MS medium supplemented with 0.25mg/1 BAP, 0.1 mg/1 NAA and 0.1 mg/1 Gibberellic acid (GA 3 . To overcome the delay in shoot initiation using meristem culture, nodal cuttings of sweet potato were used as explants and cultured on MS medium amended with 0.3 - 0.9mg/1 BAP. All explants cultured on 0.3 or 0.6mg/1 BAP developed shoots. Furthermore, liquid MS medium amended with 0.25mg/1 BAP, 0.1mg/I NAA, and 0.1mg/1 GA 3 also enhanced early shoot development from nodal cutting explants compared to solid culture. Post flask acclimatisation of meristem or nodal cutting-derived plantlets showed that meristem derived plantlets were better acclimatised than nodal cutting plants due to vigorous root development leading to higher percentage survival in pots and subsequent tuber production. Callusogenesis was achieved when leaf lobe explants were cultured on CLC/ Ipomoea medium supplemented with 1.0 - 4.0mg/1 2,4-D with 4.0mg/1 2,4-D being the optimal concentration. However, the calli were non-embryogenic and therefore could not produce embryos when transferred to 0.1mg/1 BAP amended medium but rather produced either single or multiple shoots. The highest percentage shoot (83

Combined-modality therapy for patients with regional nodal metastases from melanoma

International Nuclear Information System (INIS)

Ballo, Matthew T.; Ross, Merrick I.; Cormier, Janice N.; Myers, Jeffrey N.; Lee, Jeffrey E.; Gershenwald, Jeffrey E.; Hwu, Patrick; Zagars, Gunar K.

2006-01-01

Purpose: To evaluate the outcome and patterns of failure for patients with nodal metastases from melanoma treated with combined-modality therapy. Methods and Materials: Between 1983 and 2003, 466 patients with nodal metastases from melanoma were managed with lymphadenectomy and radiation, with or without systemic therapy. Surgery was a therapeutic procedure for clinically apparent nodal disease in 434 patients (regionally advanced nodal disease). Adjuvant radiation was generally delivered with a hypofractionated regimen. Adjuvant systemic therapy was delivered to 154 patients. Results: With a median follow-up of 4.2 years, 252 patients relapsed and 203 patients died of progressive disease. The actuarial 5-year disease-specific, disease-free, and distant metastasis-free survival rates were 49%, 42%, and 44%, respectively. By multivariate analysis, increasing number of involved lymph nodes and primary ulceration were associated with an inferior 5-year actuarial disease-specific and distant metastasis-free survival. Also, the number of involved lymph nodes was associated with the development of brain metastases, whereas thickness was associated with lung metastases, and primary ulceration was associated with liver metastases. The actuarial 5-year regional (in-basin) control rate for all patients was 89%, and on multivariate analysis there were no patient or disease characteristics associated with inferior regional control. The risk of lymphedema was highest for those patients with groin lymph node metastases. Conclusions: Although regional nodal disease can be satisfactorily controlled with lymphadenectomy and radiation, the risk of distant metastases and melanoma death remains high. A management approach to these patients that accounts for the competing risks of distant metastases, regional failure, and long-term toxicity is needed

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

An Experiment of Robust Parallel Algorithm for the Eigenvalue problem of a Multigroup Neutron Diffusion based on modified FETI-DP

Energy Technology Data Exchange (ETDEWEB)

Chang, Jonghwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

Parallelization of Monte Carlo simulation is widely adpoted. There are also several parallel algorithms developed for the SN transport theory using the parallel wave sweeping algorithm and for the CPM using parallel ray tracing. For practical purpose of reactor physics application, the thermal feedback and burnup effects on the multigroup cross section should be considered. In this respect, the domain decomposition method(DDM) is suitable for distributing the expensive cross section calculation work. Parallel transport code and diffusion code based on the Raviart-Thomas mixed finite element method was developed. However most of the developed methods rely on the heuristic convergence of flux and current at the domain interfaces. Convergence was not attained in some cases. Mechanical stress computation community has also work on the DDM to solve the stress-strain equation using the finite element methods. The most successful domain decomposition method in terms of robustness is FETI-DP. We have modified the original FETI-DP to solve the eigenvalue problem for the multigroup diffusion problem in this study.

On the calculation of multi-group fission spectrum vectors

International Nuclear Information System (INIS)

Mueller, E.Z.

1984-05-01

In this report, the problem of calculating fission spectrum vectors in a consistent manner is formulated. The practical implications of using fission spectrum vectors in multi-group transport calculations are also addressed. The significance of the weighting spectra used for the calculation of fission spectrum vectors is illustrated for the case of a simple neutronic assembly

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

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.

Multi-group diffusion perturbation calculation code. PERKY (2002)

Energy Technology Data Exchange (ETDEWEB)

Iijima, Susumu; Okajima, Shigeaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-12-01

Perturbation calculation code based on the diffusion theory ''PERKY'' is designed for nuclear characteristic analyses of fast reactor. The code calculates reactivity worth on the multi-group diffusion perturbation theory in two or three dimensional core model and kinetics parameters such as effective delayed neutron fraction, prompt neutron lifetime and absolute reactivity scale factor ({rho}{sub 0} {delta}k/k) for FCA experiments. (author)

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.

Multi-level iteration optimization for diffusive critical calculation

International Nuclear Information System (INIS)

Li Yunzhao; Wu Hongchun; Cao Liangzhi; Zheng Youqi

2013-01-01

In nuclear reactor core neutron diffusion calculation, there are usually at least three levels of iterations, namely the fission source iteration, the multi-group scattering source iteration and the within-group iteration. Unnecessary calculations occur if the inner iterations are converged extremely tight. But the convergence of the outer iteration may be affected if the inner ones are converged insufficiently tight. Thus, a common scheme suit for most of the problems was proposed in this work to automatically find the optimized settings. The basic idea is to optimize the relative error tolerance of the inner iteration based on the corresponding convergence rate of the outer iteration. Numerical results of a typical thermal neutron reactor core problem and a fast neutron reactor core problem demonstrate the effectiveness of this algorithm in the variational nodal method code NODAL with the Gauss-Seidel left preconditioned multi-group GMRES algorithm. The multi-level iteration optimization scheme reduces the number of multi-group and within-group iterations respectively by a factor of about 1-2 and 5-21. (authors)

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.

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.

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.)

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

Encapsulation of nodal cuttings and shoot tips for storage and exchange of cassava germplasm.

Science.gov (United States)

Danso, K E; Ford-Lloyd, B V

2003-04-01

We report the encapsulation of in vitro-derived nodal cuttings or shoot tips of cassava in 3% calcium alginate for storage and germplasm exchange purposes. Shoot regrowth was not significantly affected by the concentration of sucrose in the alginate matrix while root formation was. In contrast, increasing the sucrose concentration in the calcium chloride polymerisation medium significantly reduced regrowth from encapsulated nodal cuttings of accession TME 60444. Supplementing the alginate matrix with increased concentrations of 6-benzylaminopurine and alpha-naphthaleneacetic acid enhanced complete plant regrowth within 2 weeks. Furthermore, plant regrowth by encapsulated nodal cuttings and shoot tips was significantly affected by the duration of the storage period as shoot recovery decreased from almost 100% to 73.3% for encapsulated nodal cuttings and 94.4% to 60% for shoot tips after 28 days of storage. The high frequency of plant regrowth from alginate-coated micropropagules coupled with high viability percentage after 28 days of storage is highly encouraging for the exchange of cassava genetic resources. Such encapsulated micropropagules could be used as an alternative to synthetic seeds derived from somatic embryos.

Reference calculations on critical assemblies with Apollo2 code working with a fine multigroup mesh

International Nuclear Information System (INIS)

Aggery, A.

1999-12-01

The objective of this thesis is to add to the multigroup transport code APOLLO2 the capability to perform deterministic reference calculations, for any type of reactor, using a very fine energy mesh of several thousand groups. This new reference tool allows us to validate the self-shielding model used in industrial applications, to perform depletion calculations, differential effects calculations, critical buckling calculations or to evaluate precisely data required by the self shielding model. At its origin, APOLLO2 was designed to perform routine calculations with energy meshes around one hundred groups. That is why, in the current format of cross sections libraries, almost each value of the multigroup energy transfer matrix is stored. As this format is not convenient for a high number of groups (concerning memory size), we had to search out a new format for removal matrices and consequently to modify the code. In the new format we found, only some values of removal matrices are kept (these values depend on a reconstruction precision choice), the other ones being reconstructed by a linear interpolation, what reduces the size of these matrices. Then we had to show that APOLLO2 working with a fine multigroup mesh had the capability to perform reference calculations on any assembly geometry. For that, we successfully carried out the validation with several calculations for which we compared APOLLO2 results (obtained with the universal mesh of 11276 groups) to results obtained with Monte Carlo codes (MCNP, TRIPOLI4). Physical analysis led with this new tool have been very fruitful and show a great potential for such an R and D tool. (author)

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

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.

Resolution of the multigroup scattering equation in a one-dimensional geometry and subsidiary calculations: the MUDE code; Resolution de l'equation multigroupe de la diffusion dans une geometrie a une dimension et calculs annexes: code MUDE

Energy Technology Data Exchange (ETDEWEB)

Bore, C; Dandeu, Y; Saint-Amand, Ch [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1965-07-01

MUDE is a nuclear code written in FORTRAN II for IBM 7090-7094. It resolves a system of difference equations approximating to the one-dimensional multigroup neutron scattering problem. More precisely, this code makes it possible to: 1. Calculate the critical condition of a reactor (k{sub eff}, critical radius, critical composition) and the corresponding fluxes; 2. Calculate the associated fluxes and various subsidiary results; 3. Carry out perturbation calculations; 4. Study the propagation of fluxes at a distance; 5. Estimate the relative contributions of the cross sections (macroscopic or microscopic); 6. Study the changes with time of the composition of the reactor. (authors) [French] MUDE est un code nucleaire ecrit en FORTRAN II pour IBM 7090-7094. Il resout un systeme d'equations aux differences approchant le probleme de diffusion neutronique multigroupe a une dimension. Plus precisement ce code permet de: 1. Calculer la condition critique d'un reacteur (k{sub eff}, rayon critique, composition critique) et les flux correspondants; 2. Calculer les flux adjoints et divers resultats connexes; 3. Effectuer des calculs de perturbation; 4. Etudier la propagation des flux a longue distance; 5. Ponderer des sections efficaces (macroscopiques ou microscopiques); 6. Etudier l'evolution de la composition du reacteur au cours du temps. (auteurs)

A multilevel in space and energy solver for multigroup diffusion eigenvalue problems

Directory of Open Access Journals (Sweden)

Ben C. Yee

2017-09-01

Full Text Available In this paper, we present a new multilevel in space and energy diffusion (MSED method for solving multigroup diffusion eigenvalue problems. The MSED method can be described as a PI scheme with three additional features: (1 a grey (one-group diffusion equation used to efficiently converge the fission source and eigenvalue, (2 a space-dependent Wielandt shift technique used to reduce the number of PIs required, and (3 a multigrid-in-space linear solver for the linear solves required by each PI step. In MSED, the convergence of the solution of the multigroup diffusion eigenvalue problem is accelerated by performing work on lower-order equations with only one group and/or coarser spatial grids. Results from several Fourier analyses and a one-dimensional test code are provided to verify the efficiency of the MSED method and to justify the incorporation of the grey diffusion equation and the multigrid linear solver. These results highlight the potential efficiency of the MSED method as a solver for multidimensional multigroup diffusion eigenvalue problems, and they serve as a proof of principle for future work. Our ultimate goal is to implement the MSED method as an efficient solver for the two-dimensional/three-dimensional coarse mesh finite difference diffusion system in the Michigan parallel characteristics transport code. The work in this paper represents a necessary step towards that goal.

Specifications for a two-dimensional multi-group scattering code: ALCI; Specification d'un code de diffusion multigroupe a deux dimensions: ALCI

Energy Technology Data Exchange (ETDEWEB)

Bayard, J P; Guillou, A; Lago, B; Bureau du Colombier, M J; Guillou, G; Vasseur, Ch [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1965-02-01

This report describes the specifications of the ALCI programme. This programme resolves the system of difference equations similar to the homogeneous problem of multigroup neutron scattering, with two dimensions in space, in the three geometries XY, RZ, R{theta}. It is possible with this method to calculate geometric and composition criticalities and also to calculate the accessory problem on demand. The maximum number of points dealt with is 6000. The maximum permissible number of groups is 12. The internal iterations are treated by the method of alternating directions. The external iterations are accelerated using the extrapolation method due to Tchebychev. (authors) [French] Ce rapport decrit les specifications du programme ALCI. Ce programme resout le systeme d'equations aux differences approchant le probleme homogene de la diffusion neutronique multigroupe, a deux dimensions d'espace, dans les trois geometries XY, RZ, R{theta}. Il permet des calculs de criticalite geometrique et de composition et calcule sur demande le probleme adjoint. Le nombre maximum de points traites est de 6000. Le nombre maximum de groupes permis est de 12. Les iterations interieure sont traitees par la methode des directions alternees. Les iterations exterieures sont accelerees par la methode d'extrapolation de Tchebychev. (auteurs)

Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

Energy Technology Data Exchange (ETDEWEB)

Chiang, Min-Han; Wang, Jui-Yu [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Sheu, Rong-Jiun, E-mail: rjsheu@mx.nthu.edu.tw [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Liu, Yen-Wan Hsueh [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China)

2014-05-01

The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects.

Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

International Nuclear Information System (INIS)

Chiang, Min-Han; Wang, Jui-Yu; Sheu, Rong-Jiun; Liu, Yen-Wan Hsueh

2014-01-01

The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects

An analytical multigroup benchmark for (n, γ) and (n, n', γ) verification of diffusion theory algorithms

International Nuclear Information System (INIS)

Ganapol, B.D.

2011-01-01

Highlights: → Coupled neutron and gamma transport is considered in the multigroup diffusion approximation. → The model accommodates fission, up- and down-scattering and common neutron-gamma interactions. → The exact solution to the diffusion equation in a heterogeneous media of any number of regions is found. → The solution is shown to parallel the one-group case in a homogeneous medium. → The discussion concludes with a heterogeneous, 2 fuel-plate 93.2% enriched reactor fuel benchmark demonstration. - Abstract: The angular flux for the 'rod model' describing coupled neutron/gamma (n, γ) diffusion has a particularly straightforward analytical representation when viewed from the perspective of a one-group homogeneous medium. Cast in the form of matrix functions of a diagonalizable matrix, the solution to the multigroup equations in heterogeneous media is greatly simplified. We shall show exactly how the one-group homogeneous medium solution leads to the multigroup solution.

Approximate albedo boundary conditions for energy multigroup X,Y-geometry discrete ordinates nuclear global calculations

Energy Technology Data Exchange (ETDEWEB)

Silva, Davi J.M.; Nunes, Carlos E.A.; Alves Filho, Hermes; Barros, Ricardo C., E-mail: davijmsilva@yahoo.com.br, E-mail: ceanunes@yahoo.com.br, E-mail: rcbarros@pq.cnpq.br [Secretaria Municipal de Educacao de Itaborai, RJ (Brazil); Universidade Estacio de Sa (UNESA), Rio de Janeiro, RJ (Brazil); Universidade do Estado do Rio de Janeiro (UERJ), Novra Friburgo, RJ (Brazil). Instituto Politecnico. Departamento de Modelagem Computacional

2017-11-01

Discussed here is the accuracy of approximate albedo boundary conditions for energy multigroup discrete ordinates (S{sub N}) eigenvalue problems in two-dimensional rectangular geometry for criticality calculations in neutron fission reacting systems, such as nuclear reactors. The multigroup (S{sub N}) albedo matrix substitutes approximately the non-multiplying media around the core, e.g., baffle and reflector, as we neglect the transverse leakage terms within these non-multiplying regions. Numerical results to a typical model problem are given to illustrate the accuracy versus the computer running time. (author)

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

Program to solve the multigroup discrete ordinates transport equation in (x,y,z) geometry

International Nuclear Information System (INIS)

Lathrop, K.D.

1976-04-01

Numerical formulations and programming algorithms are given for the THREETRAN computer program which solves the discrete ordinates, multigroup transport equation in (x,y,z) geometry. An efficient, flexible, and general data-handling strategy is derived to make use of three hierarchies of storage: small core memory, large core memory, and disk file. Data management, input instructions, and sample problem output are described. A six-group, S 4 , 18 502 mesh point, 2 800 zone, k/sub eff/ calculation of the ZPPR-4 critical assembly required 144 min of CDC-7600 time to execute to a convergence tolerance of 5 x 10 -4 and gave results in good qualitative agreement with experiment and other calculations. 6 references

On efficiently computing multigroup multi-layer neutron reflection and transmission conditions

International Nuclear Information System (INIS)

Abreu, Marcos P. de

2007-01-01

In this article, we present an algorithm for efficient computation of multigroup discrete ordinates neutron reflection and transmission conditions, which replace a multi-layered boundary region in neutron multiplication eigenvalue computations with no spatial truncation error. In contrast to the independent layer-by-layer algorithm considered thus far in our computations, the algorithm here is based on an inductive approach developed by the present author for deriving neutron reflection and transmission conditions for a nonactive boundary region with an arbitrary number of arbitrarily thick layers. With this new algorithm, we were able to increase significantly the computational efficiency of our spectral diamond-spectral Green's function method for solving multigroup neutron multiplication eigenvalue problems with multi-layered boundary regions. We provide comparative results for a two-group reactor core model to illustrate the increased efficiency of our spectral method, and we conclude this article with a number of general remarks. (author)

Angular finite volume method for solving the multigroup transport equation with piecewise average scattering cross sections

International Nuclear Information System (INIS)

Calloo, A.; Vidal, J.F.; Le Tellier, R.; Rimpault, G.

2011-01-01

This paper deals with the solving of the multigroup integro-differential form of the transport equation for fine energy group structure. In that case, multigroup transfer cross sections display strongly peaked shape for light scatterers and the current Legendre polynomial expansion is not well-suited to represent them. Furthermore, even if considering an exact scattering cross sections representation, the scattering source in the discrete ordinates method (also known as the Sn method) being calculated by sampling the angular flux at given directions, may be wrongly computed due to lack of angular support for the angular flux. Hence, following the work of Gerts and Matthews, an angular finite volume solver has been developed for 2D Cartesian geometries. It integrates the multigroup transport equation over discrete volume elements obtained by meshing the unit sphere with a product grid over the polar and azimuthal coordinates and by considering the integrated flux per solid angle element. The convergence of this method has been compared to the S_n method for a highly anisotropic benchmark. Besides, piecewise-average scattering cross sections have been produced for non-bound Hydrogen atoms using a free gas model for thermal neutrons. LWR lattice calculations comparing Legendre representations of the Hydrogen scattering multigroup cross section at various orders and piecewise-average cross sections for this same atom are carried out (while keeping a Legendre representation for all other isotopes). (author)

Semi-continuous and multigroup models in extended kinetic theory

International Nuclear Information System (INIS)

Koller, W.

2000-01-01

The aim of this thesis is to study energy discretization of the Boltzmann equation in the framework of extended kinetic theory. In case that external fields can be neglected, the semi- continuous Boltzmann equation yields a sound basis for various generalizations. Semi-continuous kinetic equations describing a three component gas mixture interacting with monochromatic photons as well as a four component gas mixture undergoing chemical reactions are established and investigated. These equations reflect all major aspects (conservation laws, equilibria, H-theorem) of the full continuous kinetic description. For the treatment of the spatial dependence, an expansion of the distribution function in terms of Legendre polynomials is carried out. An implicit finite differencing scheme is combined with the operator splitting method. The obtained numerical schemes are applied to the space homogeneous study of binary chemical reactions and to spatially one-dimensional laser-induced acoustic waves. In the presence of external fields, the developed overlapping multigroup approach (with the spline-interpolation as its extension) is well suited for numerical studies. Furthermore, two formulations of consistent multigroup approaches to the non-linear Boltzmann equation are presented. (author)

Mapping of nodal disease in locally advanced prostate cancer: Rethinking the clinical target volume for pelvic nodal irradiation based on vascular rather than bony anatomy

International Nuclear Information System (INIS)

Shih, Helen A.; Harisinghani, Mukesh; Zietman, Anthony L.; Wolfgang, John A.; Saksena, Mansi; Weissleder, Ralph

2005-01-01

Purpose: Toxicity from pelvic irradiation could be reduced if fields were limited to likely areas of nodal involvement rather than using the standard 'four-field box.' We employed a novel magnetic resonance lymphangiographic technique to highlight the likely sites of occult nodal metastasis from prostate cancer. Methods and Materials: Eighteen prostate cancer patients with pathologically confirmed node-positive disease had a total of 69 pathologic nodes identifiable by lymphotropic nanoparticle-enhanced MRI and semiquantitative nodal analysis. Fourteen of these nodes were in the para-aortic region, and 55 were in the pelvis. The position of each of these malignant nodes was mapped to a common template based on its relation to skeletal or vascular anatomy. Results: Relative to skeletal anatomy, nodes covered a diffuse volume from the mid lumbar spine to the superior pubic ramus and along the sacrum and pelvic side walls. In contrast, the nodal metastases mapped much more tightly relative to the large pelvic vessels. A proposed pelvic clinical target volume to encompass the region at greatest risk of containing occult nodal metastases would include a 2.0-cm radial expansion volume around the distal common iliac and proximal external and internal iliac vessels that would encompass 94.5% of the pelvic nodes at risk as defined by our node-positive prostate cancer patient cohort. Conclusions: Nodal metastases from prostate cancer are largely localized along the major pelvic vasculature. Defining nodal radiation treatment portals based on vascular rather than bony anatomy may allow for a significant decrease in normal pelvic tissue irradiation and its associated toxicities

MVP/GMVP Version 3. General purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods (Translated document)

International Nuclear Information System (INIS)

Nagaya, Yasunobu; Okumura, Keisuke; Sakurai, Takeshi; Mori, Takamasa

2017-03-01

In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two Monte Carlo codes MVP (continuous-energy method) and GMVP (multigroup method) have been developed at Japan Atomic Energy Agency. The codes have adopted a vectorized algorithm and have been developed for vector-type supercomputers. They also support parallel processing with a standard parallelization library MPI and thus a speed-up of Monte Carlo calculations can be achieved on general computing platforms. The first and second versions of the codes were released in 1994 and 2005, respectively. They have been extensively improved and new capabilities have been implemented. The major improvements and new capabilities are as follows: (1) perturbation calculation for effective multiplication factor, (2) exact resonant elastic scattering model, (3) calculation of reactor kinetics parameters, (4) photo-nuclear model, (5) simulation of delayed neutrons, (6) generation of group constants. This report describes the physical model, geometry description method used in the codes, new capabilities and input instructions. (author)

Social comparison and perceived breach of psychological contract: their effects on burnout in a multigroup analysis.

Science.gov (United States)

Cantisano, Gabriela Topa; Domínguez, J Francisco Morales; García, J Luis Caeiro

2007-05-01

This study focuses on the mediator role of social comparison in the relationship between perceived breach of psychological contract and burnout. A previous model showing the hypothesized effects of perceived breach on burnout, both direct and mediated, is proposed. The final model reached an optimal fit to the data and was confirmed through multigroup analysis using a sample of Spanish teachers (N = 401) belonging to preprimary, primary, and secondary schools. Multigroup analyses showed that the model fit all groups adequately.

The problem of resonance self-shielding effect in neutron multigroup calculations

International Nuclear Information System (INIS)

Wang Qingming; Huang Jinghua

1991-01-01

It is not allowed to neglect the resonance self-shielding effect in hybrid blanket and fast reactor neutron designs. The authors discussed the importance as well as the method of considering the resonance self-shielding effect in hybrid blanket and fast reactor neutron multigroup calculations

A simple nodal force distribution method in refined finite element meshes

Energy Technology Data Exchange (ETDEWEB)

Park, Jai Hak [Chungbuk National University, Chungju (Korea, Republic of); Shin, Kyu In [Gentec Co., Daejeon (Korea, Republic of); Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

2017-05-15

In finite element analyses, mesh refinement is frequently performed to obtain accurate stress or strain values or to accurately define the geometry. After mesh refinement, equivalent nodal forces should be calculated at the nodes in the refined mesh. If field variables and material properties are available at the integration points in each element, then the accurate equivalent nodal forces can be calculated using an adequate numerical integration. However, in certain circumstances, equivalent nodal forces cannot be calculated because field variable data are not available. In this study, a very simple nodal force distribution method was proposed. Nodal forces of the original finite element mesh are distributed to the nodes of refined meshes to satisfy the equilibrium conditions. The effect of element size should also be considered in determining the magnitude of the distributing nodal forces. A program was developed based on the proposed method, and several example problems were solved to verify the accuracy and effectiveness of the proposed method. From the results, accurate stress field can be recognized to be obtained from refined meshes using the proposed nodal force distribution method. In example problems, the difference between the obtained maximum stress and target stress value was less than 6 % in models with 8-node hexahedral elements and less than 1 % in models with 20-node hexahedral elements or 10-node tetrahedral elements.

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

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

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.

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.

Type-I and type-II topological nodal superconductors with s -wave interaction

Science.gov (United States)

Huang, Beibing; Yang, Xiaosen; Xu, Ning; Gong, Ming

2018-01-01

Topological nodal superconductors with protected gapless points in momentum space are generally realized based on unconventional pairings. In this work we propose a minimal model to realize these topological nodal phases with only s -wave interaction. In our model the linear and quadratic spin-orbit couplings along the two orthogonal directions introduce anisotropic effective unconventional pairings in momentum space. This model may support different nodal superconducting phases characterized by either an integer winding number in BDI class or a Z2 index in D class at the particle-hole invariant axes. In the vicinity of the nodal points the effective Hamiltonian can be described by either type-I or type-II Dirac equations, and the Lifshitz transition from type-I nodal phases to type-II nodal phases can be driven by external in-plane magnetic fields. We show that these nodal phases are robust against weak impurities, which only slightly renormalizes the momentum-independent parameters in the impurity-averaged Hamiltonian, thus these phases are possible to be realized in experiments with real semi-Dirac materials. The smoking-gun evidences to verify these phases based on scanning tunneling spectroscopy method are also briefly discussed.

COMPAR, NJOY, GROUPIE, FLANGE-2, ETOG-3, XLACS Multigroup Cross-Sections General Comparison

International Nuclear Information System (INIS)

Anaf, Jaime; Chalhoub, E.S.

1990-01-01

1 - Description of program or function: A system for comparing multigroup cross sections generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS. This system comprises the COMPAR program and interface (auxiliary) programs developed for each of the programs under consideration. These are REDCOMP for GROUPIE, FLACOMP for FLANGE-II, ETOCOMP for ETOG-3 and XLACOMP for XLACS. For the NJOY program there is RGENDF, a program developed apart from this system. It is a modular system in which the inclusion of new multigroup cross section generating program requires no more than the development of a new interface module. 2 - Method of solution: Refer to comments in main routine. 3 - Restrictions on the complexity of the problem: Refer to comments in main routine

CLUB - a multigroup integral transport theory code for lattice calculations of PHWR cells

International Nuclear Information System (INIS)

Krishnani, P.D.

1992-01-01

The computer code CLUB has been developed to calculate lattice parameters as a function of burnup for a pressurised heavy water reactor (PHWR) lattice cell containing fuel in the form of cluster. It solves the multigroup integral transport equation by the method based on combination of small scale collision probability (CP) method and large scale interface current technique. The calculations are performed by using WIMS 69 group cross section library or its condensed versions of 27 or 28 group libraries. It can also compute Keff from the given geometrical buckling in the input using multigroup diffusion theory in fundamental mode. The first order differential burnup equations can be solved by either Trapezoidal rule or Runge-Kutta method. (author). 17 refs., 2 figs

Nodalization qualification process of the PSBVVER facility for the Cathare2 thermal-hydraulic code

International Nuclear Information System (INIS)

Del Nevo, A.; Araneo, D.; D'Auria, F.; Galassi, G.

2004-01-01

The present document deals with the nodalization qualification process of the PSB-VVER test facility for Cathare2 code. PSB-VVER facility is a 1/300 volume scale model of a VVER-1000, reactor installed at Electrogorsk Research and Engineering Centre in 1998. The version V1.5b of the Cathare2 code has been used. In order to evaluate the nodalization performance, the qualifying procedure set up at the DIMNP of Pisa University (UNIPI) has been applied that foresees two qualification levels: a 'steady state' level and an 'on transient' level. After the steady state behavior check of the nodalization, it has been preformed the on transient qualification the PSB-VVER test 2. It is a 11% equivalent break in Upper Plenum with the actuation of one high pressure injection system, connected to the hot leg of the loop 4, and 4 passive systems (ECCS hydro-accumulators), connected to the outlet plenum and to the inlet chamber of the downcomer. The low-pressure injection system is not available in the test. The goal of this paper is to demonstrate that the first step of the nodalization qualification adopted for the PSB test analyses is achieved and the PSB facility input deck is available and ready to use. The quantitative accuracy of the performed calculation has been evaluated by using the FFT-BM tool developed at the University of Pisa.(author)

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.

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).

Correction of multigroup cross sections for resolved resonance interference in mixed absorbers

International Nuclear Information System (INIS)

Williams, M.L.

1982-07-01

The effect that interference between resolved resonances has on averaging multigroup cross sections is examined for thermal reactor-type problems. A simple and efficient numerical scheme is presented to correct a preprocessed multigroup library for interference effects. The procedure is implemented in a design oriented lattice physics computer code and compared with rigorous numerical calculations. The approximate method for computing resonance interference correction factors is applied to obtaining fine-group cross sections for a homogeneous uranium-plutonium mixture and a uranium oxide lattice. It was found that some fine group cross sections are changed by more than 40% due to resonance interference. The change in resonance interference correction factors due to burnup of a PWR fuel pin is examined and found to be small. The effect of resolved resonance interference on collapsed broad-group cross sections for thermal reactor calculations is discussed

Angular finite volume method for solving the multigroup transport equation with piecewise average scattering cross sections

Energy Technology Data Exchange (ETDEWEB)

Calloo, A.; Vidal, J.F.; Le Tellier, R.; Rimpault, G., E-mail: ansar.calloo@cea.fr, E-mail: jean-francois.vidal@cea.fr, E-mail: romain.le-tellier@cea.fr, E-mail: gerald.rimpault@cea.fr [CEA, DEN, DER/SPRC/LEPh, Saint-Paul-lez-Durance (France)

2011-07-01

This paper deals with the solving of the multigroup integro-differential form of the transport equation for fine energy group structure. In that case, multigroup transfer cross sections display strongly peaked shape for light scatterers and the current Legendre polynomial expansion is not well-suited to represent them. Furthermore, even if considering an exact scattering cross sections representation, the scattering source in the discrete ordinates method (also known as the Sn method) being calculated by sampling the angular flux at given directions, may be wrongly computed due to lack of angular support for the angular flux. Hence, following the work of Gerts and Matthews, an angular finite volume solver has been developed for 2D Cartesian geometries. It integrates the multigroup transport equation over discrete volume elements obtained by meshing the unit sphere with a product grid over the polar and azimuthal coordinates and by considering the integrated flux per solid angle element. The convergence of this method has been compared to the S{sub n} method for a highly anisotropic benchmark. Besides, piecewise-average scattering cross sections have been produced for non-bound Hydrogen atoms using a free gas model for thermal neutrons. LWR lattice calculations comparing Legendre representations of the Hydrogen scattering multigroup cross section at various orders and piecewise-average cross sections for this same atom are carried out (while keeping a Legendre representation for all other isotopes). (author)

A multigroup treatment of radiation transport

International Nuclear Information System (INIS)

Tahir, N.A.; Laing, E.W.; Nicholas, D.J.

1980-12-01

A multi-group radiation package is outlined which will accurately handle radiation transfer problems in laser-produced plasmas. Bremsstrahlung, recombination and line radiation are included as well as fast electron Bremsstrahlung radiation. The entire radiation field is divided into a large number of groups (typically 20), which diffuse radiation energy in real space as well as in energy space, the latter occurring via electron-radiation interaction. Using this model a radiation transport code will be developed to be incorporated into MEDUSA. This modified version of MEDUSA will be used to study radiative preheat effects in laser-compression experiments at the Central Laser Facility, Rutherford Laboratory. The model is also relevant to heavy ion fusion studies. (author)

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)

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)

Optical conductivity of three and two dimensional topological nodal-line semimetals

Science.gov (United States)

Barati, Shahin; Abedinpour, Saeed H.

2017-10-01

The peculiar shape of the Fermi surface of topological nodal-line semimetals at low carrier concentrations results in their unusual optical and transport properties. We analytically investigate the linear optical responses of three- and two-dimensional nodal-line semimetals using the Kubo formula. The optical conductivity of a three-dimensional nodal-line semimetal is anisotropic. Along the axial direction (i.e., the direction perpendicular to the nodal-ring plane), the Drude weight has a linear dependence on the chemical potential at both low and high carrier dopings. For the radial direction (i.e., the direction parallel to the nodal-ring plane), this dependence changes from linear into quadratic in the transition from low into high carrier concentration. The interband contribution into optical conductivity is also anisotropic. In particular, at large frequencies, it saturates to a constant value for the axial direction and linearly increases with frequency along the radial direction. In two-dimensional nodal-line semimetals, no interband optical transition could be induced and the only contribution to the optical conductivity arises from the intraband excitations. The corresponding Drude weight is independent of the carrier density at low carrier concentrations and linearly increases with chemical potential at high carrier doping.

Analysis of the asymmetrically expressed Ablim1 locus reveals existence of a lateral plate Nodal-independent left sided signal and an early, left-right independent role for nodal flow

Directory of Open Access Journals (Sweden)

Hilton Helen

2010-05-01

Full Text Available Abstract Background Vertebrates show clear asymmetry in left-right (L-R patterning of their organs and associated vasculature. During mammalian development a cilia driven leftwards flow of liquid leads to the left-sided expression of Nodal, which in turn activates asymmetric expression of the transcription factor Pitx2. While Pitx2 asymmetry drives many aspects of asymmetric morphogenesis, it is clear from published data that additional asymmetrically expressed loci must exist. Results A L-R expression screen identified the cytoskeletally-associated gene, actin binding lim protein 1 (Ablim1, as asymmetrically expressed in both the node and left lateral plate mesoderm (LPM. LPM expression closely mirrors that of Nodal. Significantly, Ablim1 LPM asymmetry was detected in the absence of detectable Nodal. In the node, Ablim1 was initially expressed symmetrically across the entire structure, resolving to give a peri-nodal ring at the headfold stage in a flow and Pkd2-dependent manner. The peri-nodal ring of Ablim1 expression became asymmetric by the mid-headfold stage, showing stronger right than left-sided expression. Node asymmetry became more apparent as development proceeded; expression retreated in an anticlockwise direction, disappearing first from the left anterior node. Indeed, at early somite stages Ablim1 shows a unique asymmetric expression pattern, in the left lateral plate and to the right side of the node. Conclusion Left LPM Ablim1 is expressed in the absence of detectable LPM Nodal, clearly revealing existence of a Pitx2 and Nodal-independent left-sided signal in mammals. At the node, a previously unrecognised action of early nodal flow and Pkd2 activity, within the pit of the node, influences gene expression in a symmetric manner. Subsequent Ablim1 expression in the peri-nodal ring reveals a very early indication of L-R asymmetry. Ablim1 expression analysis at the node acts as an indicator of nodal flow. Together these results make

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...

Verification of KARMA GEOM/TRPT Module with Given Multi-group Cross Sections

International Nuclear Information System (INIS)

Koo, Bon Seung; Hong, Ser Gi; Song, Jae Seung

2009-01-01

KAERI has developed a two-dimensional multigroup transport theory code KARMA (Kernel Analyzer by Ray-tracing Method for Fuel Assembly). KARMA uses CMFD (Coarse Mesh Finite Difference) accelerated MOC (Method of Characteristics) method for burnup calculation on a single fuel pin, a fuel assembly and a core consisting of rectangular array of fuel pins. KARMA code intends to be employed as a nuclear design tool for the Korean commercial pressurizer water reactor. Prior to the application to actual assembly designs, the code has to be approved by regularity agency. Therefore, it is essential that the reliability of KARMA code should be sufficiently evaluated against well-defined benchmark problems. In this paper, verification of GEOM/TRPT modules of KARMA was performed to confirm a reliability of the KARMA transport solution via comparisons with Monte Carlo calculations by using a consistent set of multi-group macroscopic cross-sections

Scalable Multi-group Key Management for Advanced Metering Infrastructure

OpenAIRE

Benmalek , Mourad; Challal , Yacine; Bouabdallah , Abdelmadjid

2015-01-01

International audience; Advanced Metering Infrastructure (AMI) is composed of systems and networks to incorporate changes for modernizing the electricity grid, reduce peak loads, and meet energy efficiency targets. AMI is a privileged target for security attacks with potentially great damage against infrastructures and privacy. For this reason, Key Management has been identified as one of the most challenging topics in AMI development. In this paper, we propose a new Scalable multi-group key ...

Optimal calculational schemes for solving multigroup photon transport problem

International Nuclear Information System (INIS)

Dubinin, A.A.; Kurachenko, Yu.A.

1987-01-01

A scheme of complex algorithm for solving multigroup equation of radiation transport is suggested. The algorithm is based on using the method of successive collisions, the method of forward scattering and the spherical harmonics method, and is realized in the FORAP program (FORTRAN, BESM-6 computer). As an example the results of calculating reactor photon transport in water are presented. The considered algorithm being modified may be used for solving neutron transport problems

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.

ANOVA-HDMR structure of the higher order nodal diffusion solution

International Nuclear Information System (INIS)

Bokov, P. M.; Prinsloo, R. H.; Tomasevic, D. I.

2013-01-01

Nodal diffusion methods still represent a standard in global reactor calculations, but employ some ad-hoc approximations (such as the quadratic leakage approximation) which limit their accuracy in cases where reference quality solutions are sought. In this work we solve the nodal diffusion equations utilizing the so-called higher-order nodal methods to generate reference quality solutions and to decompose the obtained solutions via a technique known as High Dimensional Model Representation (HDMR). This representation and associated decomposition of the solution provides a new formulation of the transverse leakage term. The HDMR structure is investigated via the technique of Analysis of Variance (ANOVA), which indicates why the existing class of transversely-integrated nodal methods prove to be so successful. Furthermore, the analysis leads to a potential solution method for generating reference quality solutions at a much reduced calculational cost, by applying the ANOVA technique to the full higher order solution. (authors)

Adaptive solution of the multigroup diffusion equation on irregular structured grids using a conforming finite element method formulation

International Nuclear Information System (INIS)

Ragusa, J. C.

2004-01-01

In this paper, a method for performing spatially adaptive computations in the framework of multigroup diffusion on 2-D and 3-D Cartesian grids is investigated. The numerical error, intrinsic to any computer simulation of physical phenomena, is monitored through an a posteriori error estimator. In a posteriori analysis, the computed solution itself is used to assess the accuracy. By efficiently estimating the spatial error, the entire computational process is controlled through successively adapted grids. Our analysis is based on a finite element solution of the diffusion equation. Bilinear test functions are used. The derived a posteriori error estimator is therefore based on the Hessian of the numerical solution. (authors)

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)

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)

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)

Prophylactic Level VII Nodal Dissection as a Prognostic Factor in Papillary Thyroid Carcinoma: a Pilot Study of 27 Patients.

Science.gov (United States)

Fayek, Ihab Samy

2015-01-01

Prognostic value of prophylactic level VII nodal dissection in papillary thyroid carcinoma has been highlighted. A total of 27 patients with papillary thyroid carcinoma with N0 neck underwent total thyroidectomy with level VI and VII nodal dissection through same collar neck incision. Multicentricity, bilaterality, extrathyroidal extension, level VI and VII lymph nodes were studied as separate and independent prognostic factors for DFS at 24 months. 21 females and 6 males with a mean age of 34.6 years old, tumor size was 5-24 mm. (mean 12.4 mm.), multicentricity in 11 patients 2-4 foci (mean 2.7), bilaterality in 8 patients and extrathyroidal extension in 8 patients. Dissected level VI LNs 2-8 (mean 5 LNs) and level VII LNs 1-4 (mean 1.9). Metastatic level VI LNs 0-3 (mean 1) and level VII LNs 0-2 (mean 0.5). Follow-up from 6-51 months (mean 25.6) with 7 patients showed recurrence (3 local and 4 distant). Cumulative DFS at 24 months was 87.8% and was significantly affected in relation to bilaterality (p-valueVII positive ((p-valueVII nodal involvement. Level VII prophylactic nodal dissection is an important and integral prognostic factor in papillary thyroid carcinoma. A larger multicenter study is crucial to reach a satisfactory conclusion about the necessity and safety of this approach.

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

Development of 3D multi-group neutron diffusion code for hexagonal geometry

International Nuclear Information System (INIS)

Sun Wei; Wang Kan; Ni Dongyang; Li Qing

2013-01-01

Based on the theory of new flux expansion nodal method to solve the neutron diffusion equations, the intra-nodal fluence rate distribution was expanded in a series of analytic basic functions for each group. In order to improve the accuracy of calculation result, continuities of neutron fluence rate and current were utilized across the nodal surfaces. According to the boundary conditions, the iteration method was adopted to solve the diffusion equation, where inner iteration speedup method is Gauss-Seidel method and outer is Lyusternik-Wagner. A new speedup method (one-outer-iteration and multi-inner-iteration method) was proposed according to the characteristic that the convergence speed of multiplication factor is faster than that of neutron fluence rate and the update of inner iteration matrix is slow. Based on the proposed model, the code HANDF-D was developed and tested by 3D two-group vver440 benchmark, experiment 2 of HFETR, 3D four-group thermal reactor benchmark, and 3D seven-group fast reactor benchmark. The numerical results show that HANDF-D can predict accurately the multiplication factor and nodal powers. (authors)

Cilia are required for asymmetric nodal induction in the sea urchin embryo.

Science.gov (United States)

Tisler, Matthias; Wetzel, Franziska; Mantino, Sabrina; Kremnyov, Stanislav; Thumberger, Thomas; Schweickert, Axel; Blum, Martin; Vick, Philipp

2016-08-23

Left-right (LR) organ asymmetries are a common feature of metazoan animals. In many cases, laterality is established by a conserved asymmetric Nodal signaling cascade during embryogenesis. In most vertebrates, asymmetric nodal induction results from a cilia-driven leftward fluid flow at the left-right organizer (LRO), a ciliated epithelium present during gastrula/neurula stages. Conservation of LRO and flow beyond the vertebrates has not been reported yet. Here we study sea urchin embryos, which use nodal to establish larval LR asymmetry as well. Cilia were found in the archenteron of embryos undergoing gastrulation. Expression of foxj1 and dnah9 suggested that archenteron cilia were motile. Cilia were polarized to the posterior pole of cells, a prerequisite of directed flow. High-speed videography revealed rotating cilia in the archenteron slightly before asymmetric nodal induction. Removal of cilia through brief high salt treatments resulted in aberrant patterns of nodal expression. Our data demonstrate that cilia - like in vertebrates - are required for asymmetric nodal induction in sea urchin embryos. Based on these results we argue that the anterior archenteron represents a bona fide LRO and propose that cilia-based symmetry breakage is a synapomorphy of the deuterostomes.

The application of modern nodal methods to PWR reactor physics analysis

International Nuclear Information System (INIS)

Knight, M.P.

1988-06-01

The objective of this research is to develop efficient computational procedures for PWR reactor calculations, based on modern nodal methods. The analytic nodal method, which is characterised by the use of exact exponential expansions in transverse-integrated equations, is implemented within an existing finite-difference code. This shows considerable accuracy and efficiency on standard benchmark problems, very much in line with existing experience with nodal methods., Assembly powers can be calculated to within 2.0% with just one mesh per assembly. (author)

MC2-2, Calculation of Fast Neutron Spectra and Multigroup Cross-Sections from ENDF/B Data

International Nuclear Information System (INIS)

2001-01-01

1 - Description of program or function: MC 2 -2 solves the neutron slowing-down equations using basic neutron data derived from ENDF/B data files to determine fundamental mode spectra for use in generating multigroup neutron cross sections. The current edition includes the ability to treat all ENDF/B-V and -VI data representations. It accommodates high-order P scattering representations and provides numerous capabilities such as isotope mixing, delayed neutron processing, free-format input, and flexibility in output data selection. This edition supersedes previous releases of the MC22 program and the earlier MC2 program. Improved physics algorithms and increased computational efficiency are incorporated. Input data files required by MC2-2 may be generated from ENDF/B data by the code ETOE-2. The hyper-fine-group integral transport theory module of MC2-2, RABANL, is an improved version of the RABBLE/RABID codes. Many of the MC2-2 modules are used in the SDX code. 2 - Methods: The extended transport P1, B1, consistent P1, and consistent B1 fundamental mode ultra-fine-group equations are solved using continuous slowing-down theory and multigroup methods. Fast and accurate resonance integral methods are used in the narrow resonance resolved and unresolved resonance treatments. A fundamental mode homogeneous unit cell calculation is performed using either a multigroup or a continuous slowing-down treatment. Multigroup neutron homogeneous cross sections are generated in an ISOTXS format for an arbitrary group structure. A hyper-fine-group integral transport slowing down calculation (RABANL) is available as an option. RABANL performs a homogeneous or heterogeneous (pin or slab) unit cell calculation over the resonance region (resolved and unresolved) and generates multigroup neutron cross sections in an ISOTXS format. Neutron cross sections are generated by RABANL for the homogeneous unit cell and for each heterogeneous region in the pin or slab unit cell calculation

Achievement and qualification of multigroup cross-section library for light water reactor calculation

International Nuclear Information System (INIS)

Gastaldi, B.

1986-07-01

This study intends to improve then to check on integral experiments, the calculation of the main neutronic parameters in light water moderated lattices: Uranium 238 capture and consequently Plutonium 239 build-up, multiplication factor, temperature coefficient. The first part of this work concerns the resonant reaction rate calculation method implemented in the APOLLO code, the so-called LIVOLANT and JEANPIERRE formalism. The errors introduced by the corresponding assumptions are quantified and we propose substitution methods which avoid large biases and supply satisfactory results. The second part is dedicated to the cross-section evaluation of uranium major isotopes and to the achievement of APOLLO multigroup cross-sections. This cross-section set takes into considerations on the one hand the recent differential information and the other hand the various integral information obtained in the French Atomic Energy Commission facilities. The nuclear data file (JEF abd ENDF/B5) processing, for multigroup and self-shielded cross-sections achieving enable us to check the new THEMIS computer code. In the last part, the experimental validation of the proposed procedure (accurate formalism mutuel shielding and new multigroup library) is presented. This qualification is based on the reinterpretation of critical experiments performed in the EOLE reactor at Cadarache and spent fuel analysis. The corresponding results demonstrate that our propositions provide improvements on the computation of the PWR neutronic parameters; calculation-experiment discrepancies are now consistent with experimental uncertainty margins. 46 refs; 31 figs; 23 tabl [fr

Prognostic impact of the level of nodal involvement: retrospective analysis of patients with advanced oral squamous cell carcinoma.

Science.gov (United States)

Murakami, R; Nakayama, H; Semba, A; Hiraki, A; Nagata, M; Kawahara, K; Shiraishi, S; Hirai, T; Uozumi, H; Yamashita, Y

2017-01-01

We retrospectively evaluated the prognostic impact of the level of nodal involvement in patients with advanced oral squamous cell carcinoma (SCC). Between 2005 and 2010, 105 patients with clinical stage III or IV oral SCC had chemoradiotherapy preoperatively. Clinical (cN) and pathological nodal (pN) involvement was primarily at levels Ib and II. We defined nodal involvement at levels Ia and III-V as anterior and inferior extensions, respectively, and recorded such findings as extensive. With respect to pretreatment variables (age, clinical stage, clinical findings of the primary tumour, and nodal findings), univariate analysis showed that extensive cN was the only significant factor for overall survival (hazard ratio [HR], 3.27; 95% CI 1.50 to 7.13; p=0.001). Univariate analysis showed that all pN findings, including the nodal classification (invaded nodes, multiple, and contralateral) and extensive involvement were significant, and multivariate analysis confirmed that extensive pN (HR 4.71; 95% CI 1.85 to 11.97; p=0.001) and multiple pN (HR 2.59; 95% CI 1.10 to 6.09; p=0.029) were independent predictors of overall survival. Assessment based on the level of invaded neck nodes may be a better predictor of survival than the current nodal classification. Copyright © 2016 The British Association of Oral and Maxillofacial Surgeons. Published by Elsevier Ltd. All rights reserved.

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

Comparison of treatment outcomes between involved-field and elective nodal irradiation in limited-stage small cell lung cancer

International Nuclear Information System (INIS)

Han, Tae-Jin; Kim, Hak-Jae; Wu, Hong-Gyun; Heo, Dae-Seog; Kim, Young-Whan; Lee, Se-Hoon

2012-01-01

The present study was performed to assess the usefulness of involved-field irradiation and the impact of 18 F-fluorodeoxyglucose-positron emission tomography-based staging on treatment outcomes in limited-stage small cell lung cancer. Eighty patients who received definitive chemoradiotherapy for limited-stage small cell lung cancer were retrospectively analyzed. Fifty patients were treated with involved-field irradiation, which means that the radiotherapy portal includes only clinically identifiable tumors. The other 30 patients were irradiated with a comprehensive portal, including uninvolved mediastinal and/or supraclavicular lymph nodes, so-called elective nodal irradiation. No significant difference was seen in clinical factors between the two groups. At a median follow-up of 27 months (range, 5-75 months), no significant differences were observed in 3 year overall survival (44.6 vs. 54.1%, P=0.220) and 3 year progression-free survival (24.4 vs. 42.8%, P=0.133) between the involved-field irradiation group and the elective nodal irradiation group, respectively. For patients who did not undergo positron emission tomography scans, 3 year overall survival (29.3 vs. 56.3%, P=0.022) and 3 year progression-free survival (11.0 vs. 50.0%, P=0.040) were significantly longer in the elective nodal irradiation group. Crude incidences of isolated nodal failure were 6.0% in the involved-field irradiation group and 0% in the elective nodal irradiation group, respectively. All isolated nodal failures were developed in patients who had not undergone positron emission tomography scans in their initial work-ups. If patients did not undergo positron emission tomography-based staging, the omission of elective nodal irradiation resulted in impaired survival outcomes and raised the risk of isolated nodal failure. Therefore, involved-field irradiation for limited-stage small cell lung cancer might be reasonable only with positron emission tomography scan implementation. (author)

Three-dimensional h-adaptivity for the multigroup neutron diffusion equations

KAUST Repository

Wang, Yaqi

2009-04-01

Adaptive mesh refinement (AMR) has been shown to allow solving partial differential equations to significantly higher accuracy at reduced numerical cost. This paper presents a state-of-the-art AMR algorithm applied to the multigroup neutron diffusion equation for reactor applications. In order to follow the physics closely, energy group-dependent meshes are employed. We present a novel algorithm for assembling the terms coupling shape functions from different meshes and show how it can be made efficient by deriving all meshes from a common coarse mesh by hierarchic refinement. Our methods are formulated using conforming finite elements of any order, for any number of energy groups. The spatial error distribution is assessed with a generalization of an error estimator originally derived for the Poisson equation. Our implementation of this algorithm is based on the widely used Open Source adaptive finite element library deal.II and is made available as part of this library\\'s extensively documented tutorial. We illustrate our methods with results for 2-D and 3-D reactor simulations using 2 and 7 energy groups, and using conforming finite elements of polynomial degree up to 6. © 2008 Elsevier Ltd. All rights reserved.

In Vitro Regeneration of Shoots From Nodal Explants of Dendrobium Chrysotoxum Lindl

Directory of Open Access Journals (Sweden)

Kaur Saranjeet

2017-06-01

Full Text Available Transverse sections (2 mm thickness of stem-nodes from in vitro raised seedlings had morphogenic potential on semisolid and liquid Murashige and Skoog medium supplemented with cytokinins N6-benzyladenine (BA 4.44 μM, furfurylaminopurine (KIN 4.65 μM and auxin α-naphthalene acetic acid (NAA 5.37 μM individually and in combinations. The regeneration response was influenced by both the type of growth regulator and physical state of the medium. The explants produced either shoot buds on cytokinincontaining media or protocorm-like bodies (PLBs on NAA containing media both solid and liquid. More neo-formations were produced on liquid media, especially those containing only NAA. They were formed at nodal and inter-nodal regions. The secondary buds were produced on the surface of primary PLBs. The plantlets were developed on MS medium containing banana homogenate 50 g·dm-3. The current study is the first ever report on successful regeneration of Dendrobium chrysotoxum from stem-node segments.

NUMERICAL SOLUTION OF SINGULAR INVERSE NODAL PROBLEM BY USING CHEBYSHEV POLYNOMIALS

OpenAIRE

NEAMATY, ABDOLALI; YILMAZ, EMRAH; AKBARPOOR, SHAHRBANOO; DABBAGHIAN, ABDOLHADI

2017-01-01

In this study, we consider Sturm-Liouville problem in two cases: the first case having no singularity and the second case having a singularity at zero. Then, we calculate the eigenvalues and the nodal points and present the uniqueness theorem for the solution of the inverse problem by using a dense subset of the nodal points in two given cases. Also, we use Chebyshev polynomials of the first kind for calculating the approximate solution of the inverse nodal problem in these cases. Finally, we...

Características morfológicas y posibles implicaciones clínicas de las arterias nodales Morphological characteristics and potential clinical implications of nodal arteries

Directory of Open Access Journals (Sweden)

Luis E Ballesteros

2010-12-01

Full Text Available La expresión morfológica de las arterias nodales es relevante en el diagnóstico y manejo de eventos clínicos y en abordajes quirúrgicos del corazón. Se estudiaron 88 arterias nodales de corazones obtenidos como material de autopsia. Las arterias coronarias se inyectaron con resina poliéster pigmentada de color rojo. Se registraron las formas de presentación de las arterias nodales y sus características morfométricas. La arteria del nodo sinoatrial se originó de la coronaria derecha en 52 casos (59,1%, de la circunfleja en 33 corazones (37,35% y de ambas en 3 (3,4%. Su calibre proximal fue de 1,31 mm (± 0,3, correspondiente a las arterias originadas de la coronaria derecha de 1,25 mm (± 0,3 mientras que las que se originaron de la arteria circunfleja obtuvieron un calibre de 1,42 mm (± 0,3, siendo esta diferencia significativa (p= 0,01. Se originó con mayor frecuencia en el tercio anteromedial, tanto de la coronaria derecha como de la circunfleja (54,6% y 61,2% respectivamente. En su segmento final cruzó por delante de la desembocadura de la vena cava superior en la mayoría de los casos (44%, mientras que en 22 corazones (24,5% cursó alrededor de la cava. Se observó arteria en forma de «S» en 14 casos (15,9% del total de la muestra y 42,4% de las originadas de la arteria circunfleja. La arteria del nodo atrioventricular se originó del segmento en «U» invertida de la coronaria derecha, al nivel de la cruz cardiaca, en 81 corazones (92%, y presentó un calibre proximal de 1,06 mm (± 0,22. Con relación al calibre y al origen se evidencian hallazgos que coinciden con estudios previos. Se destaca la alta prevalencia de la arteria en forma de «S» y de la trayectoria de la arteria sinoatrial alrededor de la vena cava superior.The morphological expression of nodal arteries is important in the diagnosis and management of cardiac clinical events and surgical approaches. 88 nodal arteries of hearts obtained from autopsies were

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.

The impact of nodal tumour burden on lymphoscintigraphic imaging in patients with melanomas

Energy Technology Data Exchange (ETDEWEB)

Kretschmer, Lutz; Bertsch, Hans Peter; Hellriegel, Simin; Thoms, Kai-Martin; Schoen, Michael Peter [Georg August University of Goettingen, Department of Dermatology, Venereology and Allergology, Goettingen (Germany); Bardzik, Pawel; Meller, Johannes; Sahlmann, Carsten Oliver [Georg-August-University of Goettingen, Department of Nuclear Medicine, Goettingen (Germany)

2014-10-15

To retrospectively study the influence of nodal tumour burden on lymphoscintigraphic imaging in 509 consecutive patients with melanomas. Bidirectional lymphatic drainage, the clear depiction of an afferent lymphatic vessel, time to depiction of the first sentinel lymph node (SLN) and number of depicted and excised nodes were recorded. Nodal tumour load was classified as SLN-negative, SLN micrometastases or macrometastases. In the overall population, using multivariate regression analysis, a short SLN depiction time was significantly associated with the depiction of a greater number of radioactive nodes, a short distance between the primary tumour site and the nodal basin, younger age and lower nodal tumour burden. The proportion of patients with clear depiction of an afferent lymphatic vessel depended on the nodal tumour load (46 % in SLN-negative patients, 57 % in SLN positive patients, and 69 % in patients with macrometastases; P = 0.009). Macrometastasis was significantly associated with delayed depiction of the first radioactive node and a greater number of depicted hotspots. In patients with clinically nonsuspicious nodes, i.e. the classical target group for SLN biopsy, clear depiction of an afferent vessel was significantly associated with a higher number of SLNs during dynamic acquisition, SLN micrometastasis and a higher overall number of metastatic lymph nodes after SLN biopsy plus completion lymphadenectomy. The excision of more than two SLNs did not increase the metastasis detection rate. In patients with bidirectional or tridirectional lymphatic drainage, the SLN positivity rates for the first, second and third basin were 25.4 %, 11.7 % and 0.0 %, respectively (P = 0.002). In patients with clinically nonsuspicious lymph nodes, clear depiction of an afferent lymph vessel may be a sign of micrometastasis. Macrometastasis is associated with prominent afferent vessels, delayed depiction of the first radioactive node and a higher number of depicted hotspots

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.

The LAW Library -- A multigroup cross-section library for use in radioactive waste analysis calculations

International Nuclear Information System (INIS)

Greene, N.M.; Arwood, J.W.; Wright, R.Q.; Parks, C.V.

1994-08-01

The 238-group LAW Library is a new multigroup neutron cross-section library based on ENDF/B-V data, with five sets of data taken from ENDF/B-VI ( 14 N 7 , 15 N 7 , 16 O 8 , 154Eu 63 , and 155 Eu 63 ). These five nuclides are included because the new evaluations are thought to be superior to those in Version 5. The LAW Library contains data for over 300 materials and will be distributed by the Radiation Shielding Information Center, located at Oak Ridge National Laboratory. It was generated for use in neutronics calculations required in radioactive waste analyses, although it has equal utility in any study requiring multigroup neutron cross sections

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

Finally! A valid test of configural invariance using permutation in multigroup CFA

NARCIS (Netherlands)

Jorgensen, T.D.; Kite, B.A.; Chen, P.-Y.; Short, S.D.; van der Ark, L.A.; Wiberg, M.; Culpepper, S.A.; Douglas, J.A.; Wang, W.-C.

2017-01-01

In multigroup factor analysis, configural measurement invariance is accepted as tenable when researchers either (a) fail to reject the null hypothesis of exact fit using a χ2 test or (b) conclude that a model fits approximately well enough, according to one or more alternative fit indices (AFIs).

Churchill regulates cell movement and mesoderm specification by repressing Nodal signaling

Directory of Open Access Journals (Sweden)

Mentzer Laura

2007-11-01

Full Text Available Abstract Background Cell movements are essential to the determination of cell fates during development. The zinc-finger transcription factor, Churchill (ChCh has been proposed to regulate cell fate by regulating cell movements during gastrulation in the chick. However, the mechanism of action of ChCh is not understood. Results We demonstrate that ChCh acts to repress the response to Nodal-related signals in zebrafish. When ChCh function is abrogated the expression of mesodermal markers is enhanced while ectodermal markers are expressed at decreased levels. In cell transplant assays, we observed that ChCh-deficient cells are more motile than wild-type cells. When placed in wild-type hosts, ChCh-deficient cells often leave the epiblast, migrate to the germ ring and are later found in mesodermal structures. We demonstrate that both movement of ChCh-compromised cells to the germ ring and acquisition of mesodermal character depend on the ability of the donor cells to respond to Nodal signals. Blocking Nodal signaling in the donor cells at the levels of Oep, Alk receptors or Fast1 inhibited migration to the germ ring and mesodermal fate change in the donor cells. We also detect additional unusual movements of transplanted ChCh-deficient cells which suggests that movement and acquisition of mesodermal character can be uncoupled. Finally, we demonstrate that ChCh is required to limit the transcriptional response to Nodal. Conclusion These data establish a broad role for ChCh in regulating both cell movement and Nodal signaling during early zebrafish development. We show that chch is required to limit mesodermal gene expression, inhibit Nodal-dependant movement of presumptive ectodermal cells and repress the transcriptional response to Nodal signaling. These findings reveal a dynamic role for chch in regulating cell movement and fate during early development.

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)

A new diffusion nodal method based on analytic basis function expansion

International Nuclear Information System (INIS)

Noh, J.M.; Cho, N.Z.

1993-01-01

The transverse integration procedure commonly used in most advanced nodal methods results in some limitations. The first is that the transverse leakage term that appears in the transverse integration procedure must be appropriately approximated. In most advanced nodal methods, this term is expanded in a quadratic polynomial. The second arises when reconstructing the pinwise flux distribution within a node. The available one-dimensional flux shapes from nodal calculation in each spatial direction cannot be used directly in the flux reconstruction. Finally, the transverse leakage defined for a hexagonal node becomes so complicated as not to be easily handled and contains nonphysical singular terms. In this paper, a new nodal method called the analytic function expansion nodal (AFEN) method is described for both the rectangular geometry and the hexagonal geometry in order to overcome these limitations. This method does not solve the transverse-integrated one-dimensional diffusion equations but instead solves directly the original multidimensional diffusion equation within a node. This is a accomplished by expanding the solution (or the intranodal homogeneous flux distribution) in terms of nonseparable analytic basis functions satisfying the diffusion equation at any point in the node

RTk/SN Solutions of the Two-Dimensional Multigroup Transport Equations in Hexagonal Geometry

International Nuclear Information System (INIS)

Valle, Edmundo del; Mund, Ernest H.

2004-01-01

This paper describes an extension to the hexagonal geometry of some weakly discontinuous nodal finite element schemes developed by Hennart and del Valle for the two-dimensional discrete ordinates transport equation in quadrangular geometry. The extension is carried out in a way similar to the extension to the hexagonal geometry of nodal element schemes for the diffusion equation using a composite mapping technique suggested by Hennart, Mund, and del Valle. The combination of the weakly discontinuous nodal transport scheme and the composite mapping is new and is detailed in the main section of the paper. The algorithm efficiency is shown numerically through some benchmark calculations on classical problems widely referred to in the literature

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

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.)

Error quantification of the axial nodal diffusion kernel of the DeCART code

International Nuclear Information System (INIS)

Cho, J. Y.; Kim, K. S.; Lee, C. C.

2006-01-01

This paper is to quantify the transport effects involved in the axial nodal diffusion kernel of the DeCART code. The transport effects are itemized into three effects, the homogenization, the diffusion, and the nodal effects. A five pin model consisting of four fuel pins and one non-fuel pin is demonstrated to quantify the transport effects. The transport effects are analyzed for three problems, the single pin (SP), guide tube (GT) and control rod (CR) problems by replacing the non-fuel pin with the fuel pin, a guide-tube and a control rod pins, respectively. The homogenization and diffusion effects are estimated to be about -4 and -50 pcm for the eigenvalue, and less than 2 % for the node power. The nodal effect on the eigenvalue is evaluated to be about -50 pcm in the SP and GT problems, and +350 pcm in the CR problem. Regarding the node power, this effect induces about a 3 % error in the SP and GT problems, and about a 20 % error in the CR problem. The large power error in the CR problem is due to the plane thickness, and it can be decreased by using the adaptive plane size. From the error quantification, it is concluded that the homogenization and the diffusion effects are not controllable if DeCART maintains the diffusion kernel for the axial solution, but the nodal effect is controllable by introducing the adaptive plane size scheme. (authors)

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

RGENDF - An interface program between the NJOY code and codes using multigroup cross-sections

International Nuclear Information System (INIS)

Chalhoub, E.S.; Anaf, J.

1988-02-01

An interface program for reformatting multigroup cross-section libraries generated by NJOY into ENDF/B-V format and the EXPANDA, PFCOND and COMPAR input formats is presented. (author). 7 refs, 1 fig., 1 tab

F-18 fluorodeoxyglucose positron emission tomography in the mediastinal nodal staging of non-small cell lung carcinoma

International Nuclear Information System (INIS)

Berlangieri, S.U.; Scott, A.M.; Knight, S.; Fitt, G.J.; Hess, E.M.; Pathmaraj, K.; Hennessy, O.F.; Tochon-Danguy, H.J.; Chan, J.G.; Egan, G.F.; Sinclair, R.A.; Clarke, C.P.; McKay, W.J.; St Vincents Hospital, Fitzroy, VIC

1998-01-01

Full text: Positron emission tomography (PET) using F-18 fluorodeoxyglucose (FDG), as a metabolic tumour marker, has been proposed for staging of oncological disease. To determine its role in the mediastinal staging of lung cancer, a prospective comparison of FDG PET with surgery was performed in patients with suspected non-small cell lung carcinoma. The analysis group consists of 70 patients, 49 men and 21 women, mean age 64 yrs (range 41-83 yrs). The PET study was acquired on a Siemens 951/31R scanner over 3 bed positions, 45 minutes following 400MBq FDG. The emission scan was attenuation corrected using measured transmission data. The FDG PET were interpreted by a nuclear physician blinded to the clinical data and the results of the patients' CT scan. On PET, nodes were graded qualitatively on a 5 point scale with scores 4 or greater, positive for tumour involvement. Surgical specimens were obtained in all patients by thoracotomy or mediastinoscopy. The PET metabolic studies and pathology were mapped according to the American Thoracic Society nodal classification resulting in a total of 277 nodal stations evaluated. The PET studies analysed N2 or N3 tumour involvement by nodal station in comparison to histology of pathological specimens or direct visual assessment of the nodal stations at surgery. All patients had proven non-small cell lung carcinoma, except two, in whom, a tissue confirmation of the suspected diagnosis was not attained. PET excluded tumour in 237 of 246 nodal stations (specificity 96%). PET correctly identified 23 of 31 nodal stations with disease (sensitivity 74%). PET correctly staged 260 of 277 nodal stations (accuracy 94%) for disease. FDG PET is an accurate non-invasive functional imaging modality for the mediastinal staging of non-small cell lung cancer and has an important clinical role in the preoperative staging of lung cancer patients

Hamiltonian circuited simulations in reactor physics

International Nuclear Information System (INIS)

Rio Hirowati Shariffudin

2002-01-01

In the assessment of suitability of reactor designs and in the investigations into reactor safety, the steady state of a nuclear reactor has to be studied carefully. The analysis can be done through mockup designs but this approach costs a lot of money and consumes a lot of time. A less expensive approach is via simulations where the reactor and its neutron interactions are modelled mathematically. Finite difference discretization of the diffusion operator has been used to approximate the steady state multigroup neutron diffusion equations. The steps include the outer scheme which estimates the resulting right hand side of the matrix equation, the group scheme which calculates the upscatter problem and the inner scheme which solves for the flux for a particular group. The Hamiltonian circuited simulations for the inner iterations of the said neutron diffusion equation enable the effective use of parallel computing, especially where the solutions of multigroup neutron diffusion equations involving two or more space dimensions are required. (Author)

Multigroup cross section library; WIMS library

International Nuclear Information System (INIS)

Kannan, Umasankari

2000-01-01

The WIMS library has been extensively used in thermal reactor calculations. This multigroup constants library was originally developed from the UKNDL in the late 60's and has been updated in 1986. This library has been distributed with the WIMS-D code by NEA data bank. The references to WIMS library in literature are the 'old' which is the original as developed by the AEA Winfrith and the 'new' which is the current 1986 WIMS library. IAEA has organised a CRP where a new and fully updated WIMS library will soon be available. This paper gives an overview of the definitions of the group constants that go into any basic nuclear data library used for reactor calculations. This paper also outlines the contents of the WIMS library and some of its shortcomings

LOLA SYSTEM: A code block for nodal PWR simulation. Part. II - MELON-3, CONCON and CONAXI Codes

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 MELON-3, CONCON and CONAXI codes, which are part of the core calculation system by nodal theory in one group, called LOLA SYSTEM. These auxiliary codes, provide some of the input data for the main module SIMULA-3; these are, the reactivity correlations constants, the albe does and the transport factors. (Author) 7 refs

LOLA SYSTEM: A code block for nodal PWR simulation. Part. II - MELON-3, CONCON and CONAXI Codes

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 MELON-3, CONCON and CONAXI codes, which are part of the core calculation system by nodal theory in one group, called LOLA SYSTEM. These auxiliary codes, provide some of the input data for the main module SIMULA-3; these are, the reactivity correlations constants, the albe does and the transport factors. (Author) 7 refs.

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

Routine use of standard breast MRI compared to axillary ultrasound for differentiating between no, limited and advanced axillary nodal disease in newly diagnosed breast cancer patients

Energy Technology Data Exchange (ETDEWEB)

Nijnatten, T.J.A. van, E-mail: Thiemovn@gmail.com [Department of Radiology and Nuclear Medicine, Maastricht University Medical Center+, Maastricht (Netherlands); Department of Surgery, Maastricht University Medical Center+, Maastricht (Netherlands); GROW – School for Oncology and Developmental Biology, Maastricht University Medical Center+, Maastricht (Netherlands); Ploumen, E.H. [Department of Radiology and Nuclear Medicine, Maastricht University Medical Center+, Maastricht (Netherlands); Department of Surgery, Maastricht University Medical Center+, Maastricht (Netherlands); Schipper, RJ. [Department of Radiology and Nuclear Medicine, Maastricht University Medical Center+, Maastricht (Netherlands); Department of Surgery, Maastricht University Medical Center+, Maastricht (Netherlands); GROW – School for Oncology and Developmental Biology, Maastricht University Medical Center+, Maastricht (Netherlands); Department of Surgery, Catharina Hospital, Eindhoven (Netherlands); Goorts, B. [Department of Radiology and Nuclear Medicine, Maastricht University Medical Center+, Maastricht (Netherlands); Department of Surgery, Maastricht University Medical Center+, Maastricht (Netherlands); GROW – School for Oncology and Developmental Biology, Maastricht University Medical Center+, Maastricht (Netherlands); Andriessen, E.H. [Department of Radiology and Nuclear Medicine, Maastricht University Medical Center+, Maastricht (Netherlands); Department of Surgery, Maastricht University Medical Center+, Maastricht (Netherlands); Vanwetswinkel, S.; Schavemaker, M. [Department of Radiology and Nuclear Medicine, Maastricht University Medical Center+, Maastricht (Netherlands); Nelemans, P. [Department of Epidemiology, Maastricht University Medical Center+, Maastricht (Netherlands); Vries, B. de [Department of Pathology, Zuyderland Hospital, Heerlen (Netherlands); and others

2016-12-15

Objectives: To compare standard breast MRI to dedicated axillary ultrasound (with or without tissue sampling) for differentiating between no, limited and advanced axillary nodal disease in breast cancer patients. Methods: All patients who underwent breast MRI and dedicated axillary ultrasound between 2009 and 2014 were eligible. Exclusion criteria were recurrent disease, neoadjuvant systemic therapy and not receiving completion axillary lymph node dissection after positive sentinel lymph node biopsy (SLNB). Two radiologists independently reassessed all MRI exams. Axillary ultrasound findings were retrospectively collected. Probability of advanced axillary nodal disease (pN2-3) given clinically node negative (cN0) or limited (cN1) findings was calculated, with corresponding negative predictive value (NPV) to exclude pN2-3 and positive predictive value (PPV) to identify axillary nodal disease. Histopathology served as gold standard. Results: A total of 377 cases resulted in 81.4% no, 14.4% limited and 4.2% advanced axillary nodal disease at final histopathology. Probability of pN2-3 given cN0 for breast MRI and axillary ultrasound was 0.7–0.9% versus 1.5% and probability of pN2-3 given cN1 was 11.6–15.4% versus 29.0%. When cN1 on breast MRI was observed, PPV to identify positive axillary nodal disease was 50.7% and 59.0%. Conclusions: Evaluation of axillary nodal status on standard breast MRI is comparable to dedicated axillary ultrasound in breast cancer patients. In patients who underwent preoperative standard breast MRI, axillary ultrasound is only required in case of suspicious nodal findings on MRI.

Variational P1 approximations of general-geometry multigroup transport problems

International Nuclear Information System (INIS)

Rulko, R.P.; Tomasevic, D.; Larsen, E.W.

1995-01-01

A variational approximation is developed for general-geometry multigroup transport problems with arbitrary anisotropic scattering. The variational principle is based on a functional that approximates a reaction rate in a subdomain of the system. In principle, approximations that result from this functional ''optimally'' determine such reaction rates. The functional contains an arbitrary parameter α and requires the approximate solutions of a forward and an adjoint transport problem. If the basis functions for the forward and adjoint solutions are chosen to be linear functions of the angular variable Ω, the functional yields the familiar multigroup P 1 equations for all values of α. However, the boundary conditions that result from the functional depend on α. In particular, for problems with vacuum boundaries, one obtains the conventional mixed boundary condition, but with an extrapolation distance that depends continuously on α. The choice α = 0 yields a generalization of boundary conditions derived earlier by Federighi and Pomraning for a more limited class of problems. The choice α = 1 yields a generalization of boundary conditions derived previously by Davis for monoenergetic problems. Other boundary conditions are obtained by choosing different values of α. The authors discuss this indeterminancy of α in conjunction with numerical experiments

Contribution to the solution of the multigroup Boltzmann equation by the determinist methods and the Monte Carlo method; Contribution a la resolution de l`equation de Bolztmann en multigroupe par les methodes deterministes et Monte-Carlo

Energy Technology Data Exchange (ETDEWEB)

Li, M

1998-08-01

In this thesis, two methods for solving the multigroup Boltzmann equation have been studied: the interface-current method and the Monte Carlo method. A new version of interface-current (IC) method has been develop in the TDT code at SERMA, where the currents of interface are represented by piecewise constant functions in the solid angle space. The convergence of this method to the collision probability (CP) method has been tested. Since the tracking technique is used for both the IC and CP methods, it is necessary to normalize he collision probabilities obtained by this technique. Several methods for this object have been studied and implemented in our code, we have compared their performances and chosen the best one as the standard choice. The transfer matrix treatment has been a long-standing difficulty for the multigroup Monte Carlo method: when the cross-sections are converted into multigroup form, important negative parts will appear in the angular transfer laws represented by low-order Legendre polynomials. Several methods based on the preservation of the first moments, such as the discrete angles methods and the equally-probable step function method, have been studied and implemented in the TRIMARAN-II code. Since none of these codes has been satisfactory, a new method, the non equally-probably step function method, has been proposed and realized in our code. The comparisons for these methods have been done in several aspects: the preservation of the moments required, the calculation of a criticality problem and the calculation of a neutron-transfer in water problem. The results have showed that the new method is the best one in all these comparisons, and we have proposed that it should be a standard choice for the multigroup transfer matrix. (author) 76 refs.

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 ...

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

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)

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)

TGF-β promotes glioma cell growth via activating Nodal expression through Smad and ERK1/2 pathways

International Nuclear Information System (INIS)

Sun, Jing; Liu, Su-zhi; Lin, Yan; Cao, Xiao-pan; Liu, Jia-ming

2014-01-01

Highlights: •TGF-β promoted Nodal expression in glioma cells. •TGF-β promoted Nodal expression via activating Smad and ERK1/2 pathways. •TGF-β promotes glioma cell growth via activating Nodal expression. -- Abstract: While there were certain studies focusing on the mechanism of TGF-β promoting the growth of glioma cells, the present work revealed another novel mechanism that TGF-β may promote glioma cell growth via enhancing Nodal expression. Our results showed that Nodal expression was significantly upregulated in glioma cells when TGF-β was added, whereas the TGF-β-induced Nodal expression was evidently inhibited by transfection Smad2 or Smad3 siRNAs, and the suppression was especially significant when the Smad3 was downregulated. Another, the attenuation of TGF-β-induced Nodal expression was observed with blockade of the ERK1/2 pathway also. Further detection of the proliferation, apoptosis, and invasion of glioma cells indicated that Nodal overexpression promoted the proliferation and invasion of tumor cells and inhibited their apoptosis, resembling the effect of TGF-β addition. Downregulation of Nodal expression via transfection Nodal-specific siRNA in the presence of TGF-β weakened the promoting effect of the latter on glioma cells growth, and transfecting Nodal siRNA alone in the absence of exogenous TGF-β more profoundly inhibited the growth of glioma cells. These results demonstrated that while both TGF-β and Nodal promoted glioma cells growth, the former might exert such effect by enhancing Nodal expression, which may form a new target for glioma therapy

Topological crystalline superconductivity and second-order topological superconductivity in nodal-loop materials

Science.gov (United States)

Shapourian, Hassan; Wang, Yuxuan; Ryu, Shinsei

2018-03-01

We study the intrinsic fully gapped odd-parity superconducting order in doped nodal-loop materials with a torus-shaped Fermi surface. We show that the mirror symmetry, which protects the nodal loop in the normal state, also protects the superconducting state as a topological crystalline superconductor. As a result, the surfaces preserving the mirror symmetry host gapless Majorana cones. Moreover, for a Weyl-loop system (twofold degenerate at the nodal loop), the surfaces that break the mirror symmetry (those parallel to the bulk nodal loop) contribute a Chern (winding) number to the quasi-two-dimensional system in a slab geometry, which leads to a quantized thermal Hall effect and a single Majorana zero mode bound at a vortex line penetrating the system. This Chern number can be viewed as a higher-order topological invariant, which supports hinge modes in a cubic sample when mirror symmetry is broken. For a Dirac-loop system (fourfold degenerate at the nodal loop), the fully gapped odd-parity state can be either time-reversal symmetry-breaking or symmetric, similar to the A and B phases of 3He. In a slab geometry, the A phase has a Chern number two, while the B phase carries a nontrivial Z2 invariant. We discuss the experimental relevance of our results to nodal-loop materials such as CaAgAs.

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.

Status of multigroup cross-section data for shielding applications

International Nuclear Information System (INIS)

Roussin, R.W.; Maskewitz, B.F.; Trubey, D.K.

1983-01-01

Multigroup cross-section libraries for shielding applications in formats for direct use in discrete ordinates or Monte Carlo codes have long been a part of the Data Library Collection (DLC) of the Radiation Shielding Information Center (RSIC). In recent years libraries in more flexible and comprehensive formats, which allow the user to derive his own problem-dependent sets, have been added to the collection. The current status of both types is described, as well as projections for adding data libraries based on ENDF/B-V

Nodal-dependent mesendoderm specification requires the combinatorial activities of FoxH1 and Eomesodermin.

Directory of Open Access Journals (Sweden)

Christopher E Slagle

2011-05-01

Full Text Available Vertebrate mesendoderm specification requires the Nodal signaling pathway and its transcriptional effector FoxH1. However, loss of FoxH1 in several species does not reliably cause the full range of loss-of-Nodal phenotypes, indicating that Nodal signals through additional transcription factors during early development. We investigated the FoxH1-dependent and -independent roles of Nodal signaling during mesendoderm patterning using a novel recessive zebrafish FoxH1 mutation called midway, which produces a C-terminally truncated FoxH1 protein lacking the Smad-interaction domain but retaining DNA-binding capability. Using a combination of gel shift assays, Nodal overexpression experiments, and genetic epistasis analyses, we demonstrate that midway more accurately represents a complete loss of FoxH1-dependent Nodal signaling than the existing zebrafish FoxH1 mutant schmalspur. Maternal-zygotic midway mutants lack notochords, in agreement with FoxH1 loss in other organisms, but retain near wild-type expression of markers of endoderm and various nonaxial mesoderm fates, including paraxial and intermediate mesoderm and blood precursors. We found that the activity of the T-box transcription factor Eomesodermin accounts for specification of these tissues in midway embryos. Inhibition of Eomesodermin in midway mutants severely reduces the specification of these tissues and effectively phenocopies the defects seen upon complete loss of Nodal signaling. Our results indicate that the specific combinations of transcription factors available for signal transduction play critical and separable roles in determining Nodal pathway output during mesendoderm patterning. Our findings also offer novel insights into the co-evolution of the Nodal signaling pathway, the notochord specification program, and the chordate branch of the deuterostome family of animals.

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

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.

Global dynamics of multi-group SEI animal disease models with indirect transmission

International Nuclear Information System (INIS)

Wang, Yi; Cao, Jinde

2014-01-01

A challenge to multi-group epidemic models in mathematical epidemiology is the exploration of global dynamics. Here we formulate multi-group SEI animal disease models with indirect transmission via contaminated water. Under biologically motivated assumptions, the basic reproduction number R 0 is derived and established as a sharp threshold that completely determines the global dynamics of the system. In particular, we prove that if R 0 <1, the disease-free equilibrium is globally asymptotically stable, and the disease dies out; whereas if R 0 >1, then the endemic equilibrium is globally asymptotically stable and thus unique, and the disease persists in all groups. Since the weight matrix for weighted digraphs may be reducible, the afore-mentioned approach is not directly applicable to our model. For the proofs we utilize the classical method of Lyapunov, graph-theoretic results developed recently and a new combinatorial identity. Since the multiple transmission pathways may correspond to the real world, the obtained results are of biological significance and possible generalizations of the model are also discussed

Three-dimensional conformal radiation may deliver considerable dose of incidental nodal irradiation in patients with early stage node-negative non-small cell lung cancer when the tumor is large and centrally located

International Nuclear Information System (INIS)

Zhao Lujun; Chen Ming; Haken, Randall ten; Chetty, Indrin; Chapet, Olivier; Hayman, James A.; Kong Fengming

2007-01-01

Background and purpose: To determine the dose to regional nodal stations in patients with T 1-3 N 0 M 0 non-small cell lung cancer (NSCLC) treated with three-dimensional conformal radiation therapy (3DCRT) without intentional elective nodal irradiation (ENI). Materials and methods: Twenty-three patients with medically inoperable T 1-3 N 0 M 0 NSCLC were treated with 3DCRT without ENI. Hilar and mediastinal nodal regions were contoured on planning CT. The prescription dose was normalized to 70 Gy. Equivalent uniform dose (EUD) and other dosimetric parameters (e.g., V 40 ) were calculated for each nodal station. Results: The median EUD for the whole group ranged from 0.4 to 4.4 Gy for all elective nodal regions. Gross tumor volume (GTV) and the relationship between GTV and hilum were significantly correlated with irradiation dose to ipsilateral hilar nodal regions (P 3 (diameter ∼ 4 cm) and or having any overlap with hilum, the median EUDs were 9.6, 22.6, and 62.9 Gy for ipsilateral lower paratracheal, subcarinal, and ipsilateral hilar regions, respectively. The corresponding median V 40 were 32.5%, 39.3%, and 97.6%, respectively. Conclusions: Although incidental nodal irradiation dose is low in the whole group, the dose to high-risk nodal regions is considerable in patients with T 1-3 N 0 NSCLC when the primary is large and/or centrally located

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)

Power transmission charges based on nodal pricing which considers restriction on power transmission; Soden setsuyaku wo koryoshita nodaru pricing ni motozuku soden ryokin

Energy Technology Data Exchange (ETDEWEB)

Okada, K.; Asano, H. [Central Research Institute of Electric Power Industry, Tokyo (Japan); Matsukawa, I. [Musashi University, Tokyo (Japan)

1997-01-30

Power transmission charges were derived by using nodal pricing, and a discussion was given on what effects are given on system conditions, nodal price and consignment charge by how coordination points of independent power producers (IPP) and power demand are handled. A test model having six nodes (busbars) and eleven branches (transmission lines) was used. Since demands of the same kind are hypothesized to be coordinated in this simulation, the total nodal price becomes an equivalent value if there is no restrictions in transmission line current. If the transmission restrictions are taken into consideration, demand amounts at each node are so adjusted that excess current in a transmission line exceeding the transmission capacity will be eliminated. Thus, the demand-supply balancing amount in the entire system becomes smaller than when restrictions are not considered. As a result of the analysis, the IPP coordination points have possibilities to cause congestion (overload current) in the system, raise nodal price at each point, and sharply raise the consignment charge. It was found that an effect may also occur to a node depending on position of demand generation. 6 refs., 3 figs., 7 tabs.

Application of the SPH method in nodal diffusion analyses of SFR cores

Energy Technology Data Exchange (ETDEWEB)

Nikitin, Evgeny; Fridman, Emil [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Div. Reactor Safety; Mikityuk, K. [Paul Scherrer Institut, Villigen (Switzerland)

2016-07-01

The current study investigated the potential of the SPH method, applied to correct the few-group XS produced by Serpent, to further improve the accuracy of the nodal diffusion solutions. The procedure for the generation of SPH-corrected few-group XS is presented in the paper. The performance of the SPH method was tested on a large oxide SFR core from the OECD/NEA SFR benchmark. The reference SFR core was modeled with the DYN3D and PARCS nodal diffusion codes using the SPH-corrected few-group XS generated by Serpent. The nodal diffusion results obtained with and without SPH correction were compared to the reference full-core Serpent MC solution. It was demonstrated that the application of the SPH method improves the accuracy of the nodal diffusion solutions, particularly for the rodded core state.

A spectrum correction method for fuel assembly rehomogenization

International Nuclear Information System (INIS)

Lee, Kyung Taek; Cho, Nam Zin

2004-01-01

To overcome the limitation of existing homogenization methods based on the single assembly calculation with zero current boundary condition, we propose a new rehomogenization method, named spectrum correction method (SCM), consisting of the multigroup energy spectrum approximation by spectrum correction and the condensed two-group heterogeneous single assembly calculations with non-zero current boundary condition. In SCM, the spectrum shifting phenomena caused by current across assembly interfaces are considered by the spectrum correction at group condensation stage at first. Then, heterogeneous single assembly calculations with two-group cross sections condensed by using corrected multigroup energy spectrum are performed to obtain rehomogenized nodal diffusion parameters, i.e., assembly-wise homogenized cross sections and discontinuity factors. To evaluate the performance of SCM, it was applied to the analytic function expansion nodal (AFEN) method and several test problems were solved. The results show that SCM can reduce the errors significantly both in multiplication factors and assembly averaged power distributions

Second order time evolution of the multigroup diffusion and P1 equations for radiation transport

International Nuclear Information System (INIS)

Olson, Gordon L.

2011-01-01

Highlights: → An existing multigroup transport algorithm is extended to be second-order in time. → A new algorithm is presented that does not require a grey acceleration solution. → The two algorithms are tested with 2D, multi-material problems. → The two algorithms have comparable computational requirements. - Abstract: An existing solution method for solving the multigroup radiation equations, linear multifrequency-grey acceleration, is here extended to be second order in time. This method works for simple diffusion and for flux-limited diffusion, with or without material conduction. A new method is developed that does not require the solution of an averaged grey transport equation. It is effective solving both the diffusion and P 1 forms of the transport equation. Two dimensional, multi-material test problems are used to compare the solution methods.

Analysis of nodal coverage utilizing image guided radiation therapy for primary gynecologic tumor volumes

Energy Technology Data Exchange (ETDEWEB)

Ahmed, Faisal [University of Utah School of Medicine, Salt Lake City, UT (United States); Loma Linda University Medical Center, Department of Radiation Oncology, Loma Linda, CA (United States); Sarkar, Vikren; Gaffney, David K.; Salter, Bill [Department of Radiation Oncology, University of Utah, Salt Lake City, UT (United States); Poppe, Matthew M., E-mail: matthew.poppe@hci.utah.edu [Department of Radiation Oncology, University of Utah, Salt Lake City, UT (United States)

2016-10-01

Purpose: To evaluate radiation dose delivered to pelvic lymph nodes, if daily Image Guided Radiation Therapy (IGRT) was implemented with treatment shifts based on the primary site (primary clinical target volume [CTV]). Our secondary goal was to compare dosimetric coverage with patient outcomes. Materials and methods: A total of 10 female patients with gynecologic malignancies were evaluated retrospectively after completion of definitive intensity-modulated radiation therapy (IMRT) to their pelvic lymph nodes and primary tumor site. IGRT consisted of daily kilovoltage computed tomography (CT)-on-rails imaging fused with initial planning scans for position verification. The initial plan was created using Varian's Eclipse treatment planning software. Patients were treated with a median radiation dose of 45 Gy (range: 37.5 to 50 Gy) to the primary volume and 45 Gy (range: 45 to 64.8 Gy) to nodal structures. One IGRT scan per week was randomly selected from each patient's treatment course and re-planned on the Eclipse treatment planning station. CTVs were recreated by fusion on the IGRT image series, and the patient's treatment plan was applied to the new image set to calculate delivered dose. We evaluated the minimum, maximum, and 95% dose coverage for primary and nodal structures. Reconstructed primary tumor volumes were recreated within 4.7% of initial planning volume (0.9% to 8.6%), and reconstructed nodal volumes were recreated to within 2.9% of initial planning volume (0.01% to 5.5%). Results: Dosimetric parameters averaged less than 10% (range: 1% to 9%) of the original planned dose (45 Gy) for primary and nodal volumes on all patients (n = 10). For all patients, ≥99.3% of the primary tumor volume received ≥ 95% the prescribed dose (V95%) and the average minimum dose was 96.1% of the prescribed dose. In evaluating nodal CTV coverage, ≥ 99.8% of the volume received ≥ 95% the prescribed dose and the average minimum dose was 93%. In

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.

HELIOS/DRAGON/NESTLE codes' simulation of the Gentilly-2 loss of class 4 power event

International Nuclear Information System (INIS)

Sarsour, H.N.; Turinsky, P.J.; Rahnema, F.; Mosher, S.; Serghiuta, D.; Marleau, G.; Courau, T.

2002-01-01

[4] based upon HELIOS generated cross sections. Core thermal-hydraulic conditions, zone controller levels, and control device positions as a function of time are all taken from the CERBERUS/SOPHT-G2 simulation of this event. To address sensitivity to spatial discretization treatment, NESTLE core simulator predictions based upon the finite difference method (FDM), nodal expansion method (NM) without utilizing assembly discontinuity factors (ADF), and NM with utilizing ADF will be contrasted. Note that NESTLE Version 5 has the option to solve the two-group neutron diffusion equation via the nodal expansion method, employing a quartic polynomial flux expansion and quadratic transverse leakage representation in solving the 1-D transverse integrated diffusion equation. (author)

MC2-2: a code to calculate fast neutron spectra and multigroup cross sections

International Nuclear Information System (INIS)

Henryson, H. II; Toppel, B.J.; Stenberg, C.G.

1976-06-01

MC 2 -2 is a program to solve the neutron slowing down problem using basic neutron data derived from the ENDF/B data files. The spectrum calculated by MC 2 -2 is used to collapse the basic data to multigroup cross sections for use in standard reactor neutronics codes. Four different slowing down formulations are used by MC 2 -2: multigroup, continuous slowing down using the Goertzel-Greuling or Improved Goertzel-Greuling moderating parameters, and a hyper-fine-group integral transport calculation. Resolved and unresolved resonance cross sections are calculated accounting for self-shielding, broadening and overlap effects. This document provides a description of the MC 2 -2 program. The physics and mathematics of the neutron slowing down problem are derived and detailed information is provided to aid the MC 2 -2 user in preparing input for the program and implementation of the program on IBM 370 or CDC 7600 computers

Young Adults’ Attitude Towards Advertising: a multi-group analysis by ethnicity

OpenAIRE

Hiram Ting; Ernest Cyril de Run; Ramayah Thurasamy

2015-01-01

Objective – This study aims to investigate the attitude of Malaysian young adults towards advertising. How this segment responds to advertising, and how ethnic/cultural differences moderate are assessed.Design/methodology/approach – A quantitative questionnaire is used to collect data at two universities. Purposive sampling technique is adopted to ensure the sample represents the actual population. Structural equation modelling (SEM) and multi-group analysis (MGA) are utilized in analysis.Fin...

MicroRNA expression in nodal and extranodal Diffuse Large B-cell Lymphoma

DEFF Research Database (Denmark)

Mandrup, Charlotte; Petersen, Anders; Højfeldt, Anne Dirks

MicroRNA expression in nodal and extranodal Diffuse Large B-cell Lymphoma   C. Mandrup1, A. Petersen1, A. D. Hoejfeldt1, H. F. Thomsen1, J. Madsen1, J. Dahlgaard1, P. Johansen2, A. Bukh1, K. Dybkaer1 and H. E Johnsen1. 1Department of Hematology, 2Pathological Institute, Aalborg Hospital, Aarhus...... University Hospital, Aalborg, Denmark Introduction: The aim of this project was to analyse microRNA (miRNA) expression in nodal and extranodal diffuse large B-cell lymphoma (DLBCL). Manifestation at diagnosis may be nodal and/or extranodal. At present, there are no known determinants for none...... of the manifestations, and no way to predict the potential progression from nodal to extranodal disease. miRNA are small regulatory RNA molecules with core function to repress/cleave sequence complementary mRNA targets. Abnormalities in miRNA genetics and expression are known to affect initiation and development...

Mining the multigroup-discrete ordinates algorithm for high quality solutions

International Nuclear Information System (INIS)

Ganapol, B.D.; Kornreich, D.E.

2005-01-01

A novel approach to the numerical solution of the neutron transport equation via the discrete ordinates (SN) method is presented. The new technique is referred to as 'mining' low order (SN) numerical solutions to obtain high order accuracy. The new numerical method, called the Multigroup Converged SN (MGCSN) algorithm, is a combination of several sequence accelerators: Romberg and Wynn-epsilon. The extreme accuracy obtained by the method is demonstrated through self consistency and comparison to the independent semi-analytical benchmark BLUE. (authors)

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

A geometrically exact beam element based on the absolute nodal coordinate formulation

International Nuclear Information System (INIS)

Gerstmayr, Johannes; Matikainen, Marko K.; Mikkola, Aki M.

2008-01-01

In this study, Reissner's classical nonlinear rod formulation, as implemented by Simo and Vu-Quoc by means of the large rotation vector approach, is implemented into the framework of the absolute nodal coordinate formulation. The implementation is accomplished in the planar case accounting for coupled axial, bending, and shear deformation. By employing the virtual work of elastic forces similarly to Simo and Vu-Quoc in the absolute nodal coordinate formulation, the numerical results of the formulation are identical to those of the large rotation vector formulation. It is noteworthy, however, that the material definition in the absolute nodal coordinate formulation can differ from the material definition used in Reissner's beam formulation. Based on an analytical eigenvalue analysis, it turns out that the high frequencies of cross section deformation modes in the absolute nodal coordinate formulation are only slightly higher than frequencies of common shear modes, which are present in the classical large rotation vector formulation of Simo and Vu-Quoc, as well. Thus, previous claims that the absolute nodal coordinate formulation is inefficient or would lead to ill-conditioned finite element matrices, as compared to classical approaches, could be refuted. In the introduced beam element, locking is prevented by means of reduced integration of certain parts of the elastic forces. Several classical large deformation static and dynamic examples as well as an eigenvalue analysis document the equivalence of classical nonlinear rod theories and the absolute nodal coordinate formulation for the case of appropriate material definitions. The results also agree highly with those computed in commercial finite element codes

Amplification and protein overexpression of cyclin D1: Predictor of occult nodal metastasis in early oral cancer.

Science.gov (United States)

Noorlag, Rob; Boeve, Koos; Witjes, Max J H; Koole, Ronald; Peeters, Ton L M; Schuuring, Ed; Willems, Stefan M; van Es, Robert J J

2017-02-01

Accurate nodal staging is pivotal for treatment planning in early (stage I-II) oral cancer. Unfortunately, current imaging modalities lack sensitivity to detect occult nodal metastases. Chromosomal region 11q13, including genes CCND1, Fas-associated death domain (FADD), and CTTN, is often amplified in oral cancer with nodal metastases. However, evidence in predicting occult nodal metastases is limited. In 158 patients with early tongue and floor of mouth (FOM) squamous cell carcinomas, both CCND1 amplification and cyclin D1, FADD, and cortactin protein expression were correlated with occult nodal metastases. CCND1 amplification and cyclin D1 expression correlated with occult nodal metastases. Cyclin D1 expression was validated in an independent multicenter cohort, confirming the correlation with occult nodal metastases in early FOM cancers. Cyclin D1 is a predictive biomarker for occult nodal metastases in early FOM cancers. Prospective research on biopsy material should confirm these results before implementing its use in routine clinical practice. © 2016 Wiley Periodicals, Inc. Head Neck 39: 326-333, 2017. © 2016 Wiley Periodicals, Inc.

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.

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)

Recognizing nodal marginal zone lymphoma: recent advances and pitfalls. A systematic review

Science.gov (United States)

van den Brand, Michiel; van Krieken, J. Han J.M.

2013-01-01

The diagnosis of nodal marginal zone lymphoma is one of the remaining problem areas in hematopathology. Because no established positive markers exist for this lymphoma, it is frequently a diagnosis of exclusion, making distinction from other low-grade B-cell lymphomas difficult or even impossible. This systematic review summarizes and discusses the current knowledge on nodal marginal zone lymphoma, including clinical features, epidemiology and etiology, histology, and cytogenetic and molecular features. In particular, recent advances in diagnostics and pathogenesis are discussed. New immunohistochemical markers have become available that could be used as positive markers for nodal marginal zone lymphoma. These markers could be used to ensure more homogeneous study groups in future research. Also, recent gene expression studies and studies describing specific gene mutations have provided clues to the pathogenesis of nodal marginal zone lymphoma, suggesting deregulation of the nuclear factor kappa B pathway. Nevertheless, nodal marginal zone lymphoma remains an enigmatic entity, requiring further study to define its pathogenesis to allow an accurate diagnosis and tailored treatment. However, recent data indicate that it is not related to splenic or extranodal lymphoma, and that it is also not related to lymphoplasmacytic lymphoma. Thus, even though the diagnosis is not always easy, it is clearly a separate entity. PMID:23813646

Improved stiffness confinement method within the coarse mesh finite difference framework for efficient spatial kinetics calculation

International Nuclear Information System (INIS)

Park, Beom Woo; Joo, Han Gyu

2015-01-01

Highlights: • The stiffness confinement method is combined with multigroup CMFD with SENM nodal kernel. • The systematic methods for determining the shape and amplitude frequencies are established. • Eigenvalue problems instead of fixed source problems are solved in the transient calculation. • It is demonstrated that much larger time step sizes can be used with the SCM–CMFD method. - Abstract: An improved Stiffness Confinement Method (SCM) is formulated within the framework of the coarse mesh finite difference (CMFD) formulation for efficient multigroup spatial kinetics calculation. The algorithm for searching for the amplitude frequency that makes the dynamic eigenvalue unity is developed in a systematic way along with the methods for determining the shape and precursor frequencies. A nodal calculation scheme is established within the CMFD framework to incorporate the cross section changes due to thermal feedback and dynamic frequency update. The conditional nodal update scheme is employed such that the transient calculation is performed mostly with the CMFD formulation and the CMFD parameters are conditionally updated by intermittent nodal calculations. A quadratic representation of amplitude frequency is introduced as another improvement. The performance of the improved SCM within the CMFD framework is assessed by comparing the solution accuracy and computing times for the NEACRP control rod ejection benchmark problems with those obtained with the Crank–Nicholson method with exponential transform (CNET). It is demonstrated that the improved SCM is beneficial for large time step size calculations with stability and accuracy enhancement

Nodal involvement in Hodgkin disease and non-Hodgkin lymphoma assessed by magnetic resonance

International Nuclear Information System (INIS)

Tesoro Tess, J.D.; Balzarini, L.; Ceglia, E.; Petrillo, R.; Musumeci, R.

1990-01-01

Magnetic Resonance Imaging (MRI) demonstrates a good capability in distinguishing nodal involvement in hodgkin disease and nonhodgkin lymphoma both in the chest and in the retroperitoneal areas the initial presentation of the disease. However CT and lymphangiography demonstrated comparable or superior values of accuracy and sensitivity. (H.W.) 4 refs.; 2 tabs

Parallel computation of multigroup reactivity coefficient using iterative method

Science.gov (United States)

Susmikanti, Mike; Dewayatna, Winter

2013-09-01

One of the research activities to support the commercial radioisotope production program is a safety research target irradiation FPM (Fission Product Molybdenum). FPM targets form a tube made of stainless steel in which the nuclear degrees of superimposed high-enriched uranium. FPM irradiation tube is intended to obtain fission. The fission material widely used in the form of kits in the world of nuclear medicine. Irradiation FPM tube reactor core would interfere with performance. One of the disorders comes from changes in flux or reactivity. It is necessary to study a method for calculating safety terrace ongoing configuration changes during the life of the reactor, making the code faster became an absolute necessity. Neutron safety margin for the research reactor can be reused without modification to the calculation of the reactivity of the reactor, so that is an advantage of using perturbation method. The criticality and flux in multigroup diffusion model was calculate at various irradiation positions in some uranium content. This model has a complex computation. Several parallel algorithms with iterative method have been developed for the sparse and big matrix solution. The Black-Red Gauss Seidel Iteration and the power iteration parallel method can be used to solve multigroup diffusion equation system and calculated the criticality and reactivity coeficient. This research was developed code for reactivity calculation which used one of safety analysis with parallel processing. It can be done more quickly and efficiently by utilizing the parallel processing in the multicore computer. This code was applied for the safety limits calculation of irradiated targets FPM with increment Uranium.

Extra-nodal extension is a significant prognostic factor in lymph node positive breast cancer.

Directory of Open Access Journals (Sweden)

Sura Aziz

Full Text Available Presence of lymph node (LN metastasis is a strong prognostic factor in breast cancer, whereas the importance of extra-nodal extension and other nodal tumor features have not yet been fully recognized. Here, we examined microscopic features of lymph node metastases and their prognostic value in a population-based cohort of node positive breast cancer (n = 218, as part of the prospective Norwegian Breast Cancer Screening Program NBCSP (1996-2009. Sections were reviewed for the largest metastatic tumor diameter (TD-MET, nodal afferent and efferent vascular invasion (AVI and EVI, extra-nodal extension (ENE, number of ENE foci, as well as circumferential (CD-ENE and perpendicular (PD-ENE diameter of extra-nodal growth. Number of positive lymph nodes, EVI, and PD-ENE were significantly increased with larger primary tumor (PT diameter. Univariate survival analysis showed that several features of nodal metastases were associated with disease-free (DFS or breast cancer specific survival (BCSS. Multivariate analysis demonstrated an independent prognostic value of PD-ENE (with 3 mm as cut-off value in predicting DFS and BCSS, along with number of positive nodes and histologic grade of the primary tumor (for DFS: P = 0.01, P = 0.02, P = 0.01, respectively; for BCSS: P = 0.02, P = 0.008, P = 0.02, respectively. To conclude, the extent of ENE by its perpendicular diameter was independently prognostic and should be considered in line with nodal tumor burden in treatment decisions of node positive breast cancer.

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

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.

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

Application of the RT-0 nodal methods and NRMPO matrix-response to the cycles 1 and 2 of the LVC; Aplicacion de los metodos nodal RT-0 y matriz respuesta NRMPO a los ciclos 1 y 2 de la CLV

Energy Technology Data Exchange (ETDEWEB)

Delfin L, A.; Hernandez L, H.; Alonso V, G. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

2005-07-01

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)

Conception and development of an adaptive energy mesher for multigroup library generation of the transport codes

International Nuclear Information System (INIS)

Mosca, P.

2009-12-01

The deterministic transport codes solve the stationary Boltzmann equation in a discretized energy formalism called multigroup. The transformation of continuous data in a multigroup form is obtained by averaging the highly variable cross sections of the resonant isotopes with the solution of the self-shielding models and the remaining ones with the coarse energy spectrum of the reactor type. So far the error of such an approach could only be evaluated retrospectively. To remedy this, we studied in this thesis a set of methods to control a priori the accuracy and the cost of the multigroup transport computation. The energy mesh optimisation is achieved using a two step process: the creation of a reference mesh and its optimized condensation. In the first stage, by refining locally and globally the energy mesh, we seek, on a fine energy mesh with subgroup self-shielding, a solution equivalent to a reference solver (Monte Carlo or pointwise deterministic solver). In the second step, once fixed the number of groups, depending on the acceptable computational cost, and chosen the most appropriate self-shielding models to the reactor type, we look for the best bounds of the reference mesh minimizing reaction rate errors by the particle swarm optimization algorithm. This new approach allows us to define new meshes for fast reactors as accurate as the currently used ones, but with fewer groups. (author)

A code system to generate multigroup cross-sections using basic data

International Nuclear Information System (INIS)

Garg, S.B.; Kumar, Ashok

1978-01-01

For the neutronic studies of nuclear reactors, multigroup cross-sections derived from the basic energy point data are needed. In order to carry out the design based studies, these cross-sections should also incorporate the temperature and fuel concentration effects. To meet these requirements, a code system comprising of RESRES, UNRES, FIGERO, INSCAT, FUNMO, AVER1 and BGPONE codes has been adopted. The function of each of these codes is discussed. (author)

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.

Nodal enhances the activity of FoxO3a and its synergistic interaction with Smads to regulate cyclin G2 transcription in ovarian cancer cells.

Science.gov (United States)

Fu, G; Peng, C

2011-09-15

Nodal, a member of the transforming growth factor-β superfamily, has been recently shown to suppress cell proliferation and to stimulate the expression of cyclin G2 (CCNG2) in human epithelial ovarian cancer cells. However, the precise mechanisms underlying these events are not fully understood. In this study, we investigated the transcriptional regulation of CCNG2 by the Nodal signaling pathway. In ovarian cancer cells, overexpression of Nodal or its receptors, activin receptor-like kinase 7 (ALK7) or ALK4, resulted in an increase in the CCNG2 promoter activity. Several putative Forkhead box class O (FoxO)3a-binding sites are present in the human CCNG2 promoter and overexpression of FoxO3a enhanced the CCNG2 promoter activity. The functional FoxO3a-binding element (FBE) was mapped to a proximal region located between -398 and -380 bp (FBE1) through deletion and mutation analyses, as well as chromatin immunoprecipitation (IP) assay. Interestingly, mutation of the FBE1 not only abolished the effect of FoxO3a, but also blocked Nodal-induced CCNG2 transcription. Nodal stimulated FoxO3a mRNA and protein expression through the canonical Smad pathway and suppressed FoxO3a inactivation by inhibiting AKT activity. Silencing of FoxO3a using small interfering RNA significantly reduced the effect of Nodal on the CCNG2 promoter activity. On the other hand, overexpression of Smad2 and Smad3 enhanced the FoxO3a-induced CCNG2 promoter activity whereas knockdown of Smad4 blocked the activity of FoxO3a. Furthermore, IP assays revealed that FoxO3a formed complexes with Smad proteins and that Nodal enhanced the binding of FoxO3a to the CCNG2 promoter. Finally, silencing of FoxO3a reversed the inhibitory effect of Nodal on cell proliferation. Taken together, these findings demonstrated that Nodal signaling promotes CCNG2 transcription by upregulating FoxO3a expression, inhibiting FoxO3a phosphorylation and enhancing its synergistic interaction with Smads. These results also suggest

Intra nodal reconstruction of the numerical solution generated by the spectro nodal constant for Sn problems of eigenvalues in two-dimensional rectangular geometry

International Nuclear Information System (INIS)

Menezes, Welton Alves de

2009-01-01

In this dissertation the spectral nodal method SD-SGF-CN, cf. spectral diamond - spectral Green's function - constant nodal, is used to determine the angular fluxes averaged along the edges of the homogenized nodes in heterogeneous domains. Using these results, we developed an algorithm for the reconstruction of the node-edge average angular fluxes within the nodes of the spatial grid set up on the domain, since more localized numerical solutions are not generated by coarse-mesh numerical methods. Numerical results are presented to illustrate the accuracy of the algorithm we offer. (author)

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.

cmpXLatt: Westinghouse automated testing tool for nodal cross section models

International Nuclear Information System (INIS)

Guimaraes, Petri Forslund; Rönnberg, Kristian

2011-01-01

The procedure for evaluating the merits of different nodal cross section representation models is normally both cumbersome and time consuming, and includes many manual steps when preparing appropriate benchmark problems. Therefore, a computer tool called cmpXLatt has been developed at Westinghouse in order to facilitate the process of performing comparisons between nodal diffusion theory results and corresponding transport theory results on a single node basis. Due to the large number of state points that can be evaluated by cmpXLatt, a systematic and comprehensive way of performing verification and validation of nodal cross section models is provided. This paper presents the main features of cmpXLatt and demonstrates the benefits of using cmpXLatt in a real life application. (author)

ERRORJ, Multigroup covariance matrices generation from ENDF-6 format

International Nuclear Information System (INIS)

Chiba, Go

2007-01-01

1 - Description of program or function: ERRORJ produces multigroup covariance matrices from ENDF-6 format following mainly the methods of the ERRORR module in NJOY94.105. New version differs from previous version in the following features: Additional features in ERRORJ with respect to the NJOY94.105/ERRORR module: - expands processing for the covariance matrices of resolved and unresolved resonance parameters; - processes average cosine of scattering angle and fission spectrum; - treats cross-correlation between different materials and reactions; - accepts input of multigroup constants with various forms (user input, GENDF, etc.); - outputs files with various formats through utility NJOYCOVX (COVERX format, correlation matrix, relative error and standard deviation); - uses a 1% sensitivity method for processing of resonance parameters; - ERRORJ can process the JENDL-3.2 and 3.3 covariance matrices. Additional features of the version 2 with respect to the previous version of ERRORJ: - Since the release of version 2, ERRORJ has been modified to increase its reliability and stability, - calculation of the correlation coefficients in the resonance region, - Option for high-speed calculation is implemented, - Perturbation amount is optimised in a sensitivity calculation, - Effect of the resonance self-shielding can be considered, - a compact covariance format (LCOMP=2) proposed by N. M. Larson can be read. Additional features of the version 2.2.1 with respect to the previous version of ERRORJ: - Several routines were modified to reduce calculation time. The new one needs shorter calculation time (50-70%) than the old version without changing results. - In the U-233 and Pu-241 files of JENDL-3.3 an inconsistency between resonance parameters in MF=32 and those in MF=2 was corrected. NEA-1676/06: This version differs from the previous one (NEA-1676/05) in the following: ERRORJ2.2.1 was modified to treat the self-shielding effect accurately. NEA-1676/07: This version

Extra-nodal lymphoma. A survey of Japan lymphoma radiation therapy group

International Nuclear Information System (INIS)

Oguchi, Masahiko; Ikeda, Hiroshi; Nakamura, Shigeo

2002-01-01

The purpose of this study was to examine, retrospectively, national-wide clinical data of patients with localized extranodal non-Hodgkin's lymphoma (NHL) who were treated by radiation therapy with or without chemotherapy. The survey was carried out at 25 radiation oncology institutions in Japan in 1998. In 1999, according to the Revised European American Lymphoma (REAL) classification, central pathological review conducted at Aichi cancer center was carried out for the data from 7 radiation oncology institutions. The 5-year progression free survival rates (PFS) were calculated to identify prognostic factors. Survey: Data from 1, 141 patients with stage I and II NHL were recruited from 1988 through 1992. Of them, 787 patients, who were treated using definitive radiotherapy with or without chemotherapy for intermediate and high-grade lymphomas in Working Formulation, constituted the core of this study. Primary tumors arose mainly from extra-nodal organs (71%) in the head and neck (Waldeyer's ring: 41%, thyroid gland: 7%, nasal cavities: 5%, oral cavities: 4%, sinus: 3%, orbital structures: 3%, skin: 2% and etc.). The median age of 60 years for patients with extra-nodal NHL was higher than that of 56 years for patients with nodal NHL (p<0.01). Female were dominant in incidence of extra-nodal NHL arising from the thyroid gland, skin and gastrointestinal tract. The percentage of stage I to the extra-nodal NHL from orbit, sino-nasal presentation was higher than that of other NHLs. The percentage of stage II to the extra-nodal NHL from Waldeyer's ring and thyroid gland was higher than that of other NHLs. Central pathological review was carried out for pathological data from 79 patients (Waldeyer's ring: 45, thyroid gland: 19, sinonasal cavities: 15). Of these, diffuse large B cell lymphoma (DLBCL) composed 63% of all patients, mucosa associated lyumphoid tissue lymphoma (MALT-L): 16%, Natural Killer/T cell lymphoma (NK/T-L): 11%, and mantle cell lymphoma: 5% in REAL

MPI version of NJOY and its application to multigroup cross-section generation

Energy Technology Data Exchange (ETDEWEB)

Alpan, A.; Haghighat, A.

1999-07-01

Multigroup cross-section libraries are needed in performing neutronics calculations. These libraries are referred to as broad-group libraries. The number of energy groups and group structure are highly dependent on the application and/or user's objectives. For example, for shielding calculations, broad-group libraries such as SAILOR and BUGLE with 47-neutron and 20-gamma energy groups are used. The common procedure to obtain a broad-group library is a three-step process: (1) processing pointwise ENDF (PENDF) format cross sections; (2) generating fine-group cross sections; and (3) collapsing fine-group cross sections to broad-group. The NJOY code is used to prepare fine-group cross sections by processing pointwise ENDF data. The code has several modules, each one performing a specific task. For instance, the module RECONR performs linearization and reconstruction of the cross sections, and the module GROUPR generates multigroup self-shielded cross sections. After fine-group, i.e., groupwise ENDF (GENDF), cross sections are produced, cross sections are self-shielded, and a one-dimensional transport calculation is performed to obtain flux spectra at specific regions in the model. These fluxes are then used as weighting functions to collapse the fine-group cross sections to obtain a broad-group cross-section library. The third step described is commonly performed by the AMPX code system. SMILER converts NJOY GENDF filed to AMPX master libraries, AJAX collects the master libraries. BONAMI performs self-shielding calculations, NITAWL converts the AMPX master library to a working library, XSDRNPM performs one-dimensional transport calculations, and MALOCS collapses fine-group cross sections to broad-group. Finally, ALPO is used to generate ANISN format libraries. In this three-step procedure, generally NJOY requires the largest amount of CPU time. This time varies depending on the user's specified parameters for each module, such as reconstruction tolerances

MPI version of NJOY and its application to multigroup cross-section generation

International Nuclear Information System (INIS)

Alpan, A.; Haghighat, A.

1999-01-01

Multigroup cross-section libraries are needed in performing neutronics calculations. These libraries are referred to as broad-group libraries. The number of energy groups and group structure are highly dependent on the application and/or user's objectives. For example, for shielding calculations, broad-group libraries such as SAILOR and BUGLE with 47-neutron and 20-gamma energy groups are used. The common procedure to obtain a broad-group library is a three-step process: (1) processing pointwise ENDF (PENDF) format cross sections; (2) generating fine-group cross sections; and (3) collapsing fine-group cross sections to broad-group. The NJOY code is used to prepare fine-group cross sections by processing pointwise ENDF data. The code has several modules, each one performing a specific task. For instance, the module RECONR performs linearization and reconstruction of the cross sections, and the module GROUPR generates multigroup self-shielded cross sections. After fine-group, i.e., groupwise ENDF (GENDF), cross sections are produced, cross sections are self-shielded, and a one-dimensional transport calculation is performed to obtain flux spectra at specific regions in the model. These fluxes are then used as weighting functions to collapse the fine-group cross sections to obtain a broad-group cross-section library. The third step described is commonly performed by the AMPX code system. SMILER converts NJOY GENDF filed to AMPX master libraries, AJAX collects the master libraries. BONAMI performs self-shielding calculations, NITAWL converts the AMPX master library to a working library, XSDRNPM performs one-dimensional transport calculations, and MALOCS collapses fine-group cross sections to broad-group. Finally, ALPO is used to generate ANISN format libraries. In this three-step procedure, generally NJOY requires the largest amount of CPU time. This time varies depending on the user's specified parameters for each module, such as reconstruction tolerances, temperatures

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

A boundary integral equation for boundary element applications in multigroup neutron diffusion theory

International Nuclear Information System (INIS)

Ozgener, B.

1998-01-01

A boundary integral equation (BIE) is developed for the application of the boundary element method to the multigroup neutron diffusion equations. The developed BIE contains no explicit scattering term; the scattering effects are taken into account by redefining the unknowns. Boundary elements of the linear and constant variety are utilised for validation of the developed boundary integral formulation

[Method for optimal sensor placement in water distribution systems with nodal demand uncertainties].

Science.gov (United States)

Liu, Shu-Ming; Wu, Xue; Ouyang, Le-Yan

2013-08-01

The notion of identification fitness was proposed for optimizing sensor placement in water distribution systems. Nondominated Sorting Genetic Algorithm II was used to find the Pareto front between minimum overlap of possible detection times of two events and the best probability of detection, taking nodal demand uncertainties into account. This methodology was applied to an example network. The solutions show that the probability of detection and the number of possible locations are not remarkably affected by nodal demand uncertainties, but the sources identification accuracy declines with nodal demand uncertainties.

Application of nonlinear nodal diffusion method for a small research reactor

International Nuclear Information System (INIS)

Jaradat, Mustafa K.; Alawneh, Luay M.; Park, Chang Je; Lee, Byungchul

2014-01-01

Highlights: • We applied nonlinear unified nodal method for 10 MW IAEA MTR benchmark problem. • TRITION–NEWT system was used to obtain two-group burnup dependent cross sections. • The criticality and power distribution compared with reference (IAEA-TECDOC-233). • Comparison between different fuel materials was conducted. • Satisfactory results were provided using UNM for MTR core calculations. - Abstract: Nodal diffusion methods are usually used for LWR calculations and rarely used for research reactor calculations. A unified nodal method with an implementation of the coarse mesh finite difference acceleration was developed for use in plate type research reactor calculations. It was validated for two PWR benchmark problems and then applied for IAEA MTR benchmark problem for static calculations to check the validity and accuracy of the method. This work was conducted to investigate the unified nodal method capability to treat material testing reactor cores. A 10 MW research reactor core is considered with three calculation cases for low enriched uranium fuel depending on the core burnup status of fresh, beginning-of-life, and end-of-life cores. The validation work included criticality calculations, flux distribution, and power distribution; in addition, a comparison between different fuel materials with the same uranium content was conducted. The homogenized two-group cross sections were generated using the TRITON–NEWT system. The results were compared with a reference, which was taken from IAEA-TECDOC-233. The unified nodal method provides satisfactory results for an all-rod out case, and the three-dimensional, two-group diffusion model can be considered accurate enough for MTR core calculations

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)

Omitting elective nodal irradiation during thoracic irradiation in limited-stage small cell lung cancer--evidence from a phase II trial.

Science.gov (United States)

Colaco, Rovel; Sheikh, Hamid; Lorigan, Paul; Blackhall, Fiona; Hulse, Paul; Califano, Raffaele; Ashcroft, Linda; Taylor, Paul; Thatcher, Nicholas; Faivre-Finn, Corinne

2012-04-01

Omitting elective nodal irradiation (ENI) in limited-stage disease small cell lung cancer (LD-SCLC) is expected to result in smaller radiation fields. We report on data from a randomised phase II trial that omitted ENI in patients receiving concurrent chemo-radiotherapy for LD-SCLC. 38 patients with LD-SCLC were randomised to receive once-daily (66 Gy in 33 fractions) or twice-daily (45 Gy in 30 fractions) radiotherapy (RT). 3D-conformal RT was given concurrently with cisplatin and etoposide starting with the second cycle of a total of four cycles. The gross tumour volume was defined as primary tumour with involved lymph nodes (nodes ≥1 cm in short axis) identifiable with CT imaging. ENI was not used. Six recurrence patterns were identified: recurrence within planning target volume (PTV) only, recurrence within PTV+regional nodal recurrence and/or distant recurrence, isolated nodal recurrence outside PTV, nodal recurrence outside PTV+distant recurrence, distant metastases only and no recurrence. At median follow-up 16.9 months, 31/38 patients were evaluable and 14/31 patients had relapsed. There were no isolated nodal recurrences. Eight patients relapsed with intra-thoracic disease: 2 within PTV only, 4 within PTV and distantly and 2 with nodal recurrence outside PTV plus distant metastases. Rates of grade 3+ acute oesophagitis and pneumonitis in the 31 evaluable patients were 23 and 3% respectively. In our study of LD-SCLC, omitting ENI based on CT imaging was not associated with a high risk of isolated nodal recurrence, although further prospective studies are needed to confirm this. Routine ENI omission will be further evaluated prospectively in the ongoing phase III CONVERT trial (NCT00433563). Copyright © 2011 Elsevier Ireland Ltd. All rights reserved.

Development of one-energy group, two-dimensional, frequency dependent detector adjoint function based on the nodal method

International Nuclear Information System (INIS)

Khericha, Soli T.

2000-01-01

One-energy group, two-dimensional computer code was developed to calculate the response of a detector to a vibrating absorber in a reactor core. A concept of local/global components, based on the frequency dependent detector adjoint function, and a nodalization technique were utilized. The frequency dependent detector adjoint functions presented by complex equations were expanded into real and imaginary parts. In the nodalization technique, the flux is expanded into polynomials about the center point of each node. The phase angle and the magnitude of the one-energy group detector adjoint function were calculated for a detector located in the center of a 200x200 cm reactor using a two-dimensional nodalization technique, the computer code EXTERMINATOR, and the analytical solution. The purpose of this research was to investigate the applicability of a polynomial nodal model technique to the calculations of the real and the imaginary parts of the detector adjoint function for one-energy group two-dimensional polynomial nodal model technique. From the results as discussed earlier, it is concluded that the nodal model technique can be used to calculate the detector adjoint function and the phase angle. Using the computer code developed for nodal model technique, the magnitude of one energy group frequency dependent detector adjoint function and the phase angle were calculated for the detector located in the center of a 200x200 cm homogenous reactor. The real part of the detector adjoint function was compared with the results obtained from the EXTERMINATOR computer code as well as the analytical solution based on a double sine series expansion using the classical Green's Function solution. The values were found to be less than 1% greater at 20 cm away from the source region and about 3% greater closer to the source compared to the values obtained from the analytical solution and the EXTERMINATOR code. The currents at the node interface matched within 1% of the average

Usefulness of FDG PET for nodal staging using a dual head coincidence camera in patients with lung cancer

International Nuclear Information System (INIS)

Yoon, Seok Nam; Park, Chan H.; Lee, Myoung Hoon; Hwang, Kyung Hoon; Hwang, Kyung Hoon

2001-01-01

Staging of lung cancer requires an accurate evaluation of the mediastinum. Positron imaging with dual head cameras may be not as sensitive as dedicated PET. Therefore, the purpose of the study was to evaluated the usefulness of F-18 FDG coincidence (CoDe) PET using a dual-head gamma camera in the nodal staging of the lung cancer. CoDe-PET studies were performed in 51 patients with histologically proven non small cell lung cancer. CoDe-PET began 60 minutes after the injection of 111-185 MBq of F-18 FDG. CoDe-PET was performed using a dual-head gamma camera equipped with coincidence detection circuitry (Elscints Varicam, Haifa, lsrael). There was no attenuation correction made and reconstruction was done using a filtered back-projection. Surgery was performed in 49 patients CoDe-PET studies were evaluated visually. Any focal increased uptake was considered abnormal. The nodal stating of CoDe-PET studies were evaluated visually. Any focal increased uptake was considered abnormal. The nodal staging of CoDe-PET and of CT were compared with the nodal stating of surgical (49) and mediastinoscopical (2) pathology. All primary lung lesions were hypermetabolic and easily visualized. Compared with surgical nodal staging as a gold standard, false positives occurred in 13 CoDe PET and 17 CT studies and false negative occurred in 5 CoDe-PET and 4 CT studies. Assessment of lymph node involvement by CoDe-PET depicted a sensitivity of 67%, specificity of 64% and accuracy of 65%. CT revealed a sensitivity of 73%, specificity of 53% and accuracy of 59% in the assessment of lymph node involvement. The detection of primary lesions were 100% but nodal staging was suboptimal for routine clinical use. This is mainly due to limited resolution of our system

Qualification of a full plant nodalization for the prediction of the core exit temperature through a scaling methodology

Energy Technology Data Exchange (ETDEWEB)

Freixa, J., E-mail: jordi.freixa-terradas@upc.edu; Martínez-Quiroga, V., E-mail: victor.martinez.quiroga@upc.edu; Reventós, F., E-mail: francesc.reventos@upc.edu

2016-11-15

Highlights: • Core exit temperature is used in PWRs as an indication of core heat up. • Qualification of full scale nuclear reactors by means of a scaling methodology. • Scaling of RELAP5 calculations to full scale power plants. - Abstract: System codes and their necessary power plant nodalizations are an essential step in thermal hydraulic safety analysis. In order to assess the safety of a particular power plant, in addition to the validation and verification of the code, the nodalization of the system needs to be qualified. Since most existing experimental data come from scaled-down facilities, any qualification process must therefore address scale considerations. The Group of Thermal Hydraulic Studies at Technical University of Catalonia has developed a scaling-up methodology (SCUP) for the qualification of full-scale nodalizations through a systematic procedure based on the extrapolation of post-test simulations of Integral Test Facility experiments. In the present work, the SCUP methodology will be employed to qualify the nodalization of the AscóNPP, a Pressurized Water Reactor (PWR), for the reproduction of an important safety phenomenon which is the effectiveness of the Core Exit Temperature (CET) as an Accident Management (AM) indicator. Given the difficulties in placing measurements in the core region, CET measurements are used as a criterion for the initiation of safety operational procedures during accidental conditions in PWR. However, the CET response has some limitation in detecting inadequate core cooling simply because the measurement is not taken in the position where the cladding exposure occurs. In order to apply the SCUP methodology, the OECD/NEA ROSA-2 Test 3, an SBLOCA in the hot leg, has been selected as a starting point. This experiment was conducted at the Large Scale Test Facility (LSTF), a facility operated by the Japanese Atomic Energy Agency (JAEA) and was focused on the assessment of the effectiveness of AM actions triggered by

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.

COMPAR: system to compare multigroup cross sections generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 AND XLACS

International Nuclear Information System (INIS)

Anaf, J.; Chalhoub, E.S.

1987-11-01

A system, composed by the computer programs COMPAR and its interfaces, developed for comparing multigroup cross sections calculated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS, is presented. (author)

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

Improvements in EBR-2 core depletion calculations

International Nuclear Information System (INIS)

Finck, P.J.; Hill, R.N.; Sakamoto, S.

1991-01-01

The need for accurate core depletion calculations in Experimental Breeder Reactor No. 2 (EBR-2) is discussed. Because of the unique physics characteristics of EBR-2, it is difficult to obtain accurate and computationally efficient multigroup flux predictions. This paper describes the effect of various conventional and higher order schemes for group constant generation and for flux computations; results indicate that higher-order methods are required, particularly in the outer regions (i.e. the radial blanket). A methodology based on Nodal Equivalence Theory (N.E.T.) is developed which allows retention of the accuracy of a higher order solution with the computational efficiency of a few group nodal diffusion solution. The application of this methodology to three-dimensional EBR-2 flux predictions is demonstrated; this improved methodology allows accurate core depletion calculations at reasonable cost. 13 refs., 4 figs., 3 tabs

REX1-87, Multigroup Neutron Cross-Sections from ENDF/B

International Nuclear Information System (INIS)

Gopalakrishnan, V.; Ganesan, S.

1988-01-01

1 - Description of program or function: The program calculates self- shielding factors for reactor applications from a pre-processed (linearized) evaluated nuclear data file in the ENDF/B format. 2 - Method of solution: Bondarenko definition of multigroup self- shielding factors invoking narrow resonance treatment is used. 3 - Restrictions on the complexity of the problem: a) Maximum no. of energy group is 620. b) Only the built-in forms of the weighting functions can be chosen. c) The program is strictly limited to resolved resonance region from physical considerations

Pelvic Nodal Dosing With Registration to the Prostate: Implications for High-Risk Prostate Cancer Patients Receiving Stereotactic Body Radiation Therapy

International Nuclear Information System (INIS)

Kishan, Amar U.; Lamb, James M.; Jani, Shyam S.; Kang, Jung J.; Steinberg, Michael L.; King, Christopher R.

2015-01-01

Purpose: To determine whether image guidance with rigid registration (RR) to intraprostatic markers (IPMs) yields acceptable coverage of the pelvic lymph nodes in the context of a stereotactic body radiation therapy (SBRT) regimen. Methods and Materials: Four to seven kilovoltage cone-beam CTs (CBCTs) from 12 patients with high-risk prostate cancer were analyzed, allowing approximation of an SBRT regimen. The nodal clinical target volume (CTV N ) and bladder were contoured on all kilovoltage CBCTs. The V 100 CTV N , expressed as a ratio to the same parameter on the initial plan, and the magnitude of translational shift between RR to the IPMs versus RR to the pelvic bones, were computed. The ability of a multimodality bladder filling protocol to minimize bladder height variation was assessed in a separate cohort of 4 patients. Results: Sixty-five CBCTs were assessed. The average V 100 CTV N was 92.6%, but for a subset of 3 patients the average was 80.0%, compared with 97.8% for the others (P<.0001). The average overall and superior–inferior axis magnitudes of the bony-to-fiducial translations were significantly larger in the subgroup with suboptimal nodal coverage (8.1 vs 3.9 mm and 5.8 vs 2.4 mm, respectively; P<.0001). Relative bladder height changes were also significantly larger in the subgroup with suboptimal nodal coverage (42.9% vs 18.5%; P<.05). Use of a multimodality bladder-filling protocol minimized bladder height variation (P<.001). Conclusion: A majority of patients had acceptable nodal coverage after RR to IPMs, even when approximating SBRT. However, a subset of patients had suboptimal nodal coverage. These patients had large bony-to-fiducial translations and large variations in bladder height. Nodal coverage should be excellent if the superior–inferior axis bony-to-fiducial translation and the relative bladder height change (both easily measured on CBCT) are kept to a minimum. Implementation of a strict bladder filling protocol may achieve this

The Maternal Maverick/GDF15-like TGF-β Ligand Panda Directs Dorsal-Ventral Axis Formation by Restricting Nodal Expression in the Sea Urchin Embryo

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Haillot, Emmanuel; Molina, Maria Dolores; Lapraz, François; Lepage, Thierry

2015-01-01

Specification of the dorsal-ventral axis in the highly regulative sea urchin embryo critically relies on the zygotic expression of nodal, but whether maternal factors provide the initial spatial cue to orient this axis is not known. Although redox gradients have been proposed to entrain the dorsal-ventral axis by acting upstream of nodal, manipulating the activity of redox gradients only has modest consequences, suggesting that other factors are responsible for orienting nodal expression and defining the dorsal-ventral axis. Here we uncover the function of Panda, a maternally provided transforming growth factor beta (TGF-β) ligand that requires the activin receptor-like kinases (Alk) Alk3/6 and Alk1/2 receptors to break the radial symmetry of the embryo and orient the dorsal-ventral axis by restricting nodal expression. We found that the double inhibition of the bone morphogenetic protein (BMP) type I receptors Alk3/6 and Alk1/2 causes a phenotype dramatically more severe than the BMP2/4 loss-of-function phenotype, leading to extreme ventralization of the embryo through massive ectopic expression of nodal, suggesting that an unidentified signal acting through BMP type I receptors cooperates with BMP2/4 to restrict nodal expression. We identified this ligand as the product of maternal Panda mRNA. Double inactivation of panda and bmp2/4 led to extreme ventralization, mimicking the phenotype caused by inactivation of the two BMP receptors. Inhibition of maternal panda mRNA translation disrupted the early spatial restriction of nodal, leading to persistent massive ectopic expression of nodal on the dorsal side despite the presence of Lefty. Phylogenetic analysis indicates that Panda is not a prototypical BMP ligand but a member of a subfamily of TGF-β distantly related to Inhibins, Lefty, and TGF-β that includes Maverick from Drosophila and GDF15 from vertebrates. Indeed, overexpression of Panda does not appear to directly or strongly activate phosphoSmad1

The Maternal Maverick/GDF15-like TGF-β Ligand Panda Directs Dorsal-Ventral Axis Formation by Restricting Nodal Expression in the Sea Urchin Embryo.

Science.gov (United States)

Haillot, Emmanuel; Molina, Maria Dolores; Lapraz, François; Lepage, Thierry

2015-01-01

Specification of the dorsal-ventral axis in the highly regulative sea urchin embryo critically relies on the zygotic expression of nodal, but whether maternal factors provide the initial spatial cue to orient this axis is not known. Although redox gradients have been proposed to entrain the dorsal-ventral axis by acting upstream of nodal, manipulating the activity of redox gradients only has modest consequences, suggesting that other factors are responsible for orienting nodal expression and defining the dorsal-ventral axis. Here we uncover the function of Panda, a maternally provided transforming growth factor beta (TGF-β) ligand that requires the activin receptor-like kinases (Alk) Alk3/6 and Alk1/2 receptors to break the radial symmetry of the embryo and orient the dorsal-ventral axis by restricting nodal expression. We found that the double inhibition of the bone morphogenetic protein (BMP) type I receptors Alk3/6 and Alk1/2 causes a phenotype dramatically more severe than the BMP2/4 loss-of-function phenotype, leading to extreme ventralization of the embryo through massive ectopic expression of nodal, suggesting that an unidentified signal acting through BMP type I receptors cooperates with BMP2/4 to restrict nodal expression. We identified this ligand as the product of maternal Panda mRNA. Double inactivation of panda and bmp2/4 led to extreme ventralization, mimicking the phenotype caused by inactivation of the two BMP receptors. Inhibition of maternal panda mRNA translation disrupted the early spatial restriction of nodal, leading to persistent massive ectopic expression of nodal on the dorsal side despite the presence of Lefty. Phylogenetic analysis indicates that Panda is not a prototypical BMP ligand but a member of a subfamily of TGF-β distantly related to Inhibins, Lefty, and TGF-β that includes Maverick from Drosophila and GDF15 from vertebrates. Indeed, overexpression of Panda does not appear to directly or strongly activate phosphoSmad1

Concomitant occurrence of sinus histiocytosis with massive lymphadenopathy and nodal marginal zone lymphoma.

Science.gov (United States)

Pang, Changlee S; Grier, David D; Beaty, Michael W

2011-03-01

Sinus histiocytosis with massive lymphadenopathy (SHML), also known as Rosai-Dorfman disease, is a rare self-limiting disorder of histiocytes with unknown etiology. Sinus histiocytosis with massive lymphadenopathy is most common in children and young adults and is characterized by painless lymphadenopathy. Histologically there is a proliferation of sinus histiocytes with lymphophagocytosis or emperipolesis. On rare occasions, SHML has been associated with lymphoma, usually involving different anatomic sites and developing at different times. We report a case of concomitant SHML and nodal marginal zone lymphoma involving the same lymph node without involvement of other nodal or extranodal sites. The presence of concomitant SHML within the lymph node involved by nodal marginal zone lymphoma may represent the responsiveness of SHML histiocytes to B-cell-derived cytokines in lymphoproliferative disorders. To our knowledge, this is the first description of concomitant occurrence of SHML and nodal marginal zone lymphoma.

A block-iterative nodal integral method for forced convection problems

International Nuclear Information System (INIS)

Decker, W.J.; Dorning, J.J.

1992-01-01

A new efficient iterative nodal integral method for the time-dependent two- and three-dimensional incompressible Navier-Stokes equations has been developed. Using the approach introduced by Azmy and Droning to develop nodal mehtods with high accuracy on coarse spatial grids for two-dimensional steady-state problems and extended to coarse two-dimensional space-time grids by Wilson et al. for thermal convection problems, we have developed a new iterative nodal integral method for the time-dependent Navier-Stokes equations for mechanically forced convection. A new, extremely efficient block iterative scheme is employed to invert the Jacobian within each of the Newton-Raphson iterations used to solve the final nonlinear discrete-variable equations. By taking advantage of the special structure of the Jacobian, this scheme greatly reduces memory requirements. The accuracy of the overall method is illustrated by appliying it to the time-dependent version of the classic two-dimensional driven cavity problem of computational fluid dynamics

HCN4 ion channel function is required for early events that regulate anatomical left-right patterning in a nodal and lefty asymmetric gene expression-independent manner.

Science.gov (United States)

Pai, Vaibhav P; Willocq, Valerie; Pitcairn, Emily J; Lemire, Joan M; Paré, Jean-François; Shi, Nian-Qing; McLaughlin, Kelly A; Levin, Michael

2017-10-15

Laterality is a basic characteristic of all life forms, from single cell organisms to complex plants and animals. For many metazoans, consistent left-right asymmetric patterning is essential for the correct anatomy of internal organs, such as the heart, gut, and brain; disruption of left-right asymmetry patterning leads to an important class of birth defects in human patients. Laterality functions across multiple scales, where early embryonic, subcellular and chiral cytoskeletal events are coupled with asymmetric amplification mechanisms and gene regulatory networks leading to asymmetric physical forces that ultimately result in distinct left and right anatomical organ patterning. Recent studies have suggested the existence of multiple parallel pathways regulating organ asymmetry. Here, we show that an isoform of the hyperpolarization-activated cyclic nucleotide-gated (HCN) family of ion channels (hyperpolarization-activated cyclic nucleotide-gated channel 4, HCN4) is important for correct left-right patterning. HCN4 channels are present very early in Xenopus embryos. Blocking HCN channels ( I h currents) with pharmacological inhibitors leads to errors in organ situs. This effect is only seen when HCN4 channels are blocked early (pre-stage 10) and not by a later block (post-stage 10). Injections of HCN4-DN (dominant-negative) mRNA induce left-right defects only when injected in both blastomeres no later than the 2-cell stage. Analysis of key asymmetric genes' expression showed that the sidedness of Nodal , Lefty , and Pitx2 expression is largely unchanged by HCN4 blockade, despite the randomization of subsequent organ situs, although the area of Pitx2 expression was significantly reduced. Together these data identify a novel, developmental role for HCN4 channels and reveal a new Nodal-Lefty-Pitx2 asymmetric gene expression-independent mechanism upstream of organ positioning during embryonic left-right patterning. © 2017. Published by The Company of Biologists Ltd.

Survey of computer codes which produce multigroup data from ENDF/B-IV

International Nuclear Information System (INIS)

Greene, N.M.

1975-01-01

The features of three code systems that produce multigroup neutron data are contrasted. This includes the ETOE-2/MC 2 -2/SDX, MINX/SPHINX and AMPX code packages. These systems all contain a fairly extensive set of processing capabilities with the current evaluated nuclear data files--ENDF/B. They were designed with different goals and applications in mind. This paper discusses some of their differences and the implications for particular situations

Exclusion of elective nodal irradiation is associated with minimal elective nodal failure in non-small cell lung cancer

Directory of Open Access Journals (Sweden)

Cox James D

2009-01-01

Full Text Available Abstract Background Controversy still exists regarding the long-term outcome of patients whose uninvolved lymph node stations are not prophylactically irradiated for non-small cell lung cancer (NSCLC treated with definitive radiotherapy. To determine the frequency of elective nodal failure (ENF and in-field failure (IFF, we examined a large cohort of patients with NSCLC staged with positron emission tomography (PET/computed tomography (CT and treated with 3-dimensional conformal radiotherapy (3D-CRT that excluded uninvolved lymph node stations. Methods We retrospectively reviewed the records of 115 patients with non-small cell lung cancer treated at our institution with definitive radiation therapy with or without concurrent chemotherapy (CHT. All patients were treated with 3D-CRT, including nodal regions determined by CT or PET to be disease involved. Concurrent platinum-based CHT was administered for locally advanced disease. Patients were analyzed in follow-up for survival, local regional recurrence, and distant metastases (DM. Results The median follow-up time was 18 months (3 to 44 months among all patients and 27 months (6 to 44 months among survivors. The median overall survival, 2-year actuarial overall survival and disease-free survival were 19 months, 38%, and 28%, respectively. The majority of patients died from DM, the overall rate of which was 36%. Of the 31 patients with local regional failure, 26 (22.6% had IFF, 5 (4.3% had ENF and 2 (1.7% had isolated ENF. For 88 patients with stage IIIA/B, the frequencies of IFF, any ENF, isolated ENF, and DM were 23 (26%, 3 (9%, 1 (1.1% and 36 (40.9%, respectively. The comparable rates for the 22 patients with early stage node-negative disease (stage IA/IB were 3 (13.6%, 1(4.5%, 0 (0%, and 5 (22.7%, respectively. Conclusion We observed only a 4.3% recurrence of any ENF and a 1.7% recurrence of isolated ENF in patients with NSCLC treated with definitive 3D-CRT without prophylactic irradiation of

Exclusion of elective nodal irradiation is associated with minimal elective nodal failure in non-small cell lung cancer

International Nuclear Information System (INIS)

Sulman, Erik P; Komaki, Ritsuko; Klopp, Ann H; Cox, James D; Chang, Joe Y

2009-01-01

Controversy still exists regarding the long-term outcome of patients whose uninvolved lymph node stations are not prophylactically irradiated for non-small cell lung cancer (NSCLC) treated with definitive radiotherapy. To determine the frequency of elective nodal failure (ENF) and in-field failure (IFF), we examined a large cohort of patients with NSCLC staged with positron emission tomography (PET)/computed tomography (CT) and treated with 3-dimensional conformal radiotherapy (3D-CRT) that excluded uninvolved lymph node stations. We retrospectively reviewed the records of 115 patients with non-small cell lung cancer treated at our institution with definitive radiation therapy with or without concurrent chemotherapy (CHT). All patients were treated with 3D-CRT, including nodal regions determined by CT or PET to be disease involved. Concurrent platinum-based CHT was administered for locally advanced disease. Patients were analyzed in follow-up for survival, local regional recurrence, and distant metastases (DM). The median follow-up time was 18 months (3 to 44 months) among all patients and 27 months (6 to 44 months) among survivors. The median overall survival, 2-year actuarial overall survival and disease-free survival were 19 months, 38%, and 28%, respectively. The majority of patients died from DM, the overall rate of which was 36%. Of the 31 patients with local regional failure, 26 (22.6%) had IFF, 5 (4.3%) had ENF and 2 (1.7%) had isolated ENF. For 88 patients with stage IIIA/B, the frequencies of IFF, any ENF, isolated ENF, and DM were 23 (26%), 3 (9%), 1 (1.1%) and 36 (40.9%), respectively. The comparable rates for the 22 patients with early stage node-negative disease (stage IA/IB) were 3 (13.6%), 1(4.5%), 0 (0%), and 5 (22.7%), respectively. We observed only a 4.3% recurrence of any ENF and a 1.7% recurrence of isolated ENF in patients with NSCLC treated with definitive 3D-CRT without prophylactic irradiation of uninvolved lymph node stations. Thus

From Fourier Transforms to Singular Eigenfunctions for Multigroup Transport

International Nuclear Information System (INIS)

Ganapol, B.D.

2001-01-01

A new Fourier transform approach to the solution of the multigroup transport equation with anisotropic scattering and isotropic source is presented. Through routine analytical continuation, the inversion contour is shifted from the real line to produce contributions from the poles and cuts in the complex plane. The integrand along the branch cut is then recast in terms of matrix continuum singular eigenfunctions, demonstrating equivalence of Fourier transform inversion and the singular eigenfunction expansion. The significance of this paper is that it represents the initial step in revealing the intimate connection between the Fourier transform and singular eigenfunction approaches as well as serves as a basis for a numerical algorithm

PROF-DD, Generator of Multigroup Cross-Sections Library DDX for MORSE-DD, ANISN-DD, DOT-DD

International Nuclear Information System (INIS)

Mori, Takamasa; Nakagawa, Masayuki; Ishiguro, Yukio

2002-01-01

1 - Description of program or function: The code system PROF-DD generates a multi-group double-differential cross section library DDX from evaluated data in ENDF/B-IV or ENDF/B-V format. The system consists of the following five modules: PROF-DDX is the main module of the system. It calculates the multigroup DDX and stores them on a master PDS file. MCFILEF generates a control file for PROF-DDX, which contains energy group and angle bin structures. SPINPTF prepares an input data file for PROF-DDX by combining the control file with other input data. DDXLIBMK edits a DDX library from the master PDS file for transport calculations. RESENDD performs resonance cross section and Doppler broadening calculations. 2 - Restrictions on the complexity of the problem: The numbers of energy groups and angle bins are less than 150 and 40, respectively

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

Effect of interfractional shoulder motion on low neck nodal targets for patients treated using volume modulated arc therapy (VMAT

Directory of Open Access Journals (Sweden)

Kevin Casey

2014-03-01

Full Text Available Purpose: To quantify the dosimetric impact of interfractional shoulder motion on targets in the low neck for head and neck patients treated with volume modulated arc therapy (VMAT.Methods: Three patients with head and neck cancer were selected. All three required treatment to nodal regions in the low neck in addition to the primary tumor site. The patients were immobilized during simulation and treatment with a custom thermoplastic mask covering the head and shoulders. One VMAT plan was created for each patient utilizing two full 360° arcs and a second plan was created consisting of two superior VMAT arcs matched to an inferior static AP supraclavicular field. A CT-on-rails alignment verification was performed weekly during each patient’s treatment course. The weekly CT images were registered to the simulation CT and the target contours were deformed and applied to the weekly CT. The two VMAT plans were copied to the weekly CT datasets and recalculated to obtain the dose to the deformed low neck contours.Results: The average observed shoulder position shift in any single dimension relative to simulation was 2.5 mm. The maximum shoulder shift observed in a single dimension was 25.7 mm. Low neck target mean doses, normalized to simulation and averaged across all weekly recalculations were 0.996, 0.991, and 1.033 (Full VMAT plan and 0.986, 0.995, and 0.990 (Half-Beam VMAT plan for the three patients, respectively. The maximum observed deviation in target mean dose for any individual weekly recalculation was 6.5%, occurring with the Full VMAT plan for Patient 3.Conclusion: Interfractional variation in dose to low neck nodal regions was quantified for three head and neck patients treated with VMAT. Mean dose was 3.3% higher than planned for one patient using a Full VMAT plan. A Half-Beam technique is likely a safer choice when treating the supraclavicular region with VMAT.-------------------------------------------Cite this article as: Casey K

MVP/GMVP 2: general purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods

International Nuclear Information System (INIS)

Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa; Nakagawa, Masayuki

2005-06-01

In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two vectorized Monte Carlo codes MVP and GMVP have been developed at JAERI. MVP is based on the continuous energy model and GMVP is on the multigroup model. Compared with conventional scalar codes, these codes achieve higher computation speed by a factor of 10 or more on vector super-computers. Both codes have sufficient functions for production use by adopting accurate physics model, geometry description capability and variance reduction techniques. The first version of the codes was released in 1994. They have been extensively improved and new functions have been implemented. The major improvements and new functions are (1) capability to treat the scattering model expressed with File 6 of the ENDF-6 format, (2) time-dependent tallies, (3) reaction rate calculation with the pointwise response function, (4) flexible source specification, (5) continuous-energy calculation at arbitrary temperatures, (6) estimation of real variances in eigenvalue problems, (7) point detector and surface crossing estimators, (8) statistical geometry model, (9) function of reactor noise analysis (simulation of the Feynman-α experiment), (10) arbitrary shaped lattice boundary, (11) periodic boundary condition, (12) parallelization with standard libraries (MPI, PVM), (13) supporting many platforms, etc. This report describes the physical model, geometry description method used in the codes, new functions and how to use them. (author)

Face centered cubic SnSe as a Z2 trivial Dirac nodal line material

OpenAIRE

Tateishi, Ikuma; Matsuura, Hiroyasu

2018-01-01

The presence of Dirac nodal line in the time-reversal and inversion symmetric system is dictated by Z2 index when spin-orbit interaction is absent. With the first principles calculation, we show that the Dirac nodal line can emerge in Z2 trivial material by calculating the band structure of SnSe of face centered cubic lattice as an example and it becomes a topological crystalline insulator when spin-orbit interaction is taken into account. We clarify the origin of the Dirac nodal line by obta...

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)...

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

Solution of multi-group diffusion equation in x-y-z geometry by finite Fourier transformation

International Nuclear Information System (INIS)

Kobayashi, Keisuke

1975-01-01

The multi-group diffusion equation in three-dimensional x-y-z geometry is solved by finite Fourier transformation. Applying the Fourier transformation to a finite region with constant nuclear cross sections, the fluxes and currents at the material boundaries are obtained in terms of the Fourier series. Truncating the series after the first term, and assuming that the source term is piecewise linear within each mesh box, a set of coupled equations is obtained in the form of three-point equations for each coordinate. These equations can be easily solved by the alternative direction implicit method. Thus a practical procedure is established that could be applied to replace the currently used difference equation. This equation is used to solve the multi-group diffusion equation by means of the source iteration method; and sample calculations for thermal and fast reactors show that the present method yields accurate results with a smaller number of mesh points than the usual finite difference equations. (auth.)

Development of an environment-insensitive PWR radial reflector model applicable to modern nodal reactor analysis method

International Nuclear Information System (INIS)

Mueller, E.M.

1989-05-01

This research is concerned with the development and analysis of methods for generating equivalent nodal diffusion parameters for the radial reflector of a PWR. The requirement that the equivalent reflector data be insensitive to changing core conditions is set as a principle objective. Hence, the environment dependence of the currently most reputable nodal reflector models, almost all of which are based on the nodal equivalence theory homgenization methods of Koebke and Smith, is investigated in detail. For this purpose, a special 1-D nodal equivalence theory reflector model, called the NGET model, is developed and used in 1-D and 2-D numerical experiments. The results demonstrate that these modern radial reflector models exhibit sufficient sensitivity to core conditions to warrant the development of alternative models. A new 1-D nodal reflector model, which is based on a novel combination of the nodal equivalence theory and the response matrix homogenization methods, is developed. Numerical results varify that this homogenized baffle/reflector model, which is called the NGET-RM model, is highly insensitive to changing core conditions. It is also shown that the NGET-RM model is not inferior to any of the existing 1-D nodal reflector models and that it has features which makes it an attractive alternative model for multi-dimensional reactor analysis. 61 refs., 40 figs., 36 tabs

A Hybrid Interpolation Method for Geometric Nonlinear Spatial Beam Elements with Explicit Nodal Force

Directory of Open Access Journals (Sweden)

Huiqing Fang

2016-01-01

Full Text Available Based on geometrically exact beam theory, a hybrid interpolation is proposed for geometric nonlinear spatial Euler-Bernoulli beam elements. First, the Hermitian interpolation of the beam centerline was used for calculating nodal curvatures for two ends. Then, internal curvatures of the beam were interpolated with a second interpolation. At this point, C1 continuity was satisfied and nodal strain measures could be consistently derived from nodal displacement and rotation parameters. The explicit expression of nodal force without integration, as a function of global parameters, was founded by using the hybrid interpolation. Furthermore, the proposed beam element can be degenerated into linear beam element under the condition of small deformation. Objectivity of strain measures and patch tests are also discussed. Finally, four numerical examples are discussed to prove the validity and effectivity of the proposed beam element.

Nodal Clearance Rate and Long-Term Efficacy of Individualized Sentinel Node–Based Pelvic Intensity Modulated Radiation Therapy for High-Risk Prostate Cancer

Energy Technology Data Exchange (ETDEWEB)

Müller, Arndt-Christian, E-mail: arndt-christian.mueller@med.uni-tuebingen.de [Department of Radiation Oncology, Eberhard Karls University, Tübingen (Germany); Eckert, Franziska; Paulsen, Frank; Zips, Daniel [Department of Radiation Oncology, Eberhard Karls University, Tübingen (Germany); Stenzl, Arnulf; Schilling, David [Department of Urology, Eberhard Karls University, Tübingen (Germany); Alber, Markus [Department of Oncology, Aarhus University, Aarhus (Denmark); Bares, Roland [Department of Nuclear Medicine and Clinical Molecular Imaging, Eberhard Karls University, Tübingen (Germany); Martus, Peter [Institute for Clinical Epidemiology and Applied Biometry, Eberhard Karls University, Tübingen (Germany); Weckermann, Dorothea [Department of Urology, Klinikum Augsburg, Augsburg (Germany); Belka, Claus; Ganswindt, Ute [Department of Radiation Oncology, Ludwig-Maximilians-University, Munich (Germany)

2016-02-01

Purpose: To assess the efficacy of individual sentinel node (SN)-guided pelvic intensity modulated radiation therapy (IMRT) by determining nodal clearance rate [(n expected nodal involvement − n observed regional recurrences)/n expected nodal involvement] in comparison with surgically staged patients. Methods and Materials: Data on 475 high-risk prostate cancer patients were examined. Sixty-one consecutive patients received pelvic SN-based IMRT (5 × 1.8 Gy/wk to 50.4 Gy [pelvic nodes + individual SN] and an integrated boost with 5 × 2.0 Gy/wk to 70.0 Gy to prostate + [base of] seminal vesicles) and neo-/adjuvant long-term androgen deprivation therapy; 414 patients after SN–pelvic lymph node dissection were used to calculate the expected nodal involvement rate for the radiation therapy sample. Biochemical control and overall survival were estimated for the SN-IMRT patients using the Kaplan-Meier method. The expected frequency of nodal involvement in the radiation therapy group was estimated by imputing frequencies of node-positive patients in the surgical sample to the pattern of Gleason, prostate-specific antigen, and T category in the radiation therapy sample. Results: After a median follow-up of 61 months, 5-year OS after SN-guided IMRT reached 84.4%. Biochemical control according to the Phoenix definition was 73.8%. The nodal clearance rate of SN-IMRT reached 94%. Retrospective follow-up evaluation is the main limitation. Conclusions: Radiation treatment of pelvic nodes individualized by inclusion of SNs is an effective regional treatment modality in high-risk prostate cancer patients. The pattern of relapse indicates that the SN-based target volume concept correctly covers individual pelvic nodes. Thus, this SN-based approach justifies further evaluation, including current dose-escalation strategies to the prostate in a larger prospective series.

Application of the visual system analyzer (ViSA): simulation of the steam generator tube rupture event at Ulchin unit 4

International Nuclear Information System (INIS)

Lee, S.W.; Kim, K.D.; Hwang, M.K.; Jeong, J.J.

2004-01-01

Korea Atomic Energy Research Institute (KAERI) has developed the Visual System Analyzer (ViSA) based on the best-estimate (B-E) codes, MARS and RETRAN-3D. The key features of ViSA are: (1) The use of the same input and the same level of accuracy as the original codes is guaranteed (2) Users can design their own plant mimic by a drag-and-drop from the provided indicators (3) The on-line interactive control enables users to simulate the operator's actions (4) The nodalization window is designed to display the transient temperature and void distributions. ViSA is composed of two parts; the B-E code with plant input and the Graphic User Interface (GUI) that includes the plant mimic and an interactive control function, etc. The calculation results of the B-E code are transferred to a user via the GUI and a user can apply the operator action to the B-E code using an interactive control function. Therefore, it is not necessary to prepare complex control input data to simulate the various manual operations which may occur during the plant transient. In this study, the Steam Generator Tube Rupture (SGTR) Accident, which occurred at Ulchin Unit 4 in April 2002, has been simulated using ViSA and the simulation results are compared with the measured plant data. The RETRAN-3D plant input data used in this simulation is a genetic input deck prepared for the simulation from a normal operation condition to a Small-Break LOCA. From the results of the SGTR simulation, we found that the GUI functions of ViSA and the input data for Ulchin Unit 4 have enough effectiveness and soundness. (author)

The statistics of the points where nodal lines intersect a reference curve

International Nuclear Information System (INIS)

Aronovitch, Amit; Smilansky, Uzy

2007-01-01

We study the intersection points of a fixed planar curve Γ with the nodal set of a translationally invariant and isotropic Gaussian random field Ψ(r) and the zeros of its normal derivative across the curve. The intersection points form a discrete random process which is the object of this study. The field probability distribution function is completely specified by the correlation G(|r - r'|) = (Ψ(r)Ψ(r')). Given an arbitrary G(|r - r'|), we compute the two-point correlation function of the point process on the line, and derive other statistical measures (repulsion, rigidity) which characterize the short- and long-range correlations of the intersection points. We use these statistical measures to quantitatively characterize the complex patterns displayed by various kinds of nodal networks. We apply these statistics in particular to nodal patterns of random waves and of eigenfunctions of chaotic billiards. Of special interest is the observation that for monochromatic random waves, the number variance of the intersections with long straight segments grows like Lln L, as opposed to the linear growth predicted by the percolation model, which was successfully used to predict other long-range nodal properties of that field

An approach to model reactor core nodalization for deterministic safety analysis

International Nuclear Information System (INIS)

Salim, Mohd Faiz; Samsudin, Mohd Rafie; Mamat Ibrahim, Mohd Rizal; Roslan, Ridha; Sadri, Abd Aziz; Farid, Mohd Fairus Abd

2016-01-01

Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH 1.6 , stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D ® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M

An approach to model reactor core nodalization for deterministic safety analysis

Energy Technology Data Exchange (ETDEWEB)

Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my; Samsudin, Mohd Rafie, E-mail: rafies@tnb.com.my [Nuclear Energy Department, Regulatory Economics & Planning Division, Tenaga Nasional Berhad (Malaysia); Mamat Ibrahim, Mohd Rizal, E-mail: m-rizal@nuclearmalaysia.gov.my [Prototypes & Plant Development Center, Malaysian Nuclear Agency (Malaysia); Roslan, Ridha, E-mail: ridha@aelb.gov.my; Sadri, Abd Aziz [Nuclear Installation Divisions, Atomic Energy Licensing Board (Malaysia); Farid, Mohd Fairus Abd [Reactor Technology Center, Malaysian Nuclear Agency (Malaysia)

2016-01-22

Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH{sub 1.6}, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D{sup ®} computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

An approach to model reactor core nodalization for deterministic safety analysis

Science.gov (United States)

Salim, Mohd Faiz; Samsudin, Mohd Rafie; Mamat @ Ibrahim, Mohd Rizal; Roslan, Ridha; Sadri, Abd Aziz; Farid, Mohd Fairus Abd

2016-01-01

Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH1.6, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

Specifications for a two-dimensional multi-group scattering code: ALCI

International Nuclear Information System (INIS)

Bayard, J.P.; Guillou, A.; Lago, B.; Bureau du Colombier, M.J.; Guillou, G.; Vasseur, Ch.

1965-02-01

This report describes the specifications of the ALCI programme. This programme resolves the system of difference equations similar to the homogeneous problem of multigroup neutron scattering, with two dimensions in space, in the three geometries XY, RZ, RΘ. It is possible with this method to calculate geometric and composition criticalities and also to calculate the accessory problem on demand. The maximum number of points dealt with is 6000. The maximum permissible number of groups is 12. The internal iterations are treated by the method of alternating directions. The external iterations are accelerated using the extrapolation method due to Tchebychev. (authors) [fr

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

. In this geometry nodal, continuous and discontinuous schemes were used. For the continuos schemes, only the Bi Quadratic (BiQ) and the Bi Cubic (BiC) were considered. In the case of the discontinuous ones two nodal schemes were considered, namely the Discontinuous Bi Linear (DBiL) and Discontinuous Bi Quadratic (DBiQ). The nodal schemes applied use from 4 up to 16 interpolation parameters per cell. These schemes are-defined for a set D{sub c} of interpolation parameters and a polynomial space S{sub h} corresponding to each one of the nodal schemes considered. All these four nodal hybrid schemes were implemented in a computer program called TNHXY starting from the computer program TNXY developed in previous thesis works. Several subroutines wae added to calculate the average neutron flux for each cell and for each energy group, generating two versions, one for the continuous schemes and one for the discontinuous schemes. For this geometry, two benchmark problems of the ANL-7416 document were analyzed. They are 7x7 BWR fuel assemblies, one without control rod and the other one with control rod. The computer program was also applied to a MOX assembly proposed by the Nuclear Energy Agency and it is considered as a reference problem. The results obtained for the one-dimensional problems using TNX for the effective multiplication factor were compared with the ones obtained with the code ANISN/PC. TNX code shows a faster convergence within four significant figures for the case with no control rod and three significant figures for the case with control rod (using double precision). These results suggest TNX is a very useful tool for this kind of calculations. For X Y geometry, the results obtained with TNHXY were compared with those calculated with the code TWOTRAN. To get these results, several spatial (1x1, 2x2, 4x4 per cell) and angular meshes (S{sub 2}, S{sub 4}, S{sub 6}, and S{sub 8}) were used. The results for the problem with no control rod were practically the same

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)

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)

Stability analysis of multi-group deterministic and stochastic epidemic models with vaccination rate

International Nuclear Information System (INIS)

Wang Zhi-Gang; Gao Rui-Mei; Fan Xiao-Ming; Han Qi-Xing

2014-01-01

We discuss in this paper a deterministic multi-group MSIR epidemic model with a vaccination rate, the basic reproduction number ℛ 0 , a key parameter in epidemiology, is a threshold which determines the persistence or extinction of the disease. By using Lyapunov function techniques, we show if ℛ 0 is greater than 1 and the deterministic model obeys some conditions, then the disease will prevail, the infective persists and the endemic state is asymptotically stable in a feasible region. If ℛ 0 is less than or equal to 1, then the infective disappear so the disease dies out. In addition, stochastic noises around the endemic equilibrium will be added to the deterministic MSIR model in order that the deterministic model is extended to a system of stochastic ordinary differential equations. In the stochastic version, we carry out a detailed analysis on the asymptotic behavior of the stochastic model. In addition, regarding the value of ℛ 0 , when the stochastic system obeys some conditions and ℛ 0 is greater than 1, we deduce the stochastic system is stochastically asymptotically stable. Finally, the deterministic and stochastic model dynamics are illustrated through computer simulations. (general)

Collapsing of multigroup cross sections in optimization problems solved by means of the maximum principle of Pontryagin

International Nuclear Information System (INIS)

Anton, V.

1979-05-01

A new formulation of multigroup cross section collapsing based on the conservation of point or zone value of hamiltonian is presented. This attempt is proper to optimization problems solved by means of maximum principle of Pontryagin. (author)

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)

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

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

. In this geometry nodal, continuous and discontinuous schemes were used. For the continuos schemes, only the Bi Quadratic (BiQ) and the Bi Cubic (BiC) were considered. In the case of the discontinuous ones two nodal schemes were considered, namely the Discontinuous Bi Linear (DBiL) and Discontinuous Bi Quadratic (DBiQ). The nodal schemes applied use from 4 up to 16 interpolation parameters per cell. These schemes are-defined for a set D{sub c} of interpolation parameters and a polynomial space S{sub h} corresponding to each one of the nodal schemes considered. All these four nodal hybrid schemes were implemented in a computer program called TNHXY starting from the computer program TNXY developed in previous thesis works. Several subroutines wae added to calculate the average neutron flux for each cell and for each energy group, generating two versions, one for the continuous schemes and one for the discontinuous schemes. For this geometry, two benchmark problems of the ANL-7416 document were analyzed. They are 7x7 BWR fuel assemblies, one without control rod and the other one with control rod. The computer program was also applied to a MOX assembly proposed by the Nuclear Energy Agency and it is considered as a reference problem. The results obtained for the one-dimensional problems using TNX for the effective multiplication factor were compared with the ones obtained with the code ANISN/PC. TNX code shows a faster convergence within four significant figures for the case with no control rod and three significant figures for the case with control rod (using double precision). These results suggest TNX is a very useful tool for this kind of calculations. For X Y geometry, the results obtained with TNHXY were compared with those calculated with the code TWOTRAN. To get these results, several spatial (1x1, 2x2, 4x4 per cell) and angular meshes (S{sub 2}, S{sub 4}, S{sub 6}, and S{sub 8}) were used. The results for the problem with no control rod were practically the same

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)

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)

Development of an object oriented nodal code using the refined AFEN derived from the method of component decomposition

International Nuclear Information System (INIS)

Noh, J. M.; Yoo, J. W.; Joo, H. K.

2004-01-01

In this study, we invented a method of component decomposition to derive the systematic inter-nodal coupled equations of the refined AFEN method and developed an object oriented nodal code to solve the derived coupled equations. The method of component decomposition decomposes the intra-nodal flux expansion of a nodal method into even and odd components in three dimensions to reduce the large coupled linear system equation into several small single equations. This method requires no additional technique to accelerate the iteration process to solve the inter-nodal coupled equations, since the derived equations can automatically act as the coarse mesh re-balance equations. By utilizing the object oriented programming concepts such as abstraction, encapsulation, inheritance and polymorphism, dynamic memory allocation, and operator overloading, we developed an object oriented nodal code that can facilitate the input/output and the dynamic control of the memories, and can make the maintenance easy. (authors)

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.

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.

Segregated nodal domains of two-dimensional multispecies Bose-Einstein condensates

Science.gov (United States)

Chang, Shu-Ming; Lin, Chang-Shou; Lin, Tai-Chia; Lin, Wen-Wei

2004-09-01

In this paper, we study the distribution of m segregated nodal domains of the m-mixture of Bose-Einstein condensates under positive and large repulsive scattering lengths. It is shown that components of positive bound states may repel each other and form segregated nodal domains as the repulsive scattering lengths go to infinity. Efficient numerical schemes are created to confirm our theoretical results and discover a new phenomenon called verticillate multiplying, i.e., the generation of multiple verticillate structures. In addition, our proposed Gauss-Seidel-type iteration method is very effective in that it converges linearly in 10-20 steps.

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)

Optimizing bone morphogenic protein 4-mediated human embryonic stem cell differentiation into trophoblast-like cells using fibroblast growth factor 2 and transforming growth factor-β/activin/nodal signalling inhibition.

Science.gov (United States)

Koel, Mariann; Võsa, Urmo; Krjutškov, Kaarel; Einarsdottir, Elisabet; Kere, Juha; Tapanainen, Juha; Katayama, Shintaro; Ingerpuu, Sulev; Jaks, Viljar; Stenman, Ulf-Hakan; Lundin, Karolina; Tuuri, Timo; Salumets, Andres

2017-09-01

Several studies have demonstrated that human embryonic stem cells (hESC) can be differentiated into trophoblast-like cells if exposed to bone morphogenic protein 4 (BMP4) and/or inhibitors of fibroblast growth factor 2 (FGF2) and the transforming growth factor beta (TGF-β)/activin/nodal signalling pathways. The goal of this study was to investigate how the inhibitors of these pathways improve the efficiency of hESC differentiation when compared with basic BMP4 treatment. RNA sequencing was used to analyse the effects of all possible inhibitor combinations on the differentiation of hESC into trophoblast-like cells over 12 days. Genes differentially expressed compared with untreated cells were identified at seven time points. Additionally, expression of total human chorionic gonadotrophin (HCG) and its hyperglycosylated form (HCG-H) were determined by immunoassay from cell culture media. We showed that FGF2 inhibition with BMP4 activation up-regulates syncytiotrophoblast-specific genes (CGA, CGB and LGALS16), induces several molecular pathways involved in embryo implantation and triggers HCG-H production. In contrast, inhibition of the TGF-β/activin/nodal pathway decreases the ability of hESC to form trophoblast-like cells. Information about the conditions needed for hESC differentiation toward trophoblast-like cells helps us to find an optimal model for studying the early development of human trophoblasts in normal and in complicated pregnancy. Copyright © 2017 Reproductive Healthcare Ltd. Published by Elsevier Ltd. All rights reserved.

A study of the literature on nodal methods in reactor physics calculations

International Nuclear Information System (INIS)

Van de Wetering, T.F.H.

1993-01-01

During the last few decades several calculation methods have been developed for the three-dimensional analysis of a reactor core. A literature survey was carried out to gain insights in the starting points and method of operation of the advanced nodal methods. These methods are applied in reactor core analyses of large nuclear power reactors, because of their high computing speed. The so-called Nodal-Expansion method is described in detail

Hydrogen transport in a toroidal plasma using multigroup discrete-ordinates methodology

International Nuclear Information System (INIS)

Wienke, B.R.; Miller, W.F. Jr.; Seed, T.J.

1979-01-01

Neutral hydrogen transport in a fully ionized two-dimensional tokamak plasma was examined using discrete ordinates and contrasted with earlier analyses. In particular, curvature effects induced by toroidal geometries and ray effects caused by possible source localization were investigated. From an overview of the multigroup discrete-ordinates approximation, methodology in two-dimensional cylindrical geometry is detailed, mesh and plasma zoning procedures are sketched, and the piecewise polynomial solution algorithm on a triangular domain is obtained. Toroidal effects and comparisons as related to reaction rates and perticle spectra are examined for various model and source configurations

Multigroup constants for charged particle elastic nuclear (plus interference) scattering of light isotopes

International Nuclear Information System (INIS)

Cullen, D.E.; Perkins, S.T.

1977-01-01

Multi-group averaged reaction rates and transfer matrices were calculated for charged particle induced elastic nuclear (plus interference) scattering. Results are presented using a ten group structure for all twenty-five permutations of projectile and target for the following charged particles: p, d, t, 3 He and alpha. Transfer matrices are presented in a simplified form for both incident projectile and the knock-ons; these matrices explicitly conserve energy

Results of modeling advanced BWR fuel designs using CASMO-4

International Nuclear Information System (INIS)

Knott, D.; Edenius, M.

1996-01-01

Advanced BWR fuel designs from General Electric, Siemens and ABB-Atom have been analyzed using CASMO-4 and compared against fission rate distributions and control rod worths from MCNP. Included in the analysis were fuel storage rack configurations and proposed mixed oxide (MOX) designs. Results are also presented from several cycles of SIMULATE-3 core follow analysis, using nodal data generated by CASMO-4, for cycles in transition from 8x8 designs to advanced fuel designs. (author)

Interplay between short-range correlated disorder and Coulomb interaction in nodal-line semimetals

Science.gov (United States)

Wang, Yuxuan; Nandkishore, Rahul M.

2017-09-01

In nodal-line semimetals, Coulomb interactions and short-range correlated disorder are both marginal perturbations to the clean noninteracting Hamiltonian. We analyze their interplay using a weak-coupling renormalization group approach. In the clean case, the Coulomb interaction has been found to be marginally irrelevant, leading to Fermi liquid behavior. We extend the analysis to incorporate the effects of disorder. The nodal line structure gives rise to kinematical constraints similar to that for a two-dimensional Fermi surface, which plays a crucial role in the one-loop renormalization of the disorder couplings. For a twofold degenerate nodal loop (Weyl loop), we show that disorder flows to strong coupling along a unique fixed trajectory in the space of symmetry inequivalent disorder couplings. Along this fixed trajectory, all symmetry inequivalent disorder strengths become equal. For a fourfold degenerate nodal loop (Dirac loop), disorder also flows to strong coupling, however, the strengths of symmetry inequivalent disorder couplings remain different. We show that feedback from disorder reverses the sign of the beta function for the Coulomb interaction, causing the Coulomb interaction to flow to strong coupling as well. However, the Coulomb interaction flows to strong coupling asymptotically more slowly than disorder. Extrapolating our results to strong coupling, we conjecture that at low energies nodal line semimetals should be described by a noninteracting nonlinear sigma model. We discuss the relation of our results with possible many-body localization at zero temperatures in such materials.

COMPAR: A system for comparing multigroup cross-sections generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS

International Nuclear Information System (INIS)

Anaf, J.; Chalhoub, E.S.

1988-02-01

A system consisting of the COMPAR computer program and its interfaces which was developed for comparing multigroup cross-sections generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS is presented. (author). 13 refs

Some reciprocity-like relations in multi-group neutron diffusion and transport theory over bare homogeneous regions

International Nuclear Information System (INIS)

Modak, R.S.; Sahni, D.C.

1996-01-01

Some simple reciprocity-like relations that exist in multi-group neutron diffusion and transport theory over bare homogeneous regions are presented. These relations do not involve the adjoint solutions and are directly related to numerical schemes based on an explicit evaluation of the fission matrix. (author)

Delineation of Internal Mammary Nodal Target Volumes in Breast Cancer Radiation Therapy

Energy Technology Data Exchange (ETDEWEB)

Jethwa, Krishan R.; Kahila, Mohamed M. [Department of Radiation Oncology, Mayo Clinic, Rochester, Minnesota (United States); Hunt, Katie N. [Department of Radiology, Mayo Clinic, Rochester, Minnesota (United States); Brown, Lindsay C.; Corbin, Kimberly S.; Park, Sean S.; Yan, Elizabeth S. [Department of Radiation Oncology, Mayo Clinic, Rochester, Minnesota (United States); Boughey, Judy C. [Department of Surgery, Mayo Clinic, Rochester, Minnesota (United States); Mutter, Robert W., E-mail: mutter.robert@mayo.edu [Department of Radiation Oncology, Mayo Clinic, Rochester, Minnesota (United States)

2017-03-15

Purpose: The optimal clinical target volume for internal mammary (IM) node irradiation is uncertain in an era of increasingly conformal volume-based treatment planning for breast cancer. We mapped the location of gross internal mammary lymph node (IMN) metastases to identify areas at highest risk of harboring occult disease. Methods and Materials: Patients with axial imaging of IMN disease were identified from a breast cancer registry. The IMN location was transferred onto the corresponding anatomic position on representative axial computed tomography images of a patient in the treatment position and compared with consensus group guidelines of IMN target delineation. Results: The IMN location in 67 patients with 130 IMN metastases was mapped. The location was in the first 3 intercostal spaces in 102 of 130 nodal metastases (78%), whereas 18 of 130 IMNs (14%) were located caudal to the third intercostal space and 10 of 130 IMNs (8%) were located cranial to the first intercostal space. Of the 102 nodal metastases within the first 3 intercostal spaces, 54 (53%) were located within the Radiation Therapy Oncology Group consensus volume. Relative to the IM vessels, 19 nodal metastases (19%) were located medially with a mean distance of 2.2 mm (SD, 2.9 mm) whereas 29 (28%) were located laterally with a mean distance of 3.6 mm (SD, 2.5 mm). Ninety percent of lymph nodes within the first 3 intercostal spaces would have been encompassed within a 4-mm medial and lateral expansion on the IM vessels. Conclusions: In women with indications for elective IMN irradiation, a 4-mm medial and lateral expansion on the IM vessels may be appropriate. In women with known IMN involvement, cranial extension to the confluence of the IM vein with the brachiocephalic vein with or without caudal extension to the fourth or fifth interspace may be considered provided that normal tissue constraints are met.

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

One-, two- and three-dimensional transport codes using multi-group double-differential form cross sections

International Nuclear Information System (INIS)

Mori, Takamasa; Nakagawa, Masayuki; Sasaki, Makoto.

1988-11-01

We have developed a group of computer codes to realize the accurate transport calculation by using the multi-group double-differential form cross section. This type of cross section can correctly take account of the energy-angle correlated reaction kinematics. Accordingly, the transport phenomena in materials with highly anisotropic scattering are accurately calculated by using this cross section. They include the following four codes or code systems: PROF-DD : a code system to generate the multi-group double-differential form cross section library by processing basic nuclear data file compiled in the ENDF / B-IV or -V format, ANISN-DD : a one-dimensional transport code based on the discrete ordinate method, DOT-DD : a two-dimensional transport code based on the discrete ordinate method, MORSE-DD : a three-dimensional transport code based on the Monte Carlo method. In addition to these codes, several auxiliary codes have been developed to process calculated results. This report describes the calculation algorithm employed in these codes and how to use them. (author)

Improved Fiber Bragg Grating Array OFFH-CDMA System Using a Novel Frequency-Overlapping Multigroup Method

Science.gov (United States)

Peng, Wei-Ren; Lin, Wen-Piao; Chi, Sien

2006-03-01

The authors propose a novel frequency-overlapping multigroup scheme for a passive all-optical fast-frequency hopped code-division multiple-access (OFFH-CDMA) system based on fiber Bragg grating array (FBGA). In the conventional scheme, the users are assigned those codes constructed on the nonoverlapping frequency slots, and therefore the bandgaps between the adjacent gratings are wasted. To make a more efficient use of the optical spectrum, the proposed scheme divided the users into several groups, and assigned the codes, which interleaved to each other to the different groups. In addition to the higher utilization of the spectrum, the interleaved nature of the frequency allocations of different groups will make the groups less correlated and, hence, lower the multiple-access interference (MAI). The corresponding codeset and its constraints for this new scheme are also developed and analyzed. The performance of the system in terms of the correlation functions and bit error rate (BER) are given in both the conventional and the proposed schemes. The numerical results show that, with the multigroup scheme, performance is much improved compared to the conventional scheme.

Adjustement of multigroup cross sections using fast reactor integral data

International Nuclear Information System (INIS)

Renke, C.A.C.

1982-01-01

A methodology for the adjustment of multigroup cross section is presented, structured with aiming to compatibility the limitated number of measured values of integral parameters known and disponible, and the great number of cross sections to be adjusted the group of cross section used is that obtained from the Carnaval II calculation system, understanding as formular the sets of calculation methods and data bases. The adjustment is realized, using the INCOAJ computer code, developed in function of one statistical formulation, structural from the bayer considerations, taking in account the measurement processes of cross section and integral parameters defined on statistical bases. (E.G.) [pt

Validation of multigroup neutron cross sections and calculational methods for the advanced neutron source against the FOEHN critical experiments measurements

International Nuclear Information System (INIS)

Smith, L.A.; Gallmeier, F.X.; Gehin, J.C.

1995-05-01

The FOEHN critical experiment was analyzed to validate the use of multigroup cross sections and Oak Ridge National Laboratory neutronics computer codes in the design of the Advanced Neutron Source. The ANSL-V 99-group master cross section library was used for all the calculations. Three different critical configurations were evaluated using the multigroup KENO Monte Carlo transport code, the multigroup DORT discrete ordinates transport code, and the multigroup diffusion theory code VENTURE. The simple configuration consists of only the fuel and control elements with the heavy water reflector. The intermediate configuration includes boron endplates at the upper and lower edges of the fuel element. The complex configuration includes both the boron endplates and components in the reflector. Cross sections were processed using modules from the AMPX system. Both 99-group and 20-group cross sections were created and used in two-dimensional models of the FOEHN experiment. KENO calculations were performed using both 99-group and 20-group cross sections. The DORT and VENTURE calculations were performed using 20-group cross sections. Because the simple and intermediate configurations are azimuthally symmetric, these configurations can be explicitly modeled in R-Z geometry. Since the reflector components cannot be modeled explicitly using the current versions of these codes, three reflector component homogenization schemes were developed and evaluated for the complex configuration. Power density distributions were calculated with KENO using 99-group cross sections and with DORT and VENTURE using 20-group cross sections. The average differences between the measured values and the values calculated with the different computer codes range from 2.45 to 5.74%. The maximum differences between the measured and calculated thermal flux values for the simple and intermediate configurations are ∼ 13%, while the average differences are < 8%

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

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

A variational nodal diffusion method of high accuracy; Varijaciona nodalna difuziona metoda visoke tachnosti

Energy Technology Data Exchange (ETDEWEB)

Tomasevic, Dj; Altiparmarkov, D [Institut za Nuklearne Nauke Boris Kidric, Belgrade (Yugoslavia)

1988-07-01

A variational nodal diffusion method with accurate treatment of transverse leakage shape is developed and presented in this paper. Using Legendre expansion in transverse coordinates higher order quasi-one-dimensional nodal equations are formulated. Numerical solution has been carried out using analytical solutions in alternating directions assuming Legendre expansion of the RHS term. The method has been tested against 2D and 3D IAEA benchmark problem, as well as 2D CANDU benchmark problem. The results are highly accurate. The first order approximation yields to the same order of accuracy as the standard nodal methods with quadratic leakage approximation, while the second order reaches reference solution. (author)

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)

A lumped parameter method of characteristics approach and multigroup kernels applied to the subgroup self-shielding calculation in MPACT

International Nuclear Information System (INIS)

Stimpson, Shane G.; Liu, Yuxuan; Collins, Benjamin S.; Clarno, Kevin T.

2017-01-01

An essential component of the neutron transport solver is the resonance self-shielding calculation used to determine equivalence cross sections. The neutron transport code, MPACT, is currently using the subgroup self-shielding method, in which the method of characteristics (MOC) is used to solve purely absorbing fixed-source problems. Recent efforts incorporating multigroup kernels to the MOC solvers in MPACT have reduced runtime by roughly 2×. Applying the same concepts for self-shielding and developing a novel lumped parameter approach to MOC, substantial improvements have also been made to the self-shielding computational efficiency without sacrificing any accuracy. These new multigroup and lumped parameter capabilities have been demonstrated on two test cases: (1) a single lattice with quarter symmetry known as VERA (Virtual Environment for Reactor Applications) Progression Problem 2a and (2) a two-dimensional quarter-core slice known as Problem 5a-2D. From these cases, self-shielding computational time was reduced by roughly 3–4×, with a corresponding 15–20% increase in overall memory burden. An azimuthal angle sensitivity study also shows that only half as many angles are needed, yielding an additional speedup of 2×. In total, the improvements yield roughly a 7–8× speedup. Furthermore given these performance benefits, these approaches have been adopted as the default in MPACT.

A lumped parameter method of characteristics approach and multigroup kernels applied to the subgroup self-shielding calculation in MPACT

Directory of Open Access Journals (Sweden)

Shane Stimpson

2017-09-01

Full Text Available An essential component of the neutron transport solver is the resonance self-shielding calculation used to determine equivalence cross sections. The neutron transport code, MPACT, is currently using the subgroup self-shielding method, in which the method of characteristics (MOC is used to solve purely absorbing fixed-source problems. Recent efforts incorporating multigroup kernels to the MOC solvers in MPACT have reduced runtime by roughly 2×. Applying the same concepts for self-shielding and developing a novel lumped parameter approach to MOC, substantial improvements have also been made to the self-shielding computational efficiency without sacrificing any accuracy. These new multigroup and lumped parameter capabilities have been demonstrated on two test cases: (1 a single lattice with quarter symmetry known as VERA (Virtual Environment for Reactor Applications Progression Problem 2a and (2 a two-dimensional quarter-core slice known as Problem 5a-2D. From these cases, self-shielding computational time was reduced by roughly 3–4×, with a corresponding 15–20% increase in overall memory burden. An azimuthal angle sensitivity study also shows that only half as many angles are needed, yielding an additional speedup of 2×. In total, the improvements yield roughly a 7–8× speedup. Given these performance benefits, these approaches have been adopted as the default in MPACT.

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

. In this geometry nodal, continuous and discontinuous schemes were used. For the continuos schemes, only the Bi Quadratic (BiQ) and the Bi Cubic (BiC) were considered. In the case of the discontinuous ones two nodal schemes were considered, namely the Discontinuous Bi Linear (DBiL) and Discontinuous Bi Quadratic (DBiQ). The nodal schemes applied use from 4 up to 16 interpolation parameters per cell. These schemes are-defined for a set D c of interpolation parameters and a polynomial space S h corresponding to each one of the nodal schemes considered. All these four nodal hybrid schemes were implemented in a computer program called TNHXY starting from the computer program TNXY developed in previous thesis works. Several subroutines wae added to calculate the average neutron flux for each cell and for each energy group, generating two versions, one for the continuous schemes and one for the discontinuous schemes. For this geometry, two benchmark problems of the ANL-7416 document were analyzed. They are 7x7 BWR fuel assemblies, one without control rod and the other one with control rod. The computer program was also applied to a MOX assembly proposed by the Nuclear Energy Agency and it is considered as a reference problem. The results obtained for the one-dimensional problems using TNX for the effective multiplication factor were compared with the ones obtained with the code ANISN/PC. TNX code shows a faster convergence within four significant figures for the case with no control rod and three significant figures for the case with control rod (using double precision). These results suggest TNX is a very useful tool for this kind of calculations. For X Y geometry, the results obtained with TNHXY were compared with those calculated with the code TWOTRAN. To get these results, several spatial (1x1, 2x2, 4x4 per cell) and angular meshes (S 2 , S 4 , S 6 , and S 8 ) were used. The results for the problem with no control rod were practically the same as those obtained with

Development and qualification of a thermal-hydraulic nodalization for modeling station blackout accident in PSB-VVER test facility

Energy Technology Data Exchange (ETDEWEB)

Saghafi, Mahdi [Department of Energy Engineering, Sharif University of Technology, Azadi Avenue, Tehran (Iran, Islamic Republic of); Ghofrani, Mohammad Bagher, E-mail: ghofrani@sharif.edu [Department of Energy Engineering, Sharif University of Technology, Azadi Avenue, Tehran (Iran, Islamic Republic of); D’Auria, Francesco [San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa, Via Livornese 1291, San Piero a Grado, Pisa (Italy)

2016-07-15

Highlights: • A thermal-hydraulic nodalization for PSB-VVER test facility has been developed. • Station blackout accident is modeled with the developed nodalization in MELCOR code. • The developed nodalization is qualified at both steady state and transient levels. • MELCOR predictions are qualitatively and quantitatively in acceptable range. • Fast Fourier Transform Base Method is used to quantify accuracy of code predictions. - Abstract: This paper deals with the development of a qualified thermal-hydraulic nodalization for modeling Station Black-Out (SBO) accident in PSB-VVER Integral Test Facility (ITF). This study has been performed in the framework of a research project, aiming to develop an appropriate accident management support tool for Bushehr nuclear power plant. In this regard, a nodalization has been developed for thermal-hydraulic modeling of the PSB-VVER ITF by MELCOR integrated code. The nodalization is qualitatively and quantitatively qualified at both steady-state and transient levels. The accuracy of the MELCOR predictions is quantified in the transient level using the Fast Fourier Transform Base Method (FFTBM). FFTBM provides an integral representation for quantification of the code accuracy in the frequency domain. It was observed that MELCOR predictions are qualitatively and quantitatively in the acceptable range. In addition, the influence of different nodalizations on MELCOR predictions was evaluated and quantified using FFTBM by developing 8 sensitivity cases with different numbers of control volumes and heat structures in the core region and steam generator U-tubes. The most appropriate case, which provided results with minimum deviations from the experimental data, was then considered as the qualified nodalization for analysis of SBO accident in the PSB-VVER ITF. This qualified nodalization can be used for modeling of VVER-1000 nuclear power plants when performing SBO accident analysis by MELCOR code.

Pressure-induced organic topological nodal-line semimetal in the three-dimensional molecular crystal Pd (dddt) 2

Science.gov (United States)

Liu, Zhao; Wang, Haidi; Wang, Z. F.; Yang, Jinlong; Liu, Feng

2018-04-01

The nodal-line semimetal represents a class of topological materials characterized with highest band degeneracy. It is usually found in inorganic materials of high crystal symmetry or a minimum symmetry of inversion aided with accidental band degeneracy [Phys. Rev. Lett. 118, 176402 (2017), 10.1103/PhysRevLett.118.176402]. Based on first-principles band structure, Wannier charge center, and topological surface state calculations, here we predict a pressure-induced topological nodal-line semimetal in the absence of spin-orbit coupling (SOC) in the synthesized single-component 3D molecular crystal Pd (dddt) 2 . We show a Γ -centered single nodal line undulating within a narrow energy window across the Fermi level. This intriguing nodal line is generated by pressure-induced accidental band degeneracy, without protection from any crystal symmetry. When SOC is included, the fourfold degenerated nodal line is gapped and Pd (dddt) 2 becomes a strong 3D topological metal with an Z2 index of (1;000). However, the tiny SOC gap makes it still possible to detect the nodal-line properties experimentally. Our findings afford an attractive route for designing and realizing topological states in 3D molecular crystals, as they are weakly bonded through van der Waals forces with a low crystal symmetry so that their electronic structures can be easily tuned by pressure.

Role of CT/PET in predicting nodal disease in head and neck cancers

International Nuclear Information System (INIS)

Singham, S.; Iyer, G.; Clark, J.

2009-01-01

Full text:Introduction: Pre-treatment evaluation of the presence of cervical nodal metastases is important in head and neck cancers and has major prognostic implications. In this study, we aim to determine the accuracy of CT/PET as a tool for identifying such metastases. Methods: All patients from Royal Prince Alfred and Liverpool Hospitals, who underwent CT/PET for any cancer arising from the head and neck, and who underwent subsequent surgery (which included a neck dissection) within 8 weeks of the CT/PET were included. Nodal staging was undertaken by utilising imaging-based nodal classification, and comparison with pathologic data from the surgical specimen was made. PET was considered positive if the SUV was greater than 2. Results: We identified 111 patients from the above criteria. 80 of such patients were treated for squamous cell carcinoma (SCC). CT/PET identified unsuspected metastatic disease in 6 patients. Correlation of CT/PET findings and the presence of disease at the primary site: sensitivity: 98%, specificity: 93%, positive predictive value (PPV): 98% and negative predictive value (NPV): 93%. Correlating CT/PET findings with the presence of nodal disease at any level: sensitivity: 95%, specificity: 88%, PPV: 95% and NPV: 88%. CT/PET was anatomically accurate in predicting the site of metastases in 62/74 (84%). Conclusion: PET is accurate in predicting both presence of nodal metastases and the level of involvement. CT/PET should be undertaken as a pre-operative tool to assist in planning the extent of surgery required in head and neck cancers.

Risk of isolated nodal failure for non-small cell lung cancer (NSCLC) treated with the elective nodal irradiation (ENI) using 3D-conformal radiotherapy (3D-CRT) techniques--a retrospective analysis.

Science.gov (United States)

Kepka, Lucyna; Bujko, Krzysztof; Zolciak-Siwinska, Agnieszka

2008-01-01

To estimate retrospectively the rate of isolated nodal failures (INF) in NSCLC patients treated with the elective nodal irradiation (ENI) using 3D-conformal radiotherapy (3D-CRT). One hundred and eighty-five patients with I-IIIB stage treated with 3D-CRT in consecutive clinical trials differing in an extent of the ENI were analyzed. According to the extent of the ENI, two groups were distinguished: extended (n = 124) and limited (n = 61) ENI. INF was defined as regional nodal failure occurring without local progression. Cumulative Incidence of INF (CIINF) was evaluated by univariate and multivariate analysis with regard to prognostic factors. With a median follow up of 30 months, the two-year actuarial overall survival was 35%. The two-year CIINF rate was 12%. There were 16 (9%) INF, eight (6%) for extended and eight (13%) for limited ENI. In the univariate analysis bulky mediastinal disease (BMD), left side, higher N stage, and partial response to RT had a significant negative impact on the CIINF. BMD was the only independent predictor of the risk of incidence of the INF (p = 0.001). INF is more likely to occur in case of more advanced nodal status.

Intensity-Modulated Proton Therapy for Elective Nodal Irradiation and Involved-Field Radiation in the Definitive Treatment of Locally Advanced Non-Small Cell Lung Cancer: A Dosimetric Study

Science.gov (United States)

Kesarwala, Aparna H.; Ko, Christine J.; Ning, Holly; Xanthopoulos, Eric; Haglund, Karl E.; O’Meara, William P.; Simone, Charles B.; Rengan, Ramesh

2015-01-01

Background Photon involved-field radiation therapy (IFRT), the standard for locally advanced non-small cell lung cancer (LA-NSCLC), results in favorable outcomes without increased isolated nodal failures, perhaps from scattered dose to elective nodal stations. Given the high conformality of intensity-modulated proton therapy (IMPT), proton IFRT could increase nodal failures. We investigated the feasibility of IMPT for elective nodal irradiation (ENI) in LA-NSCLC. Materials and Methods IMPT IFRT plans were generated to the same total dose of 66.6–72 Gy received by 20 LA-NSCLC patients treated with photon IFRT. IMPT ENI plans were generated to 46 CGE to elective nodal (EN) planning treatment volumes (PTV) plus 24 CGE to involved field (IF)-PTVs. Results Proton IFRT and ENI both improved D95 involved field (IF)-PTV coverage by 4% (pENI. Mean esophagus dose decreased 16% with IFRT and 12% with ENI; heart V25 decreased 63% with both (all pENI. Potential decreased toxicity indicates IMPT could allow ENI while maintaining a favorable therapeutic ratio compared to photon IFRT. PMID:25604729

Testing a new multigroup inference approach to reconstructing past environmental conditions

Directory of Open Access Journals (Sweden)

Maria RIERADEVALL

2008-08-01

Full Text Available A new, quantitative, inference model for environmental reconstruction (transfer function, based for the first time on the simultaneous analysis of multigroup species, has been developed. Quantitative reconstructions based on palaeoecological transfer functions provide a powerful tool for addressing questions of environmental change in a wide range of environments, from oceans to mountain lakes, and over a range of timescales, from decades to millions of years. Much progress has been made in the development of inferences based on multiple proxies but usually these have been considered separately, and the different numeric reconstructions compared and reconciled post-hoc. This paper presents a new method to combine information from multiple biological groups at the reconstruction stage. The aim of the multigroup work was to test the potential of the new approach to making improved inferences of past environmental change by improving upon current reconstruction methodologies. The taxonomic groups analysed include diatoms, chironomids and chrysophyte cysts. We test the new methodology using two cold-environment training-sets, namely mountain lakes from the Pyrenees and the Alps. The use of multiple groups, as opposed to single groupings, was only found to increase the reconstruction skill slightly, as measured by the root mean square error of prediction (leave-one-out cross-validation, in the case of alkalinity, dissolved inorganic carbon and altitude (a surrogate for air-temperature, but not for pH or dissolved CO2. Reasons why the improvement was less than might have been anticipated are discussed. These can include the different life-forms, environmental responses and reaction times of the groups under study.

On an analytical evaluation of the flux and dominant eigenvalue problem for the steady state multi-group multi-layer neutron diffusion equation

Energy Technology Data Exchange (ETDEWEB)

Ceolin, Celina; Schramm, Marcelo; Bodmann, Bardo Ernst Josef; Vilhena, Marco Tullio Mena Barreto de [Universidade Federal do Rio Grande do Sul, Porto Alegre (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica; Bogado Leite, Sergio de Queiroz [Comissao Nacional de Energia Nuclear, Rio de Janeiro (Brazil)

2014-11-15

In this work the authors solved the steady state neutron diffusion equation for a multi-layer slab assuming the multi-group energy model. The method to solve the equation system is based on an expansion in Taylor Series resulting in an analytical expression. The results obtained can be used as initial condition for neutron space kinetics problems. The neutron scalar flux was expanded in a power series, and the coefficients were found by using the ordinary differential equation and the boundary and interface conditions. The effective multiplication factor k was evaluated using the power method. We divided the domain into several slabs to guarantee the convergence with a low truncation order. We present the formalism together with some numerical simulations.

Using Multi-Group Confirmatory Factor Analysis to Evaluate Cross-Cultural Research: Identifying and Understanding Non-Invariance

Science.gov (United States)

Brown, Gavin T. L.; Harris, Lois R.; O'Quin, Chrissie; Lane, Kenneth E.

2017-01-01

Multi-group confirmatory factor analysis (MGCFA) allows researchers to determine whether a research inventory elicits similar response patterns across samples. If statistical equivalence in responding is found, then scale score comparisons become possible and samples can be said to be from the same population. This paper illustrates the use of…

Some topics on safety analysis and accident nodalization of CAREM-25

International Nuclear Information System (INIS)

Gimenez, Marcelo O.; Zanocco, Pablo; Schlamp, Miguel A.; Ottaviani, Anahi; Garcia, Alicia

2000-01-01

The main goal of nuclear safety area in the CAREM Project Phase I, carried out during 1999, was to consolidate the safety systems design through an integral analysis of the reactor and the safety systems response to different accidental sequences. A primary circuit nodalization, including the steam generators, was done with RELAP5 code. The modeling of System 230 (absorber rods drive feed water system), System 1400 (purification and control volume system) and steam condensation on the absorber rods drive system and on RPV wall is implemented through boundary conditions. Also the Residual Heat Removal System and the Second Shutdown system are modeled. The reactor steady state at full power was calculated. The results agree quite well with design values. It can be said from the accident analysis that the nodalization responds properly. Further analysis should be done in order to qualify the nodalization and to compare benchmarks with other codes and experimental data. On the other hand, the steam dome model should be improved with more precise data about absorber rods drive system condensation, loss of heat and inner components layout. (author)

NUMERICAL MULTIGROUP TRANSIENT ANALYSIS OF SLAB NUCLEAR REACTOR WITH THERMAL FEEDBACK

Directory of Open Access Journals (Sweden)

Filip Osuský

2016-12-01

Full Text Available The paper describes a new numerical code for multigroup transient analyses with thermal feedback. The code is developed at Institute of Nuclear and Physical Engineering. It is necessary to carefully investigate transient states of fast neutron reactors, due to recriticality issues after accident scenarios. The code solves numerical diffusion equation for 1D problem with possible neutron source incorporation. Crank-Nicholson numerical method is used for the transient states. The investigated cases are describing behavior of PWR fuel assembly inside of spent fuel pool and with the incorporated neutron source for better illustration of thermal feedback.

Mapping of selected markets with Nodal pricing or similar systems. Australia, New Zealand and North American power markets

Energy Technology Data Exchange (ETDEWEB)

Mathiesen, Vivi (ed.)

2011-07-01

This report shows that the principals of nodal pricing can be implemented in different ways. A common denominator for markets with nodal pricing is a central market based nodal dispatch, where prices and flows are determined simultaneously close to real time. This stands apart from the European market design, which is based on a highly simplified version of the grid, and a physical point auction day ahead. Congestion management is handled by the TSO during the operational hour and not through the market as is the case in nodal pricing systems. Nodal pricing yields optimal dispatch and congestion management through the market, and as such an optimal utilisation of energy generation and network. However, whether this short term optimisation delivers the highest overall efficiency for the market in terms of competition in the wholesale and retail market, price discovery, possibilities for hedging, long term price signals etc. is difficult to determine. The markets investigated handle issues such as market power, risk management, investment signals and retail markets in very different ways. New Zealand and PJM are examples of markets with full nodal pricing, i.e. both generators and the demand side are exposed to nodal prices. The PJM market has more 'additional features' than the New Zealand market. Examples of these are separate capacity market to trigger investments in generation and generator price caps to deal with situations of market power. In addition PJM offers liquid and mature markets for risk management, such as aggregates of nodes where market participant can chose to be settled (rather than to be settled directly at the node). A general finding though, seems to be that risk management at peripheral nodes is challenging in nodal markets, particularly for independent retailers. In New Zealand generators and retailers were permitted to 'reintegrate' in order to cope with the nodal prices. The Australian market has central market based

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.

Numerical divergence effects of equivalence theory in the nodal expansion method

International Nuclear Information System (INIS)

Zika, M.R.; Downar, T.J.

1993-01-01

Accurate solutions of the advanced nodal equations require the use of discontinuity factors (DFs) to account for the homogenization errors that are inherent in all coarse-mesh nodal methods. During the last several years, nodal equivalence theory (NET) has successfully been implemented for the Cartesian geometry and has received widespread acceptance in the light water reactor industry. The extension of NET to other reactor types has had limited success. Recent efforts to implement NET within the framework of the nodal expansion method have successfully been applied to the fast breeder reactor. However, attempts to apply the same methods to thermal reactors such as the Modular High-Temperature Gas Reactor (MHTGR) have led to numerical divergence problems that can be attributed directly to the magnitude of the DFs. In the work performed here, it was found that the numerical problems occur in the inner and upscatter iterations of the solution algorithm. These iterations use a Gauss-Seidel iterative technique that is always convergent for problems with unity DFs. However, for an MHTGR model that requires large DFs, both the inner and upscatter iterations were divergent. Initial investigations into methods for bounding the DFs have proven unsatisfactory as a means of remedying the convergence problems. Although the DFs could be bounded to yield a convergent solution, several cases were encountered where the resulting flux solution was less accurate than the solution without DFs. For the specific case of problems without upscattering, an alternate numerical method for the inner iteration, an LU decomposition, was identified and shown to be feasible

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

A proposal for combined MRI and PET/CT interpretation criteria for preoperative nodal staging in non-small-cell lung cancer

International Nuclear Information System (INIS)

Kim, Yoo Na; Yi, Chin A.; Lee, Kyung Soo; Lee, Ho Yun; Kim, Tae Sung; Chung, Myung Jin; Kwon, O.Jung; Chung, Man Pyo; Kim, Byung-Tae; Choi, Joon Young; Kim, Seon Woo; Han, Joungho; Shim, Young Mog

2012-01-01

To determine the positive reading criteria for malignant nodes when interpreting combined MRI and PET/CT images for preoperative nodal staging in non-small-cell lung cancer (NSCLC). Forty-nine patients with biopsy-proven NSCLC underwent both PET/CT and thoracic MRI [diffusion weighted imaging (DWI)]. Each nodal station was evaluated for the presence of metastasis by applying either inclusive (positive if either one read positive) or exclusive (positive if both read positive) criteria in the combined interpretation of PET/CT and MRI. Nodal stage was confirmed pathologically. The combined diagnostic accuracy of PET/CT and MRI was determined on per-nodal station and per-patient bases and compared with that of PET/CT alone. In 49 patients, 39 (19%) of 206 nodal stations harboured malignant cells. Out of 206 nodal stations, 186 (90%) had concordant readings, while the rest (10%) had discordant readings. Inclusive criteria of combined PET/CT and MRI helped increase sensitivity for detecting nodal metastasis (69%) compared with PET/CT alone (46%; P = 0.003), while specificity was not significantly decreased. Inclusive criteria in combined MRI and PET/CT readings help improve significantly the sensitivity for detecting nodal metastasis compared with PET/CT alone and may decrease unnecessary open thoracotomy. (orig.)

TRIMARAN: a three dimensional multigroup P1 Monte Carlo code for criticality studies

International Nuclear Information System (INIS)

Ermumcu, G.; Gonnord, J.; Nimal, J.C.

1980-01-01

TRIMARAN is developed for safety analysis of nuclear components containing fissionable materials: shipping casks, storage and cooling pools, manufacture and reprocessing plants. It solves the transport equation by Monte Carlo method, in general three dimensional geometry with multigroup P1 approximation. A special representation of cross sections and numbers has been developed in order to reduce considerably the computing cost and allow this three dimensional code to compete with standard numerical program used in parametric studies

TRIMARAN: a three dimensional multigroup P1 Monte Carlo code for criticallity studies

International Nuclear Information System (INIS)

Ermuncu, G.; Gonnord, J.; Nimal, J.C.

1980-04-01

TRIMARAN is developed for safety analysis of nuclar components containing fissionnable materials: shipping casks, storage and cooling pools, manufacture and reprocessing plants. It solves the transport equation by Monte Carlo method in general three dimensional geometry with multigroup P1 approximation. A special representation of cross sections and numbers has been developed in order to reduce considerably the computing cost and allow this three dimensional code to compete with standard numerical program used in parametric studies

Assessment and comparison of different multigroup neutron cross section libraries for dosimetry purposes

International Nuclear Information System (INIS)

Erradi, L.; Karouani, K.

1994-01-01

Many multigroup neutron cross section libraries have been processed from basic evaluated nuclear data for use in neutron dosimetry, reactor shielding calculation and in the development of fusion reactors. Most of these libraries have been tested only for fission spectra and were not validated for fusion spectra. Fifteen of these libraries such as DOSCROS84, IRDF85 and ENDFB5 have been used along with the neutron spectra unfolding code SAND II to evaluate about fifteen threshold detector saturated activities. The comparison between these computed activities and the measured ones of a set of foils placed in different places along the axis of a paraffin cylinder and irradiated by 14 MeV neutrons generated by a D-T source, hence giving rise to complex spectra, leads to different types of discrepancies. The analysis of these discrepancies allows to select from these libraries the ones that can be recommended. 1 fig., 4 refs. (author)

Ancestral regulatory circuits governing ectoderm patterning downstream of Nodal and BMP2/4 revealed by gene regulatory network analysis in an echinoderm.

Directory of Open Access Journals (Sweden)

Alexandra Saudemont

2010-12-01

Full Text Available Echinoderms, which are phylogenetically related to vertebrates and produce large numbers of transparent embryos that can be experimentally manipulated, offer many advantages for the analysis of the gene regulatory networks (GRN regulating germ layer formation. During development of the sea urchin embryo, the ectoderm is the source of signals that pattern all three germ layers along the dorsal-ventral axis. How this signaling center controls patterning and morphogenesis of the embryo is not understood. Here, we report a large-scale analysis of the GRN deployed in response to the activity of this signaling center in the embryos of the Mediterranean sea urchin Paracentrotus lividus, in which studies with high spatial resolution are possible. By using a combination of in situ hybridization screening, overexpression of mRNA, recombinant ligand treatments, and morpholino-based loss-of-function studies, we identified a cohort of transcription factors and signaling molecules expressed in the ventral ectoderm, dorsal ectoderm, and interposed neurogenic ("ciliary band" region in response to the known key signaling molecules Nodal and BMP2/4 and defined the epistatic relationships between the most important genes. The resultant GRN showed a number of striking features. First, Nodal was found to be essential for the expression of all ventral and dorsal marker genes, and BMP2/4 for all dorsal genes. Second, goosecoid was identified as a central player in a regulatory sub-circuit controlling mouth formation, while tbx2/3 emerged as a critical factor for differentiation of the dorsal ectoderm. Finally, and unexpectedly, a neurogenic ectoderm regulatory circuit characterized by expression of "ciliary band" genes was triggered in the absence of TGF beta signaling. We propose a novel model for ectoderm regionalization, in which neural ectoderm is the default fate in the absence of TGF beta signaling, and suggest that the stomodeal and neural subcircuits that we

A stabilised nodal spectral element method for fully nonlinear water waves

DEFF Research Database (Denmark)

Engsig-Karup, Allan Peter; Eskilsson, C.; Bigoni, Daniele

2016-01-01

can cause severe aliasing problems and consequently numerical instability for marginally resolved or very steep waves. We show how the scheme can be stabilised through a combination of over-integration of the Galerkin projections and a mild spectral filtering on a per element basis. This effectively......We present an arbitrary-order spectral element method for general-purpose simulation of non-overturning water waves, described by fully nonlinear potential theory. The method can be viewed as a high-order extension of the classical finite element method proposed by Cai et al. (1998) [5], although...... the numerical implementation differs greatly. Features of the proposed spectral element method include: nodal Lagrange basis functions, a general quadrature-free approach and gradient recovery using global L2 projections. The quartic nonlinear terms present in the Zakharov form of the free surface conditions...

New procedure for criticality search using coarse mesh nodal methods

International Nuclear Information System (INIS)

Pereira, Wanderson F.; Silva, Fernando C. da; Martinez, Aquilino S.

2011-01-01

The coarse mesh nodal methods have as their primary goal to calculate the neutron flux inside the reactor core. Many computer systems use a specific form of calculation, which is called nodal method. In classical computing systems that use the criticality search is made after the complete convergence of the iterative process of calculating the neutron flux. In this paper, we proposed a new method for the calculation of criticality, condition which will be over very iterative process of calculating the neutron flux. Thus, the processing time for calculating the neutron flux was reduced by half compared with the procedure developed by the Nuclear Engineering Program of COPPE/UFRJ (PEN/COPPE/UFRJ). (author)

New procedure for criticality search using coarse mesh nodal methods

Energy Technology Data Exchange (ETDEWEB)

Pereira, Wanderson F.; Silva, Fernando C. da; Martinez, Aquilino S., E-mail: wneto@con.ufrj.b, E-mail: fernando@con.ufrj.b, E-mail: Aquilino@lmp.ufrj.b [Coordenacao dos Programas de Pos-Graduacao de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

2011-07-01

The coarse mesh nodal methods have as their primary goal to calculate the neutron flux inside the reactor core. Many computer systems use a specific form of calculation, which is called nodal method. In classical computing systems that use the criticality search is made after the complete convergence of the iterative process of calculating the neutron flux. In this paper, we proposed a new method for the calculation of criticality, condition which will be over very iterative process of calculating the neutron flux. Thus, the processing time for calculating the neutron flux was reduced by half compared with the procedure developed by the Nuclear Engineering Program of COPPE/UFRJ (PEN/COPPE/UFRJ). (author)

Accidente cerebrovascular isquémico asociado con ablación por radiofrecuencia de reentrada nodal Ischemic stroke associated with radio frequency ablation for nodal reentry

Directory of Open Access Journals (Sweden)

Juan C Díaz Martínez

2010-04-01

Full Text Available La taquicardia por reentrada nodal es la causa más común de taquicardia supraventricular paroxística; en aquellos pacientes en quienes el manejo farmacológico no es efectivo o deseado la ablación por radiofrecuencia es un excelente método terapéutico dada su alta tasa de curación. Aunque en términos generales dichos procedimientos son rápidos y seguros, se han descrito varias complicaciones entre las que sobresale el accidente cerebrovascular isquémico. Se presenta el caso de una paciente de 41 años con episodios de taquicardia por reentrada nodal a repetición, que fue llevada a ablación por radiofrecuencia. En el post-operatorio inmediato se evidenció déficit neurológico focal con isquemia en el territorio de la arteria cerebral media derecha, tras lo cual se realizó angiografía con intento de angioplastia y abxicimab y posteriormente infusión local de activador de plasminógeno tisular (rtPA con adecuado resultado clínico y angiográfico.Atrioventricular nodal reentry tachycardia is the most common type of paroxismal 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.

AMZ, multigroup constant library for EXPANDA code, generated by NJOY code from ENDF/B-IV

International Nuclear Information System (INIS)

Chalhoub, E.S.; Moraes, Marisa de

1985-01-01

It is described a library of multigroup constants with 70 energy groups and 37 isotopes to fast reactor calculation. The cross sections, scattering matrices and self-shielding factors were generated by NJOY code and RGENDF interface program, from ENDF/B-IV'S evaluated data. The library is edited in adequated format to be used by EXPANDA code. (M.C.K.) [pt

Status of multigroup sensitivity profiles and covariance matrices available from the radiation shielding information center

International Nuclear Information System (INIS)

Roussin, R.W.; Drischler, J.D.; Marable, J.H.

1980-01-01

In recent years multigroup sensitivity profiles and covariance matrices have been added to the Radiation Shielding Information Center's Data Library Collection (DLC). Sensitivity profiles are available in a single package. DLC-45/SENPRO, and covariance matrices are found in two packages, DLC-44/COVERX and DLC-77/COVERV. The contents of these packages are described and their availability is discussed

The effect of nodalization and temperature of reactor upper region: Sensitivity analysis for APR-1400 LBLOCA

International Nuclear Information System (INIS)

Kang, Dong Gu

2017-01-01

Highlights: • The nodalization of APR-1400 was modified to reflect the characteristic of upper region temperature. • The effect of nodalization and temperature of reactor upper region on LBLOCA consequence was evaluated. • The modification of nodalization is an essential prerequisite in APR-1400 LBLOCA analysis. - Abstract: In best estimate (BE) calculation, the definition of system nodalization is important step influencing the prediction accuracy for specific thermal-hydraulic phenomena. The upper region of reactor is defined as the region of the upper guide structure (UGS) and upper dome. It has been assumed that the temperature of upper region is close to average temperature in most large break loss of coolant accident (LBLOCA) analysis cases. However, it was recently found that the temperature of upper region of APR-1400 reactor might be little lower than or similar to hot leg temperature through the review of detailed design data. In this study, the nodalization of APR-1400 was modified to reflect the characteristic of upper region temperature, and the effect of nodalization and temperature of reactor upper region on LBLOCA consequence was evaluated by sensitivity analysis including best estimate plus uncertainty (BEPU) calculation. In basecase calculation, in case of modified version, the peak cladding temperature (PCT) in blowdown phase became higher and the blowdown quenching (or cooling) was significantly deteriorated as compared to original case, and as a result, the cladding temperature in reflood phase became higher and the final quenching was also delayed. In addition, thermal-hydraulic parameters were compared and analyzed to investigate the effect of change of upper region on cladding temperature. In BEPU analysis, the 95 percentile PCT used in current regulatory practice was increased due to the modification of upper region nodalization, and it occurred in the reflood phase unlike original case.

Multigroup P8 - elastic scattering matrices of main reactor elements

International Nuclear Information System (INIS)

Garg, S.B.; Shukla, V.K.

1979-01-01

To study the effect of anisotropic scattering phenomenon on shielding and neutronics of nuclear reactors multigroup P8-elastic scattering matrices have been generated for H, D, He, 6 Li, 7 Li, 10 B, C, N, O, Na, Cr, Fe, Ni, 233 U, 235 U, 238 U, 239 Pu, 240 Pu, 241 Pu and 242 Pu using their angular distribution, Legendre coefficient and elastic scattering cross-section data from the basic ENDF/B library. Two computer codes HSCAT and TRANS have been developed to complete this task for BESM-6 and CDC-3600 computers. These scattering matrices can be directly used as input to the transport theory codes ANISN and DOT. (auth.)

FEMSYN - a code system to solve multigroup diffusion theory equations using a variety of solution techniques. Part 4 : SYNTHD - The synthesis module

International Nuclear Information System (INIS)

Jagannathan, V.

1985-01-01

For solving the multigroup diffusion theory equations in 3-D problems in which the material properties are uniform in large segments of axial direction, the synthesis method is known to give fairly accurate results, at very low computational cost. In the code system FEMSYN, the single channel continuous flux synthesis option has been incorporated. One can generate the radial trail functions by either finite difference method (FDM) or finite element method (FEM). The axial mixing functions can also be found by either FDM or FEM. Use of FEM for both radial and axial directions is found to reduce the calculation time considerably. One can determine eigenvalue, 3-D flux and power distributions with FEMSYN. In this report, a detailed discription of the synthesis module SYNTHD is given. (author)

Multi-group dynamic quantum secret sharing with single photons

Energy Technology Data Exchange (ETDEWEB)

Liu, Hongwei [School of Science and State Key Laboratory of Information Photonics and Optical Communications, Beijing University of Posts and Telecommunications, Beijing 100876 (China); Ma, Haiqiang, E-mail: hqma@bupt.edu.cn [School of Science and State Key Laboratory of Information Photonics and Optical Communications, Beijing University of Posts and Telecommunications, Beijing 100876 (China); Wei, Kejin [School of Science and State Key Laboratory of Information Photonics and Optical Communications, Beijing University of Posts and Telecommunications, Beijing 100876 (China); Yang, Xiuqing [School of Science, Beijing Jiaotong University, Beijing 100044 (China); Qu, Wenxiu; Dou, Tianqi; Chen, Yitian; Li, Ruixue; Zhu, Wu [School of Science and State Key Laboratory of Information Photonics and Optical Communications, Beijing University of Posts and Telecommunications, Beijing 100876 (China)

2016-07-15

In this letter, we propose a novel scheme for the realization of single-photon dynamic quantum secret sharing between a boss and three dynamic agent groups. In our system, the boss can not only choose one of these three groups to share the secret with, but also can share two sets of independent keys with two groups without redistribution. Furthermore, the security of communication is enhanced by using a control mode. Compared with previous schemes, our scheme is more flexible and will contribute to a practical application. - Highlights: • A multi-group dynamic quantum secret sharing with single photons scheme is proposed. • Any one of the groups can be chosen to share secret through controlling the polarization of photons. • Two sets of keys can be shared simultaneously without redistribution.