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Sample records for nodal kinetics code

  1. 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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  3. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  5. 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)

  6. 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)

  7. Development and validation of a nodal code for core calculation

    International Nuclear Information System (INIS)

    Nowakowski, Pedro Mariano

    2004-01-01

    The code RHENO solves the multigroup three-dimensional diffusion equation using a nodal method of polynomial expansion.A comparative study has been made between this code and present internationals nodal diffusion codes, resulting that the RHENO is up to date.The RHENO has been integrated to a calculation line and has been extend to make burnup calculations.Two methods for pin power reconstruction were developed: modulation and imbedded. The modulation method has been implemented in a program, while the implementation of the imbedded method will be concluded shortly.The validation carried out (that includes experimental data of a MPR) show very good results and calculation efficiency

  8. Nodal 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

  9. Real depletion in nodal diffusion codes

    International Nuclear Information System (INIS)

    Petkov, P.T.

    2002-01-01

    The fuel depletion is described by more than one hundred fuel isotopes in the advanced lattice codes like HELIOS, but only a few fuel isotopes are accounted for even in the advanced steady-state diffusion codes. The general assumption that the number densities of the majority of the fuel isotopes depend only on the fuel burnup is seriously in error if high burnup is considered. The real depletion conditions in the reactor core differ from the asymptotic ones at the stage of lattice depletion calculations. This study reveals which fuel isotopes should be explicitly accounted for in the diffusion codes in order to predict adequately the real depletion effects in the core. A somewhat strange conclusion is that if the real number densities of the main fissionable isotopes are not explicitly accounted for in the diffusion code, then Sm-149 should not be accounted for either, because the net error in k-inf is smaller (Authors)

  10. 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

  11. 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.

  12. 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

  13. 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.

  14. 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)

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  16. On-line application of the PANTHER advanced nodal code

    International Nuclear Information System (INIS)

    Hutt, P.K.; Knight, M.P.

    1992-01-01

    Over the last few years, Nuclear Electric has developed an integrated core performance code package for both light water reactors (LWRs) and advanced gas-cooled reactors (AGRs) that can perform a comprehensive range of calculations for fuel cycle design, safety analysis, and on-line operational support for such plants. The package consists of the following codes: WIMS for lattice physics, PANTHER whole reactor nodal flux and AGR thermal hydraulics, VIPRE for LWR thermal hydraulics, and ENIGMA for fuel performance. These codes are integrated within a UNIX-based interactive system called the Reactor Physics Workbench (RPW), which provides an interactive graphic user interface and quality assurance records/data management. The RPW can also control calculational sequences and data flows. The package has been designed to run both off-line and on-line accessing plant data through the RPW

  17. 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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  19. HEXAN - a hexagonal nodal code for solving the diffusion equation

    International Nuclear Information System (INIS)

    Makai, M.

    1982-07-01

    This report describes the theory of and provides a user's manual for the HEXAN program, which is a nodal program for the solution of the few-group diffusion equation in hexagonal geometry. Based upon symmetry considerations, the theory provides an analytical solution in a homogeneous node. WWER and HTGR test problem solutions are presented. The equivalence of the finite-difference scheme and the response matrix method is proven. The properties of a symmetric node's response matrix are investigated. (author)

  20. 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.

  1. 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.

  2. A nodal Grean's function method of reactor core fuel management code, NGCFM2D

    International Nuclear Information System (INIS)

    Li Dongsheng; Yao Dong.

    1987-01-01

    This paper presents the mathematical model and program structure of the nodal Green's function method of reactor core fuel management code, NGCFM2D. Computing results of some reactor cores by NGCFM2D are analysed and compared with other codes

  3. 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.

  4. 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

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

    International Nuclear Information System (INIS)

    Kaya, S.; Yavuz, H.

    2001-01-01

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

  6. 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)

  7. SIRIUS - A one-dimensional multigroup analytic nodal diffusion theory code

    Energy Technology Data Exchange (ETDEWEB)

    Forslund, P. [Westinghouse Atom AB, Vaesteraas (Sweden)

    2000-09-01

    In order to evaluate relative merits of some proposed intranodal cross sections models, a computer code called Sirius has been developed. Sirius is a one-dimensional, multigroup analytic nodal diffusion theory code with microscopic depletion capability. Sirius provides the possibility of performing a spatial homogenization and energy collapsing of cross sections. In addition a so called pin power reconstruction method is available for the purpose of reconstructing 'heterogeneous' pin qualities. consequently, Sirius has the capability of performing all the calculations (incl. depletion calculations) which are an integral part of the nodal calculation procedure. In this way, an unambiguous numerical analysis of intranodal cross section models is made possible. In this report, the theory of the nodal models implemented in sirius as well as the verification of the most important features of these models are addressed.

  8. Inclusion of nodal option in diffusion conventional codes

    International Nuclear Information System (INIS)

    Prati, A.; Anaf, J.

    1985-01-01

    The GCMDT (Generalized Coarse Mesh Diffusion Theory) is studied to use in the 2DB diffusion conventional code. An adequate formalism for its implementation in codes of 'Mesh-Centered' is developed for retangular, triangular and hexagonal geometries. (M.C.K.) [pt

  9. 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)

  10. 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

  11. 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)

  12. 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)

  13. Implementation of burnup in FERM nodal computer code

    International Nuclear Information System (INIS)

    Yoriyaz, H.; Nakata, H.

    1986-01-01

    In this work a spatial burnup scheme and feedback effects has been implemented into the FERM [1] ('Finite Element Response Matrix') program. The spatially dependent neutronic parameters have been considered in three levels: zonewise calculation, assemblywise calculation and pointwise calculation. The results have been compared with the results obtained by CITATION [2] program and showed that the processing time in the FERM code has been hundred of times shorter and no significant difference has been observed in the assembly average power distribution. (Author) [pt

  14. 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

  15. 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)

  16. 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

  17. 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)

  18. An evaluation of nodalization/decay heat/ volatile fission product release models in ISAAC code

    Energy Technology Data Exchange (ETDEWEB)

    Song, Yong Mann; Park, Soo Yong; Kim, Dong Ha

    2003-03-01

    An ISAAC computer code, which was developed for a Level-2 PSA during 1995, has developed mainly with fundamental models for CANDU-specific severe accident progression and also the accident-analyzing experiences are limited to Level-2 PSA purposes. Hence the system nodalization model, decay model and volatile fission product release model, which are known to affect fission product behavior directly or indirectly, are evaluated to both enhance understanding for basic models and accumulate accident-analyzing experiences. As a research strategy, sensitivity studies of model parameters and sensitivity coefficients are performed. According to the results from core nodalization sensitivity study, an original 3x3 nodalization (per loop) method which groups horizontal fuel channels into 12 representative channels, is evaluated to be sufficient for an optimal scheme because detailed nodalization methods have no large effect on fuel thermal-hydraulic behavior, total accident progression and fission product behavior. As ANSI/ANS standard model for decay heat prediction after reactor trip has no needs for further model evaluation due to both wide application on accident analysis codes and good comparison results with the ORIGEN code, ISAAC calculational results of decay heat are used as they are. In addition, fission product revaporization in a containment which is caused by the embedded decay heat, is demonstrated. The results for the volatile fission product release model are analyzed. In case of early release, the IDCOR model with an in-vessel Te release option shows the most conservative results and for the late release case, NUREG-0772 model shows the most conservative results. Considering both early and late release, the IDCOR model with an in-vessel Te bound option shows mitigated conservative results.

  19. 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)

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

    Directory of Open Access Journals (Sweden)

    Tagor Malem Sembiring

    2017-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Surian Pinem

    2014-01-01

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

  2. Implementation of the kinetics in the transport code AZTRAN

    International Nuclear Information System (INIS)

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

    2017-09-01

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

  3. 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.

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

    International Nuclear Information System (INIS)

    Lee, Joo Hee

    2006-02-01

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

  5. Spatial neutron kinetic module of ROSA code

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  6. 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

  7. Benchmarking of the PHOENIX-P/ANC [Advanced Nodal Code] advanced nuclear design system

    International Nuclear Information System (INIS)

    Nguyen, T.Q.; Liu, Y.S.; Durston, C.; Casadei, A.L.

    1988-01-01

    At Westinghouse, an advanced neutronic methods program was designed to improve the quality of the predictions, enhance flexibility in designing advanced fuel and related products, and improve design lead time. Extensive benchmarking data is presented to demonstrate the accuracy of the Advanced Nodal Code (ANC) and the PHOENIX-P advanced lattice code. Qualification data to demonstrate the accuracy of ANC include comparison of key physics parameters against a fine-mesh diffusion theory code, TORTISE. Benchmarking data to demonstrate the validity of the PHOENIX-P methodologies include comparison of physics predictions against critical experiments, isotopics measurements and measured power distributions from spatial criticals. The accuracy of the PHOENIX-P/ANC Advanced Design System is demonstrated by comparing predictions of hot zero power physics parameters and hot full power core follow against measured data from operating reactors. The excellent performance of this system for a broad range of comparisons establishes the basis for implementation of these tools for core design, licensing and operational follow of PWR [pressurized water reactor] cores at Westinghouse

  8. MOSRA-Light; high speed three-dimensional nodal diffusion code for vector computers

    Energy Technology Data Exchange (ETDEWEB)

    Okumura, Keisuke [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-10-01

    MOSRA-Light is a three-dimensional neutron diffusion calculation code for X-Y-Z geometry. It is based on the 4th order polynomial nodal expansion method (NEM). As the 4th order NEM is not sensitive to mesh sizes, accurate calculation is possible by the use of coarse meshes of about 20 cm. The drastic decrease of number of unknowns in a 3-dimensional problem results in very fast computation. Furthermore, it employs newly developed computation algorithm `boundary separated checkerboard sweep method` appropriate to vector computers. This method is very efficient because the speedup factor by vectorization increases, as a scale of problem becomes larger. Speed-up factor compared to the scalar calculation is from 20 to 40 in the case of PWR core calculation. Considering the both effects by the vectorization and the coarse mesh method, total speedup factor is more than 1000 as compared with conventional scalar code with the finite difference method. MOSRA-Light can be available on most of vector or scalar computers with the UNIX or it`s similar operating systems (e.g. freeware like Linux). Users can easily install it by the help of the conversation style installer. This report contains the general theory of NEM, the fast computation algorithm, benchmark calculation results and detailed information for usage of this code including input data instructions and sample input data. (author)

  9. MOSRA-Light; high speed three-dimensional nodal diffusion code for vector computers

    International Nuclear Information System (INIS)

    Okumura, Keisuke

    1998-10-01

    MOSRA-Light is a three-dimensional neutron diffusion calculation code for X-Y-Z geometry. It is based on the 4th order polynomial nodal expansion method (NEM). As the 4th order NEM is not sensitive to mesh sizes, accurate calculation is possible by the use of coarse meshes of about 20 cm. The drastic decrease of number of unknowns in a 3-dimensional problem results in very fast computation. Furthermore, it employs newly developed computation algorithm 'boundary separated checkerboard sweep method' appropriate to vector computers. This method is very efficient because the speedup factor by vectorization increases, as a scale of problem becomes larger. Speed-up factor compared to the scalar calculation is from 20 to 40 in the case of PWR core calculation. Considering the both effects by the vectorization and the coarse mesh method, total speedup factor is more than 1000 as compared with conventional scalar code with the finite difference method. MOSRA-Light can be available on most of vector or scalar computers with the UNIX or it's similar operating systems (e.g. freeware like Linux). Users can easily install it by the help of the conversation style installer. This report contains the general theory of NEM, the fast computation algorithm, benchmark calculation results and detailed information for usage of this code including input data instructions and sample input data. (author)

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

    International Nuclear Information System (INIS)

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

    2005-02-01

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

  11. Acceleration of nodal diffusion code by Chebychev polynomial extrapolation method; Ubrzanje spoljasnjih iteracija difuzionog nodalnog proracuna Chebisevijevom ekstrapolacionom metodom

    Energy Technology Data Exchange (ETDEWEB)

    Zmijarevic, I; Tomashevic, Dj [Institut za Nuklearne Nauke Boris Kidric, Belgrade (Yugoslavia)

    1988-07-01

    This paper presents Chebychev acceleration of outer iterations of a nodal diffusion code of high accuracy. Extrapolation parameters, unique for all moments are calculated using the node integrated distribution of fission source. Sample calculations are presented indicating the efficiency of method. (author)

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  14. Reactivity Coefficient Calculation for AP1000 Reactor Using the NODAL3 Code

    Science.gov (United States)

    Pinem, Surian; Malem Sembiring, Tagor; Tukiran; Deswandri; Sunaryo, Geni Rina

    2018-02-01

    The reactivity coefficient is a very important parameter for inherent safety and stability of nuclear reactors operation. To provide the safety analysis of the reactor, the calculation of changes in reactivity caused by temperature is necessary because it is related to the reactor operation. In this paper, the temperature reactivity coefficients of fuel and moderator of the AP1000 core are calculated, as well as the moderator density and boron concentration. All of these coefficients are calculated at the hot full power condition (HFP). All neutron diffusion constant as a function of temperature, water density and boron concentration were generated by the SRAC2006 code. The core calculations for determination of the reactivity coefficient parameter are done by using NODAL3 code. The calculation results show that the fuel temperature, moderator temperature and boron reactivity coefficients are in the range between -2.613 pcm/°C to -4.657pcm/°C, -1.00518 pcm/°C to 1.00649 pcm/°C and -9.11361 pcm/ppm to -8.0751 pcm/ppm, respectively. For the water density reactivity coefficients, the positive reactivity occurs at the water temperature less than 190 °C. The calculation results show that the reactivity coefficients are accurate because the results have a very good agreement with the design value.

  15. SPQR: a Monte Carlo reactor kinetics code

    International Nuclear Information System (INIS)

    Cramer, S.N.; Dodds, H.L.

    1980-02-01

    The SPQR Monte Carlo code has been developed to analyze fast reactor core accident problems where conventional methods are considered inadequate. The code is based on the adiabatic approximation of the quasi-static method. This initial version contains no automatic material motion or feedback. An existing Monte Carlo code is used to calculate the shape functions and the integral quantities needed in the kinetics module. Several sample problems have been devised and analyzed. Due to the large statistical uncertainty associated with the calculation of reactivity in accident simulations, the results, especially at later times, differ greatly from deterministic methods. It was also found that in large uncoupled systems, the Monte Carlo method has difficulty in handling asymmetric perturbations

  16. Modification of the ANC Nodal Code for analysis of PWR assembly bow

    International Nuclear Information System (INIS)

    Franceschini, Fausto; Fetterman, Robert J.; Little, David C.

    2008-01-01

    Refueling operations at certain PWR cores have revealed fuel assemblies with assembly bow that was higher than expected. As the fuel assemblies bow, the gaps between assemblies change from the uniform nominal configuration. This causes a change in the water volume which affects neutron moderation and thereby power distribution, fuel depletion history, rod internal pressure, etc., with non-trivial impacts on the safety analysis. Westinghouse has developed a new methodology for incorporation of assembly bow in its reload safety analysis package. As part of the new process, the standard Westinghouse reactor physics tool for core analysis, the Advanced Nodal Code ANC, has been modified. The modified ANC, ANCGAP, enables explicit treatment of three-dimensional gap distributions in its neutronic calculations; its accuracy is similar to that of the standard ANC, as demonstrated through an extensive benchmark campaign conducted over a variety of fuel compositions and challenging gap configurations. These features make ANCGAP a crucial tool in the Westinghouse assembly bow package. (authors)

  17. Modification of the ANC Nodal Code for analysis of PWR assembly bow

    Energy Technology Data Exchange (ETDEWEB)

    Franceschini, Fausto; Fetterman, Robert J.; Little, David C. [Westinghouse Electric Company LLC, Pittsburgh PA (United States)

    2008-07-01

    Refueling operations at certain PWR cores have revealed fuel assemblies with assembly bow that was higher than expected. As the fuel assemblies bow, the gaps between assemblies change from the uniform nominal configuration. This causes a change in the water volume which affects neutron moderation and thereby power distribution, fuel depletion history, rod internal pressure, etc., with non-trivial impacts on the safety analysis. Westinghouse has developed a new methodology for incorporation of assembly bow in its reload safety analysis package. As part of the new process, the standard Westinghouse reactor physics tool for core analysis, the Advanced Nodal Code ANC, has been modified. The modified ANC, ANCGAP, enables explicit treatment of three-dimensional gap distributions in its neutronic calculations; its accuracy is similar to that of the standard ANC, as demonstrated through an extensive benchmark campaign conducted over a variety of fuel compositions and challenging gap configurations. These features make ANCGAP a crucial tool in the Westinghouse assembly bow package. (authors)

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

    Directory of Open Access Journals (Sweden)

    Mohammadnia Meysam

    2013-01-01

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

  19. Improvements in practical applicability of NSHEX: nodal transport calculation code for three-dimensional hexagonal-Z geometry

    International Nuclear Information System (INIS)

    Sugino, Kazuteru

    1998-07-01

    As a tool to perform a fast reactor core calculations with high accuracy, NSHEX the nodal transport calculation code for three-dimensional hexagonal-Z geometry is under development. To improve the practical applicability of NSHEX, for instance, in its application to safety analysis and commercial reactor core design studies, we investigated the basic theory used in it, improved the program performance, and evaluated its applicability to the analysis of commercial reactor cores. The current studies show the following: (1) An improvement in the treatment of radial leakage in the radial nodal coupling equation bettered calculational convergence for safety analysis calculation, so the applicability of NSHEX to safety analysis was improved. (2) As a result of comparison of results from NSHEX and the standard core calculation code, it was confirmed that there was consistency between them. (3) According to the evaluation of the effect due to the difference of calculational condition, it was found that the calculation under appropriate nodal expansion orders and Sn orders correspond to the one under most detailed condition. However further investigation is required to reduce the uncertainty in calculational results due to the treatment of high order flux moments. (4) A whole core version of NSHEX enabling calculation for any FBR core geometry has been developed, this improved general applicability of NSHEX. (5) An investigation of the applicability of the rebalance method to acceleration clarified that this improved calculational convergence and it was effective. (J.P.N.)

  20. 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)

  1. Development of a qualified nodalization for small-break LOCA transient analysis in PSB-VVER integral test facility by RELAP5 system code

    Energy Technology Data Exchange (ETDEWEB)

    Shahedi, S. [Department of Energy Engineering, Sharif University of Technology, Azadi Street, Tehran (Iran, Islamic Republic of); Jafari, J., E-mail: jalil_jafari@yahoo.co [Reactors and Accelerators R and D School, Nuclear Science and Technology Research Institute, North Kargar Street, Tehran (Iran, Islamic Republic of); Boroushaki, M. [Department of Energy Engineering, Sharif University of Technology, Azadi Street, Tehran (Iran, Islamic Republic of); D' Auria, F. [DIMNP, University of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy)

    2010-10-15

    This paper deals with development and qualification of a nodalization for modeling of the PSB-VVER integral test facility (ITF) by RELAP5/MOD3.2 code and prediction of its primary and secondary systems behaviors at steady state and transient conditions. The PSB-VVER is a full-height, 1/300 volume and power scale representation of a VVER-1000 NPP. A RELAP5 nodalization has been developed for PSB-VVER modeling and a nodalization qualification process has been applied for the developed nodalization at steady state and transient levels and a qualified nodalization has been proposed for modeling of the PSB ITF. The 11% small-break loss-of-coolant-accident (SBLOCA), i.e. rupture of one of the hydroaccumulators (HA) injection lines in the upper plenum (UP) region of reactor pressure vessel (RPV) below the hot legs (HL), inlets has been considered for nodalization qualification process. The influence of the different steam generator (SG) nodalizations on the RELAP5 results and on the nodalization qualification process has been examined. The 'steady state' qualification level includes checking the correctness of the initial and boundary conditions and geometrical fidelity. In the 'transient' qualification level, the time dependent results of the code calculation are compared with the experimental time trends from both the qualitative and quantitative point of view. For quantitative assessment of the results, a Fast Fourier Transform Based Method (FFTBM) has been used. The FFTBM was used to establish a range in which the steam generators nodalizations can vary.

  2. Core design optimization by integration of a fast 3-D nodal code in a heuristic search procedure

    Energy Technology Data Exchange (ETDEWEB)

    Geemert, R. van; Leege, P.F.A. de; Hoogenboom, J.E.; Quist, A.J. [Delft University of Technology, NL-2629 JB Delft (Netherlands)

    1998-07-01

    An automated design tool is being developed for the Hoger Onderwijs Reactor (HOR) in Delft, the Netherlands, which is a 2 MWth swimming-pool type research reactor. As a black box evaluator, the 3-D nodal code SILWER, which up to now has been used only for evaluation of predetermined core designs, is integrated in the core optimization procedure. SILWER is a part of PSl's ELCOS package and features optional additional thermal-hydraulic, control rods and xenon poisoning calculations. This allows for fast and accurate evaluation of different core designs during the optimization search. Special attention is paid to handling the in- and output files for SILWER such that no adjustment of the code itself is required for its integration in the optimization programme. The optimization objective, the safety and operation constraints, as well as the optimization procedure, are discussed. (author)

  3. Core design optimization by integration of a fast 3-D nodal code in a heuristic search procedure

    International Nuclear Information System (INIS)

    Geemert, R. van; Leege, P.F.A. de; Hoogenboom, J.E.; Quist, A.J.

    1998-01-01

    An automated design tool is being developed for the Hoger Onderwijs Reactor (HOR) in Delft, the Netherlands, which is a 2 MWth swimming-pool type research reactor. As a black box evaluator, the 3-D nodal code SILWER, which up to now has been used only for evaluation of predetermined core designs, is integrated in the core optimization procedure. SILWER is a part of PSl's ELCOS package and features optional additional thermal-hydraulic, control rods and xenon poisoning calculations. This allows for fast and accurate evaluation of different core designs during the optimization search. Special attention is paid to handling the in- and output files for SILWER such that no adjustment of the code itself is required for its integration in the optimization programme. The optimization objective, the safety and operation constraints, as well as the optimization procedure, are discussed. (author)

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

    International Nuclear Information System (INIS)

    Halilou, A.; Lounici, A.

    1981-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-03-15

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

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

    International Nuclear Information System (INIS)

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

    2006-03-01

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

  7. Development of AGNES, a kinetics code for fissile solutions, 1

    International Nuclear Information System (INIS)

    Nakajima, Ken; Ohnishi, Nobuaki

    1986-01-01

    A kinetics code for fissile solutions, AGNES (Accidentally Generated Nuclear Excursion Simulation code), has been developed. This code calculates the radiolytic gas void effect as a reactivity feedback. Physical and calculative models of the radiolytic gas void are summarized and the usage of AGNES is described. In addition, some benchmark calculations were performed and results of calculations show good agreement with those of experiments. (author)

  8. QUIC: a chemical kinetics code for use with the chemical equilibrium code QUIL

    International Nuclear Information System (INIS)

    Lunsford, J.L.

    1977-10-01

    A chemical rate kinetics code QUIC is described, along with a support code RATE. QUIC is designed to allow chemical kinetics calculations on a wide variety of chemical environments while operating in the overlay environment of the chemical equilibrium code QUIL. QUIC depends upon a rate-data library called LIBR. This library is maintained by RATE. RATE enters into the library all reactions in a standardized format. The code QUIC, operating in conjunction with QUIL, is interactive and written to be used from a remote terminal, with paging control provided. Plotted output is also available

  9. Sandia reactor kinetics codes: SAK and PK1D

    International Nuclear Information System (INIS)

    Pickard, P.S.; Odom, J.P.

    1978-01-01

    The Sandia Kinetics code (SAK) is a one-dimensional coupled thermal-neutronics transient analysis code for use in simulation of reactor transients. The time-dependent cross section routines allow arbitrary time-dependent changes in material properties. The one-dimensional heat transfer routines are for cylindrical geometry and allow arbitrary mesh structure, temperature-dependent thermal properties, radiation treatment, and coolant flow and heat-transfer properties at the surface of a fuel element. The Point Kinetics 1 Dimensional Heat Transfer Code (PK1D) solves the point kinetics equations and has essentially the same heat-transfer treatment as SAK. PK1D can address extended reactor transients with minimal computer execution time

  10. Multidimensional method of spatially coupled approximation to the transverse escape in nodal codes

    International Nuclear Information System (INIS)

    Jatuff, F.E.

    1990-01-01

    A natural extension of the polynomic development programmed in RHENO code is presented, which adds to the variable order one-dimensional functions sum, a number of terms that represent functions of production. These new terms, which provide a direct determination of transverse escapes, are calculated from the new variables coupling among nodes: the 4 fluxes in rectangle vortices (bidimensional Cartesian geometry) or the 12 fluxes half-way through the parallelepiped edges (tridimensional Cartesian geometry). (Author) [es

  11. Calculation of accurate albedo boundary conditions for three-dimensional nodal diffusion codes by the method of characteristics

    International Nuclear Information System (INIS)

    Petkov, Petko T.

    2000-01-01

    Most of the few-group three-dimensional nodal diffusion codes used for neutronics calculations of the WWER reactors use albedo type boundary conditions on the core-reflector boundary. The conventional albedo are group-to-group reflection probabilities, defined on each outer node face. The method of characteristics is used to calculate accurate albedo by the following procedure. A many-group two-dimensional heterogeneous core-reflector problem, including a sufficient part of the core and detailed description of the adjacent reflector, is solved first. From this solution the angular flux on the core-reflector boundary is calculated in all groups for all traced neutron directions. Accurate boundary conditions can be calculated for the radial, top and bottom reflectors as well as for the absorber part of the WWER-440 reactor control assemblies. The algorithm can be used to estimate also albedo, coupling outer node faces on the radial reflector in the axial direction. Numerical results for the WWER-440 reactor are presented. (Authors)

  12. MIDAS/PK code development using point kinetics model

    International Nuclear Information System (INIS)

    Song, Y. M.; Park, S. H.

    1999-01-01

    In this study, a MIDAS/PK code has been developed for analyzing the ATWS (Anticipated Transients Without Scram) which can be one of severe accident initiating events. The MIDAS is an integrated computer code based on the MELCOR code to develop a severe accident risk reduction strategy by Korea Atomic Energy Research Institute. In the mean time, the Chexal-Layman correlation in the current MELCOR, which was developed under a BWR condition, is appeared to be inappropriate for a PWR. So as to provide ATWS analysis capability to the MIDAS code, a point kinetics module, PKINETIC, has first been developed as a stand-alone code whose reference model was selected from the current accident analysis codes. In the next step, the MIDAS/PK code has been developed via coupling PKINETIC with the MIDAS code by inter-connecting several thermal hydraulic parameters between the two codes. Since the major concern in the ATWS analysis is the primary peak pressure during the early few minutes into the accident, the peak pressure from the PKINETIC module and the MIDAS/PK are compared with the RETRAN calculations showing a good agreement between them. The MIDAS/PK code is considered to be valuable for analyzing the plant response during ATWS deterministically, especially for the early domestic Westinghouse plants which rely on the operator procedure instead of an AMSAC (ATWS Mitigating System Actuation Circuitry) against ATWS. This capability of ATWS analysis is also important from the view point of accident management and mitigation

  13. 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)

  14. Benchmark studies of the gyro-Landau-fluid code and gyro-kinetic codes on kinetic ballooning modes

    Energy Technology Data Exchange (ETDEWEB)

    Tang, T. F. [Dalian University of Technology, Dalian 116024 (China); Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Xu, X. Q. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Ma, C. H. [Fusion Simulation Center, School of Physics, Peking University, Beijing (China); Bass, E. M.; Candy, J. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Holland, C. [University of California San Diego, La Jolla, California 92093-0429 (United States)

    2016-03-15

    A Gyro-Landau-Fluid (GLF) 3 + 1 model has been recently implemented in BOUT++ framework, which contains full Finite-Larmor-Radius effects, Landau damping, and toroidal resonance [Ma et al., Phys. Plasmas 22, 055903 (2015)]. A linear global beta scan has been conducted using the JET-like circular equilibria (cbm18 series), showing that the unstable modes are kinetic ballooning modes (KBMs). In this work, we use the GYRO code, which is a gyrokinetic continuum code widely used for simulation of the plasma microturbulence, to benchmark with GLF 3 + 1 code on KBMs. To verify our code on the KBM case, we first perform the beta scan based on “Cyclone base case parameter set.” We find that the growth rate is almost the same for two codes, and the KBM mode is further destabilized as beta increases. For JET-like global circular equilibria, as the modes localize in peak pressure gradient region, a linear local beta scan using the same set of equilibria has been performed at this position for comparison. With the drift kinetic electron module in the GYRO code by including small electron-electron collision to damp electron modes, GYRO generated mode structures and parity suggest that they are kinetic ballooning modes, and the growth rate is comparable to the GLF results. However, a radial scan of the pedestal for a particular set of cbm18 equilibria, using GYRO code, shows different trends for the low-n and high-n modes. The low-n modes show that the linear growth rate peaks at peak pressure gradient position as GLF results. However, for high-n modes, the growth rate of the most unstable mode shifts outward to the bottom of pedestal and the real frequency of what was originally the KBMs in ion diamagnetic drift direction steadily approaches and crosses over to the electron diamagnetic drift direction.

  15. 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.)

  16. Energy Conservation Tests of a Coupled Kinetic-kinetic Plasma-neutral Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Stotler, D. P.; Chang, C. S.; Ku, S. H.; Lang, J.; Park, G.

    2012-08-29

    A Monte Carlo neutral transport routine, based on DEGAS2, has been coupled to the guiding center ion-electron-neutral neoclassical PIC code XGC0 to provide a realistic treatment of neutral atoms and molecules in the tokamak edge plasma. The DEGAS2 routine allows detailed atomic physics and plasma-material interaction processes to be incorporated into these simulations. The spatial pro le of the neutral particle source used in the DEGAS2 routine is determined from the uxes of XGC0 ions to the material surfaces. The kinetic-kinetic plasma-neutral transport capability is demonstrated with example pedestal fueling simulations.

  17. Kinetic models of gene expression including non-coding RNAs

    Energy Technology Data Exchange (ETDEWEB)

    Zhdanov, Vladimir P., E-mail: zhdanov@catalysis.r

    2011-03-15

    In cells, genes are transcribed into mRNAs, and the latter are translated into proteins. Due to the feedbacks between these processes, the kinetics of gene expression may be complex even in the simplest genetic networks. The corresponding models have already been reviewed in the literature. A new avenue in this field is related to the recognition that the conventional scenario of gene expression is fully applicable only to prokaryotes whose genomes consist of tightly packed protein-coding sequences. In eukaryotic cells, in contrast, such sequences are relatively rare, and the rest of the genome includes numerous transcript units representing non-coding RNAs (ncRNAs). During the past decade, it has become clear that such RNAs play a crucial role in gene expression and accordingly influence a multitude of cellular processes both in the normal state and during diseases. The numerous biological functions of ncRNAs are based primarily on their abilities to silence genes via pairing with a target mRNA and subsequently preventing its translation or facilitating degradation of the mRNA-ncRNA complex. Many other abilities of ncRNAs have been discovered as well. Our review is focused on the available kinetic models describing the mRNA, ncRNA and protein interplay. In particular, we systematically present the simplest models without kinetic feedbacks, models containing feedbacks and predicting bistability and oscillations in simple genetic networks, and models describing the effect of ncRNAs on complex genetic networks. Mathematically, the presentation is based primarily on temporal mean-field kinetic equations. The stochastic and spatio-temporal effects are also briefly discussed.

  18. 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)

  19. Implementation of the kinetics in the transport code AZTRAN; Implementacion de la cinetica en el codigo de transporte AZTRAN

    Energy Technology Data Exchange (ETDEWEB)

    Duran G, J. A.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, San Pedro Zacatenco, 07738 Ciudad de Mexico (Mexico); Gomez T, A. M., E-mail: redfield1290@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2017-09-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

  1. TUTANK a two-dimensional neutron kinetics code

    International Nuclear Information System (INIS)

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

    1975-04-01

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

  2. A generalized interface module for the coupling of spatial kinetics and thermal-hydraulics codes

    Energy Technology Data Exchange (ETDEWEB)

    Barber, D.A.; Miller, R.M.; Joo, H.G.; Downar, T.J. [Purdue Univ., West Lafayette, IN (United States). Dept. of Nuclear Engineering; Wang, W. [SCIENTECH, Inc., Rockville, MD (United States); Mousseau, V.A.; Ebert, D.D. [Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research

    1999-03-01

    A generalized interface module has been developed for the coupling of any thermal-hydraulics code to any spatial kinetics code. The coupling scheme was designed and implemented with emphasis placed on maximizing flexibility while minimizing modifications to the respective codes. In this design, the thermal-hydraulics, general interface, and spatial kinetics codes function independently and utilize the Parallel Virtual Machine software to manage cross-process communication. Using this interface, the USNRC version of the 3D neutron kinetics code, PARCX, has been coupled to the USNRC system analysis codes RELAP5 and TRAC-M. RELAP5/PARCS assessment results are presented for two NEACRP rod ejection benchmark problems and an NEA/OECD main steam line break benchmark problem. The assessment of TRAC-M/PARCS has only recently been initiated, nonetheless, the capabilities of the coupled code are presented for a typical PWR system/core model.

  3. A generalized interface module for the coupling of spatial kinetics and thermal-hydraulics codes

    International Nuclear Information System (INIS)

    Barber, D.A.; Miller, R.M.; Joo, H.G.; Downar, T.J.; Mousseau, V.A.; Ebert, D.D.

    1999-01-01

    A generalized interface module has been developed for the coupling of any thermal-hydraulics code to any spatial kinetics code. The coupling scheme was designed and implemented with emphasis placed on maximizing flexibility while minimizing modifications to the respective codes. In this design, the thermal-hydraulics, general interface, and spatial kinetics codes function independently and utilize the Parallel Virtual Machine software to manage cross-process communication. Using this interface, the USNRC version of the 3D neutron kinetics code, PARCX, has been coupled to the USNRC system analysis codes RELAP5 and TRAC-M. RELAP5/PARCS assessment results are presented for two NEACRP rod ejection benchmark problems and an NEA/OECD main steam line break benchmark problem. The assessment of TRAC-M/PARCS has only recently been initiated, nonetheless, the capabilities of the coupled code are presented for a typical PWR system/core model

  4. Validation and applicability of the 3D core kinetics and thermal hydraulics coupled code SPARKLE

    International Nuclear Information System (INIS)

    Miyata, Manabu; Maruyama, Manabu; Ogawa, Junto; Otake, Yukihiko; Miyake, Shuhei; Tabuse, Shigehiko; Tanaka, Hirohisa

    2009-01-01

    The SPARKLE code is a coupled code system based on three individual codes whose physical models have already been verified and validated. Mitsubishi Heavy Industries (MHI) confirmed the coupling calculation, including data transfer and the total reactor coolant system (RCS) behavior of the SPARKLE code. The confirmation uses the OECD/NEA MSLB benchmark problem, which is based on Three Mile Island Unit 1 (TMI-1) nuclear power plant data. This benchmark problem has been used to verify coupled codes developed and used by many organizations. Objectives of the benchmark program are as follows. Phase 1 is to compare the results of the system transient code using point kinetics. Phase 2 is to compare the results of the coupled three-dimensional (3D) core kinetics code and 3D core thermal-hydraulics (T/H) code, and Phase 3 is to compare the results of the combined coupled system transient code, 3D core kinetics code, and 3D core T/H code as a total validation of the coupled calculation. The calculation results of the SPARKLE code indicate good agreement with other benchmark participants' results. Therefore, the SPARKLE code is validated through these benchmark problems. In anticipation of applying the SPARKLE code to licensing analyses, MHI and Japanese PWR utilities have established a safety analysis method regarding the calculation conditions such as power distributions, reactivity coefficients, and event-specific features. (author)

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

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

    International Nuclear Information System (INIS)

    Nigg, D.W.

    1983-09-01

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

  7. 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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  9. Benchmark test of drift-kinetic and gyrokinetic codes through neoclassical transport simulations

    International Nuclear Information System (INIS)

    Satake, S.; Sugama, H.; Watanabe, T.-H.; Idomura, Yasuhiro

    2009-09-01

    Two simulation codes that solve the drift-kinetic or gyrokinetic equation in toroidal plasmas are benchmarked by comparing the simulation results of neoclassical transport. The two codes are the drift-kinetic δf Monte Carlo code (FORTEC-3D) and the gyrokinetic full- f Vlasov code (GT5D), both of which solve radially-global, five-dimensional kinetic equation with including the linear Fokker-Planck collision operator. In a tokamak configuration, neoclassical radial heat flux and the force balance relation, which relates the parallel mean flow with radial electric field and temperature gradient, are compared between these two codes, and their results are also compared with the local neoclassical transport theory. It is found that the simulation results of the two codes coincide very well in a wide rage of plasma collisionality parameter ν * = 0.01 - 10 and also agree with the theoretical estimations. The time evolution of radial electric field and particle flux, and the radial profile of the geodesic acoustic mode frequency also coincide very well. These facts guarantee the capability of GT5D to simulate plasma turbulence transport with including proper neoclassical effects of collisional diffusion and equilibrium radial electric field. (author)

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  11. 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)

  12. 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

  13. 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.)

  14. 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

  15. Kinetic parameters evaluation of PWRs using static cell and core calculation codes

    International Nuclear Information System (INIS)

    Jahanbin, Ali; Malmir, Hessam

    2012-01-01

    Highlights: ► In this study, we have calculated effective delayed neutron fraction and prompt neutron lifetime in PWRs. ► New software has been developed to link the WIMS, BORGES and CITATION codes in Visual C computer programming language. ► This software is used for calculation of the kinetic parameters in a typical VVER-1000 and NOK Beznau reactor. ► The ratios ((β eff ) i )/((β eff ) core ) , which are the important input data for the reactivity accident analysis, are also calculated. - Abstract: In this paper, evaluation of the kinetic parameters (effective delayed neutron fraction and prompt neutron lifetime) in PWRs, using static cell and core calculation codes, is reported. A new software has been developed to link the WIMS, BORGES and CITATION codes in Visual C computer programming language. Using the WIMS cell calculation code, multigroup microscopic cross-sections and number densities of different materials can be generated in a binary file. By the use of BORGES code, these binary-form cross-sections and number densities are converted to a format readable by the CITATION core calculation code, by which the kinetic parameters can be finally obtained. This software is used for calculation of the kinetic parameters in a typical VVER-1000 and NOK Beznau reactor. The ratios ((β eff ) i )/((β eff ) core ) , which are the important input data for the reactivity accident analysis, are also calculated. Benchmarking of the results against the final safety analysis report (FSAR) of the aforementioned reactors shows very good agreements with these published documents.

  16. Space-time reactor kinetics for heterogeneous reactor structure

    Energy Technology Data Exchange (ETDEWEB)

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

    1969-11-15

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

  17. 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

  18. Development of an inelastic stress analysis code 'KINE-T' and its evaluations

    International Nuclear Information System (INIS)

    Kobatake, K.; Takahashi, S.; Suzuki, M.

    1977-01-01

    Referring to the ASME B and PVC Code Case 1592-7, the inelastic stress analysis is required for the designs of the class 1 components in elevated temperature if the results of the elastic stress analysis and/or simplified inelastic analysis do not satisfy the requirements. Authors programmed a two-dimensional axisymmetric inelastic analysis code 'KINE-T', and carried out its evaluations and an application. This FEM code is based on the incremental method and the following: elastic-plastic constitutive equation (yield condition of von Mises; flow rule of Prandtl-Reuss; Prager's hardening rule); creep constitutive equation (equation of state approach; flow rule of von Mises; strain hardening rule); the temperature dependency of the yield function is considered; solution procedure of the assembled stiffness matrix is the 'initial stress method'. After the completion of the programming, authors compared the output with not only theoretical results but also with those of the MARC code and the ANSYS code. In order to apply the code to the practical designing, authors settled a quasi-component two-dimensional axisymmetric model and a loading cycle (500 cycles). Then, an inelastic analysis and its integrity evaluation are carried out

  19. 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

  20. G4-STORK: A Geant4-based Monte Carlo reactor kinetics simulation code

    International Nuclear Information System (INIS)

    Russell, Liam; Buijs, Adriaan; Jonkmans, Guy

    2014-01-01

    Highlights: • G4-STORK is a new, time-dependent, Monte Carlo code for reactor physics applications. • G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. • G4-STORK was designed to simulate short-term fluctuations in reactor cores. • G4-STORK is well suited for simulating sub- and supercritical assemblies. • G4-STORK was verified through comparisons with DRAGON and MCNP. - Abstract: In this paper we introduce G4-STORK (Geant4 STOchastic Reactor Kinetics), a new, time-dependent, Monte Carlo particle tracking code for reactor physics applications. G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. The toolkit provides the fundamental physics models and particle tracking algorithms that track each particle in space and time. It is a framework for further development (e.g. for projects such as G4-STORK). G4-STORK derives reactor physics parameters (e.g. k eff ) from the continuous evolution of a population of neutrons in space and time in the given simulation geometry. In this paper we detail the major additions to the Geant4 toolkit that were necessary to create G4-STORK. These include a renormalization process that maintains a manageable number of neutrons in the simulation even in very sub- or supercritical systems, scoring processes (e.g. recording fission locations, total neutrons produced and lost, etc.) that allow G4-STORK to calculate the reactor physics parameters, and dynamic simulation geometries that can change over the course of simulation to illicit reactor kinetics responses (e.g. fuel temperature reactivity feedback). The additions are verified through simple simulations and code-to-code comparisons with established reactor physics codes such as DRAGON and MCNP. Additionally, G4-STORK was developed to run a single simulation in parallel over many processors using MPI (Message Passing Interface) pipes

  1. Viriato: a Fourier-Hermite spectral code for strongly magnetised fluid-kinetic plasma dynamics

    Science.gov (United States)

    Loureiro, Nuno; Dorland, William; Fazendeiro, Luis; Kanekar, Anjor; Mallet, Alfred; Zocco, Alessandro

    2015-11-01

    We report on the algorithms and numerical methods used in Viriato, a novel fluid-kinetic code that solves two distinct sets of equations: (i) the Kinetic Reduced Electron Heating Model equations [Zocco & Schekochihin, 2011] and (ii) the kinetic reduced MHD (KRMHD) equations [Schekochihin et al., 2009]. Two main applications of these equations are magnetised (Alfvnénic) plasma turbulence and magnetic reconnection. Viriato uses operator splitting to separate the dynamics parallel and perpendicular to the ambient magnetic field (assumed strong). Along the magnetic field, Viriato allows for either a second-order accurate MacCormack method or, for higher accuracy, a spectral-like scheme. Perpendicular to the field Viriato is pseudo-spectral, and the time integration is performed by means of an iterative predictor-corrector scheme. In addition, a distinctive feature of Viriato is its spectral representation of the parallel velocity-space dependence, achieved by means of a Hermite representation of the perturbed distribution function. A series of linear and nonlinear benchmarks and tests are presented, with focus on 3D decaying kinetic turbulence. Work partially supported by Fundação para a Ciência e Tecnologia via Grants UID/FIS/50010/2013 and IF/00530/2013.

  2. 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

  3. Preliminary Development of the MARS/FREK Spatial Kinetics Coupled System Code for Square Fueled Fast Reactor Applications

    International Nuclear Information System (INIS)

    Bae, Moo Hoon; Joo, Han Gyu

    2009-01-01

    Incorporation of a three-dimensional (3-D) reactor kinetics model into a system thermal-hydraulic (T/H) code enhances the capability to perform realistic analyses of the core neutronic behavior and the plant system dynamics which are coupled each other. For this advantage, several coupled system T/H and spatial kinetics codes, such as RELAP/PARCS, RELAP5/ PANBOX, and MARS/MASTER have been developed. These codes, however, so far limited to LWR applications. The objective of this work is to develop such a coupled code for fast reactor applications. Particularly, applications to lead-bismuth eutectic (LBE) cooled fast reactor are of interest which employ open square lattices. A fast reactor kinetics code applicable to square fueled cores called FREK is coupled the LBE version of the MARS code. The MARS/MASTER coupled code is used as the reference for the integration. The coupled code MARS/FREK is examined for a conceptual reactor called P-DEMO which is being developed by NUTRECK. In order to check the validity of the coupled code, however, the OECD MSLB benchmark exercise III calculation is solved first

  4. 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)

  5. GARLIC-B. A digital code for real-time calculation of the transient behaviour of nodal and global core and plant parameters of BWR nuclear power plants

    International Nuclear Information System (INIS)

    Ercan, Y.; Hoeld, A.; Lupas, O.

    1982-04-01

    A program description of the code GARLIC-B is given. The code is based on a nonlinear transient model for BWR nuclear power plants which consist of a 3D-core, a top plenum, steam removal and feed water systems and a downcomer with main coolant recirculation pumps. The core is subdivided into a number of superboxes and flow channels with different coolant mass flow rates. Subcooled boiling within these channels has an important reactivity feed back effect and has to be taken also into account. The code computes the local and global core and plant transient situation as dependent on both the inherent core dynamics and external control actions, i.e., disturbances such as motions of control rod banks, changes of mass flow rates of coolant, feed water and steam outlet. The case of a pressure-controlled reactor operation is also considered. (orig./GL) [de

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

    Energy Technology Data Exchange (ETDEWEB)

    Hartmann, Christoph Oliver

    2016-06-13

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

  7. Recalculating the steady state conditions of the V-1000 zero-power facility at Kurchatov Institute using Monte Carlo and nodal diffusion codes

    Energy Technology Data Exchange (ETDEWEB)

    Sahlberg, Ville [VTT Technical Research Centre of Finland Ltd, VTT (Finland)

    2017-09-15

    Continuous-energy Monte Carlo reactor physics code Serpent 2 was used to model the critical steady state conditions measured in V-1000 zero-power critical facility at Kurchatov Institute (KI), Moscow in 1990-1992. The Serpent 2 results were compared to measurements and Serpent 2 was used to generate group constants for reactor dynamics code HEXTRAN. The results of a HEXTRAN calculation of the steady state were compared to Serpent 2. The relative power density distribution of the SERPENT2 calculations compared with the measurements was within the statistical accuracy. The comparison of HEXTRAN and Serpent 2 node-wise relative power density distributions showed an accuracy of ±10%.

  8. 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

  9. Study of gamma propagation by using the ZEUS mono-kinetic code

    International Nuclear Information System (INIS)

    Vergnaud, Therese

    1969-10-01

    As studies of radiation protection often require the knowledge of heating due to capture gamma of thermal neutrons, the authors report an attempt of assessment of neutrons and γ propagation by using the same code (Zeus) which computes particle scattering by implementing a mono-kinetic Monte Carlo method. With this method, it is possible to study rather complex geometries and gamma source distributions directly obtained by a previous calculation of thermal neutrons. However, this method is not suitable for the study of energy degradation of gamma rays during their propagation. An approximate shock law is used to take shock-induced energy loss into account. This method is tested for different materials or media (either light like water and aluminium, or heavy like iron). Results are compared with those obtained by Goldstein with the method of moments. Results obtained by using Zeus are discussed: some of them appear to be over-estimated [fr

  10. SurfKin: an ab initio kinetic code for modeling surface reactions.

    Science.gov (United States)

    Le, Thong Nguyen-Minh; Liu, Bin; Huynh, Lam K

    2014-10-05

    In this article, we describe a C/C++ program called SurfKin (Surface Kinetics) to construct microkinetic mechanisms for modeling gas-surface reactions. Thermodynamic properties of reaction species are estimated based on density functional theory calculations and statistical mechanics. Rate constants for elementary steps (including adsorption, desorption, and chemical reactions on surfaces) are calculated using the classical collision theory and transition state theory. Methane decomposition and water-gas shift reaction on Ni(111) surface were chosen as test cases to validate the code implementations. The good agreement with literature data suggests this is a powerful tool to facilitate the analysis of complex reactions on surfaces, and thus it helps to effectively construct detailed microkinetic mechanisms for such surface reactions. SurfKin also opens a possibility for designing nanoscale model catalysts. Copyright © 2014 Wiley Periodicals, Inc.

  11. 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)

  12. 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.

  13. Development of a coupled dynamics code with transport theory capability and application to accelerator driven systems transients

    International Nuclear Information System (INIS)

    Cahalan, J.E.; Ama, T.; Palmiotti, G.; Taiwo, T.A.; Yang, W.S.

    2000-01-01

    The VARIANT-K and DIF3D-K nodal spatial kinetics computer codes have been coupled to the SAS4A and SASSYS-1 liquid metal reactor accident and systems analysis codes. SAS4A and SASSYS-1 have been extended with the addition of heavy liquid metal (Pb and Pb-Bi) thermophysical properties, heat transfer correlations, and fluid dynamics correlations. The coupling methodology and heavy liquid metal modeling additions are described. The new computer code suite has been applied to analysis of neutron source and thermal-hydraulics transients in a model of an accelerator-driven minor actinide burner design proposed in an OECD/NEA/NSC benchmark specification. Modeling assumptions and input data generation procedures are described. Results of transient analyses are reported, with emphasis on comparison of P1 and P3 variational nodal transport theory results with nodal diffusion theory results, and on significance of spatial kinetics effects

  14. 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.

  15. VULCAN: An Open-source, Validated Chemical Kinetics Python Code for Exoplanetary Atmospheres

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, Shang-Min; Grosheintz, Luc; Kitzmann, Daniel; Heng, Kevin [University of Bern, Center for Space and Habitability, Sidlerstrasse 5, CH-3012, Bern (Switzerland); Lyons, James R. [Arizona State University, School of Earth and Space Exploration, Bateman Physical Sciences, Tempe, AZ 85287-1404 (United States); Rimmer, Paul B., E-mail: shang-min.tsai@space.unibe.ch, E-mail: kevin.heng@csh.unibe.ch, E-mail: jimlyons@asu.edu [University of St. Andrews, School of Physics and Astronomy, St. Andrews, KY16 9SS (United Kingdom)

    2017-02-01

    We present an open-source and validated chemical kinetics code for studying hot exoplanetary atmospheres, which we name VULCAN. It is constructed for gaseous chemistry from 500 to 2500 K, using a reduced C–H–O chemical network with about 300 reactions. It uses eddy diffusion to mimic atmospheric dynamics and excludes photochemistry. We have provided a full description of the rate coefficients and thermodynamic data used. We validate VULCAN by reproducing chemical equilibrium and by comparing its output versus the disequilibrium-chemistry calculations of Moses et al. and Rimmer and Helling. It reproduces the models of HD 189733b and HD 209458b by Moses et al., which employ a network with nearly 1600 reactions. We also use VULCAN to examine the theoretical trends produced when the temperature–pressure profile and carbon-to-oxygen ratio are varied. Assisted by a sensitivity test designed to identify the key reactions responsible for producing a specific molecule, we revisit the quenching approximation and find that it is accurate for methane but breaks down for acetylene, because the disequilibrium abundance of acetylene is not directly determined by transport-induced quenching, but is rather indirectly controlled by the disequilibrium abundance of methane. Therefore we suggest that the quenching approximation should be used with caution and must always be checked against a chemical kinetics calculation. A one-dimensional model atmosphere with 100 layers, computed using VULCAN, typically takes several minutes to complete. VULCAN is part of the Exoclimes Simulation Platform (ESP; exoclime.net) and publicly available at https://github.com/exoclime/VULCAN.

  16. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

  18. 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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  1. 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.

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

    International Nuclear Information System (INIS)

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

    1998-03-01

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

  3. Nodal methods in numerical reactor calculations

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  4. Nodal methods in numerical reactor calculations

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

  5. Kinetics Parameters of VVER-1000 Core with 3 MOX Lead Test Assemblies To Be Used for Accident Analysis Codes

    International Nuclear Information System (INIS)

    Pavlovitchev, A.M.

    2000-01-01

    The present work is a part of Joint U.S./Russian Project with Weapons-Grade Plutonium Disposition in VVER Reactor and presents the neutronics calculations of kinetics parameters of VVER-1000 core with 3 introduced MOX LTAs. MOX LTA design has been studied in [1] for two options of MOX LTA: 100% plutonium and of ''island'' type. As a result, zoning i.e. fissile plutonium enrichments in different plutonium zones, has been defined. VVER-1000 core with 3 introduced MOX LTAs of chosen design has been calculated in [2]. In present work, the neutronics data for transient analysis codes (RELAP [3]) has been obtained using the codes chain of RRC ''Kurchatov Institute'' [5] that is to be used for exploitation neutronics calculations of VVER. Nowadays the 3D assembly-by-assembly code BIPR-7A and 2D pin-by-pin code PERMAK-A, both with the neutronics constants prepared by the cell code TVS-M, are the base elements of this chain. It should be reminded that in [6] TVS-M was used only for the constants calculations of MOX FAs. In current calculations the code TVS-M has been used both for UOX and MOX fuel constants. Besides, the volume of presented information has been increased and additional explications have been included. The results for the reference uranium core [4] are presented in Chapter 2. The results for the core with 3 MOX LTAs are presented in Chapter 3. The conservatism that is connected with neutronics parameters and that must be taken into account during transient analysis calculations, is discussed in Chapter 4. The conservative parameters values are considered to be used in 1-point core kinetics models of accident analysis codes

  6. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  8. Code Coupling via Jacobian-Free Newton-Krylov Algorithms with Application to Magnetized Fluid Plasma and Kinetic Neutral Models

    Energy Technology Data Exchange (ETDEWEB)

    Joseph, Ilon [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2014-05-27

    Jacobian-free Newton-Krylov (JFNK) algorithms are a potentially powerful class of methods for solving the problem of coupling codes that address dfferent physics models. As communication capability between individual submodules varies, different choices of coupling algorithms are required. The more communication that is available, the more possible it becomes to exploit the simple sparsity pattern of the Jacobian, albeit of a large system. The less communication that is available, the more dense the Jacobian matrices become and new types of preconditioners must be sought to efficiently take large time steps. In general, methods that use constrained or reduced subsystems can offer a compromise in complexity. The specific problem of coupling a fluid plasma code to a kinetic neutrals code is discussed as an example.

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  10. A comparison of Nodal methods in neutron diffusion calculations

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  12. Joint reconstruction of dynamic PET activity and kinetic parametric images using total variation constrained dictionary sparse coding

    Science.gov (United States)

    Yu, Haiqing; Chen, Shuhang; Chen, Yunmei; Liu, Huafeng

    2017-05-01

    Dynamic positron emission tomography (PET) is capable of providing both spatial and temporal information of radio tracers in vivo. In this paper, we present a novel joint estimation framework to reconstruct temporal sequences of dynamic PET images and the coefficients characterizing the system impulse response function, from which the associated parametric images of the system macro parameters for tracer kinetics can be estimated. The proposed algorithm, which combines statistical data measurement and tracer kinetic models, integrates a dictionary sparse coding (DSC) into a total variational minimization based algorithm for simultaneous reconstruction of the activity distribution and parametric map from measured emission sinograms. DSC, based on the compartmental theory, provides biologically meaningful regularization, and total variation regularization is incorporated to provide edge-preserving guidance. We rely on techniques from minimization algorithms (the alternating direction method of multipliers) to first generate the estimated activity distributions with sub-optimal kinetic parameter estimates, and then recover the parametric maps given these activity estimates. These coupled iterative steps are repeated as necessary until convergence. Experiments with synthetic, Monte Carlo generated data, and real patient data have been conducted, and the results are very promising.

  13. 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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  15. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  17. 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

  18. 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

  19. Tripoli-4, a three-dimensional poly-kinetic particle transport Monte-Carlo code

    International Nuclear Information System (INIS)

    Both, J.P.; Lee, Y.K.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B.; Soldevila, M.

    2003-01-01

    In this updated of the Monte-Carlo transport code Tripoli-4, we list and describe its current main features. The code computes coupled neutron-photon propagation as well as the electron-photon cascade shower. While providing the user with common biasing techniques, it also implements an automatic weighting scheme. Tripoli-4 enables the user to compute the following physical quantities: a flux, a multiplication factor, a current, a reaction rate, a dose equivalent rate as well as deposit of energy and recoil energies. For each interesting physical quantity, a Monte-Carlo simulation offers different types of estimators. Tripoli-4 has support for execution in parallel mode. Special features and applications are also presented

  20. Kinetic---a system code for analyzing nuclear thermal propulsion rocket engine transients

    International Nuclear Information System (INIS)

    Schmidt, E.; Lazareth, O.; Ludewig, H.

    1993-01-01

    A system code suitable for analyzing Nuclear Thermal Propulsion (NTP) rocket engines is described in this paper. The code consists of a point reactor model and nodes to describe the fluid dynamics and heat transfer mechanism. Feedback from the fuel, coolant, moderator and reflector are allowed for, and the control of the reactor is by motion of controls element (drums or rods). The worth of the control element and feedback coefficients are predetermined. Separate models for the turbo-pump assembly (TPA) and nozzle are also included. The model to be described in this paper is specific for the Particle Bed Reactor (PBR). An illustrative problem is solved. This problem consists of a PBR operating in a blowdown mode

  1. KINETIC: A system code for analyzing Nuclear thermal propulsion rocket engine transients

    Science.gov (United States)

    Schmidt, E.; Lazareth, O.; Ludewig, H.

    1993-07-01

    A system code suitable for analyzing Nuclear Thermal Propulsion (NTP) rocket engines is described in this paper. The code consists of a point reactor model and nodes to describe the fluid dynamics and heat transfer mechanism. Feedback from the fuel coolant, moderator and reflector are allowed for, and the control of the reactor is by motion of control elements (drums or rods). The worth of the control clement and feedback coefficients are predetermined. Separate models for the turbo-pump assembly (TPA) and nozzle are also included. The model to be described in this paper is specific for the Particle Bed Reactor (PBR). An illustrative problem is solved. This problem consists of a PBR operating in a blowdown mode.

  2. Tripoli-4, a three-dimensional poly-kinetic particle transport Monte-Carlo code

    Energy Technology Data Exchange (ETDEWEB)

    Both, J P; Lee, Y K; Mazzolo, A; Peneliau, Y; Petit, O; Roesslinger, B; Soldevila, M [CEA Saclay, Dir. de l' Energie Nucleaire (DEN/DM2S/SERMA/LEPP), 91 - Gif sur Yvette (France)

    2003-07-01

    In this updated of the Monte-Carlo transport code Tripoli-4, we list and describe its current main features. The code computes coupled neutron-photon propagation as well as the electron-photon cascade shower. While providing the user with common biasing techniques, it also implements an automatic weighting scheme. Tripoli-4 enables the user to compute the following physical quantities: a flux, a multiplication factor, a current, a reaction rate, a dose equivalent rate as well as deposit of energy and recoil energies. For each interesting physical quantity, a Monte-Carlo simulation offers different types of estimators. Tripoli-4 has support for execution in parallel mode. Special features and applications are also presented.

  3. FX2-TH: a two-dimensional nuclear reactor kinetics code with thermal-hydraulic feedback

    International Nuclear Information System (INIS)

    Shober, R.A.; Daly, T.A.; Ferguson, D.R.

    1978-10-01

    FX2-TH is a two-dimensional, time-dependent nuclear reactor kinetics program with thermal and hydraulic feedback. The neutronics model used is multigroup neutron diffusion theory. The following geometry options are available: x, r, x-y, r-z, theta-r, and triangular. FX2-TH contains two basic thermal and hydraulic models: a simple adiabatic fuel temperature calculation, and a more detailed model consisting of an explicit representation of a fuel pin, gap, clad, and coolant. FX2-TH allows feedback effects from both fuel temperature (Doppler) and coolant temperature (density) changes. FX2-TH will calculate a consistent set of steady state conditions by iterating between the neutronics and thermal-hydraulics until convergence is reached. The time-dependent calculation is performed by the use of the improved quasistatic method. A disk editing capability is available. FX2-TH is operational on IBM system 360 or 370 computers and on the CDC 7600

  4. Global and kinetic MHD simulation by the Gpic-MHD code

    International Nuclear Information System (INIS)

    Naitou, Hiroshi; Yamada, Yusuke; Kajiwara, Kenji; Lee, Wei-li; Tokuda, Shinji; Yagi, Masatoshi

    2011-01-01

    In order to implement large-scale and high-beta tokamak simulation, a new algorithm of the electromagnetic gyrokinetic PIC (particle-in-cell) code was proposed and installed on the Gpic-MHD code [Gyrokinetic PIC code for magnetohydrodynamic (MHD) simulation]. In the new algorithm, the vortex equation and the generalized ohm's law along the magnetic field are derived from the basic equations of the gyrokinetic Vlasov, Poisson, and Ampere system and are used to describe the spatio-temporal evolution of the field quantities of the electrostatic potential φ and the longitudinal component of the vector potential A z . Particle information is mainly used to estimate second order moments in the generalized ohm's law. Because the lower order moments of the charge density and the longitudinal current density are not used explicitly to determine φ and A z , the numerical noise induced by the discreteness of particle quantities reduces drastically. Another advantage of the algorithm is that the longitudinal induced electric field, E Tz =-∂A z /∂t, is explicitly estimated by the generalized ohm's law and used in the equations of motion. The particle velocities along the magnetic field are used (v z -formulation) instead of generalized momentums (p z -formulation), hence there is no problem of 'cancellation', which appear when estimating A z from the Ampere's law in the p z -formulation. The successful simulation of the collisionless internal kink mode by new Gpic-MHD with the realistic values of the large-scale and high-beta, revealed the usefulness of the new algorithm. (author)

  5. The fusion code XGC: Enabling kinetic study of multi-scale edge turbulent transport in ITER

    Energy Technology Data Exchange (ETDEWEB)

    D' Azevedo, Eduardo [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Abbott, Stephen [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Koskela, Tuomas [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Worley, Patrick [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ku, Seung-Hoe [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Ethier, Stephane [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Yoon, Eisung [Rensselaer Polytechnic Inst., Troy, NY (United States); Shephard, Mark [Rensselaer Polytechnic Inst., Troy, NY (United States); Hager, Robert [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Lang, Jianying [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Intel Corporation, Santa Clara, CA (United States); Choi, Jong [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Podhorszki, Norbert [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Klasky, Scott [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Parashar, Manish [Rutgers Univ., Piscataway, NJ (United States); Chang, Choong-Seock [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)

    2017-01-01

    The XGC fusion gyrokinetic code combines state-of-the-art, portable computational and algorithmic technologies to enable complicated multiscale simulations of turbulence and transport dynamics in ITER edge plasma on the largest US open-science computer, the CRAY XK7 Titan, at its maximal heterogeneous capability, which have not been possible before due to a factor of over 10 shortage in the time-to-solution for less than 5 days of wall-clock time for one physics case. Frontier techniques such as nested OpenMP parallelism, adaptive parallel I/O, staging I/O and data reduction using dynamic and asynchronous applications interactions, dynamic repartitioning.

  6. 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

  7. Nodal in computerized control systems of accelerators

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  8. The reactor kinetics code tank: a validation against selected SPERT-1b experiments

    International Nuclear Information System (INIS)

    Ellis, R.J.

    1990-01-01

    The two-dimensional space-time analysis code TANK is being developed for the simulation of transient behaviour in the MAPLE class of research reactors. MAPLE research reactor cores are compact, light-water-cooled and -moderated, with a high degree of forced subcooling. The SPERT-1B(24/32) reactor core had many similarities to MAPLE-X10, and the results of the SPERT transient experiments are well documented. As a validation of TANK, a series of simulations of certain SPERT reactor transients was undertaken. Special features were added to the TANK code to model reactors with plate-type fuel and to allow for the simulation of rapid void production. The results of a series of super-prompt-critical reactivity step-insertion transient simulations are presented. The selected SPERT transients were all initiated from low power, at ambient temperatures, and with negligible coolant flow. Th results of the TANK simulations are in good agreement with the trends in the experimental SPERT data

  9. 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

  10. 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

  11. HELIOS-CR - A 1-D radiation-magnetohydrodynamics code with inline atomic kinetics modeling

    International Nuclear Information System (INIS)

    MacFarlane, J.J.; Golovkin, I.E.; Woodruff, P.R.

    2006-01-01

    HELIOS-CR is a user-oriented 1D radiation-magnetohydrodynamics code to simulate the dynamic evolution of laser-produced plasmas and z-pinch plasmas. It includes an in-line collisional-radiative (CR) model for computing non-LTE atomic level populations at each time step of the hydrodynamics simulation. HELIOS-CR has been designed for ease of use, and is well-suited for experimentalists, as well as graduate and undergraduate student researchers. The energy equations employed include models for laser energy deposition, radiation from external sources, and high-current discharges. Radiative transport can be calculated using either a multi-frequency flux-limited diffusion model, or a multi-frequency, multi-angle short characteristics model. HELIOS-CR supports the use of SESAME equation of state (EOS) tables, PROPACEOS EOS/multi-group opacity data tables, and non-LTE plasma properties computed using the inline CR modeling. Time-, space-, and frequency-dependent results from HELIOS-CR calculations are readily displayed with the HydroPLOT graphics tool. In addition, the results of HELIOS simulations can be post-processed using the SPECT3D Imaging and Spectral Analysis Suite to generate images and spectra that can be directly compared with experimental measurements. The HELIOS-CR package runs on Windows, Linux, and Mac OSX platforms, and includes online documentation. We will discuss the major features of HELIOS-CR, and present example results from simulations

  12. 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)

  13. 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)

  14. SPANDOM - source projection analytic nodal discrete ordinates method

    International Nuclear Information System (INIS)

    Kim, Tae Hyeong; Cho, Nam Zin

    1994-01-01

    We describe a new discrete ordinates nodal method for the two-dimensional transport equation. We solve the discrete ordinates equation analytically after the source term is projected and represented in polynomials. The method is applied to two fast reactor benchmark problems and compared with the TWOHEX code. The results indicate that the present method accurately predicts not only multiplication factor but also flux distribution

  15. 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%

  16. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  18. The nodal discrete-ordinate transport calculation of anisotropy scattering problem in three-dimensional cartesian geometry

    International Nuclear Information System (INIS)

    Wu Hongchun; Xie Zhongsheng; Zhu Xuehua

    1994-01-01

    The nodal discrete-ordinate transport calculating model of anisotropy scattering problem in three-dimensional cartesian geometry is given. The computing code NOTRAN/3D has been encoded and the satisfied conclusion is gained

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

    International Nuclear Information System (INIS)

    Ikeda, Hideaki; Takeda, Toshikazu

    2001-01-01

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

  20. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-15

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

  2. Biological dose estimation for charged-particle therapy using an improved PHITS code coupled with a microdosimetric kinetic model

    International Nuclear Information System (INIS)

    Sato, Tatsuhiko; Watanabe, Ritsuko; Kase, Yuki; Niita, Koji; Sihver, Lembit

    2009-01-01

    -particle therapy was established using the improved PHITS coupled with a microdosimetric kinetic (MK) model. The detailed procedures for improving the PHITS code and establishing the biological-dose-estimation method were presented elsewhere. Hence, this report focuses on describing their capabilities to optimize the treatment planning of charged particle therapy, thereby maximizing the therapeutic effect on tumor while minimizing unintended harmful effects on surrounding normal tissues. (author)

  3. 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)

  4. SCRAM reactivity calculations with the KIKO3D code

    International Nuclear Information System (INIS)

    Hordosy, G.; Kerszturi, A.; Maraczy, Cs.; Temesvari, E.

    1999-01-01

    Discrepancies between calculated static reactivities and measured reactivities evaluated with reactivity meters led to investigating SCRAM with the KIKO3D dynamic code, The time and space dependent neutron flux in the reactor core during the rod drop measurement was calculated by the KIKO3D nodal diffusion code. For calculating the ionisation chamber signals the Green function technique was applied. The Green functions of ionisation chambers were evaluated via solving the neutron transport equation in the reflector regions with the MCNP Monte Carlo code. The detector signals during asymmetric SCRAM measurements were calculated and compared with measured data using the inverse point kinetics transformation. The sufficient agreement validates the KIKO3D code to determine the reactivities after SCRAM. (Authors)

  5. Comparison of 'system thermal-hydraulics-3 dimensional reactor kinetics' coupled calculations using the MARS 1D and 3D modules and the MASTER code

    International Nuclear Information System (INIS)

    Jung, J. J.; Joo, H. K.; Lee, W. J.; Ji, S. K.; Jung, B. D.

    2002-01-01

    KAERI has developed the coupled 'system thermal-hydraulics - 3 dimensional reactor kinetics' code, MARS/MASTER since 1998. However, there is a limitation in the existing MARS/MASTER code; that is, to perform the coupled calculations using MARS/MASTER, we have to utilize the hydrodynamic model and the heat structure model of the MARS '3D module'. In some transients, reactor kinetics behavior is strongly multi-dimensional, but core thermal-hydraulic behavior remains in one-dimensional manner. For efficient analysis of such transients, we coupled the MARS 1D module with MASTER. The new feature has been assessed by the 'OECD NEA Main Steam Line Break (MSLB) benchmark exercise III' simulations

  6. 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)

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  8. 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)

  9. 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.

  10. 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

  11. 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.

  12. Investigation of 3D spatial effect on point kinetics estimation of the thermal hydraulics code RELAP for the analysis of MSLB accident of KK-NP

    International Nuclear Information System (INIS)

    Bera, S.; Pradhan, S.K.; Dubey, S.K.; Gupta, S.K.

    2011-01-01

    In general safety analyses of design basis accident of NPPs are being carried out using system thermal hydraulics code like RELAP. In RELAP, power is calculated based on point kinetics approximation, which virtually ignores the space and energy dependence of neutron flux. To include the space and energy dependence of neutron flux, three-dimensional neutronics code TRIHEXFA has been externally coupled with RELAP through interface program, TRIHEXFA-RELAP Interface Program (TRIP). Calculation methodology of TRIP program is based on adiabatic approximation. In the adiabatic approximation the neutron flux is being factored into spatial and amplitude part. Spatial part of flux is slowly varying with time whereas amplitude part is strongly varying function. The RELAP controls the transient time steps. Transient time is divided into several major and minor time steps. Minor time step is the sub-step of major time step. Thermal hydraulics and neutronics data are exchanged at each major time step. Spatial part of neutron flux has been updated at each major time step using TRIHEXFA code. But amplitude part of the neutron flux is calculated at each minor time step using RELAP code. Convergence of results of the coupled code, TRIP has been checked through coupling time step descritization study. This study determines the minimum coupling time step. Transient concerning VVER-1000 Main Steam Line Break, MSLB has been considered to investigate the space-time effect on point kinetics. MSLB occurs as a consequence of the rupture of one steam line upstream of main steam line isolation valves. Reference design and data from Kudankulam Nuclear Power Plant (KK-NPP) are used for the analysis. From this investigation it is found that TRIP significantly overestimates the maximum reactor power against uncoupled RELAP result. The time of scram also occur six seconds earlier in TRIP calculation compared to the RELAP. This exercise has also shown a proof of principle that coupling 3D

  13. Development of a kinetics analysis code for fuel solution combined with thermal-hydraulics analysis code PHOENICS and analysis of natural-cooling characteristic test of TRACY. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Shouichi; Yamane, Yuichi; Miyoshi, Yoshinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Since exact information is not always acquired in the criticality accident of fuel solution, parametric survey calculations are required for grasping behaviors of the thermal-hydraulics. On the other hand, the practical methods of the calculation with can reduce the computation time with allowable accuracy will be also required, since the conventional method takes a long calculation time. In order to fulfill the requirement, a two-dimensional (R-Z) nuclear-kinetics analysis code considering thermal-hydraulic based on the multi-region kinetic equations with one-group neutron energy was created by incorporating with the thermal-hydraulics analysis code PHOENICS for all-purpose use the computation time of the code was shortened by separating time mesh intervals of the nuclear- and heat-calculations from that of the hydraulics calculation, and by regulating automatically the time mesh intervals in proportion to power change rate. A series of analysis were performed for the natural-cooling characteristic test using TRACY in which the power changed slowly for 5 hours after the transient power resulting from the reactivity insertion of a 0.5 dollar. It was found that the code system was able to calculate within the limit of practical time, and acquired the prospect of reproducing the experimental values considerably for the power and temperature change. (author)

  14. 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

  15. SHOVAV-JUEL. A one dimensional space-time kinetic code for pebble-bed high-temperature reactors with temperature and Xenon feedback

    International Nuclear Information System (INIS)

    Nabbi, R.; Meister, G.; Finken, R.; Haben, M.

    1982-09-01

    The present report describes the modelling basis and the structure of the neutron kinetics-code SHOVAV-Juel. Information for users is given regarding the application of the code and the generation of the input data. SHOVAV-Juel is a one-dimensional space-time-code based on a multigroup diffusion approach for four energy groups and six groups of delayed neutrons. It has been developed for the analysis of the transient behaviour of high temperature reactors with pebble-bed core. The reactor core is modelled by horizontal segments to which different materials compositions can be assigned. The temperature dependence of the reactivity is taken into account by using temperature dependent neutron cross sections. For the simulation of transients in an extended time range the time dependence of the reactivity absorption by Xenon-135 is taken into account. (orig./RW)

  16. Present Status of GNF New Nodal Simulator

    International Nuclear Information System (INIS)

    Iwamoto, T.; Tamitani, M.; Moore, B.

    2001-01-01

    This paper presents core simulator consolidation work done at Global Nuclear Fuel (GNF). The unified simulator needs to supercede the capabilities of past simulator packages from the original GNF partners: GE, Hitachi, and Toshiba. At the same time, an effort is being made to produce a simulation package that will be a state-of-the-art analysis tool when released, in terms of the physics solution methodology and functionality. The core simulator will be capable and qualified for (a) high-energy cycles in the U.S. markets, (b) mixed-oxide (MOX) introduction in Japan, and (c) high-power density plants in Europe, etc. The unification of the lattice physics code is also in progress based on a transport model with collision probability methods. The AETNA core simulator is built upon the PANAC11 software base. The goal is to essentially replace the 1.5-energy-group model with a higher-order multigroup nonlinear nodal solution capable of the required modeling fidelity, while keeping highly automated library generation as well as functionality. All required interfaces to PANAC11 will be preserved, which minimizes the impact on users and process automation. Preliminary results show statistical accuracy improvement over the 1.5-group model

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

    International Nuclear Information System (INIS)

    Hebert, Alain

    2008-01-01

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

  18. 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)

  19. Approximate Schur complement preconditioning of the lowest order nodal discretizations

    Energy Technology Data Exchange (ETDEWEB)

    Moulton, J.D.; Ascher, U.M. [Univ. of British Columbia, Vancouver, British Columbia (Canada); Morel, J.E. [Los Alamos National Lab., NM (United States)

    1996-12-31

    Particular classes of nodal methods and mixed hybrid finite element methods lead to equivalent, robust and accurate discretizations of 2nd order elliptic PDEs. However, widespread popularity of these discretizations has been hindered by the awkward linear systems which result. The present work exploits this awkwardness, which provides a natural partitioning of the linear system, by defining two optimal preconditioners based on approximate Schur complements. Central to the optimal performance of these preconditioners is their sparsity structure which is compatible with Dendy`s black box multigrid code.

  20. LSENS: A General Chemical Kinetics and Sensitivity Analysis Code for homogeneous gas-phase reactions. Part 1: Theory and numerical solution procedures

    Science.gov (United States)

    Radhakrishnan, Krishnan

    1994-01-01

    LSENS, the Lewis General Chemical Kinetics and Sensitivity Analysis Code, has been developed for solving complex, homogeneous, gas-phase chemical kinetics problems and contains sensitivity analysis for a variety of problems, including nonisothermal situations. This report is part 1 of a series of three reference publications that describe LENS, provide a detailed guide to its usage, and present many example problems. Part 1 derives the governing equations and describes the numerical solution procedures for the types of problems that can be solved. The accuracy and efficiency of LSENS are examined by means of various test problems, and comparisons with other methods and codes are presented. LSENS is a flexible, convenient, accurate, and efficient solver for chemical reaction problems such as static system; steady, one-dimensional, inviscid flow; reaction behind incident shock wave, including boundary layer correction; and perfectly stirred (highly backmixed) reactor. In addition, the chemical equilibrium state can be computed for the following assigned states: temperature and pressure, enthalpy and pressure, temperature and volume, and internal energy and volume. For static problems the code computes the sensitivity coefficients of the dependent variables and their temporal derivatives with respect to the initial values of the dependent variables and/or the three rate coefficient parameters of the chemical reactions.

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

    International Nuclear Information System (INIS)

    Ichikawa, Ryoko; Masuhara, Yasuhiro; Kasahara, Fumio

    2012-01-01

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

  2. 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...

  3. Analysis of the OECD main steam line break benchmark using ANC-K/MIDAC code

    International Nuclear Information System (INIS)

    Aoki, Shigeaki; Tahara, Yoshihisa; Suemura, Takayuki; Ogawa, Junto

    2004-01-01

    A three-dimensional (3D) neutronics and thermal-and-hydraulics (T/H) coupling code ANC-K/MIDAC has been developed. It is the combination of the 3D nodal kinetic code ANC-K and the 3D drift flux thermal hydraulic code MIDAC. In order to verify the adequacy of this code, we have performed several international benchmark problems. In this paper, we show the calculation results of ''OECD Main Steam Line Break Benchmark (MSLB benchmark)'', which gives the typical local power peaking problem. And we calculated the return-to-power scenario of the Phase II problem. The comparison of the results shows the very good agreement of important core parameters between the ANC-K/MIDAC and other participant codes. (author)

  4. Adaption, validation and application of advanced codes with 3-dimensional neutron kinetics for accident analysis calculations - STC with Bulgaria

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.; Mittag, S.; Rohde, U.; Seidel, A.; Panayotov, D.; Ilieva, B.

    2001-08-01

    In the frame of a project on scientific-technical co-operation funded by BMBF/BMWi, the program code DYN3D and the coupled code ATHLET-DYN3D have been transferred to the Institute for Nuclear Research and Nuclear Energy (INRNE) Sofia. The coupled code represents an implementation of the 3D core model DYN3D developed by FZR into the GRS thermal-hydraulics code system ATHLET. For the purpose of validation of these codes, a measurement data base about a start-up experiment obtained at the unit 6 of Kozloduy NPP (VVER-1000/V-320) has been generated. The results of performed validation calculations were compared with measurement values from the data base. A simplified model for estimation of cross flow mixing between fuel assemblies has been implemented into the program code DYN3D by Bulgarian experts. Using this cross flow model, transient processes with asymmetrical boundary conditions can be analysed more realistic. The validation of the implemented model were performed with help of comparison calculations between modified DYD3D code and thermal-hydraulics code COBRA-4I, and also on the base of the collected measurement data from Kozloduy NPP. (orig.) [de

  5. 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.

  6. 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.

  7. 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

  8. 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.

  9. 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.

  10. Development of the coupled 'system thermal-hydraulics, 3D reactor kinetics, and hot channel' analysis capability of the MARS code

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, J. J.; Chung, B. D.; Lee, W.J

    2005-02-01

    The subchannel analysis capability of the MARS 3D module has been improved. Especially, the turbulent mixing and void drift models for flow mixing phenomena in rod bundles have been assessed using some well-known rod bundle test data. Then, the subchannel analysis feature was combined to the existing coupled 'system Thermal-Hydraulics (T/H) and 3D reactor kinetics' calculation capability of MARS. These features allow the coupled 'system T/H, 3D reactor kinetics, and hot channel' analysis capability and, thus, realistic simulations of hot channel behavior as well as global system T/H behavior. In this report, the MARS code features for the coupled analysis capability are described first. The code modifications relevant to the features are also given. Then, a coupled analysis of the Main Steam Line Break (MSLB) is carried out for demonstration. The results of the coupled calculations are very reasonable and realistic, and show these methods can be used to reduce the over-conservatism in the conventional safety analysis.

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  13. 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

  14. Code Cactus; Code Cactus

    Energy Technology Data Exchange (ETDEWEB)

    Fajeau, M; Nguyen, L T; Saunier, J [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

    1966-09-01

    This code handles the following problems: -1) Analysis of thermal experiments on a water loop at high or low pressure; steady state or transient behavior; -2) Analysis of thermal and hydrodynamic behavior of water-cooled and moderated reactors, at either high or low pressure, with boiling permitted; fuel elements are assumed to be flat plates: - Flowrate in parallel channels coupled or not by conduction across plates, with conditions of pressure drops or flowrate, variable or not with respect to time is given; the power can be coupled to reactor kinetics calculation or supplied by the code user. The code, containing a schematic representation of safety rod behavior, is a one dimensional, multi-channel code, and has as its complement (FLID), a one-channel, two-dimensional code. (authors) [French] Ce code permet de traiter les problemes ci-dessous: 1. Depouillement d'essais thermiques sur boucle a eau, haute ou basse pression, en regime permanent ou transitoire; 2. Etudes thermiques et hydrauliques de reacteurs a eau, a plaques, a haute ou basse pression, ebullition permise: - repartition entre canaux paralleles, couples on non par conduction a travers plaques, pour des conditions de debit ou de pertes de charge imposees, variables ou non dans le temps; - la puissance peut etre couplee a la neutronique et une representation schematique des actions de securite est prevue. Ce code (Cactus) a une dimension d'espace et plusieurs canaux, a pour complement Flid qui traite l'etude d'un seul canal a deux dimensions. (auteurs)

  15. Development of Regulatory Audit Core Safety Code : COREDAX

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Chae Yong; Jo, Jong Chull; Roh, Byung Hwan [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Jae Jun; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2005-07-01

    Korea Institute of Nuclear Safety (KINS) has developed a core neutronics simulator, COREDAX code, for verifying core safety of SMART-P reactor, which is technically supported by Korea Advanced Institute of Science and Technology (KAIST). The COREDAX code would be used for regulatory audit calculations of 3- dimendional core neutronics. The COREDAX code solves the steady-state and timedependent multi-group neutron diffusion equation in hexagonal geometry as well as rectangular geometry by analytic function expansion nodal (AFEN) method. AFEN method was developed at KAIST, and it was internationally verified that its accuracy is excellent. The COREDAX code is originally programmed based on the AFEN method. Accuracy of the code on the AFEN method was excellent for the hexagonal 2-dimensional problems, but there was a need for improvement for hexagonal-z 3-dimensional problems. Hence, several solution routines of the AFEN method are improved, and finally the advanced AFEN method is created. COREDAX code is based on the advanced AFEN method . The initial version of COREDAX code is to complete a basic framework, performing eigenvalue calculations and kinetics calculations with thermal-hydraulic feedbacks, for audit calculations of steady-state core design and reactivity-induced accidents of SMART-P reactor. This study describes the COREDAX code for hexagonal geometry.

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

    International Nuclear Information System (INIS)

    Jacqmin, R.P.

    1991-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-12-10

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Jacqmin, R.P.

    1991-12-10

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

  19. 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.

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

    International Nuclear Information System (INIS)

    Peng Muzhang; Zhang Quan; Wang Guoli; Zhang Yuman

    1988-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-03-01

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

  2. Complex models of nodal nuclear data

    International Nuclear Information System (INIS)

    Dufek, Jan

    2011-01-01

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

  3. 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.

  4. 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...

  5. Improvements and validation of the transient analysis code MOREL for molten salt reactors

    International Nuclear Information System (INIS)

    Zhuang Kun; Zheng Youqi; Cao Liangzhi; Hu Tianliang; Wu Hongchun

    2017-01-01

    The liquid fuel salt used in the molten salt reactors (MSRs) serves as the fuel and coolant simultaneously. On the one hand, the delayed neutron precursors circulate in the whole primary loop and part of them decay outside the core. On the other hand, the fission heat is carried off directly by the fuel flow. These two features require new analysis method with the coupling of fluid flow, heat transfer and neutronics. In this paper, the recent update of MOREL code is presented. The update includes: (1) the improved quasi-static method for the kinetics equation with convection term is developed. (2) The multi-channel thermal hydraulic model is developed based on the geometric feature of MSR. (3) The Variational Nodal Method is used to solve the neutron diffusion equation instead of the original analytic basis functions expansion nodal method. The update brings significant improvement on the efficiency of MOREL code. And, the capability of MOREL code is extended for the real core simulation with feedback. The numerical results and experiment data gained from molten salt reactor experiment (MSRE) are used to verify and validate the updated MOREL code. The results agree well with the experimental data, which prove the new development of MOREL code is correct and effective. (author)

  6. Analysis of the three dimensional core kinetics NESTLE code coupling with the advanced thermo-hydraulic code systems, RELAP5/SCDAPSIM and its application to the Laguna Verde Central reactor

    International Nuclear Information System (INIS)

    Salazar C, J.H.; Nunez C, A.; Chavez M, C.

    2004-01-01

    The objective of the written present is to propose a methodology for the joining of the codes RELAP5/SCDAPSIM and NESTLE. The development of this joining will be carried out inside a doctoral program of Engineering in Energy with nuclear profile of the Ability of Engineering of the UNAM together with the National Commission of Nuclear Security and Safeguards (CNSNS). The general purpose of this type of developments, is to have tools that are implemented by multiple programs or codes such a that systems or models of the three-dimensional kinetics of the core can be simulated and those of the dynamics of the reactor (water heater-hydraulics). In the past, by limitations for the calculation of the complete answer of both systems, the developed models they were carried out for separate, putting a lot of emphasis in one but neglecting the other one. These methodologies, calls of better estimate, will be good to the nuclear industry to evaluate, with more high grades of detail, the designs of the nuclear power plant (for modifications to those already existent or for new concepts in the designs of advanced reactors), besides analysing events (transitory and have an accident), among other applications. The coupled system was applied to design studies and investigation of the Laguna Verde Nuclear power plant (CNLV). (Author)

  7. Power feedback effects in the LEM code

    International Nuclear Information System (INIS)

    Kromar, M.

    1999-01-01

    The nodal diffusion code LEM has been extended with the power feedback option. Thermohydraulic and neutronic coupling is covered with the Reactivity Coefficient Method. Presented are results of the code testing. Verification is done on the typical non-uprated NPP Krsko reload cycles. Results show that the code fulfill objectives arising in the process of reactor core analysis.(author)

  8. 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)

  9. 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)

  10. Kinetic instabilities of thin current sheets: Results of two-and-one-half-dimensional Vlasov code simulations

    International Nuclear Information System (INIS)

    Silin, I.; Buechner, J.

    2003-01-01

    Nonlinear triggering of the instability of thin current sheets is investigated by two-and-one-half- dimensional Vlasov code simulations. A global drift-resonant instability (DRI) is found, which results from the lower-hybrid-drift waves penetrating from the current sheet edges to the center where they resonantly interact with unmagnetized ions. This resonant nonlinear instability grows faster than a Kelvin-Helmholtz instability obtained in previous studies. The DRI is either asymmetric or symmetric mode or a combination of the two, depending on the relative phase of the lower-hybrid-drift waves at the edges of the current sheet. With increasing particle mass ratio the wavenumber of the fastest-growing mode increases as kL z ∼(m i /m e ) 1/2 /2 and the growth rate of the DRI saturates at a finite level

  11. 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

  12. 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)

  13. Comparison of PANTHER nodal solutions in hexagonal-z geometry

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  14. Final Report, Nuclear Energy Research Initiative (NERI) Project: An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model

    International Nuclear Information System (INIS)

    Anistratov, Dmitriy Y.; Adams, Marvin L.; Palmer, Todd S.; Smith, Kord S.; Clarno, Kevin; Hikaru Hiruta; Razvan Nes

    2003-01-01

    OAK (B204) Final Report, NERI Project: ''An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model'' The present generation of reactor analysis methods uses few-group nodal diffusion approximations to calculate full-core eigenvalues and power distributions. The cross sections, diffusion coefficients, and discontinuity factors (collectively called ''group constants'') in the nodal diffusion equations are parameterized as functions of many variables, ranging from the obvious (temperature, boron concentration, etc.) to the more obscure (spectral index, moderator temperature history, etc.). These group constants, and their variations as functions of the many variables, are calculated by assembly-level transport codes. The current methodology has two main weaknesses that this project addressed. The first weakness is the diffusion approximation in the full-core calculation; this can be significantly inaccurate at interfaces between different assemblies. This project used the nodal diffusion framework to implement nodal quasidiffusion equations, which can capture transport effects to an arbitrary degree of accuracy. The second weakness is in the parameterization of the group constants; current models do not always perform well, especially at interfaces between unlike assemblies. The project developed a theoretical foundation for parameterization and homogenization models and used that theory to devise improved models. The new models were extended to tabulate information that the nodal quasidiffusion equations can use to capture transport effects in full-core calculations

  15. Development of an advanced code system for fast-reactor transient analysis

    International Nuclear Information System (INIS)

    Konstantin Mikityuk; Sandro Pelloni; Paul Coddington

    2005-01-01

    FAST (Fast-spectrum Advanced Systems for power production and resource management) is a recently approved PSI activity in the area of fast spectrum core and safety analysis with emphasis on generic developments and Generation IV systems. In frames of the FAST project we will study both statics and transients core physics, reactor system behaviour and safety; related international experiments. The main current goal of the project is to develop unique analytical and code capability for core and safety analysis of critical (and sub-critical) fast spectrum systems with an initial emphasis on a gas cooled fast reactors. A structure of the code system is shown on Fig. 1. The main components of the FAST code system are 1) ERANOS code for preparation of basic x-sections and their partial derivatives; 2) PARCS transient nodal-method multi-group neutron diffusion code for simulation of spatial (3D) neutron kinetics in hexagonal and square geometries; 3) TRAC/AAA code for system thermal hydraulics; 4) FRED transient model for fuel thermal-mechanical behaviour; 5) PVM system as an interface between separate parts of the code system. The paper presents a structure of the code system (Fig. 1), organization of interfaces and data exchanges between main parts of the code system, examples of verification and application of separate codes and the system as a whole. (authors)

  16. 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.

  17. 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.

  18. 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

  19. 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.

  20. 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.

  1. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lawrence, R.D.

    1983-03-01

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

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

    International Nuclear Information System (INIS)

    Lawrence, R.D.

    1983-03-01

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

  4. 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)

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

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.

    2003-12-01

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

  6. Acceleration of the FERM nodal program

    International Nuclear Information System (INIS)

    Nakata, H.

    1985-01-01

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

  7. 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

  8. 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

  9. 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

  10. Depletion Calculations for MTR Core Using MCNPX and Multi-Group Nodal Diffusion Methods

    International Nuclear Information System (INIS)

    Jaradata, Mustafa K.; Park, Chang Je; Lee, Byungchul

    2013-01-01

    In order to maintain a self-sustaining steady-state chain reaction, more fuel than is necessary in order to maintain a steady state chain reaction must be loaded. The introduction of this excess fuel increases the net multiplication capability of the system. In this paper MCNPX and multi-group nodal diffusion theory will be used for depletion calculations for MTR core. The eigenvalue and power distribution in the core will be compared for different burnup. Multi-group nodal diffusion theory with combination of NEWT-TRITON system was used to perform depletion calculations for 3Χ3 MTR core. 2G and 6G approximations were used and compared with MCNPX results for 2G approximation the maximum difference from MCNPX was 40 mk and for 6G approximation was 6 mk which is comparable to the MCNPX results. The calculated power using nodal code was almost the same MCNPX results. Finally the results of the multi-group nodal theory were acceptable and comparable to the calculated using MCNPX

  11. 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...

  12. New sparse matrix solver in the KIKO3D 3-dimensional reactor dynamics code

    International Nuclear Information System (INIS)

    Panka, I.; Kereszturi, A.; Hegedus, C.

    2005-01-01

    The goal of this paper is to present a more effective method Bi-CGSTAB for accelerating the large sparse matrix equation solution in the KIKO3D code. This equation system is obtained by using the factorization of the improved quasi static (IQS) method for the time dependent nodal kinetic equations. In the old methodology standard large sparse matrix techniques were considered, where Gauss-Seidel preconditioning and a GMRES-type solver were applied. The validation of KIKO3D using Bi-CGSTAB has been performed by solving of a VVER-1000 kinetic benchmark problem. Additionally, the convergence characteristics were investigated in given macro time steps of Control Rod Ejection transients. The results have been obtained by the old GMRES and new Bi-CGSTAB methods are compared. (author)

  13. 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)

  14. Topological surface states in nodal superconductors.

    Science.gov (United States)

    Schnyder, Andreas P; Brydon, Philip M R

    2015-06-24

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

  15. Twisted vector bundles on pointed nodal curves

    Indian Academy of Sciences (India)

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

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

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

    International Nuclear Information System (INIS)

    Ferreira, C.R.

    1984-01-01

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

  17. An analytical nodal method for time-dependent one-dimensional discrete ordinates problems

    International Nuclear Information System (INIS)

    Barros, R.C. de

    1992-01-01

    In recent years, relatively little work has been done in developing time-dependent discrete ordinates (S N ) computer codes. Therefore, the topic of time integration methods certainly deserves further attention. In this paper, we describe a new coarse-mesh method for time-dependent monoenergetic S N transport problesm in slab geometry. This numerical method preserves the analytic solution of the transverse-integrated S N nodal equations by constants, so we call our method the analytical constant nodal (ACN) method. For time-independent S N problems in finite slab geometry and for time-dependent infinite-medium S N problems, the ACN method generates numerical solutions that are completely free of truncation errors. Bsed on this positive feature, we expect the ACN method to be more accurate than conventional numerical methods for S N transport calculations on coarse space-time grids

  18. Reconstruction of pin burnup characteristics from nodal calculations in hexagonal geometry

    International Nuclear Information System (INIS)

    Yang, W.S.; Finck, P.J.; Khalil, H.S.

    1990-01-01

    A reconstruction method has been developed for recovering pin burnup characteristics from fuel cycle calculations performed in hexagonal-z geometry using the nodal diffusion option of the DIF3D/REBUS-3 code system. Intra-modal distributions of group fluxes, nuclide densities, power density, burnup, and fluence are efficiently computed using polynomial shapes constrained to satisfy nodal information. The accuracy of the method has been tested by performing several numerical benchmark calculations and by comparing predicted local burnups to values measured for experimental assemblies in EBR-11. The results indicate that the reconstruction methods are quite accurate, yielding maximum errors in power and nuclide densities that are less than 2% for driver assemblies and typically less than 5% for blanket assemblies. 14 refs., 2 figs., 5 tabs

  19. 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

  20. 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

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

    International Nuclear Information System (INIS)

    Aragones, J.M.

    2008-01-01

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

  2. 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.

  3. Contribution to the study by a Monte-Carlo method of mono-kinetic particles propagation in a cavity simulated by rectangular cubes. CODE Cupidon 2 (version 29); Contribution a l'etude par une methode de Monte-Carlo de la propagation de particules monocinetiques dans une enceinte schematisee par un assemblage de parallelepipedes rectangles. Programme Cupidon 2 (version 29)

    Energy Technology Data Exchange (ETDEWEB)

    Duco-Sivagnanam, J

    1967-07-01

    The Cupidon 2 CODE aims to calculate the mono-kinetic neutrons flux in an assembly of cubes cavities jointed by rectangular holes. This report is a partial description of the code Cupidon 2 which explains the calculation procedure: data entry, code limits...). (A.L.B.)

  4. Radiotherapy of adult nodal non Hodgkin's lymphoma

    International Nuclear Information System (INIS)

    Gamen, G.; Thirion, P.

    1999-01-01

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

  5. Error estimation for variational nodal calculations

    International Nuclear Information System (INIS)

    Zhang, H.; Lewis, E.E.

    1998-01-01

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

  6. A coarse-mesh nodal method-diffusive-mesh finite difference method

    International Nuclear Information System (INIS)

    Joo, H.; Nichols, W.R.

    1994-01-01

    Modern nodal methods have been successfully used for conventional light water reactor core analyses where the homogenized, node average cross sections (XSs) and the flux discontinuity factors (DFs) based on equivalence theory can reliably predict core behavior. For other types of cores and other geometries characterized by tightly-coupled, heterogeneous core configurations, the intranodal flux shapes obtained from a homogenized nodal problem may not accurately portray steep flux gradients near fuel assembly interfaces or various reactivity control elements. This may require extreme values of DFs (either very large, very small, or even negative) to achieve a desired solution accuracy. Extreme values of DFs, however, can disrupt the convergence of the iterative methods used to solve for the node average fluxes, and can lead to a difficulty in interpolating adjacent DF values. Several attempts to remedy the problem have been made, but nothing has been satisfactory. A new coarse-mesh nodal scheme called the Diffusive-Mesh Finite Difference (DMFD) technique, as contrasted with the coarse-mesh finite difference (CMFD) technique, has been developed to resolve this problem. This new technique and the development of a few-group, multidimensional kinetics computer program are described in this paper

  7. 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 ...

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  9. 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

  10. Simulation of A Main Steam Line Break Accident Using the Coupled 'System Thermal-Hydraulics, 3D reactor Kinetics, and Hot Channel' Analysis Capability of MARS 3.0

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jae Jun; Chung, Bub Dong

    2005-09-15

    For realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients. In this work, a main steam line break (MSLB) accident is simulated using the coupled 'system T-H, 3-D reactor kinetics, and hot channel analysis' feature of the MARS code. Two coupled calculations are performed for demonstration. First, a coupled calculation of the 'system T-H and 3-D reactor kinetics' with a refined core T-H nodalization is carried out to obtain global core power and local departure from nucleate boiling (DNB) ratio (DNBR) behaviors. Next, for a more accurate DNBR prediction, another coupled calculation with subchannel meshes for the hot channels is performed. The results of the coupled calculations are very reasonable and consistent so that these can be used to remove the excessive conservatism in the conventional safety analysis.

  11. Simulation of A Main Steam Line Break Accident Using the Coupled 'System Thermal-Hydraulics, 3D reactor Kinetics, and Hot Channel' Analysis Capability of MARS 3.0

    International Nuclear Information System (INIS)

    Jeong, Jae Jun; Chung, Bub Dong

    2005-09-01

    For realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients. In this work, a main steam line break (MSLB) accident is simulated using the coupled 'system T-H, 3-D reactor kinetics, and hot channel analysis' feature of the MARS code. Two coupled calculations are performed for demonstration. First, a coupled calculation of the 'system T-H and 3-D reactor kinetics' with a refined core T-H nodalization is carried out to obtain global core power and local departure from nucleate boiling (DNB) ratio (DNBR) behaviors. Next, for a more accurate DNBR prediction, another coupled calculation with subchannel meshes for the hot channels is performed. The results of the coupled calculations are very reasonable and consistent so that these can be used to remove the excessive conservatism in the conventional safety analysis

  12. Topological surface states in nodal superconductors

    International Nuclear Information System (INIS)

    Schnyder, Andreas P; Brydon, Philip M R

    2015-01-01

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

  13. Nodal aberration theory applied to freeform surfaces

    Science.gov (United States)

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

    2014-12-01

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

  14. Development of a coupled code system based on system transient code, RETRAN, and 3-D neutronics code, MASTER

    International Nuclear Information System (INIS)

    Kim, K. D.; Jung, J. J.; Lee, S. W.; Cho, B. O.; Ji, S. K.; Kim, Y. H.; Seong, C. K.

    2002-01-01

    A coupled code system of RETRAN/MASTER has been developed for best-estimate simulations of interactions between reactor core neutron kinetics and plant thermal-hydraulics by incorporation of a 3-D reactor core kinetics analysis code, MASTER into system transient code, RETRAN. The soundness of the consolidated code system is confirmed by simulating the MSLB benchmark problem developed to verify the performance of a coupled kinetics and system transient codes by OECD/NEA

  15. International training program: 3D S.UN.COP - Scaling, uncertainty and 3D thermal-hydraulics/neutron-kinetics coupled codes seminar

    International Nuclear Information System (INIS)

    Petruzzi, A.; D'Auria, F.; Bajs, T.; Reventos, F.

    2006-01-01

    Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers and vendors, nuclear fuel companies, research organizations, consulting companies, and technical support organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the 'user effect' and stems from the limitations embedded in the codes as well as from the limited capability of the analysts to use the codes. Code user training and qualification is an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. The 3D S.UN.COP 2005 (Scaling, Uncertainty and 3D COuPled code calculations) seminar has been organized by University of Pisa and University of Zagreb as follow-up of the proposal to IAEA for the Permanent Training Course for System Code Users (D'Auria, 1998). It was recognized that such a course represented both a source of continuing education for current code users and a means for current code users to enter the formal training structure of a proposed 'permanent' stepwise approach to user training. The seminar-training was successfully held with the participation of 19 persons coming from 9 countries and 14 different institutions (universities, vendors, national laboratories and regulatory bodies). More than 15 scientists were involved in the organization of the seminar, presenting theoretical aspects of the proposed methodologies and holding the training and the final examination. A certificate (LA Code User grade) was released

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

    International Nuclear Information System (INIS)

    Grundmann, U.

    1995-11-01

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

  17. Three Mile Island Unit 1 Main Steam Line Break Three-Dimensional Neutronics/Thermal-Hydraulics Analysis: Application of Different Coupled Codes

    International Nuclear Information System (INIS)

    D'Auria, Francesco; Moreno, Jose Luis Gago; Galassi, Giorgio Maria; Grgic, Davor; Spadoni, Antonino

    2003-01-01

    A comprehensive analysis of the double ended main steam line break (MSLB) accident assumed to occur in the Babcock and Wilcox Three Mile Island Unit 1 (TMI-1) has been carried out at the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione of the University of Pisa, Italy, in cooperation with the University of Zagreb, Croatia. The overall activity has been completed within the framework of the participation in the Organization for Economic Cooperation and Development-Committee on the Safety of Nuclear Installations-Nuclear Science Committee pressurized water reactor MSLB benchmark.Thermal-hydraulic system codes (various versions of Relap5), three-dimensional (3-D) neutronics codes (Parcs, Quabbox, and Nestle), and one subchannel code (Cobra) have been adopted for the analysis. Results from the following codes (or code versions) are assumed as reference:1. Relap5/mod3.2.2, beta version, coupled with the 3-D neutron kinetics Parcs code parallel virtual machine (PVM) coupling2. Relap5/mod3.2.2, gamma version, coupled with the 3-D neutron kinetics Quabbox code (direct coupling)3. Relap5/3D code coupled with the 3-D neutron kinetics Nestle code.The influence of PVM and of direct coupling is also discussed.Boundary and initial conditions of the system, including those relevant to the fuel status, have been supplied by Pennsylvania State University in cooperation with GPU Nuclear Corporation (the utility, owner of TMI) and the U.S. Nuclear Regulatory Commission. The comparison among the results obtained by adopting the same thermal-hydraulic nodalization and the coupled code version is discussed in this paper.The capability of the control rods to recover the accident has been demonstrated in all the cases as well as the capability of all the codes to predict the time evolution of the assigned transient. However, one stuck control rod caused some 'recriticality' or 'return to power' whose magnitude is largely affected by boundary and initial conditions

  18. Aztheca Code

    International Nuclear Information System (INIS)

    Quezada G, S.; Espinosa P, G.; Centeno P, J.; Sanchez M, H.

    2017-09-01

    This paper presents the Aztheca code, which is formed by the mathematical models of neutron kinetics, power generation, heat transfer, core thermo-hydraulics, recirculation systems, dynamic pressure and level models and control system. The Aztheca code is validated with plant data, as well as with predictions from the manufacturer when the reactor operates in a stationary state. On the other hand, to demonstrate that the model is applicable during a transient, an event occurred in a nuclear power plant with a BWR reactor is selected. The plant data are compared with the results obtained with RELAP-5 and the Aztheca model. The results show that both RELAP-5 and the Aztheca code have the ability to adequately predict the behavior of the reactor. (Author)

  19. Solving two-dimensions heat conduction problem for fuel elements in reactor by nodal green's function method

    International Nuclear Information System (INIS)

    Tang Jian; Peng Muzhang; Cao Dongxing

    1989-01-01

    A new numerical method-nodal green's function method is used for solving heat conduction function. A heat conduction problem in cylindrical geometry with axial conduction is solved in this paper. The Kirchhoff transformation is used to deal with the problem with temperature dependent conductivity. Therefor, the calculation for the function is simplified. On the basis of the formulas developed, the code named NGMEFC is programmed. A sample problem which has been calculated by the code COBRA-IV is chosen as checking. A good agreement between both codes is achieved. The calculation shows that the calculation efficiency of the nodel green's function method is much higher than that of finite difference method

  20. Determination of power distribution in reactor with nodal expansion method; Izrachun porazdelitve mochi v reaktorju z metodo nodalne ekspanzije

    Energy Technology Data Exchange (ETDEWEB)

    Kromar, M; Trkov, A [Institut Jozef Stefan, Ljubljana (Yugoslavia); Pregl, G [Tehnishka Fakulteta Maribor Univ. (Yugoslavia)

    1988-07-01

    Nodal expansion method (NEM) is one of the advanced coarse-mesh methods based on integral form of few-group diffusion equation. NEM can be characterized by high accuracy and computational efficiency. Method was tested by development of computer code NEXT. Validation of the code was performed by calculation of 2-D and 3-D IAEA benchmark problem. NEXT was compared with codes based on other methods (finite differences, finite elements) and has been found to be accurate as well as fast. (author)

  1. Introduction of SCIENCE code package

    International Nuclear Information System (INIS)

    Lu Haoliang; Li Jinggang; Zhu Ya'nan; Bai Ning

    2012-01-01

    The SCIENCE code package is a set of neutronics tools based on 2D assembly calculations and 3D core calculations. It is made up of APOLLO2F, SMART and SQUALE and used to perform the nuclear design and loading pattern analysis for the reactors on operation or under construction of China Guangdong Nuclear Power Group. The purpose of paper is to briefly present the physical and numerical models used in each computation codes of the SCIENCE code pack age, including the description of the general structure of the code package, the coupling relationship of APOLLO2-F transport lattice code and SMART core nodal code, and the SQUALE code used for processing the core maps. (authors)

  2. Intercomparison of the finite difference and nodal discrete ordinates and surface flux transport methods for a LWR pool-reactor benchmark problem in X-Y geometry

    International Nuclear Information System (INIS)

    O'Dell, R.D.; Stepanek, J.; Wagner, M.R.

    1983-01-01

    The aim of the present work is to compare and discuss the three of the most advanced two dimensional transport methods, the finite difference and nodal discrete ordinates and surface flux method, incorporated into the transport codes TWODANT, TWOTRAN-NODAL, MULTIMEDIUM and SURCU. For intercomparison the eigenvalue and the neutron flux distribution are calculated using these codes in the LWR pool reactor benchmark problem. Additionally the results are compared with some results obtained by French collision probability transport codes MARSYAS and TRIDENT. Because the transport solution of this benchmark problem is close to its diffusion solution some results obtained by the finite element diffusion code FINELM and the finite difference diffusion code DIFF-2D are included

  3. 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

  4. 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.

  5. WWER reactor physics code applications

    International Nuclear Information System (INIS)

    Gado, J.; Kereszturi, A.; Gacs, A.; Telbisz, M.

    1994-01-01

    The coupled steady-state reactor physics and thermohydraulic code system KARATE has been developed and applied for WWER-1000 and WWER-440 operational calculations. The 3 D coupled kinetic code KIKO3D has been developed and validated for WWER-440 accident analysis applications. The coupled kinetic code SMARTA developed by VTT Helsinki has been applied for WWER-440 accident analysis. The paper gives a summary of the experience in code development and application. (authors). 10 refs., 2 tabs., 5 figs

  6. International training program in support of safety analysis. 3D S.UN.COP-scaling uncertainty and 3D thermal-hydraulics/neutron-kinetics coupled codes seminars

    International Nuclear Information System (INIS)

    Petruzzi, Alessandro; D'Auria, Francesco; Bajs, Tomislav; Reventos, Francesc; Hassan, Yassin

    2007-01-01

    Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers and vendors, nuclear fuel companies, research organizations, consulting companies, and technical support organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the user effect' and stems from the limitations embedded in the codes as well as from the limited capability of the analysis to use the codes. Code user training and qualification is an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. The 3D S.UN.COP (Scaling, Uncertainty and 3D COuPled code calculations) seminars have been organized as follow-up of the proposal to IAEA for the Permanent Training Course for System Code Users. Six seminars have been held at University of Pisa (2003, 2004), at The Pennsylvania State University (2004), at University of Zagreb (2005), at the School of Industrial Engineering of Barcelona (January-February 2006) and in Buenos Aires, Argentina (October 2006), being this last one requested by ARN (Autoridad Regulatoria Nuclear), NA-SA (Nucleoelectrica Argentina S.A) and CNEA (Comision Nacional de Energia Atomica). It was recognized that such courses represented both a source of continuing education for current code users and a mean for current code users to enter the formal training structure of a proposed 'permanent' stepwise approach to user training. The 3D S.UN.COP 2006 in Barcelona was successfully held with the attendance of 33

  7. International Training Program in Support of Safety Analysis: 3D S.UN.COP-Scaling, Uncertainty and 3D Thermal-Hydraulics/Neutron-Kinetics Coupled Codes Seminars

    International Nuclear Information System (INIS)

    Petruzzi, Alessandro; D'Auria, Francesco; Bajs, Tomislav; Reventos, Francesc

    2006-01-01

    Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers and vendors, nuclear fuel companies, research organizations, consulting companies, and technical support organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the 'user effect' and stems from the limitations embedded in the codes as well as from the limited capability of the analysts to use the codes. Code user training and qualification is an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. The 3D S.UN.COP (Scaling, Uncertainty and 3D COuPled code calculations) seminars have been organized as follow-up of the proposal to IAEA for the Permanent Training Course for System Code Users [1]. Five seminars have been held at University of Pisa (2003, 2004), at The Pennsylvania State University (2004), at University of Zagreb (2005) and at the School of Industrial Engineering of Barcelona (2006). It was recognized that such courses represented both a source of continuing education for current code users and a mean for current code users to enter the formal training structure of a proposed 'permanent' stepwise approach to user training. The 3D S.UN.COP 2006 was successfully held with the attendance of 33 participants coming from 18 countries and 28 different institutions (universities, vendors, national laboratories and regulatory bodies). More than 30 scientists (coming from 13 countries and 23 different institutions) were

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Duerigen, Susan

    2013-05-15

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

  10. OECD/NEZ Main Steam Line Break Benchmark Problem Exercise I Simulation Using the SPACE Code with the Point Kinetics Model

    International Nuclear Information System (INIS)

    Kim, Yohan; Kim, Seyun; Ha, Sangjun

    2014-01-01

    The Safety and Performance Analysis Code for Nuclear Power Plants (SPACE) has been developed in recent years by the Korea Nuclear Hydro and Nuclear Power Co. (KHNP) through collaborative works with other Korean nuclear industries. The SPACE is a best-estimated two-phase three-field thermal-hydraulic analysis code to analyze the safety and performance of pressurized water reactors (PWRs). The SPACE code has sufficient features to replace outdated vendor supplied codes and to be used for the safety analysis of operating PWRs and the design of advanced reactors. As a result of the second phase of the development, the 2.14 version of the code was released through the successive various V and V works. The topical reports on the code and related safety analysis methodologies have been prepared for license works. In this study, the OECD/NEA Main Steam Line Break (MSLB) Benchmark Problem Exercise I was simulated as a V and V work. The results were compared with those of the participants in the benchmark project. The OECD/NEA MSLB Benchmark Problem Exercise I was simulated using the SPACE code. The results were compared with those of the participants in the benchmark project. Through the simulation, it was concluded that the SPACE code can effectively simulate PWR MSLB accidents

  11. International Training Program: 3D S. Un. Cop - Scaling, Uncertainty and 3D Thermal-Hydraulics/Neutron-Kinetics Coupled Codes Seminar

    International Nuclear Information System (INIS)

    Pertuzzi, A.; D'Auria, F.; Bajs, T.; Reventos, F.

    2006-01-01

    Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers and vendors, nuclear fuel companies, research organizations, consulting companies, and technical support organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the 'user effect' and stems from the limitations embedded in the codes as well as from the limited capability of the analysts to use the codes. Code user training and qualification is an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. The 3D S.UN.COP (Scaling, Uncertainty and 3D COuPled code calculations) seminars have been organized as follow-up of the proposal to IAEA for the Permanent Training Course for System Code Users (D'Auria, 1998). Four seminars have been held at University of Pisa (2003, 2004), at The Pennsylvania State University (2004) and at University of Zagreb (2005). It was recognized that such courses represented both a source of continuing education for current code users and a mean for current code users to enter the formal training structure of a proposed 'permanent' stepwise approach to user training. The 3D S.UN.COP 2005 was successfully held with the participation of 19 persons coming from 9 countries and 14 different institutions (universities, vendors, national laboratories and regulatory bodies). More than 15 scientists were involved in the organization of the seminar, presenting theoretical aspects of the proposed methodologies and

  12. Modelling horizontal steam generator with ATHLET. Verification of different nodalization schemes and implementation of verified constitutive equations

    Energy Technology Data Exchange (ETDEWEB)

    Beliaev, J.; Trunov, N.; Tschekin, I. [OKB Gidropress (Russian Federation); Luther, W. [GRS Garching (Germany); Spolitak, S. [RNC-KI (Russian Federation)

    1995-12-31

    Currently the ATHLET code is widely applied for modelling of several Power Plants of WWER type with horizontal steam generators. A main drawback of all these applications is the insufficient verification of the models for the steam generator. This paper presents the nodalization schemes for the secondary side of the steam generator, the results of stationary calculations, and preliminary comparisons to experimental data. The consideration of circulation in the water inventory of the secondary side is proved to be necessary. (orig.). 3 refs.

  13. Modelling horizontal steam generator with ATHLET. Verification of different nodalization schemes and implementation of verified constitutive equations

    Energy Technology Data Exchange (ETDEWEB)

    Beliaev, J; Trunov, N; Tschekin, I [OKB Gidropress (Russian Federation); Luther, W [GRS Garching (Germany); Spolitak, S [RNC-KI (Russian Federation)

    1996-12-31

    Currently the ATHLET code is widely applied for modelling of several Power Plants of WWER type with horizontal steam generators. A main drawback of all these applications is the insufficient verification of the models for the steam generator. This paper presents the nodalization schemes for the secondary side of the steam generator, the results of stationary calculations, and preliminary comparisons to experimental data. The consideration of circulation in the water inventory of the secondary side is proved to be necessary. (orig.). 3 refs.

  14. 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

  15. 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

  16. 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)...

  17. 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.

  18. 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

  19. 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.

  20. Kinetic equation solution by inverse kinetic method

    International Nuclear Information System (INIS)

    Salas, G.

    1983-01-01

    We propose a computer program (CAMU) which permits to solve the inverse kinetic equation. The CAMU code is written in HPL language for a HP 982 A microcomputer with a peripheral interface HP 9876 A ''thermal graphic printer''. The CAMU code solves the inverse kinetic equation by taking as data entry the output of the ionization chambers and integrating the equation with the help of the Simpson method. With this program we calculate the evolution of the reactivity in time for a given disturbance

  1. Contributions to the validation of advanced codes for accident analysis calculations with 3-dimensional neutron kinetics. STC with the Ukraine. Final report

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.; Rohde, U.; Khalimonchuk, V.; Kuchin, A.; Seidel, A.

    2000-10-01

    In the frame of a project of scientific-technical cooperation funded by BMBF/BMWi, the coupled code ATHLET-DYN3D has been transferred to the Scientific and Technical Centre on Nuclear and Radiation Safety Kiev (Ukraine). This program code represents an implementation of the 3D core model DYN3D developed by FZR into the GRS thermohydraulics code system ATHLET. For the purpose of validation of this coupled code, a measurement data base has been generated. In the data base suitable experimental data for operational transients from NPPs are collected. The data collection and documentation was performed in accordance with a directive about requirements to measurement data for code validation, which has been elaborated within the project. The validation calculations have been performed for two selected transients. The results of these calculations were compared with measurement values from the data base. The function of the code DYN3D was expanded with a subroutine for reactivity coefficients calculation. Using this modification of the code DYN3D, investigations of reactivity contributions on different operational processes can be performed. (orig.) [de

  2. 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

  3. 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

  4. 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.

  5. 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.

  6. 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.

  7. Development of an Analytic Nodal Diffusion Solver in Multi-groups for 3D Reactor Cores with Rectangular or Hexagonal Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Lozano, Juan Andres; Aragones, Jose Maria; Garcia-Herranz, Nuria [Universidad Politecnica de Madrid, 28006 Jose Gutierrez Abascal 2, Madrid (Spain)

    2008-07-01

    More accurate modelling of physical phenomena involved in present and future nuclear reactors requires a multi-scale and multi-physics approach. This challenge can be accomplished by the coupling of best-estimate core-physics, thermal-hydraulics and multi-physics solvers. In order to make viable that coupling, the current trends in reactor simulations are along the development of a new generation of tools based on user-friendly, modular, easily linkable, faster and more accurate codes to be integrated in common platforms. These premises are in the origin of the NURESIM Integrated Project within the 6. European Framework Program, which is envisaged to provide the initial step towards a Common European Standard Software Platform for nuclear reactors simulations. In the frame of this project and to reach the above-mentioned goals, a 3-D multigroup nodal solver for neutron diffusion calculations called ANDES (Analytic Nodal Diffusion Equation Solver) has been developed and tested in-depth in this Thesis. ANDES solves the steady-state and time-dependent neutron diffusion equation in three-dimensions and any number of energy groups, utilizing the Analytic Coarse-Mesh Finite-Difference (ACMFD) scheme to yield the nodal coupling equations. It can be applied to both Cartesian and triangular-Z geometries, so that simulations of LWR as well as VVER, HTR and fast reactors can be performed. The solver has been implemented in a fully encapsulated way, enabling it as a module to be readily integrated in other codes and platforms. In fact, it can be used either as a stand-alone nodal code or as a solver to accelerate the convergence of whole core pin-by-pin code systems. Verification of performance has shown that ANDES is a code with high order definition for whole core realistic nodal simulations. In this paper, the methodology developed and involved in ANDES is presented. (authors)

  8. Subchannel analysis of a boiloff experiment by a system thermalhydraulic code

    International Nuclear Information System (INIS)

    Bousbia-Salah, A.; D'Auria, F.

    2001-01-01

    This paper presents the results of system thermalhydraulic code using the sub-channel analysis approach in predicting the Neptun boil off experiments. This approach will be suitable for further works in view of coupling the system code with a 3D neutron kinetic one. The boil off tests were conducted in order to simulate the consequences of loss of coolant inventory leading to uncovery and heat up of fuel elements of a nuclear reactor core. In this framework, the Neptun low pressure test No5002, which is a good repeat experiment, is considered. The calculations were carried out using the system transient analysis code Relap5/Mod3.2. A detailed nodalization of the Neptun test section was developed. A reference case was run, and the overall data comparison shows good agreement between calculated and experimental thermalhydraulic parameters. A series of sensitivity analyses were also performed in order to assess the code prediction capabilities. The obtained results were almost satisfactory, this demonstrates, as well, the reasonable success of the subchannel analysis approach adopted in the present context for a system thermalhydraulic code.(author)

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

    Directory of Open Access Journals (Sweden)

    Jorge Pérez Mañes

    2014-01-01

    Full Text Available The Institute for Neutron Physics and Reactor Technology (INR at the Karlsruhe Institute of Technology (KIT is investigating the application of the meso- and microscale analysis for the prediction of local safety parameters for light water reactors (LWR. By applying codes like CFD (computational fluid dynamics and SP3 (simplified transport reactor dynamics it is possible to describe the underlying phenomena in a more accurate manner than by the nodal/coarse 1D thermal hydraulic coupled codes. By coupling the transport (SP3 based neutron kinetics (NK code DYN3D with NEPTUNE-CFD, within a parallel MPI-environment, the NHESDYN platform is created. The newly developed system will allow high fidelity simulations of LWR fuel assemblies and cores. In NHESDYN, a heat conduction solver, SYRTHES, is coupled to NEPTUNE-CFD. The driver module of NHESDYN controls the sequence of execution of the solvers as well as the communication between the solvers based on MPI. In this paper, the main features of NHESDYN are discussed and the proof of the concept is done by solving a single pin problem. The prediction capability of NHESDYN is demonstrated by a code-to-code comparison with the DYNSUB code. Finally, the future developments and validation efforts are highlighted.

  10. 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.

  11. 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

  12. 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.)

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

    International Nuclear Information System (INIS)

    Noh, Hae Man

    1995-02-01

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

  14. 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.

  15. 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.

  16. 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)

  17. 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

  18. 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.

  19. 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.).

  20. 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)

  1. 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)

  2. 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

  3. Oyster Creek cycle 10 nodal model parameter optimization study using PSMS

    International Nuclear Information System (INIS)

    Dougher, J.D.

    1987-01-01

    The power shape monitoring system (PSMS) is an on-line core monitoring system that uses a three-dimensional nodal code (NODE-B) to perform nodal power calculations and compute thermal margins. The PSMS contains a parameter optimization function that improves the ability of NODE-B to accurately monitor core power distributions. This functions iterates on the model normalization parameters (albedos and mixing factors) to obtain the best agreement between predicted and measured traversing in-core probe (TIP) reading on a statepoint-by-statepoint basis. Following several statepoint optimization runs, an average set of optimized normalization parameters can be determined and can be implemented into the current or subsequent cycle core model for on-line core monitoring. A statistical analysis of 19 high-power steady-state state-points throughout Oyster Creek cycle 10 operation has shown a consistently poor virgin model performance. The normalization parameters used in the cycle 10 NODE-B model were based on a cycle 8 study, which evaluated only Exxon fuel types. The introduction of General Electric (GE) fuel into cycle 10 (172 assemblies) was a significant fuel/core design change that could have altered the optimum set of normalization parameters. Based on the need to evaluate a potential change in the model normalization parameters for cycle 11 and in an attempt to account for the poor cycle 10 model performance, a parameter optimization study was performed

  4. INGEN: a general-purpose mesh generator for finite element codes

    International Nuclear Information System (INIS)

    Cook, W.A.

    1979-05-01

    INGEN is a general-purpose mesh generator for two- and three-dimensional finite element codes. The basic parts of the code are surface and three-dimensional region generators that use linear-blending interpolation formulas. These generators are based on an i, j, k index scheme that is used to number nodal points, construct elements, and develop displacement and traction boundary conditions. This code can generate truss elements (2 modal points); plane stress, plane strain, and axisymmetry two-dimensional continuum elements (4 to 8 nodal points); plate elements (4 to 8 nodal points); and three-dimensional continuum elements (8 to 21 nodal points). The traction loads generated are consistent with the element generated. The expansion--contraction option is of special interest. This option makes it possible to change an existing mesh such that some regions are refined and others are made coarser than the original mesh. 9 figures

  5. 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

  6. Development of a thermo-kinetic diffusion model for UO2 and (U,Pu)O2 oxide fuels using the DICTRA code

    International Nuclear Information System (INIS)

    Moore, Emily Elaine

    2013-01-01

    Uranium dioxide is the most widely used nuclear fuel for light water reactors, while some countries including France make use of the uranium-plutonium (U,Pu)O 2±x mixed oxide (MOX). The MOX is also considered for future use in the Gen IV reactors, of which the sodium cooled fast reactor (SFR) is of current research interest. Both oxides exhibit a large range of non-stoichiometry due to various oxidative states of uranium and plutonium metal. Thermo-physical properties of the fuel strongly depend on deviations in composition and temperature. Extreme temperature gradients (800 K) between the center (2300 K)and periphery of the MOX fuel pellet expose a central void due to the migration and subsequent redistribution of the fuel-elements. To gain insight into the restructuring, which occurs during the fuel lifetime as well as possible accident scenarios the thermodynamic and kinetic behavior, is crucial. A comprehensive evaluation of these properties can be incorporated in computational models to describe fuel behavior over large temperature and compositions ranges, providing a predictive tool that is applicable to other parts of the fuel cycle, such as optimizing the sintering conditions for manufacturing. Atomic transport especially in UO 2 is widely treated in the experimental and computational materials communities. The current understanding of diffusion properties is limited by the stoichiometric deviations inherent to the fuel. The difficulty is apparent in experimental settings as controlling the oxygen content is problematic. Defects (interstitial and vacancy) associated with the stoichiometric deviations of the oxides facilitate the diffusion process and is of interest in regards to the restructuring of the fuel. Experimental data is widely available; however, coherence between the evaluated diffusion coefficients is not always evident. Existing computational models based on the migration of defects are often based on atomistic level simulations. A complete

  7. The Nodal Location of Metastases in Melanoma Sentinel Lymph Nodes

    DEFF Research Database (Denmark)

    Riber-Hansen, Rikke; Nyengaard, Jens; Hamilton-Dutoit, Stephen

    2009-01-01

    BACKGROUND: The design of melanoma sentinel lymph node (SLN) histologic protocols is based on the premise that most metastases are found in the central parts of the nodes, but the evidence for this belief has never been thoroughly tested. METHODS: The nodal location of melanoma metastases in 149...

  8. Real-time control of power systems using nodal prices

    NARCIS (Netherlands)

    Jokic, A.; Lazar, M.; Bosch, van den P.P.J.

    2009-01-01

    This article presents a novel control scheme for achieving optimal power balancing and congestion management in electrical power systems via nodal prices. We develop a dynamic controller that guarantees economically optimal steady-state operation while respecting all line flow constraints in

  9. 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.

  10. Hyoid bone chondrosarcoma with cervical nodal metastasis: A case ...

    African Journals Online (AJOL)

    Background: Hyoid bone chondrosarcoma is a very rare condition. This study presents a case report of low-grade chondrosarcoma of hyoid bone with cervical nodal metastasis. The study also presents preoperative radiological investigations, pathological examination and the follow-up of the case. Case presentation: A 42 ...

  11. Note on the nodal line of the p-Laplacian

    Directory of Open Access Journals (Sweden)

    Abdel R. El Amrouss

    2006-09-01

    Full Text Available In this paper, we prove that the length of the nodal line of the eigenfunctions associated to the second eigenvalue of the problem $$ -Delta_p u = lambda ho (x |u|^{p-2}u quad hbox{in } Omega $$ with the Dirichlet conditions is not bounded uniformly with respect to the weight.

  12. A nodal method based on the response-matrix method

    International Nuclear Information System (INIS)

    Cunha Menezes Filho, A. da; Rocamora Junior, F.D.

    1983-02-01

    A nodal approach based on the Response-Matrix method is presented with the purpose of investigating the possibility of mixing two different allocations in the same problem. It is found that the use of allocation of albedo combined with allocation of direct reflection produces good results for homogeneous fast reactor configurations. (Author) [pt

  13. CRY 1AB trangenic cowpea obtained by nodal electroporation ...

    African Journals Online (AJOL)

    Electroporation-mediated genetic transformation was used to introduce Cry 1 Ab insecticidal gene into cowpea. Nodal buds were electroporated in planta with a plasmid carrying the Cry 1Ab and antibiotic resistance npt II genes driven by a 35S CaMV promoter. T1 seeds derived from electroporated branches were selected ...

  14. Validation of Neutron Calculation Codes and Models by means of benchmark cases in the frame of the Binational Commission of Nuclear Energy. Kinetic Parameters, Temperature Coefficients and Power Distribution

    International Nuclear Information System (INIS)

    Dos Santos, Adimir; Siqueira, Paulo de Tarso D.; Andrade e Silva, Graciete Simões; Grant, Carlos; Tarazaga, Ariel E.; Barberis, Claudia

    2013-01-01

    In year 2008 the Atomic Energy National Commission (CNEA) of Argentina, and the Brazilian Institute of Energetic and Nuclear Research (IPEN), under the frame of Nuclear Energy Argentine Brazilian Agreement (COBEN), among many others, included the project “Validation and Verification of Calculation Methods used for Research and Experimental Reactors . At this time, it was established that the validation was to be performed with models implemented in the deterministic codes HUEMUL and PUMA (cell and reactor codes) developed by CNEA and those ones implemented in MCNP by CNEA and IPEN. The necessary data for these validations would correspond to theoretical-experimental reference cases in the research reactor IPEN/MB-01 located in São Paulo, Brazil. The staff of the group Reactor and Nuclear Power Studies (SERC) of CNEA, from the argentine side, performed calculations with deterministic models (HUEMUL-PUMA) and probabilistic methods (MCNP) modeling a great number of physical situations of de reactor, which previously have been studied and modeled by members of the Center of Nuclear Engineering of the IPEN, whose results were extensively provided to CNEA. In this paper results of comparison of calculated and experimental results for temperature coefficients, kinetic parameters and fission rates spatial distributions are shown. (author)

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  16. 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...

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

    International Nuclear Information System (INIS)

    Minguez, E.

    1982-01-01

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

  18. Application of the HGPT methodology of reactor operation problems with a nodal mixed method

    International Nuclear Information System (INIS)

    Baudron, A.M.; Bruna, G.B.; Gandini, A.; Lautard, J.J.; Monti, S.; Pizzigati, G.

    1998-01-01

    The heuristically based generalized perturbation theory (HGPT), to first and higher order, applied to the neutron field of a reactor system, is discussed in relation to quasistatic problems. This methodology is of particular interest in reactor operation. In this application it may allow an on-line appraisal of the main physical responses of the reactor system when subject to alterations relevant to normal system exploitation, e.g. control rod movement, and/or soluble boron concentration changes to be introduced, for instance, for compensating power level variations following electrical network demands. In this paper, after describing the main features of the theory, its implementation into the diffusion, 3D mixed dual nodal code MINOS of the SAPHYR system is presented. The results from a small scale investigation performed on a simplified PWR system corroborate the validity of the methodology proposed

  19. The Nodal Polynomial Expansion method to solve the multigroup diffusion equations

    International Nuclear Information System (INIS)

    Ribeiro, R.D.M.

    1983-03-01

    The methodology of the solutions of the multigroup diffusion equations and uses the Nodal Polynomial Expansion Method is covered. The EPON code was developed based upon the above mentioned method for stationary state, rectangular geometry, one-dimensional or two-dimensional and for one or two energy groups. Then, one can study some effects such as the influence of the baffle on the thermal flux by calculating the flux and power distribution in nuclear reactors. Furthermore, a comparative study with other programs which use Finite Difference (CITATION and PDQ5) and Finite Element (CHD and FEMB) Methods was undertaken. As a result, the coherence, feasibility, speed and accuracy of the methodology used were demonstrated. (Author) [pt

  20. A two-dimensional, semi-analytic expansion method for nodal calculations

    International Nuclear Information System (INIS)

    Palmtag, S.P.

    1995-08-01

    Most modern nodal methods used today are based upon the transverse integration procedure in which the multi-dimensional flux shape is integrated over the transverse directions in order to produce a set of coupled one-dimensional flux shapes. The one-dimensional flux shapes are then solved either analytically or by representing the flux shape by a finite polynomial expansion. While these methods have been verified for most light-water reactor applications, they have been found to have difficulty predicting the large thermal flux gradients near the interfaces of highly-enriched MOX fuel assemblies. A new method is presented here in which the neutron flux is represented by a non-seperable, two-dimensional, semi-analytic flux expansion. The main features of this method are (1) the leakage terms from the node are modeled explicitly and therefore, the transverse integration procedure is not used, (2) the corner point flux values for each node are directly edited from the solution method, and a corner-point interpolation is not needed in the flux reconstruction, (3) the thermal flux expansion contains hyperbolic terms representing analytic solutions to the thermal flux diffusion equation, and (4) the thermal flux expansion contains a thermal to fast flux ratio term which reduces the number of polynomial expansion functions needed to represent the thermal flux. This new nodal method has been incorporated into the computer code COLOR2G and has been used to solve a two-dimensional, two-group colorset problem containing uranium and highly-enriched MOX fuel assemblies. The results from this calculation are compared to the results found using a code based on the traditional transverse integration procedure

  1. 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

  2. Simulation of radio nuclide migration in crystalline rock under influence of matrix diffusion and sorption kinetics: Code development and pre-assessment of migration experiment

    International Nuclear Information System (INIS)

    Woerman, A.; Xu Shulan

    1996-04-01

    The overall objective of the present study is to illuminate how spatial variability in rock chemistry in combination with spatial variability in matrix diffusion affects the radio nuclide migration along single fractures in crystalline rock. Models for ground water flow and transport of radio nuclides in a single fracture with micro-fissures have been formulated on the basis of generally accepted physical and chemical principles. Limits for the validity of the models are stated. The model equations are solved by combining finite differences and finite element methods in a computer code package. The computational package consists of three parts, namely, a stochastic field generator, a sub-program that solves the flow problem and a sub-program that solves the transport problem in a single fracture with connecting micro-fissures. Migration experiments have been pre-assessed by simulations of breakthrough curves for a constant concentration change at the upstream boundary. Breakthrough curves are sensitive to variations of parameters, such as, fracture aperture, porosity, distribution coefficient and advection velocity. The impact of matrix diffusion and sorption is manifested in terms of a retention of radionuclides causing a prolonged breakthrough. Heterogeneous sorption was characterized with a variable distribution coefficient for which the coefficient of variation CV(K d )=1 and the integral scale of an exponential covariance function is one tenth of the drill core's length. Simulated breakthrough curves for the heterogeneous sorption case have a relative variance of 3% in comparison to that of homogeneous case. An appropriate experimental set up for investigation of the effect of matrix diffusion and sorption on radio nuclide migration experiments would be an aperture less than 1 mm and porosity larger than 0.5%. 36 refs, 19 figs

  3. CINETHICA - Core accident analysis code

    International Nuclear Information System (INIS)

    Nakata, H.

    1989-10-01

    A computer program for nuclear accident analysis has been developed based on the point-kinetics approximation and one-dimensional heat transfer model for reactivity feedback calculation. Hansen's method/1/ were used for the kinetics equation solution and explicit Euler method were adopted for the thermohidraulic equations. The results were favorably compared to those from the GAPOTKIN Code/2/. (author) [pt

  4. Nuclear safety code study

    Energy Technology Data Exchange (ETDEWEB)

    Hu, H.H.; Ford, D.; Le, H.; Park, S.; Cooke, K.L.; Bleakney, T.; Spanier, J.; Wilburn, N.P.; O' Reilly, B.; Carmichael, B.

    1981-01-01

    The objective is to analyze an overpower accident in an LMFBR. A simplified model of the primary coolant loop was developed in order to understand the instabilities encountered with the MELT III and SAS codes. The computer programs were translated for switching to the IBM 4331. Numerical methods were investigated for solving the neutron kinetics equations; the Adams and Gear methods were compared. (DLC)

  5. 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)

  6. 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)

  7. 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

  8. On the nodal structure of atomic and molecular Wigner functions

    International Nuclear Information System (INIS)

    Dahl, J.P.; Schmider, H.

    1996-01-01

    In previous work on the phase-space representation of quantum mechanics, we have presented detailed pictures of the electronic one-particle reduced Wigner function for atoms and small molecules. In this communication, we focus upon the nodal structure of the function. On the basis of the simplest systems, we present an expression which relates the oscillatory decay of the Wigner function solely to the dot product of the position and momentum vector, if both arguments are large. We then demonstrate the regular behavior of nodal patterns for the larger systems. For the molecular systems, an argument analogous to the open-quotes bond-oscillatory principleclose quotes for momentum densities links the nuclear framework to an additional oscillatory term in momenta parallel to bonds. It is shown that these are visible in the Wigner function in terms of characteristic nodes

  9. Topological and trivial magnetic oscillations in nodal loop semimetals

    Science.gov (United States)

    Oroszlány, László; Dóra, Balázs; Cserti, József; Cortijo, Alberto

    2018-05-01

    Nodal loop semimetals are close descendants of Weyl semimetals and possess a topologically dressed band structure. We argue by combining the conventional theory of magnetic oscillation with topological arguments that nodal loop semimetals host coexisting topological and trivial magnetic oscillations. These originate from mapping the topological properties of the extremal Fermi surface cross sections onto the physics of two dimensional semi-Dirac systems, stemming from merging two massless Dirac cones. By tuning the chemical potential and the direction of magnetic field, a sharp transition is identified from purely trivial oscillations, arising from the Landau levels of a normal two dimensional (2D) electron gas, to a phase where oscillations of topological and trivial origin coexist, originating from 2D massless Dirac and semi-Dirac points, respectively. These could in principle be directly identified in current experiments.

  10. Ischemic stroke associated with radio frequency ablation for nodal reentry

    International Nuclear Information System (INIS)

    Diaz M, Juan C; Duran R, Carlos E; Perafan B, Pablo; Pava M, Luis F

    2010-01-01

    Atrioventricular nodal reentry tachycardia is the most common type of paroxysmal supraventricular tachycardia. In those patients in whom drug therapy is not effective or not desired, radio frequency ablation is an excellent therapeutic method. Although overall these procedures are fast and safe, several complications among which ischemic stroke stands out, have been reported. We present the case of a 41 year old female patient with repetitive episodes of tachycardia due to nodal reentry who was treated with radiofrequency ablation. Immediately after the procedure she presented focal neurologic deficit consistent with ischemic stroke in the right medial cerebral artery territory. Angiography with angioplastia and abxicimab was performed and then tissue plasminogen activator (rtPA) was locally infused, with appropriate clinical and angiographic outcome.

  11. 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

  12. Post test analysis of TEPSS tests -P2-, -P3-, -P5- and -P7- using the system code RELAP5/MOD 3.2

    International Nuclear Information System (INIS)

    Luebbesmeyer, D.

    2000-01-01

    For the PANDA-Test-Facility (TEPSS configuration) post-test calculations and analyses have been performed for experiment -P2- (Early Start), -P3- (PCC start up), -P5- (Symmetric case, Two PCCs only) and -P7- (Severe Accident). Post test calculations have been performed with the system code RELAP5/Mod 3.2 using two different nodalization of the PANDA facility namely a basis nodalization and a much reduced one. The general trend of the calculations can be summarised: RELAP5/Mod3.2 calculated the general trends of the experiments sufficiently accurate; Using the reduced nodalization the results seem to be slightly more accurate than for the basic nodalization; On the other hand, calculations based on the reduced nodalization are not significantly faster than those with basic nodalization; The mass error is in the order of 200 to 900 kg. (author)

  13. 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

  14. 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.

  15. Kinetic Typography

    DEFF Research Database (Denmark)

    van Leeuwen, Theo; Djonov, Emilia

    2014-01-01

    After discussing broad cultural drivers behind the development of kinetic typography, the chapter outlines an approach to analysing kinetic typography which is based on Halliday's theory of transitivity, as applied by Kress and Van Leeuwen to visual images.......After discussing broad cultural drivers behind the development of kinetic typography, the chapter outlines an approach to analysing kinetic typography which is based on Halliday's theory of transitivity, as applied by Kress and Van Leeuwen to visual images....

  16. Analysis of the three dimensional core kinetics NESTLE code coupling with the advanced thermo-hydraulic code systems, RELAP5/SCDAPSIM and its application to the Laguna Verde Central reactor; Analisis para el acoplamiento del codigo NESTLE para la cinetica tridimensional del nucleo al codigo avanzado de sistemas termo-hidraulicos, RELAP5/SCDAPSIM y su aplicacion al reactor de la CNLV

    Energy Technology Data Exchange (ETDEWEB)

    Salazar C, J H; Nunez C, A [CNSNS, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D.F. (Mexico); Chavez M, C [UNAM, Facultad de Ingenieria, DEPFI Campus Morelos (Mexico)

    2004-07-01

    The objective of the written present is to propose a methodology for the joining of the codes RELAP5/SCDAPSIM and NESTLE. The development of this joining will be carried out inside a doctoral program of Engineering in Energy with nuclear profile of the Ability of Engineering of the UNAM together with the National Commission of Nuclear Security and Safeguards (CNSNS). The general purpose of this type of developments, is to have tools that are implemented by multiple programs or codes such a that systems or models of the three-dimensional kinetics of the core can be simulated and those of the dynamics of the reactor (water heater-hydraulics). In the past, by limitations for the calculation of the complete answer of both systems, the developed models they were carried out for separate, putting a lot of emphasis in one but neglecting the other one. These methodologies, calls of better estimate, will be good to the nuclear industry to evaluate, with more high grades of detail, the designs of the nuclear power plant (for modifications to those already existent or for new concepts in the designs of advanced reactors), besides analysing events (transitory and have an accident), among other applications. The coupled system was applied to design studies and investigation of the Laguna Verde Nuclear power plant (CNLV). (Author)

  17. Solution of the transport equation in stationary state, in one and two dimensions, for BWR assemblies using nodal methods; Solucion de la ecuacion de transporte en estado estacionario, en 1 y 2 dimensiones, para ensambles tipo BWR usando metodos nodales

    Energy Technology Data Exchange (ETDEWEB)

    Xolocostli M, J V

    2002-07-01

    . 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

  18. 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

  19. 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

  20. Nodal methods for calculating nuclear reactor transients, control rod patterns, and fuel pin powers

    International Nuclear Information System (INIS)

    Cho, Byungoh.

    1990-01-01

    Nodal methods which are used to calculate reactor transients, control rod patterns, and fuel pin powers are investigated. The 3-D nodal code, STORM, has been modified to perform these calculations. Several numerical examples lead to the following conclusions: (1) By employing a thermal leakage-to-absorption ratio (TLAR) approximation for the spatial shape of the thermal fluxes for the 3-D Langenbuch-Maurer-Werner (LMW) and the superprompt critical transient problems, the convergence of the conventional two-group scheme is accelerated. (2) By employing the steepest-ascent hill climbing search with heuristic strategies, Optimum Control Rod Pattern Searcher (OCRPS) is developed for solving control rod positioning problem in BWRs. Using the method of approximation programming the objective function and the nuclear and thermal-hydraulic constraints are modified as heuristic functions that guide the search. The test calculations have demonstrated that, for the first cycle of the Edwin Hatch Unit number-sign 2 reactor, OCRPS shows excellent performance for finding a series of optimum control rod patterns for six burnup steps during the operating cycle. (3) For the modified two-dimensional EPRI-9R problem, the least square second-order polynomial flux expansion method was demonstrated to be computationally about 30 times faster than a fine-mesh finite difference calculation in order to achieve comparable accuracy for pin powers. The basic assumption of this method is that the reconstructed flux can be expressed as a product of an assembly form function and a second-order polynomial function

  1. 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...

  2. 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)

  3. 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

  4. A Nodal-independent and tissue-intrinsic mechanism controls heart-looping chirality

    Science.gov (United States)

    Noël, Emily S.; Verhoeven, Manon; Lagendijk, Anne Karine; Tessadori, Federico; Smith, Kelly; Choorapoikayil, Suma; den Hertog, Jeroen; Bakkers, Jeroen

    2013-11-01

    Breaking left-right symmetry in bilateria is a major event during embryo development that is required for asymmetric organ position, directional organ looping and lateralized organ function in the adult. Asymmetric expression of Nodal-related genes is hypothesized to be the driving force behind regulation of organ laterality. Here we identify a Nodal-independent mechanism that drives asymmetric heart looping in zebrafish embryos. In a unique mutant defective for the Nodal-related southpaw gene, preferential dextral looping in the heart is maintained, whereas gut and brain asymmetries are randomized. As genetic and pharmacological inhibition of Nodal signalling does not abolish heart asymmetry, a yet undiscovered mechanism controls heart chirality. This mechanism is tissue intrinsic, as explanted hearts maintain ex vivo retain chiral looping behaviour and require actin polymerization and myosin II activity. We find that Nodal signalling regulates actin gene expression, supporting a model in which Nodal signalling amplifies this tissue-intrinsic mechanism of heart looping.

  5. Elsevier Trophoblast Research Award lecture: The multifaceted role of Nodal signaling during mammalian reproduction.

    Science.gov (United States)

    Park, C B; Dufort, D

    2011-03-01

    Nodal, a secreted signaling protein in the transforming growth factor-beta (TGF-β) superfamily, has established roles in vertebrate development. However, components of the Nodal signaling pathway are also expressed at the maternal-fetal interface and have been implicated in many processes of mammalian reproduction. Emerging evidence indicates that Nodal and its extracellular inhibitor Lefty are expressed in the uterus and complex interactions between the two proteins mediate menstruation, decidualization and embryo implantation. Furthermore, several studies have shown that Nodal from both fetal and maternal sources may regulate trophoblast cell fate and facilitate placentation as both embryonic and uterine-specific Nodal knockout mouse strains exhibit disrupted placenta morphology. Here we review the established and prospective roles of Nodal signaling in facilitating successful pregnancy, including recent evidence supporting a potential link to parturition and preterm birth. Copyright © 2011 Elsevier Ltd. All rights reserved.

  6. 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

    A loss of electrical power occurred at Gentilly-2 in September of 1995 while the station was operating at full power. There was an unexpectedly rapid core power increase initiated by the drainage of the zone controllers and accelerated by coolant boiling. The core transient was terminated by Shutdown System No 1 (SDS1) tripping when the out-of-core ion chambers exceeded the 10%/sec high rate of power increase trip setpoint at 1.29 sec. This resulted in the station automatically shutting down within 2 sec of event initiation. In the first 2 sec, 26 of the 58 SDS1 and SDS2 in-core flux detectors reached there overpower trip (ROPT) setpoints. The peak reactor power reached approximately 110%FP. Reference 1 presented detailed results of the simulations performed with coupled thermalhydraulics and 3D neutron kinetics codes, SOPHT-G2 and the CERBERUS module of RFSP, and the various adjustments of these codes and plant representation that were needed to obtain the neutronic response observed in 1995. The purposes of this paper are to contrast a simulation prediction of the peak prompt core thermal power transient versus experimental estimate, and to note the impact of spatial discretization approach utilized on the prompt core thermal power transient and the channel power distribution as a function of time. In addition, adequacy of the time-step sizes employed and sensitivity to core's transient thermal-hydraulics conditions are studied. The work presented in this paper has been performed as part of a project sponsored by the Canadian Nuclear Safety Commission (CNSC). The purpose of the project was to gather information and assess the accuracy of best estimate methods using calculation methods and codes developed independently from the CANDU industry. The simulation of the accident was completed using the NESTLE core simulator, employing cross sections generated by the HELIOS lattice physics code, and incremental cross sections generated by the DRAGON lattice physics code

  7. Need for higher order polynomial basis for polynomial nodal methods employed in LWR calculations

    International Nuclear Information System (INIS)

    Taiwo, T.A.; Palmiotti, G.

    1997-01-01

    The paper evaluates the accuracy and efficiency of sixth order polynomial solutions and the use of one radial node per core assembly for pressurized water reactor (PWR) core power distributions and reactivities. The computer code VARIANT was modified to calculate sixth order polynomial solutions for a hot zero power benchmark problem in which a control assembly along a core axis is assumed to be out of the core. Results are presented for the VARIANT, DIF3D-NODAL, and DIF3D-finite difference codes. The VARIANT results indicate that second order expansion of the within-node source and linear representation of the node surface currents are adequate for this problem. The results also demonstrate the improvement in the VARIANT solution when the order of the polynomial expansion of the within-node flux is increased from fourth to sixth order. There is a substantial saving in computational time for using one radial node per assembly with the sixth order expansion compared to using four or more nodes per assembly and fourth order polynomial solutions. 11 refs., 1 tab

  8. A finite element code for electric motor design

    Science.gov (United States)

    Campbell, C. Warren

    1994-01-01

    FEMOT is a finite element program for solving the nonlinear magnetostatic problem. This version uses nonlinear, Newton first order elements. The code can be used for electric motor design and analysis. FEMOT can be embedded within an optimization code that will vary nodal coordinates to optimize the motor design. The output from FEMOT can be used to determine motor back EMF, torque, cogging, and magnet saturation. It will run on a PC and will be available to anyone who wants to use it.

  9. TITAN: an advanced three-dimensional coupled neutronic/thermal-hydraulics code for light water nuclear reactor core analysis

    International Nuclear Information System (INIS)

    Griggs, D.P.; Kazimi, M.S.; Henry, A.F.

    1984-06-01

    The three-dimensional nodal neutronics code QUANDRY and the three-dimensional two-fluid thermal-hydraulics code THERMIT are combined into TITAN. Steady-state and transient coupling methodologies based upon a tandem structure were devised and implemented. Additional models for nuclear feedback, equilibrium xenon and direct moderator heating were added. TITAN was tested using a boiling water two channel problem and the coupling methodologies were shown to be effective. Simulated turbine trip transients and several control rod withdrawal transients were analyzed with good results. Sensitivity studies indicated that the time-step size can affect transient results significantly. TITAN was also applied to a quarter core PWR problem based on a real reactor geometry. The steady-state results were compared to a solution produced by MEKIN-B and poor agreement between the horizontal power shapes was found. Calculations with various mesh spacings showed that the mesh spacings in the MEKIN-B analysis were too large to produce accurate results with a finite difference method. The TITAN results were shown to be reasonable. A pair of control rod ejection accidents were also analyzed with TITAN. A comparison of the TITAN PWR control rod ejection results with results from coupled point kinetics/thermal-hydraulics analyses showed that the point kinetics method used (adiabatic method for control rod reactivities, steady-state flux shape for core-averaged reactivity feedback) underpredicted the power excursion in one case and overpredicted it in the other. It was therefore concluded that point kinetics methods should be used with caution and that three-dimensional codes like TITAN are superior for analyzing PWR control rod ejection transients

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

    International Nuclear Information System (INIS)

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

    1994-06-01

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

  11. 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

  12. Mechanism of polyuria and natriuresis in atrioventricular nodal tachycardia.

    Science.gov (United States)

    Canepa-Anson, R; Williams, M; Marshall, J; Mitsuoka, T; Lightman, S; Sutton, R

    1984-01-01

    A woman with tachycardia associated with polyuria was investigated. Electrophysiological analysis showed that the tachycardia was an atrioventricular nodal re-entrant tachycardia. Programmed stimulation was then used to provoke and sustain the tachycardia for 40 minutes. Polyuria, with an appreciable increase in free water clearance, was observed. This was associated with reduction in plasma and urinary arginine vasopressin concentrations. Appreciable natriuresis also developed. These results support the hypothesis that the polyuria with increased free water clearance and the natriuresis occurring during sustained tachycardia in man are due to inhibition of secretion of vasopressin and the release of natriuretic factor. PMID:6434116

  13. 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.

  14. 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.

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

    Directory of Open Access Journals (Sweden)

    Dougan Scott T

    2007-03-01

    Full Text Available Abstract Background The vertebrate body plan is generated during gastrulation with the formation of the three germ layers. Members of the Nodal-related subclass of the TGF-β superfamily induce and pattern the mesoderm and endoderm in all vertebrates. In zebrafish, two nodal-related genes, called squint and cyclops, are required in a dosage-dependent manner for the formation of all derivatives of the mesoderm and endoderm. These genes are expressed dynamically during the blastula stages and may have different roles at different times. This question has been difficult to address because conditions that alter the timing of nodal-related gene expression also change Nodal levels. We utilized a pharmacological approach to conditionally inactivate the ALK 4, 5 and 7 receptors during the blastula stages without disturbing earlier signaling activity. This permitted us to directly examine when Nodal signals specify cell types independently of dosage effects. Results We show that two drugs, SB-431542 and SB-505124, completely block the response to Nodal signals when added to embryos after the mid-blastula transition. By blocking Nodal receptor activity at later stages, we demonstrate that Nodal signaling is required from the mid-to-late blastula period to specify sequentially, the somites, notochord, blood, Kupffer's vesicle, hatching gland, heart, and endoderm. Blocking Nodal signaling at late times prevents specification of cell types derived from the embryo margin, but not those from more animal regions. This suggests a linkage between cell fate and length of exposure to Nodal signals. Confirming this, cells exposed to a uniform Nodal dose adopt progressively more marginal fates with increasing lengths of exposure. Finally, cell fate specification is delayed in squint mutants and accelerated when Nodal levels are elevated. Conclusion We conclude that (1 Nodal signals are most active during the mid-to-late blastula stages, when nodal-related gene

  16. Development Status of Diffusion Code RAST-K 2.0 at UNIST

    Energy Technology Data Exchange (ETDEWEB)

    Park, Minyong; Zheng, Youqi; Choe, Jiwon; Zhang, Peng; Lee, Deokjung [UNIST, Ulsan (Korea, Republic of); Lee, Eunki; Shin, Hocheol [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The non-linear scheme was used based on the 2-group CMFD and a three dimensional multi -group unified nodal method (UNM). To consider the history effects, the main heavy isotopes were tracked by micro-depletion module using CRAM. The simplified 1-D single channel thermal hydraulic solver from nTACER is implemented. The θ method was adopted in the transient calculation. To get detailed pin-wise power and burnup distribution, Pin power reconstruction module was implemented. Also automatic control logic to calculate MTC, FTC, control rod worth was implemented. To perform multicycle analysis, restart and shuffling/rotation module has been implemented. To link between CASMO-4E and RAST-K 2.0, CATORA (CASMO TO RAST-K 2.0) code was developed. Unlike the other diffusion codes, RAST-K 2.0 depletion module uses CRAM and extended depletion chain for fission products. Most lattice codes give cumulative fission yield of Pm-149 without considering Pm-148 and Pm-149 capture reaction which will lead to the increase of Sm-149 number density. This paper reports the status of RAST-K 2.0 code development at UNIST. The new code applies a new kernel based on the two-node UNM with CMFD, and θ method for kinetic calculation. Also, the microdepletion calculation is used to consider the history effects. And other modules and functions also implemented such as pin power reconstruction, branch calculation, restart, multi-cycle, and 1-D single channel T/H solver.

  17. Reactor kinetics methods development. Final report

    International Nuclear Information System (INIS)

    Hansen, K.F.; Henry, A.F.

    1978-01-01

    This report is a qualitative summary of research conducted at MIT from 1967 to 1977 in the area of reactor kinetics methods. The objectives of the research were to find methods of integration of various mathematical models of nuclear reactor transients. From the beginning the work was aimed at numerical integration methods. Specific areas of research, discussed in more detail following, included: integration of multigroup diffusion theory models by finite difference and finite element methods; response matrix and nodal methods; coarse-mesh homogenization; and special treatment of boundary conditions

  18. Solution of the transport equation in stationary state, in one and two dimensions, for BWR assemblies using nodal methods; Solucion de la ecuacion de transporte en estado estacionario, en 1 y 2 dimensiones, para ensambles tipo BWR usando metodos nodales

    Energy Technology Data Exchange (ETDEWEB)

    Xolocostli M, J.V

    2002-07-01

    . 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

  19. Exact boundary controllability of nodal profile for quasilinear hyperbolic systems

    CERN Document Server

    Li, Tatsien; Gu, Qilong

    2016-01-01

    This book provides a comprehensive overview of the exact boundary controllability of nodal profile, a new kind of exact boundary controllability stimulated by some practical applications. This kind of controllability is useful in practice as it does not require any precisely given final state to be attained at a suitable time t=T by means of boundary controls, instead it requires the state to exactly fit any given demand (profile) on one or more nodes after a suitable time t=T by means of boundary controls. In this book we present a general discussion of this kind of controllability for general 1-D first order quasilinear hyperbolic systems and for general 1-D quasilinear wave equations on an interval as well as on a tree-like network using a modular-structure construtive method, suggested in LI Tatsien's monograph "Controllability and Observability for Quasilinear Hyperbolic Systems"(2010), and we establish a complete theory on the local exact boundary controllability of nodal profile for 1-D quasilinear hyp...

  20. Nodal Diffusion Burnable Poison Treatment for Prismatic Reactor Cores

    International Nuclear Information System (INIS)

    Ougouag, A.M.; Ferrer, R.M.

    2010-01-01

    The prismatic block version of the High Temperature Reactor (HTR) considered as a candidate Very High Temperature Reactor (VHTR)design may use burnable poison pins in locations at some corners of the fuel blocks (i.e., assembly equivalent structures). The presence of any highly absorbing materials, such as these burnable poisons, within fuel blocks for hexagonal geometry, graphite-moderated High Temperature Reactors (HTRs) causes a local inter-block flux depression that most nodal diffusion-based method have failed to properly model or otherwise represent. The location of these burnable poisons near vertices results in an asymmetry in the morphology of the assemblies (or blocks). Hence the resulting inadequacy of traditional homogenization methods, as these 'spread' the actually local effect of the burnable poisons throughout the assembly. Furthermore, the actual effect of the burnable poison is primarily local with influence in its immediate vicinity, which happens to include a small region within the same assembly as well as similar regions in the adjacent assemblies. Traditional homogenization methods miss this artifact entirely. This paper presents a novel method for treating the local effect of the burnable poison explicitly in the context of a modern nodal method.

  1. Hereditary bone dysplasia with pathological fractures and nodal osteoarthropathy

    International Nuclear Information System (INIS)

    Arendse, Regan; Brink, Paul; Beighton, Peter

    2009-01-01

    A father and daughter both had multiple pathological fractures and nodal osteoarthropathy. The father, aged 50 years, had at least 20 healed fractures of the axial and appendicular skeleton, sustained by minor trauma over his 50-year lifespan, many of which had been surgically fixed prior to his first presentation to us. Fractures of the clavicles, thoracic cage and long bones of the arms and legs, had healed with malalignment and deformity. Healed fractures were complicated by ankylosis of the cervical vertebrae and both elbows. He also had osteoarthritis of the hands, with exuberant osteophytosis, and profound perceptive deafness. His general health was good, his intellect and facies were normal, and his sclerae were white. The daughter, aged 27 years, had sustained at least seven fractures of the axial and appendicular skeleton following trivial injuries, in distribution similar to those of the father. She had also experienced painful swelling of the fingers, which preceded progressive development of nodal osteoarthropathy. Her hearing was normal. In both individuals, biochemical and immunological investigations yielded normal results. It was not possible for molecular studies to be undertaken. Pedigree data were consistent with autosomal dominant transmission, and this disorder appeared to be a previously undocumented heritable skeletal dysplasia. (orig.)

  2. Hereditary bone dysplasia with pathological fractures and nodal osteoarthropathy

    Energy Technology Data Exchange (ETDEWEB)

    Arendse, Regan [University of Stellenbosch, Department of Medicine, Tygerberg Hospital, Stellenbosch (South Africa); University of Cape Town, Division of Rheumatology, Groote Schuur Hospital, Cape Town (South Africa); Brink, Paul [University of Stellenbosch, Department of Medicine, Tygerberg Hospital, Stellenbosch (South Africa); Beighton, Peter [University of Cape Town, Division of Human Genetics, Faculty of Health Sciences, Cape Town (South Africa)

    2009-12-15

    A father and daughter both had multiple pathological fractures and nodal osteoarthropathy. The father, aged 50 years, had at least 20 healed fractures of the axial and appendicular skeleton, sustained by minor trauma over his 50-year lifespan, many of which had been surgically fixed prior to his first presentation to us. Fractures of the clavicles, thoracic cage and long bones of the arms and legs, had healed with malalignment and deformity. Healed fractures were complicated by ankylosis of the cervical vertebrae and both elbows. He also had osteoarthritis of the hands, with exuberant osteophytosis, and profound perceptive deafness. His general health was good, his intellect and facies were normal, and his sclerae were white. The daughter, aged 27 years, had sustained at least seven fractures of the axial and appendicular skeleton following trivial injuries, in distribution similar to those of the father. She had also experienced painful swelling of the fingers, which preceded progressive development of nodal osteoarthropathy. Her hearing was normal. In both individuals, biochemical and immunological investigations yielded normal results. It was not possible for molecular studies to be undertaken. Pedigree data were consistent with autosomal dominant transmission, and this disorder appeared to be a previously undocumented heritable skeletal dysplasia. (orig.)

  3. Dirac Magnon Nodal Loops in Quasi-2D Quantum Magnets.

    Science.gov (United States)

    Owerre, S A

    2017-07-31

    In this report, we propose a new concept of one-dimensional (1D) closed lines of Dirac magnon nodes in two-dimensional (2D) momentum space of quasi-2D quantum magnetic systems. They are termed "2D Dirac magnon nodal-line loops". We utilize the bilayer honeycomb ferromagnets with intralayer coupling J and interlayer coupling J L , which is realizable in the honeycomb chromium compounds CrX 3 (X ≡ Br, Cl, and I). However, our results can also exist in other layered quasi-2D quantum magnetic systems. Here, we show that the magnon bands of the bilayer honeycomb ferromagnets overlap for J L  ≠ 0 and form 1D closed lines of Dirac magnon nodes in 2D momentum space. The 2D Dirac magnon nodal-line loops are topologically protected by inversion and time-reversal symmetry. Furthermore, we show that they are robust against weak Dzyaloshinskii-Moriya interaction Δ DM  magnon edge modes.

  4. Physical kinetics

    International Nuclear Information System (INIS)

    Lifschitz, E.M.; Pitajewski, L.P.

    1983-01-01

    The textbook covers the subject under the following headings: kinetic gas theory, diffusion approximation, collisionless plasma, collisions within the plasma, plasma in the magnetic field, theory of instabilities, dielectrics, quantum fluids, metals, diagram technique for nonequilibrium systems, superconductors, and kinetics of phase transformations

  5. A NEM diffusion code for fuel management and time average core calculation

    International Nuclear Information System (INIS)

    Mishra, Surendra; Ray, Sherly; Kumar, A.N.

    2005-01-01

    A computer code based on Nodal expansion method has been developed for solving two groups three dimensional diffusion equation. This code can be used for fuel management and time average core calculation. Explicit Xenon and fuel temperature estimation are also incorporated in this code. TAPP-4 phase-B physics experimental results were analyzed using this code and a code based on FD method. This paper gives the comparison of the observed data and the results obtained with this code and FD code. (author)

  6. Heparin kinetics

    International Nuclear Information System (INIS)

    Swart, C.A.M. de.

    1983-01-01

    The author has studied the kinetics of heparin and heparin fractions after intravenous administration in humans and in this thesis the results of this study are reported. Basic knowledge about the physico-chemical properties of heparin and its interactions with proteins resulting in anticoagulant and lipolytic effects are discussed in a review (chapter II), which also comprises some clinical aspects of heparin therapy. In chapter III the kinetics of the anticoagulant effect are described after intravenous administration of five commercial heparin preparations. A mathematical model is presented that fits best to these kinetics. The kinetics of the anticoagulant and lipolytic effects after intravenous injection of various 35 S-radiolabelled heparin fractions and their relationship with the disappearance of the radiolabel are described in chapter IV. Chapter V gives a description of the kinetics of two radiolabels after injection of in vitro formed complexes consisting of purified, 125 I-radiolabelled antithrombin III and various 35 S-radiolabelled heparin fractions. (Auth.)

  7. International Code Assessment and Applications Program: Summary of code assessment studies concerning RELAP5/MOD2, RELAP5/MOD3, and TRAC-B

    International Nuclear Information System (INIS)

    Schultz, R.R.

    1993-12-01

    Members of the International Code Assessment Program (ICAP) have assessed the US Nuclear Regulatory Commission (USNRC) advanced thermal-hydraulic codes over the past few years in a concerted effort to identify deficiencies, to define user guidelines, and to determine the state of each code. The results of sixty-two code assessment reviews, conducted at INEL, are summarized. Code deficiencies are discussed and user recommended nodalizations investigated during the course of conducting the assessment studies and reviews are listed. All the work that is summarized was done using the RELAP5/MOD2, RELAP5/MOD3, and TRAC-B codes

  8. 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

  9. 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.

  10. Consistent Code Qualification Process and Application to WWER-1000 NPP

    International Nuclear Information System (INIS)

    Berthon, A.; Petruzzi, A.; Giannotti, W.; D'Auria, F.; Reventos, F.

    2006-01-01

    Calculation analysis by application of the system codes are performed to evaluate the NPP or the facility behavior during a postulated transient or to evaluate the code capability. The calculation analysis constitutes a process that involves the code itself, the data of the reference plant, the data about the transient, the nodalization, and the user. All these elements affect one each other and affect the results. A major issue in the use of mathematical model is constituted by the model capability to reproduce the plant or facility behavior under steady state and transient conditions. These aspects constitute two main checks that must be satisfied during the qualification process. The first of them is related to the realization of a scheme of the reference plant; the second one is related to the capability to reproduce the transient behavior. The aim of this paper is to describe the UMAE (Uncertainty Method based on Accuracy Extrapolation) methodology developed at University of Pisa for qualifying a nodalization and analysing the calculated results and to perform the uncertainty evaluation of the system code by the CIAU code (Code with the capability of Internal Assessment of Uncertainty). The activity consists with the re-analysis of the Experiment BL-44 (SBLOCA) performed in the LOBI facility and the analysis of a Kv-scaling calculation of the WWER-1000 NPP nodalization taking as reference the test BL-44. Relap5/Mod3.3 has been used as thermal-hydraulic system code and the standard procedure adopted at University of Pisa has been applied to show the capability of the code to predict the significant aspects of the transient and to obtain a qualified nodalization of the WWER-1000 through a systematic qualitative and quantitative accuracy evaluation. The qualitative accuracy evaluation is based on the selection of Relevant Thermal-hydraulic Aspects (RTAs) and is a prerequisite to the application of the Fast Fourier Transform Based Method (FFTBM) which quantifies

  11. 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.

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  13. 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.

  14. 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)

  15. 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.

  16. Coding Partitions

    Directory of Open Access Journals (Sweden)

    Fabio Burderi

    2007-05-01

    Full Text Available Motivated by the study of decipherability conditions for codes weaker than Unique Decipherability (UD, we introduce the notion of coding partition. Such a notion generalizes that of UD code and, for codes that are not UD, allows to recover the ``unique decipherability" at the level of the classes of the partition. By tacking into account the natural order between the partitions, we define the characteristic partition of a code X as the finest coding partition of X. This leads to introduce the canonical decomposition of a code in at most one unambiguouscomponent and other (if any totally ambiguouscomponents. In the case the code is finite, we give an algorithm for computing its canonical partition. This, in particular, allows to decide whether a given partition of a finite code X is a coding partition. This last problem is then approached in the case the code is a rational set. We prove its decidability under the hypothesis that the partition contains a finite number of classes and each class is a rational set. Moreover we conjecture that the canonical partition satisfies such a hypothesis. Finally we consider also some relationships between coding partitions and varieties of codes.

  17. An Adaptive Approach to Variational Nodal Diffusion Problems

    International Nuclear Information System (INIS)

    Zhang Hui; Lewis, E.E.

    2001-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  19. Fourier-Accelerated Nodal Solvers (FANS) for homogenization problems

    Science.gov (United States)

    Leuschner, Matthias; Fritzen, Felix

    2017-11-01

    Fourier-based homogenization schemes are useful to analyze heterogeneous microstructures represented by 2D or 3D image data. These iterative schemes involve discrete periodic convolutions with global ansatz functions (mostly fundamental solutions). The convolutions are efficiently computed using the fast Fourier transform. FANS operates on nodal variables on regular grids and converges to finite element solutions. Compared to established Fourier-based methods, the number of convolutions is reduced by FANS. Additionally, fast iterations are possible by assembling the stiffness matrix. Due to the related memory requirement, the method is best suited for medium-sized problems. A comparative study involving established Fourier-based homogenization schemes is conducted for a thermal benchmark problem with a closed-form solution. Detailed technical and algorithmic descriptions are given for all methods considered in the comparison. Furthermore, many numerical examples focusing on convergence properties for both thermal and mechanical problems, including also plasticity, are presented.

  20. Nodal domains on isospectral quantum graphs: the resolution of isospectrality?

    International Nuclear Information System (INIS)

    Band, Ram; Shapira, Talia; Smilansky, Uzy

    2006-01-01

    We present and discuss isospectral quantum graphs which are not isometric. These graphs are the analogues of the isospectral domains in R 2 which were introduced recently in Gordon et al (1992 Bull. Am. Math. Soc. 27 134-8), Chapman (1995 Am. Math. Mon. 102 124), Buser et al (1994 Int. Math. Res. Not. 9 391-400), Okada and Shudo (2001 J. Phys. A: Math. Gen. 34 5911-22), Jakobson et al (2006 J. Comput. Appl. Math. 194 141-55) and Levitin et al (2006 J. Phys. A: Math. Gen. 39 2073-82)) all based on Sunada's construction of isospectral domains (Sunada T 1985 Ann. Math. 121 196-86). After presenting some of the properties of these graphs, we discuss a few examples which support the conjecture that by counting the nodal domains of the corresponding eigenfunctions one can resolve the isospectral ambiguity

  1. INTERMITTENT ANTIARYTHMIC THERAPY OF ARIOVENTICULAR NODAL REENTRY TACHYCARDIA IN CHILDREN

    Directory of Open Access Journals (Sweden)

    Boris Djindjic

    2008-04-01

    Full Text Available Until recent advances in pharmacology and clinical cardiology regarding farmacodynamics of antiarrhythmic drugs and their efficiency in patients with refractory paroxysmal supraventricular tachycardia, chronic prophylactic therapy was the only treatment option for patients refusing catheter ablation. Another treatment option, also known by eponym “pill in pocket” have been shown to be equally useful and efficacious.The aim of our study was prospective examination of children with refractory atrioventricular nodal reentry tachycardia (AVNRT who were withdrawn from chronic antiarrhythmic prophylactic therapy and started with intermittent oral beta blocker treatment (propranolol at dosage 1 mg/kg - max 80 mg.Twelve children (8 boys and 4 girls with AVNRT were included in the study. Four children did not have arrhythmia during first six months after withdrawal and 7 were successfully treated without complication.Intermittent antiarrhythmic therapy in children with AVNRT could be very efficacious and useful treatment option which significantly improves their quality of life.

  2. 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.

  3. Ultrasound beam characteristics of a symmetric nodal origami based array

    Science.gov (United States)

    Bilgunde, Prathamesh N.; Bond, Leonard J.

    2018-04-01

    Origami-the ancient art of paper folding-is being explored in acoustics for effective focusing of sound. In this short communication, we present a numerical investigation of beam characteristics for an origami based ultrasound array. A spatial re-configuration of array elements is performed based upon the symmetric nodal origami. The effect of fold angle on the ultrasound beam is evaluated using frequency domain and transient finite element analysis. It was found that increase in the fold angle reduces near field length by 58% and also doubles the beam intensity as compared to the linear array. Transient analysis also indicated 80% reduction in the -6dB beam width, which can improve the lateral resolution of phased array. Such a spatially re-configurable array could potentially be used in the future to reduce the cost of electronics in the phased array instrumentation.

  4. Space-angle approximations in the variational nodal method

    International Nuclear Information System (INIS)

    Lewis, E. E.; Palmiotti, G.; Taiwo, T.

    1999-01-01

    The variational nodal method is formulated such that the angular and spatial approximations maybe examined separately. Spherical harmonic, simplified spherical harmonic, and discrete ordinate approximations are coupled to the primal hybrid finite element treatment of the spatial variables. Within this framework, two classes of spatial trial functions are presented: (1) orthogonal polynomials for the treatment of homogeneous nodes and (2) bilinear finite subelement trial functions for the treatment of fuel assembly sized nodes in which fuel-pin cell cross sections are represented explicitly. Polynomial and subelement trial functions are applied to benchmark water-reactor problems containing MOX fuel using spherical harmonic and simplified spherical harmonic approximations. The resulting accuracy and computing costs are compared

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-12-31

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

  6. Contemporary Management of Recurrent Nodal Disease in Differentiated Thyroid Carcinoma

    Science.gov (United States)

    Na’ara, Shorook; Amit, Moran; Fridman, Eran; Gil, Ziv

    2016-01-01

    Differentiated thyroid carcinoma (DTC) comprises over 90% of thyroid tumors and includes papillary and follicular carcinomas. Patients with DTC have an excellent prognosis, with a 10-year survival rate of over 90%. However, the risk of recurrent tumor ranges between 5% and 30% within 10 years of the initial diagnosis. Cervical lymph node disease accounts for the majority of recurrences and in most cases is detected during follow-up by ultrasound or elevated levels of serum thyroglobulin. Recurrent disease is accompanied by increased morbidity. The mainstay of treatment of nodal recurrence is surgical management. We provide an overview of the literature addressing surgical management of recurrent or persistent lymph node disease in patients with DTC. PMID:26886954

  7. Error Estimation and Accuracy Improvements in Nodal Transport Methods

    International Nuclear Information System (INIS)

    Zamonsky, O.M.

    2000-01-01

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

  8. Kinetic Interface

    DEFF Research Database (Denmark)

    2009-01-01

    A kinetic interface for orientation detection in a video training system is disclosed. The interface includes a balance platform instrumented with inertial motion sensors. The interface engages a participant's sense of balance in training exercises.......A kinetic interface for orientation detection in a video training system is disclosed. The interface includes a balance platform instrumented with inertial motion sensors. The interface engages a participant's sense of balance in training exercises....

  9. MARS Code in Linux Environment

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Moon Kyu; Bae, Sung Won; Jung, Jae Joon; Chung, Bub Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    The two-phase system analysis code MARS has been incorporated into Linux system. The MARS code was originally developed based on the RELAP5/MOD3.2 and COBRA-TF. The 1-D module which evolved from RELAP5 alone could be applied for the whole NSSS system analysis. The 3-D module developed based on the COBRA-TF, however, could be applied for the analysis of the reactor core region where 3-D phenomena would be better treated. The MARS code also has several other code units that could be incorporated for more detailed analysis. The separate code units include containment analysis modules and 3-D kinetics module. These code modules could be optionally invoked to be coupled with the main MARS code. The containment code modules (CONTAIN and CONTEMPT), for example, could be utilized for the analysis of the plant containment phenomena in a coupled manner with the nuclear reactor system. The mass and energy interaction during the hypothetical coolant leakage accident could, thereby, be analyzed in a more realistic manner. In a similar way, 3-D kinetics could be incorporated for simulating the three dimensional reactor kinetic behavior, instead of using the built-in point kinetics model. The MARS code system, developed initially for the MS Windows environment, however, would not be adequate enough for the PC cluster system where multiple CPUs are available. When parallelism is to be eventually incorporated into the MARS code, MS Windows environment is not considered as an optimum platform. Linux environment, on the other hand, is generally being adopted as a preferred platform for the multiple codes executions as well as for the parallel application. In this study, MARS code has been modified for the adaptation of Linux platform. For the initial code modification, the Windows system specific features have been removed from the code. Since the coupling code module CONTAIN is originally in a form of dynamic load library (DLL) in the Windows system, a similar adaptation method

  10. MARS Code in Linux Environment

    International Nuclear Information System (INIS)

    Hwang, Moon Kyu; Bae, Sung Won; Jung, Jae Joon; Chung, Bub Dong

    2005-01-01

    The two-phase system analysis code MARS has been incorporated into Linux system. The MARS code was originally developed based on the RELAP5/MOD3.2 and COBRA-TF. The 1-D module which evolved from RELAP5 alone could be applied for the whole NSSS system analysis. The 3-D module developed based on the COBRA-TF, however, could be applied for the analysis of the reactor core region where 3-D phenomena would be better treated. The MARS code also has several other code units that could be incorporated for more detailed analysis. The separate code units include containment analysis modules and 3-D kinetics module. These code modules could be optionally invoked to be coupled with the main MARS code. The containment code modules (CONTAIN and CONTEMPT), for example, could be utilized for the analysis of the plant containment phenomena in a coupled manner with the nuclear reactor system. The mass and energy interaction during the hypothetical coolant leakage accident could, thereby, be analyzed in a more realistic manner. In a similar way, 3-D kinetics could be incorporated for simulating the three dimensional reactor kinetic behavior, instead of using the built-in point kinetics model. The MARS code system, developed initially for the MS Windows environment, however, would not be adequate enough for the PC cluster system where multiple CPUs are available. When parallelism is to be eventually incorporated into the MARS code, MS Windows environment is not considered as an optimum platform. Linux environment, on the other hand, is generally being adopted as a preferred platform for the multiple codes executions as well as for the parallel application. In this study, MARS code has been modified for the adaptation of Linux platform. For the initial code modification, the Windows system specific features have been removed from the code. Since the coupling code module CONTAIN is originally in a form of dynamic load library (DLL) in the Windows system, a similar adaptation method

  11. A nodal discontinuous Galerkin approach to 3-D viscoelastic wave propagation in complex geological media

    Science.gov (United States)

    Lambrecht, L.; Lamert, A.; Friederich, W.; Möller, T.; Boxberg, M. S.

    2018-03-01

    A nodal discontinuous Galerkin (NDG) approach is developed and implemented for the computation of viscoelastic wavefields in complex geological media. The NDG approach combines unstructured tetrahedral meshes with an element-wise, high-order spatial interpolation of the wavefield based on Lagrange polynomials. Numerical fluxes are computed from an exact solution of the heterogeneous Riemann problem. Our implementation offers capabilities for modelling viscoelastic wave propagation in 1-D, 2-D and 3-D settings of very different spatial scale with little logistical overhead. It allows the import of external tetrahedral meshes provided by independent meshing software and can be run in a parallel computing environment. Computation of adjoint wavefields and an interface for the computation of waveform sensitivity kernels are offered. The method is validated in 2-D and 3-D by comparison to analytical solutions and results from a spectral element method. The capabilities of the NDG method are demonstrated through a 3-D example case taken from tunnel seismics which considers high-frequency elastic wave propagation around a curved underground tunnel cutting through inclined and faulted sedimentary strata. The NDG method was coded into the open-source software package NEXD and is available from GitHub.

  12. 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.

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

    International Nuclear Information System (INIS)

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

    2015-09-01

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

  14. Aztheca Code; Codigo Aztheca

    Energy Technology Data Exchange (ETDEWEB)

    Quezada G, S.; Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico); Centeno P, J.; Sanchez M, H., E-mail: sequga@gmail.com [UNAM, Facultad de Ingenieria, Ciudad Universitaria, Circuito Exterior s/n, 04510 Ciudad de Mexico (Mexico)

    2017-09-15

    This paper presents the Aztheca code, which is formed by the mathematical models of neutron kinetics, power generation, heat transfer, core thermo-hydraulics, recirculation systems, dynamic pressure and level models and control system. The Aztheca code is validated with plant data, as well as with predictions from the manufacturer when the reactor operates in a stationary state. On the other hand, to demonstrate that the model is applicable during a transient, an event occurred in a nuclear power plant with a BWR reactor is selected. The plant data are compared with the results obtained with RELAP-5 and the Aztheca model. The results show that both RELAP-5 and the Aztheca code have the ability to adequately predict the behavior of the reactor. (Author)

  15. Geochemical computer codes. A review

    International Nuclear Information System (INIS)

    Andersson, K.

    1987-01-01

    In this report a review of available codes is performed and some code intercomparisons are also discussed. The number of codes treating natural waters (groundwater, lake water, sea water) is large. Most geochemical computer codes treat equilibrium conditions, although some codes with kinetic capability are available. A geochemical equilibrium model consists of a computer code, solving a set of equations by some numerical method and a data base, consisting of thermodynamic data required for the calculations. There are some codes which treat coupled geochemical and transport modeling. Some of these codes solve the equilibrium and transport equations simultaneously while other solve the equations separately from each other. The coupled codes require a large computer capacity and have thus as yet limited use. Three code intercomparisons have been found in literature. It may be concluded that there are many codes available for geochemical calculations but most of them require a user that us quite familiar with the code. The user also has to know the geochemical system in order to judge the reliability of the results. A high quality data base is necessary to obtain a reliable result. The best results may be expected for the major species of natural waters. For more complicated problems, including trace elements, precipitation/dissolution, adsorption, etc., the results seem to be less reliable. (With 44 refs.) (author)

  16. 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

  17. Computer code abstract: NESTLE

    International Nuclear Information System (INIS)

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

    1995-01-01

    NESTLE is a few-group neutron diffusion equation solver utilizing the nodal expansion method (NEM) for eigenvalue, adjoint, and fixed-source steady-state and transient problems. The NESTLE code solve the eigenvalue (criticality), eigenvalue adjoint, external fixed-source steady-state, and external fixed-source or eigenvalue initiated transient problems. The eigenvalue problem allows criticality searches to be completed, and the external fixed-source steady-state problem can search to achieve a specified power level. Transient problems model delayed neutrons via precursor groups. Several core properties can be input as time dependent. Two- or four-energy groups can be utilized, with all energy groups being thermal groups (i.e., upscatter exits) is desired. Core geometries modeled include Cartesian and hexagonal. Three-, two-, and one-dimensional models can be utilized with various symmetries. The thermal conditions predicted by the thermal-hydraulic model of the core are used to correct cross sections for temperature and density effects. Cross sections for temperature and density effects. Cross sections are parameterized by color, control rod state (i.e., in or out), and burnup, allowing fuel depletion to be modeled. Either a macroscopic or microscopic model may be employed

  18. 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.

  19. 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)

  20. 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.

  1. 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.

  2. Kinetics and

    Directory of Open Access Journals (Sweden)

    Mojtaba Ahmadi

    2016-11-01

    Full Text Available The aqueous degradation of Reactive Yellow 84 (RY84 by potassium peroxydisulfate (K2S2O8 has been studied in laboratory scale experiments. The effect of the initial concentrations of potassium peroxydisulfate and RY84, pH and temperature on RY84 degradation were also examined. Experimental data were analyzed using first and second-order kinetics. The degradation kinetics of RY84 of the potassium peroxydisulfate process followed the second-order reaction kinetics. These rate constants have an extreme values similar to of 9.493 mM−1min−1 at a peroxydisulfate dose of 4 mmol/L. Thermodynamic parameters such as activation (Ea and Gibbs free energy (ΔG° were also evaluated. The negative value of ΔGo and Ea shows the spontaneous reaction natural conditions and exothermic nature.

  3. 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...

  4. 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...

  5. 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.

  6. [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.

  7. 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)

  8. Regional nodal relapse in surgically staged Merkel cell carcinoma

    Energy Technology Data Exchange (ETDEWEB)

    Hoeller, Ulrike; Mueller, Thomas; Schubert, Tina; Budach, Volker; Ghadjar, Pirus [Charite Universitaetsmedizin Berlin, Department of Radiation Oncology, Berlin (Germany); Brenner, Winfried [Charite Universitaetsmedizin Berlin, Department of Nuclear Medicine, Berlin (Germany); Kiecker, Felix [Charite Universitaetsmedizin Berlin, Department of Dermatology, Berlin (Germany); Schicke, Bernd [Tumor Center Berlin, Berlin (Germany); Haase, Oliver [Charite Universitaetsmedizin Berlin, Department of Surgery, Berlin (Germany)

    2014-10-08

    The nodal relapse pattern of surgically staged Merkel cell carcinoma (MCC) with/without elective nodal radiotherapy (RT) was studied in a single institution. A total of 51 patients with MCC, 33 % UICC stage I, 14 % II, 53 % III (4 lymph node metastases of unknown primary) were eligible. All patients had surgical staging: 23 patients sentinel node biopsy (SNB), 22 patients SNB followed by lymphadenectomy (LAD) and 6 patients LAD. In all, 94 % of the primary tumors (PT) were completely resected; 57 % of patients received RT, 51 % of known PT sites, 33 % (8/24 patients) regional RT to snN0 nodes and 68 % (17/27 patients) to pN+ nodes, mean reference dose 51.5 and 50 Gy, respectively. Mean follow-up was 6 years (range 2-14 years). A total of 22 % (11/51) patients developed regional relapses (RR); the 5-year RR rate was 27 %. In snN0 sites (stage I/II), relapse occurred in 5 of 14 nonirradiated vs. none of 8 irradiated sites (p = 0.054), resulting in a 5-year RR rate of 33 % versus 0 % (p = 0.16). The crude RR rate was lower in stage I (12 %, 2/17 patients) than for stage II (43 %, 3/7 patients). In stage III (pN+), RR appeared to be less frequent in irradiated sites (18 %, 3/14 patients) compared with nonirradiated sites (33 %, 3/10 patients, p = 0.45) with 5-year RR rates of 23 % vs. 34 %, respectively. Our data suggest that adjuvant nodal RT plays a major role even if the sentinel nodes were negative. Adjuvant RT of the lymph nodes in patients with stage IIa tumors and RT after LAD in stage III tumors is proposed and should be evaluated prospectively. (orig.) [German] Untersucht wurde das regionaere Rezidivmuster des Merkelzell-Karzinoms (MCC) nach chirurgischem Staging und stadienadaptierter Therapie. Eingeschlossen wurden 51 Patienten mit lokalisiertem MCC: 33 % hatten UICC-Stadium-I-, 14 % -II-, 53 % -III-Tumoren (davon 4 Lymphknotenmetastasen eines unbekannten Primaertumors). Alle Patienten erhielten ein chirurgisches Staging: 23 Waechterlymphknotenbiopsien (SNB

  9. Granulocyte kinetics

    International Nuclear Information System (INIS)

    Peters, A.M.; Lavender, J.P.; Saverymuttu, S.H.

    1985-01-01

    By using density gradient materials enriched with autologous plasma, the authors have been able to isolate granulocutes from other cellular elements and label them with In-111 without separation from a plasma environment. The kinetic behavior of these cells suggests that phenomena attributed to granulocyte activation are greatly reduced by this labeling. Here, they review their study of granulocyte kinetics in health and disease in hope of quantifying sites of margination and identifying principal sites of destruction. The three principle headings of the paper are distribution, life-span, and destruction

  10. Acquired experience on organizing 3D S.UN.COP: international course to support nuclear license by user training in the areas of scaling, uncertainty, and 3D thermal-hydraulics/neutron-kinetics coupled codes

    Energy Technology Data Exchange (ETDEWEB)

    Petruzzi, Alessandro; D' Auria, Francesco [University of Pisa, San Piero a Grado (Italy). Nuclear Research Group San Piero a Grado (GRNSPG); Galetti, Regina, E-mail: regina@cnen.gov.b [National Commission for Nuclear Energy (CNEN), Rio de Janeiro, RJ (Brazil); Bajs, Tomislav [University of Zagreb (Croatia). Fac. of Electrical Engineering and Computing. Dept. of Power Systems; Reventos, Francesc [Technical University of Catalonia, Barcelona (Spain). Dept. of Physics and Nuclear Engineering

    2011-07-01

    Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers, vendors, and research organizations. Computer code user represents a source of uncertainty that may significantly affect the results of system code calculations. Code user training and qualification represent an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes the experience in applying a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In addition, this paper presents the organization and the main features of the 3D S.UN.COP (scaling, uncertainty, and 3D coupled code calculations) seminars during which particular emphasis is given to practical applications in connection with the licensing process of best estimate plus uncertainty methodologies, showing the designer, utility and regulatory approaches. (author)

  11. Acquired experience on organizing 3D S.UN.COP: international course to support nuclear license by user training in the areas of scaling, uncertainty, and 3D thermal-hydraulics/neutron-kinetics coupled codes

    International Nuclear Information System (INIS)

    Petruzzi, Alessandro; D'Auria, Francesco; Galetti, Regina; Bajs, Tomislav; Reventos, Francesc

    2011-01-01

    Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers, vendors, and research organizations. Computer code user represents a source of uncertainty that may significantly affect the results of system code calculations. Code user training and qualification represent an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes the experience in applying a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In addition, this paper presents the organization and the main features of the 3D S.UN.COP (scaling, uncertainty, and 3D coupled code calculations) seminars during which particular emphasis is given to practical applications in connection with the licensing process of best estimate plus uncertainty methodologies, showing the designer, utility and regulatory approaches. (author)

  12. Analysis of PWR control rod ejection accident with the coupled code system SKETCH-INS/TRACE by incorporating pin power reconstruction model

    International Nuclear Information System (INIS)

    Nakajima, T.; Sakai, T.

    2010-01-01

    The pin power reconstruction model was incorporated in the 3-D nodal kinetics code SKETCH-INS in order to produce accurate calculation of three-dimensional pin power distributions throughout the reactor core. In order to verify the employed pin power reconstruction model, the PWR MOX/UO_2 core transient benchmark problem was analyzed with the coupled code system SKETCH-INS/TRACE by incorporating the model and the influence of pin power reconstruction model was studied. SKETCH-INS pin power distributions for 3 benchmark problems were compared with the PARCS solutions which were provided by the host organisation of the benchmark. SKETCH-INS results were in good agreement with the PARCS results. The capability of employed pin power reconstruction model was confirmed through the analysis of benchmark problems. A PWR control rod ejection benchmark problem was analyzed with the coupled code system SKETCH-INS/ TRACE by incorporating the pin power reconstruction model. The influence of pin power reconstruction model was studied by comparing with the result of conventional node averaged flux model. The results indicate that the pin power reconstruction model has significant effect on the pin powers during transient and hence on the fuel enthalpy

  13. 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)

  14. Analyses of PWR boron dilution consequences with the Arrotta code

    International Nuclear Information System (INIS)

    Johanson, E.; Cheng, H.W.; Sehgal, B.R.

    1998-03-01

    During the past few years, major attention has been paid to analyzing the issue of reactivity initiated accidents (RIAs), of which the boron dilution event is of very special interest to the countries having pressurized water reactors (PWRs) in their nuclear power delivery systems. The scenario considered is that if an inadvertent accumulation of boron free water in one loop during reactor startup operations of a PWR and the inadvertent startup of the reactor coolant pump (RCP) in the loop. This could then lead to a rapid boron dilution in the core, which can in turn give rise to a power excursion. This report is devoted to studying the potential physical and thermal hydraulic consequences of a slug of diluted coolant entering the core after one RCP start under a couple of postulated cases. The severity of the consequences of such a scenario is primarily determined by the amount of positive reactivity insertion, and they are also related to the reactivity insertion rate. Therefore, in the report, detailed calculations and analyses have been carried out from case to case by using the well-known space-time kinetics code, ARROTTA. As a result, the spatial distribution for nodal power, fuel enthalpy, fuel temperature and clad outside temperature as well as the change in core reactivity, total core power and peak fuel temperature can be provided. In general, the maximum fuel enthalpy, peak fuel temperature, and clad outside temperature, for all the cases considered in the report, do not exceed their respective routine safety limitations because of the strong Doppler effect and moderator temperature feedback, except if the safety limitations on fuel enthalpy addition for high burnup fuel are drastically reduced

  15. 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)

  16. Molecular pathogenesis of splenic and nodal marginal zone lymphoma.

    Science.gov (United States)

    Spina, Valeria; Rossi, Davide

    Genomic studies have improved our understanding of the biological basis of splenic (SMZL) and nodal (NMZL) marginal zone lymphoma by providing a comprehensive and unbiased view of the genes/pathways that are deregulated in these diseases. Consistent with the physiological involvement of NOTCH, NF-κB, B-cell receptor and toll-like receptor signaling in mature B-cells differentiation into the marginal zone B-cells, many oncogenic mutations of genes involved in these pathways have been identified in SMZL and NMZL. Beside genetic lesions, also epigenetic and post-transcriptional modifications contribute to the deregulation of marginal zone B-cell differentiation pathways in SMZL and NMZL. This review describes the progress in understanding the molecular mechanism underlying SMZL and NMZL, including molecular and post-transcriptional modifications, and discusses how information gained from these efforts has provided new insights on potential targets of diagnostic, prognostic and therapeutic relevance in SMZL and NMZL. Copyright © 2016 Elsevier Ltd. All rights reserved.

  17. Micropropagation of Calophyllum brasiliense (Cambess.) from nodal segments.

    Science.gov (United States)

    Silveira, S S; Cordeiro-Silva, R; Degenhardt-Goldbach, J; Quoirin, M

    2016-05-03

    Micropropagation of Calophyllum brasiliense Cambess. (Clusiaceae) is a way to overcome difficulties in achieving large-scale plant production, given the recalcitrant nature of the seeds, irregular fructification and absence of natural vegetative propagation of the species. Cultures were established using nodal segments 2 cm in length, obtained from 1-2 year old seedlings, maintained in a greenhouse. Mercury chloride and Plant Preservative Mixture™ were used in the surface sterilizing stage, better results being achieved with Plant Preservative Mixture™ incorporation in culture medium, at any concentration. Polyvinylpyrrolidone, activated charcoal, cysteine, ascorbic acid or citric acid were added to the culture medium to avoid oxidation. After 30 days of culture, polyvinylpirrolidone and ascorbic acid gave better results, eliminating oxidation in most explants. For shoot multiplication, benzylaminopurine was used in concentrations of 4.4 and 8.8 µM in Woody Plant Medium, resulting in an average of 4.43 and 4.68 shoots per explant, respectively, after 90 days. Indole-3-butyric acid and α-naphthalene acetic acid were used to induce root formation, reaching a maximum rooting rate of 24% with 20µM α-naphthalene acetic acid. For acclimatization. the rooted plants were transferred to Plantmax® substrate and cultured in a greenhouse, reaching 79% of survival after 30 days and 60% after one year.

  18. Relevance of regional nodal management in multimodality esophageal cancer treatment

    International Nuclear Information System (INIS)

    Wong, J.; Perez-Tamayo, C.; Takasugi, B.; Orringer, M.B.; Flint, A.; Lichter, A.S.

    1986-01-01

    A prospective study has been undertaken at the University of Michigan Hospital, where patients with distal esophageal carcinoma receive concurrent radiation therapy (3,750 cGy delivered in 15 fractions) and systemic chemotherapy (cisplatin, Velban, 5-FU), followed by blunt esophagectomy with exploration and lymph node sampling. Strict pathologic screening and handling of nodal tissue and esophagectomy specimens were analyzed. Eighteen patients with distal esophageal lesions ranging from 5 to 12 cm (average, 7 cm) detected on the initial barium swallow study have been seen to date. In three of these patients celiac axis involvement has been demonstrated on CT. All primary lesions were confirmed by biopsy. Five were found to be squamous cell carcinoma and thirteen were adenocarcinomas. One of 15 of the presently evaluable patients (5%) had microscopic involvement of a celiac node at surgery. Celiac, lesser curvature, and superior gastric nodes where all encompassed in the radiation therapy portals to the aforementioned dose. CT scan planning was done in all patients. This added volume was well tolerated by the patients without morbidity

  19. Topological Nodal Cooper Pairing in Doped Weyl Metals

    Science.gov (United States)

    Li, Yi; Haldane, F. D. M.

    2018-02-01

    We generalize the concept of Berry connection of the single-electron band structure to that of a two-particle Cooper pairing state between two Fermi surfaces with opposite Chern numbers. Because of underlying Fermi surface topology, the pairing Berry phase acquires nontrivial monopole structure. Consequently, pairing gap functions have topologically protected nodal structure as vortices in the momentum space with the total vorticity solely determined by the pair monopole charge qp. The nodes of gap function behave as the Weyl-Majorana points of the Bogoliubov-de Gennes pairing Hamiltonian. Their relation with the connection patterns of the surface modes from the Weyl band structure and the Majorana surface modes inside the pairing gap is also discussed. Under the approximation of spherical Fermi surfaces, the pairing symmetry are represented by monopole harmonic functions. The lowest possible pairing channel carries angular momentum number j =|qp|, and the corresponding gap functions are holomorphic or antiholomorphic functions on Fermi surfaces. After projected on the Fermi surfaces with nontrivial topology, all the partial-wave channels of pairing interactions acquire the monopole charge qp independent of concrete pairing mechanism.

  20. 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.

  1. 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)

  2. 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.

  3. 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...

  4. 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

  5. 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

  6. 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)

  7. 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

  8. Speaking Code

    DEFF Research Database (Denmark)

    Cox, Geoff

    Speaking Code begins by invoking the “Hello World” convention used by programmers when learning a new language, helping to establish the interplay of text and code that runs through the book. Interweaving the voice of critical writing from the humanities with the tradition of computing and software...

  9. 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.

  10. 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)

  11. 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.

  12. 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

  13. 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.

  14. 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.

  15. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  17. Solution of the transport equation in stationary state and X Y geometry, using continuous and discontinuous hybrid nodal schemes; Solucion de la ecuacion de transporte en estado estacionario y geometria X Y, usando esquemas nodales hibridos continuos y discontinuos

    Energy Technology Data Exchange (ETDEWEB)

    Xolocostli M, V.; Valle G, E. del [IPN-ESFM, 07738 Mexico D.F. (Mexico); Alonso V, G. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: xvicente@hotmail.com

    2003-07-01

    In this work it is described the development and the application of the NH-FEM schemes, Hybrid Nodal schemes using the Finite Element method in the solution of the neutron transport equation in stationary state and X Y geometry, of which two families of schemes were developed, one of which corresponds to the continuous and the other to the discontinuous ones, inside those first its are had the Bi-Quadratic Bi Q, and to the Bi-cubic BiC, while for the seconds the Discontinuous Bi-lineal DBiL and the Discontinuous Bi-quadratic DBiQ. These schemes were implemented in a program to which was denominated TNHXY, Transport of neutrons with Hybrid Nodal schemes in X Y geometry. One of the immediate applications of the schemes NH-FEM it will be in the analysis of assemblies of nuclear fuel, particularly of the BWR type. The validation of the TNHXY program was made with two test problems or benchmark, already solved by other authors with numerical techniques and to compare results. The first of them consists in an it BWR fuel assemble in an arrangement 7x7 without rod and with control rod providing numerical results. The second is a fuel assemble of mixed oxides (MOX) in an arrangement 10x10. This last problem it is known as the Benchmark problem WPPR of the NEA Data Bank and the results are compared with those of other commercial codes as HELIOS, MCNP-4B and CPM-3. (Author)

  18. Physisorption kinetics

    CERN Document Server

    Kreuzer, Hans Jürgen

    1986-01-01

    This monograph deals with the kinetics of adsorption and desorption of molecules physisorbed on solid surfaces. Although frequent and detailed reference is made to experiment, it is mainly concerned with the theory of the subject. In this, we have attempted to present a unified picture based on the master equation approach. Physisorption kinetics is by no means a closed and mature subject; rather, in writing this monograph we intended to survey a field very much in flux, to assess its achievements so far, and to give a reasonable basis from which further developments can take off. For this reason we have included many papers in the bibliography that are not referred to in the text but are of relevance to physisorption. To keep this monograph to a reasonable size, and also to allow for some unity in the presentation of the material, we had to omit a number of topics related to physisorption kinetics. We have not covered to any extent the equilibrium properties of physisorbed layers such as structures, phase tr...

  19. 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

  20. 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

  1. Radial distribution of power starting from the reactivity using nodal schemes of second and third order

    International Nuclear Information System (INIS)

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

    2003-01-01

    In this work two nodal schemes of finite element are presented, one of second and the other of third order of accurate that allow to determine the radial distribution of power starting from the corresponding reactivities.The schemes here developed were obtained taking as starting point the equation developed by Driscoll et al, the one which is based on the diffusion approach of 1-1/2 energy groups. This equation relates the power fraction of an assemble with their reactivity and with the power fractions and reactivities of the assemblies that its surround it. Driscoll and collaborators they solve in form approximate such equation supposing that the reactivity of each assemble it is but a lineal function of the burnt one of the fuel. The spatial approach carries out it with the classic technique of finite differences centered in mesh. Nevertheless that the algebraic system to which its arrive it can be solved without more considerations introduce some additional suppositions and adjustment parameters that it allows them to predict results comparable to those contributed by three dimensions analysis and this way to reduce the one obtained error when its compare their results with those of a production code like CASMO. Also in the two schemes that here are presented the same approaches of Driscoll were used being obtained errors of the one 10% and of 5% for the second schemes and third order respectively for a test case that it was built starting from data of the Cycle 1 of the Unit 1 of the Laguna Verde Nucleo electric plant. These errors its were obtained when comparing with a computer program based on the matrix response method. It is sought to have this way a quick and efficient tool for the multicycle analysis in the fuel management. However, this model presents problems in the appropriate prediction of the average burnt of the nucleus and of the burnt one by lot. (Author)

  2. 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)

  3. Solution of the transport equation in stationary state and X Y geometry, using continuous and discontinuous hybrid nodal schemes

    International Nuclear Information System (INIS)

    Xolocostli M, V.; Valle G, E. del; Alonso V, G.

    2003-01-01

    In this work it is described the development and the application of the NH-FEM schemes, Hybrid Nodal schemes using the Finite Element method in the solution of the neutron transport equation in stationary state and X Y geometry, of which two families of schemes were developed, one of which corresponds to the continuous and the other to the discontinuous ones, inside those first its are had the Bi-Quadratic Bi Q, and to the Bi-cubic BiC, while for the seconds the Discontinuous Bi-lineal DBiL and the Discontinuous Bi-quadratic DBiQ. These schemes were implemented in a program to which was denominated TNHXY, Transport of neutrons with Hybrid Nodal schemes in X Y geometry. One of the immediate applications of the schemes NH-FEM it will be in the analysis of assemblies of nuclear fuel, particularly of the BWR type. The validation of the TNHXY program was made with two test problems or benchmark, already solved by other authors with numerical techniques and to compare results. The first of them consists in an it BWR fuel assemble in an arrangement 7x7 without rod and with control rod providing numerical results. The second is a fuel assemble of mixed oxides (MOX) in an arrangement 10x10. This last problem it is known as the Benchmark problem WPPR of the NEA Data Bank and the results are compared with those of other commercial codes as HELIOS, MCNP-4B and CPM-3. (Author)

  4. Coding Labour

    Directory of Open Access Journals (Sweden)

    Anthony McCosker

    2014-03-01

    Full Text Available As well as introducing the Coding Labour section, the authors explore the diffusion of code across the material contexts of everyday life, through the objects and tools of mediation, the systems and practices of cultural production and organisational management, and in the material conditions of labour. Taking code beyond computation and software, their specific focus is on the increasingly familiar connections between code and labour with a focus on the codification and modulation of affect through technologies and practices of management within the contemporary work organisation. In the grey literature of spreadsheets, minutes, workload models, email and the like they identify a violence of forms through which workplace affect, in its constant flux of crisis and ‘prodromal’ modes, is regulated and governed.

  5. Improved quasi-static nodal green's function method

    International Nuclear Information System (INIS)

    Li Junli; Jing Xingqing; Hu Dapu

    1997-01-01

    Improved Quasi-Static Green's Function Method (IQS/NGFM) is presented, as an new kinetic method. To solve the three-dimensional transient problem, improved Quasi-Static Method is adopted to deal with the temporal problem, which will increase the time step as long as possible so as to decrease the number of times of space calculation. The time step of IQS/NGFM can be increased to 5∼10 times longer than that of Full Implicit Differential Method. In spatial calculation, the NGFM is used to get the distribution of shape function, and it's spatial mesh can be nearly 20 times larger than that of Definite Differential Method. So the IQS/NGFM is considered as an efficient kinetic method

  6. A sliding point contact model for the finite element structures code EURDYN

    International Nuclear Information System (INIS)

    Smith, B.L.

    1986-01-01

    A method is developed by which sliding point contact between two moving deformable structures may be incorporated within a lumped mass finite element formulation based on displacements. The method relies on a simple mechanical interpretation of the contact constraint in terms of equivalent nodal forces and avoids the use of nodal connectivity via a master slave arrangement or pseudo contact element. The methodology has been iplemented into the EURDYN finite element program for the (2D axisymmetric) version coupled to the hydro code SEURBNUK. Sample calculations are presented illustrating the use of the model in various contact situations. Effects due to separation and impact of structures are also included. (author)

  7. Group-decoupled multi-group pin power reconstruction utilizing nodal solution 1D flux profiles

    International Nuclear Information System (INIS)

    Yu, Lulin; Lu, Dong; Zhang, Shaohong; Wang, Dezhong

    2014-01-01

    Highlights: • A direct fitting multi-group pin power reconstruction method is developed. • The 1D nodal solution flux profiles are used as the condition. • The least square fit problem is analytically solved. • A slowing down source improvement method is applied. • The method shows good accuracy for even challenging problems. - Abstract: A group-decoupled direct fitting method is developed for multi-group pin power reconstruction, which avoids both the complication of obtaining 2D analytic multi-group flux solution and any group-coupled iteration. A unique feature of the method is that in addition to nodal volume and surface average fluxes and corner fluxes, transversely-integrated 1D nodal solution flux profiles are also used as the condition to determine the 2D intra-nodal flux distribution. For each energy group, a two-dimensional expansion with a nine-term polynomial and eight hyperbolic functions is used to perform a constrained least square fit to the 1D intra-nodal flux solution profiles. The constraints are on the conservation of nodal volume and surface average fluxes and corner fluxes. Instead of solving the constrained least square fit problem numerically, we solve it analytically by fully utilizing the symmetry property of the expansion functions. Each of the 17 unknown expansion coefficients is expressed in terms of nodal volume and surface average fluxes, corner fluxes and transversely-integrated flux values. To determine the unknown corner fluxes, a set of linear algebraic equations involving corner fluxes is established via using the current conservation condition on all corners. Moreover, an optional slowing down source improvement method is also developed to further enhance the accuracy of the reconstructed flux distribution if needed. Two test examples are shown with very good results. One is a four-group BWR mini-core problem with all control blades inserted and the other is the seven-group OECD NEA MOX benchmark, C5G7

  8. 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.

  9. 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.

  10. 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.

  11. 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

  12. 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.

  13. 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.

  14. 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.

  15. Stochastic kinetics

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  16. Code package to analyse behavior of the WWER fuel rods in normal operation: TOPRA's code

    International Nuclear Information System (INIS)

    Scheglov, A.; Proselkov, V.

    2001-01-01

    This paper briefly describes the code package intended for analysis of WWER fuel rod characteristics. The package includes two computer codes: TOPRA-1 and TOPRA-2 for full-scale fuel rod analyses; MRZ and MKK codes for analyzing the separate sections of fuel rods in r-z and r-j geometry. The TOPRA's codes are developed on the base of PIN-mod2 version and verified against experimental results obtained in MR, MIR and Halden research reactors (in the framework of SOFIT, FGR-2 and FUMEX experimental programs). Comparative analysis of calculation results and results from post-reactor examination of the WWER-440 and WWER-1000 fuel rod are also made as additional verification of these codes. To avoid the enlarging of uncertainties in fuel behavior prediction as a result of simplifying of the fuel geometry, MKK and MRZ codes are developed on the basis of the finite element method with use of the three nodal finite elements. Results obtained in the course of the code verification indicate the possibility for application of the method and TOPRA's code for simplified engineering calculations of WWER fuel rods thermal-physical parameters. An analysis of maximum relative errors for predicting of the fuel rod characteristics in the range of the accepted parameter values is also presented in the paper

  17. Tolrestat kinetics

    International Nuclear Information System (INIS)

    Hicks, D.R.; Kraml, M.; Cayen, M.N.; Dubuc, J.; Ryder, S.; Dvornik, D.

    1984-01-01

    The kinetics of tolrestat, a potent inhibitor of aldose reductase, were examined. Serum concentrations of tolrestat and of total 14 C were measured after dosing normal subjects and subjects with diabetes with 14 C-labeled tolrestat. In normal subjects, tolrestat was rapidly absorbed and disappearance from serum was biphasic. Distribution and elimination t 1/2s were approximately 2 and 10 to 12 hr, respectively, after single and multiple doses. Unchanged tolrestat accounted for the major portion of 14 C in serum. Radioactivity was rapidly and completely excreted in urine and feces in an approximate ratio of 2:1. Findings were much the same in subjects with diabetes. In normal subjects, the kinetics of oral tolrestat were independent of dose in the 10 to 800 mg range. Repetitive dosing did not result in unexpected cumulation. Tolrestat was more than 99% bound to serum protein; it did not compete with warfarin for binding sites but was displaced to some extent by high concentrations of tolbutamide or salicylate

  18. Speech coding

    Energy Technology Data Exchange (ETDEWEB)

    Ravishankar, C., Hughes Network Systems, Germantown, MD

    1998-05-08

    Speech is the predominant means of communication between human beings and since the invention of the telephone by Alexander Graham Bell in 1876, speech services have remained to be the core service in almost all telecommunication systems. Original analog methods of telephony had the disadvantage of speech signal getting corrupted by noise, cross-talk and distortion Long haul transmissions which use repeaters to compensate for the loss in signal strength on transmission links also increase the associated noise and distortion. On the other hand digital transmission is relatively immune to noise, cross-talk and distortion primarily because of the capability to faithfully regenerate digital signal at each repeater purely based on a binary decision. Hence end-to-end performance of the digital link essentially becomes independent of the length and operating frequency bands of the link Hence from a transmission point of view digital transmission has been the preferred approach due to its higher immunity to noise. The need to carry digital speech became extremely important from a service provision point of view as well. Modem requirements have introduced the need for robust, flexible and secure services that can carry a multitude of signal types (such as voice, data and video) without a fundamental change in infrastructure. Such a requirement could not have been easily met without the advent of digital transmission systems, thereby requiring speech to be coded digitally. The term Speech Coding is often referred to techniques that represent or code speech signals either directly as a waveform or as a set of parameters by analyzing the speech signal. In either case, the codes are transmitted to the distant end where speech is reconstructed or synthesized using the received set of codes. A more generic term that is applicable to these techniques that is often interchangeably used with speech coding is the term voice coding. This term is more generic in the sense that the

  19. Optimal codes as Tanner codes with cyclic component codes

    DEFF Research Database (Denmark)

    Høholdt, Tom; Pinero, Fernando; Zeng, Peng

    2014-01-01

    In this article we study a class of graph codes with cyclic code component codes as affine variety codes. Within this class of Tanner codes we find some optimal binary codes. We use a particular subgraph of the point-line incidence plane of A(2,q) as the Tanner graph, and we are able to describe ...

  20. 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

  1. 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

  2. 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.

  3. 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.

  4. 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.

  5. 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

  6. Present status on numerical algorithms and benchmark tests for point kinetics and quasi-static approximate kinetics

    International Nuclear Information System (INIS)

    Ise, Takeharu

    1976-12-01

    Review studies have been made on algorithms of numerical analysis and benchmark tests on point kinetics and quasistatic approximate kinetics computer codes to perform efficiently benchmark tests on space-dependent neutron kinetics codes. Point kinetics methods have now been improved since they can be directly applied to the factorization procedures. Methods based on Pade rational function give numerically stable solutions and methods on matrix-splitting are interested in the fact that they are applicable to the direct integration methods. An improved quasistatic (IQ) approximation is the best and the most practical method; it is numerically shown that the IQ method has a high stability and precision and the computation time which is about one tenth of that of the direct method. IQ method is applicable to thermal reactors as well as fast reactors and especially fitted for fast reactors to which many time steps are necessary. Two-dimensional diffusion kinetics codes are most practicable though there exist also three-dimensional diffusion kinetics code as well as two-dimensional transport kinetics code. On developing a space-dependent kinetics code, in any case, it is desirable to improve the method so as to have a high computing speed for solving static diffusion and transport equations. (auth.)

  7. Vocable Code

    DEFF Research Database (Denmark)

    Soon, Winnie; Cox, Geoff

    2018-01-01

    a computational and poetic composition for two screens: on one of these, texts and voices are repeated and disrupted by mathematical chaos, together exploring the performativity of code and language; on the other, is a mix of a computer programming syntax and human language. In this sense queer code can...... be understood as both an object and subject of study that intervenes in the world’s ‘becoming' and how material bodies are produced via human and nonhuman practices. Through mixing the natural and computer language, this article presents a script in six parts from a performative lecture for two persons...

  8. NSURE code

    International Nuclear Information System (INIS)

    Rattan, D.S.

    1993-11-01

    NSURE stands for Near-Surface Repository code. NSURE is a performance assessment code. developed for the safety assessment of near-surface disposal facilities for low-level radioactive waste (LLRW). Part one of this report documents the NSURE model, governing equations and formulation of the mathematical models, and their implementation under the SYVAC3 executive. The NSURE model simulates the release of nuclides from an engineered vault, their subsequent transport via the groundwater and surface water pathways tot he biosphere, and predicts the resulting dose rate to a critical individual. Part two of this report consists of a User's manual, describing simulation procedures, input data preparation, output and example test cases

  9. OSCAR-4 Code System Application to the SAFARI-1 Reactor

    International Nuclear Information System (INIS)

    Stander, Gerhardt; Prinsloo, Rian H.; Tomasevic, Djordje I.; Mueller, Erwin

    2008-01-01

    The OSCAR reactor calculation code system consists of a two-dimensional lattice code, the three-dimensional nodal core simulator code MGRAC and related service codes. The major difference between the new version of the OSCAR system, OSCAR-4, and its predecessor, OSCAR-3, is the new version of MGRAC which contains many new features and model enhancements. In this work some of the major improvements in the nodal diffusion solution method, history tracking, nuclide transmutation and cross section models are described. As part of the validation process of the OSCAR-4 code system (specifically the new MGRAC version), some of the new models are tested by comparing computational results to SAFARI-1 reactor plant data for a number of operational cycles and for varying applications. A specific application of the new features allows correct modeling of, amongst others, the movement of fuel-follower type control rods and dynamic in-core irradiation schedules. It is found that the effect of the improved control rod model, applied over multiple cycles of the SAFARI-1 reactor operation history, has a significant effect on in-cycle reactivity prediction and fuel depletion. (authors)

  10. Fuel management and core design code systems for pressurized water reactor neutronic calculations

    International Nuclear Information System (INIS)

    Ahnert, C.; Arayones, J.M.

    1985-01-01

    A package of connected code systems for the neutronic calculations relevant in fuel management and core design has been developed and applied for validation to the startup tests and first operating cycle of a 900MW (electric) PWR. The package includes the MARIA code system for the modeling of the different types of PWR fuel assemblies, the CARMEN code system for detailed few group diffusion calculations for PWR cores at operating and burnup conditions, and the LOLA code system for core simulation using onegroup nodal theory parameters explicitly calculated from the detailed solutions

  11. Two codes used in analysis of rod ejection accident for Qinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Zhu Xinguan

    1987-12-01

    Two codes were developed to analyse rod ejection accident for Qinshan Nuclear Power Plant. One was based on point model with temperature reactivity feedback. In this code, the worth of ejected rod was obtained under'adiabatic' approximation. In the other code, the Nodal Green's Function Method was used to solve space-time dependent neutron diffusion equation. Using these codes, the transient core-power have been calculated for two rod ejection cases at beginning of core-life in Qinshan Nuclear Power Plant

  12. The Aster code; Code Aster

    Energy Technology Data Exchange (ETDEWEB)

    Delbecq, J.M

    1999-07-01

    The Aster code is a 2D or 3D finite-element calculation code for structures developed by the R and D direction of Electricite de France (EdF). This dossier presents a complete overview of the characteristics and uses of the Aster code: introduction of version 4; the context of Aster (organisation of the code development, versions, systems and interfaces, development tools, quality assurance, independent validation); static mechanics (linear thermo-elasticity, Euler buckling, cables, Zarka-Casier method); non-linear mechanics (materials behaviour, big deformations, specific loads, unloading and loss of load proportionality indicators, global algorithm, contact and friction); rupture mechanics (G energy restitution level, restitution level in thermo-elasto-plasticity, 3D local energy restitution level, KI and KII stress intensity factors, calculation of limit loads for structures), specific treatments (fatigue, rupture, wear, error estimation); meshes and models (mesh generation, modeling, loads and boundary conditions, links between different modeling processes, resolution of linear systems, display of results etc..); vibration mechanics (modal and harmonic analysis, dynamics with shocks, direct transient dynamics, seismic analysis and aleatory dynamics, non-linear dynamics, dynamical sub-structuring); fluid-structure interactions (internal acoustics, mass, rigidity and damping); linear and non-linear thermal analysis; steels and metal industry (structure transformations); coupled problems (internal chaining, internal thermo-hydro-mechanical coupling, chaining with other codes); products and services. (J.S.)

  13. Cryopreservation of in vitro grown nodal segments of Rauvolfia serpentina by PVS2 vitrification.

    Science.gov (United States)

    Ray, Avik; Bhattacharya, Sabita

    2008-01-01

    This paper describes the cryopreservation by PVS2 vitrification of Rauvolfia serpentina (L.) Benth ex kurz, an important tropical medicinal plant. The effects of type and size of explants, sucrose preculture (duration and concentration) and vitrification treatment were tested. Preliminary experiments with PVS1, 2 and 3 produced shoot growth only for PVS2. When optimizing the PVS2 vitrification of nodal segments, those of 0.31 - 0.39 cm in size were better than other nodal sizes and or apices. Sucrose preculture had a positive role in survival and subsequent regrowth of the cryopreserved explants. Seven days on 0.5 M sucrose solution significantly improved the viability of nodal segments. PVS2 incubation for 45 minutes combined with a 7-day preculture gave the optimum result of 66 percent. Plantlets derived after cryopreservation resumed growth and regenerated normally.

  14. 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)

  15. 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

  16. Isospectral discrete and quantum graphs with the same flip counts and nodal counts

    Science.gov (United States)

    Juul, Jonas S.; Joyner, Christopher H.

    2018-06-01

    The existence of non-isomorphic graphs which share the same Laplace spectrum (to be referred to as isospectral graphs) leads naturally to the following question: what additional information is required in order to resolve isospectral graphs? It was suggested by Band, Shapira and Smilansky that this might be achieved by either counting the number of nodal domains or the number of times the eigenfunctions change sign (the so-called flip count) (Band et al 2006 J. Phys. A: Math. Gen. 39 13999–4014 Band and Smilansky 2007 Eur. Phys. J. Spec. Top. 145 171–9). Recent examples of (discrete) isospectral graphs with the same flip count and nodal count have been constructed by Ammann by utilising Godsil–McKay switching (Ammann private communication). Here, we provide a simple alternative mechanism that produces systematic examples of both discrete and quantum isospectral graphs with the same flip and nodal counts.

  17. NLTE4 Plasma Population Kinetics Database

    Science.gov (United States)

    SRD 159 NLTE4 Plasma Population Kinetics Database (Web database for purchase)   This database contains benchmark results for simulation of plasma population kinetics and emission spectra. The data were contributed by the participants of the 4th Non-LTE Code Comparison Workshop who have unrestricted access to the database. The only limitation for other users is in hidden labeling of the output results. Guest users can proceed to the database entry page without entering userid and password.

  18. Coding Class

    DEFF Research Database (Denmark)

    Ejsing-Duun, Stine; Hansbøl, Mikala

    Denne rapport rummer evaluering og dokumentation af Coding Class projektet1. Coding Class projektet blev igangsat i skoleåret 2016/2017 af IT-Branchen i samarbejde med en række medlemsvirksomheder, Københavns kommune, Vejle Kommune, Styrelsen for IT- og Læring (STIL) og den frivillige forening...... Coding Pirates2. Rapporten er forfattet af Docent i digitale læringsressourcer og forskningskoordinator for forsknings- og udviklingsmiljøet Digitalisering i Skolen (DiS), Mikala Hansbøl, fra Institut for Skole og Læring ved Professionshøjskolen Metropol; og Lektor i læringsteknologi, interaktionsdesign......, design tænkning og design-pædagogik, Stine Ejsing-Duun fra Forskningslab: It og Læringsdesign (ILD-LAB) ved Institut for kommunikation og psykologi, Aalborg Universitet i København. Vi har fulgt og gennemført evaluering og dokumentation af Coding Class projektet i perioden november 2016 til maj 2017...

  19. Uplink Coding

    Science.gov (United States)

    Andrews, Ken; Divsalar, Dariush; Dolinar, Sam; Moision, Bruce; Hamkins, Jon; Pollara, Fabrizio

    2007-01-01

    This slide presentation reviews the objectives, meeting goals and overall NASA goals for the NASA Data Standards Working Group. The presentation includes information on the technical progress surrounding the objective, short LDPC codes, and the general results on the Pu-Pw tradeoff.

  20. ANIMAL code

    International Nuclear Information System (INIS)

    Lindemuth, I.R.

    1979-01-01

    This report describes ANIMAL, a two-dimensional Eulerian magnetohydrodynamic computer code. ANIMAL's physical model also appears. Formulated are temporal and spatial finite-difference equations in a manner that facilitates implementation of the algorithm. Outlined are the functions of the algorithm's FORTRAN subroutines and variables

  1. Network Coding

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 15; Issue 7. Network Coding. K V Rashmi Nihar B Shah P Vijay Kumar. General Article Volume 15 Issue 7 July 2010 pp 604-621. Fulltext. Click here to view fulltext PDF. Permanent link: https://www.ias.ac.in/article/fulltext/reso/015/07/0604-0621 ...

  2. MCNP code

    International Nuclear Information System (INIS)

    Cramer, S.N.

    1984-01-01

    The MCNP code is the major Monte Carlo coupled neutron-photon transport research tool at the Los Alamos National Laboratory, and it represents the most extensive Monte Carlo development program in the United States which is available in the public domain. The present code is the direct descendent of the original Monte Carlo work of Fermi, von Neumaum, and Ulam at Los Alamos in the 1940s. Development has continued uninterrupted since that time, and the current version of MCNP (or its predecessors) has always included state-of-the-art methods in the Monte Carlo simulation of radiation transport, basic cross section data, geometry capability, variance reduction, and estimation procedures. The authors of the present code have oriented its development toward general user application. The documentation, though extensive, is presented in a clear and simple manner with many examples, illustrations, and sample problems. In addition to providing the desired results, the output listings give a a wealth of detailed information (some optional) concerning each state of the calculation. The code system is continually updated to take advantage of advances in computer hardware and software, including interactive modes of operation, diagnostic interrupts and restarts, and a variety of graphical and video aids

  3. Expander Codes

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 10; Issue 1. Expander Codes - The Sipser–Spielman Construction. Priti Shankar. General Article Volume 10 ... Author Affiliations. Priti Shankar1. Department of Computer Science and Automation, Indian Institute of Science Bangalore 560 012, India.

  4. 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.

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  6. 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.

  7. [Does nodal irradiation (clavicular and internal mammary chains) increase the toxicity of adjuvant breast radiotherapy?].

    Science.gov (United States)

    Riou, O; Bourgier, C; Fenoglietto, P; Azria, D

    2015-06-01

    Treatment volume is a major risk factor of radiation-induced toxicity. As nodal irradiation increases treatment volume, radiation toxicity should be greater. Nevertheless, scientific randomised data do not support this fact. However, a radiation-induced toxicity is possible outside tangential fields in the nodal volumes not related to breast-only treatment. Treatment should not be adapted only to the disease but personalized to the individual risk of toxicity for each patient. Copyright © 2015 Société française de radiothérapie oncologique (SFRO). Published by Elsevier SAS. All rights reserved.

  8. A self-consistent nodal method in response matrix formalism for the multigroup diffusion equations

    International Nuclear Information System (INIS)

    Malambu, E.M.; Mund, E.H.

    1996-01-01

    We develop a nodal method for the multigroup diffusion equations, based on the transverse integration procedure (TIP). The efficiency of the method rests upon the convergence properties of a high-order multidimensional nodal expansion and upon numerical implementation aspects. The discrete 1D equations are cast in response matrix formalism. The derivation of the transverse leakage moments is self-consistent i.e. does not require additional assumptions. An outstanding feature of the method lies in the linear spatial shape of the local transverse leakage for the first-order scheme. The method is described in the two-dimensional case. The method is validated on some classical benchmark problems. (author)

  9. Nodal signals mediate interactions between the extra-embryonic and embryonic tissues in zebrafish

    OpenAIRE

    Xiang, Fan; Hagos, Engda G.; Xu, Bo; Sias, Christina; Kawakami, Koichi; Burdine, Rebecca D.; Dougan, Scott T.

    2007-01-01

    In many vertebrates, extra-embryonic tissues are important signaling centers that induce and pattern the germ layers. In teleosts, the mechanism by which the extra-embryonic yolk syncytial layer (YSL) patterns the embryo is not understood. Although the Nodal-related protein Squint is expressed in the YSL, its role in this tissue is not known. We generated a series of stable transgenic lines with GFP under the control of squint genomic sequences. In all species, nodal-related genes induce thei...

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  11. 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)

  12. 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.

  13. 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

  14. The development of a transient neutron flux solution in the PANTHER code

    International Nuclear Information System (INIS)

    Hutt, P.K.; Knight, M.P.

    1990-01-01

    In the United Kingdom a new three-dimensional, two-group, homogeneous reactor diffusion code, PANTHER, has been developed for the analysis of pressurized water reactors (PWRs) and advanced gas-cooled reactors (AGRs). The code can perform a comprehensive range of calculations, steady state, depletion, and transient with either a finite difference or analytic nodal flux solution. The nodal solution allows the representation of within-node burnup variation and pin-power reconstruction in either steady-state or transient mode. Specific steady-state and transient thermal feedback modules are included for both PWRs and AGRs. The code is being developed to perform a complete range of reactor calculations from online operational support to fuel management and fault transient analysis. In the area of transient analysis, the code is currently being used for a number of PWR fault transient assessments, including rod ejection and steam-line break. In addition, work is proceeding to incorporate the PANTHER 3D nodal transient solution in the TRAC-P code. This paper outlines the development of the transient flux solutions within PANTHER

  15. 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.

  16. Panda code

    International Nuclear Information System (INIS)

    Altomare, S.; Minton, G.

    1975-02-01

    PANDA is a new two-group one-dimensional (slab/cylinder) neutron diffusion code designed to replace and extend the FAB series. PANDA allows for the nonlinear effects of xenon, enthalpy and Doppler. Fuel depletion is allowed. PANDA has a completely general search facility which will seek criticality, maximize reactivity, or minimize peaking. Any single parameter may be varied in a search. PANDA is written in FORTRAN IV, and as such is nearly machine independent. However, PANDA has been written with the present limitations of the Westinghouse CDC-6600 system in mind. Most computation loops are very short, and the code is less than half the useful 6600 memory size so that two jobs can reside in the core at once. (auth)

  17. CANAL code

    International Nuclear Information System (INIS)

    Gara, P.; Martin, E.

    1983-01-01

    The CANAL code presented here optimizes a realistic iron free extraction channel which has to provide a given transversal magnetic field law in the median plane: the current bars may be curved, have finite lengths and cooling ducts and move in a restricted transversal area; terminal connectors may be added, images of the bars in pole pieces may be included. A special option optimizes a real set of circular coils [fr

  18. High-Fidelity Kinetics and Radiation Transport for NLTE Hypersonic Flows, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — The modeling of NLTE hypersonic flows combines several disciplines: chemistry, kinetics, radiation transport, fluid mechanics, and surface science. No single code or...

  19. Stability, accuracy and numerical diffusion analysis of nodal expansion method for steady convection diffusion equation

    International Nuclear Information System (INIS)

    Zhou, Xiafeng; Guo, Jiong; Li, Fu

    2015-01-01

    Highlights: • NEMs are innovatively applied to solve convection diffusion equation. • Stability, accuracy and numerical diffusion for NEM are analyzed for the first time. • Stability and numerical diffusion depend on the NEM expansion order and its parity. • NEMs have higher accuracy than both second order upwind and QUICK scheme. • NEMs with different expansion orders are integrated into a unified discrete form. - Abstract: The traditional finite difference method or finite volume method (FDM or FVM) is used for HTGR thermal-hydraulic calculation at present. However, both FDM and FVM require the fine mesh sizes to achieve the desired precision and thus result in a limited efficiency. Therefore, a more efficient and accurate numerical method needs to be developed. Nodal expansion method (NEM) can achieve high accuracy even on the coarse meshes in the reactor physics analysis so that the number of spatial meshes and computational cost can be largely decreased. Because of higher efficiency and accuracy, NEM can be innovatively applied to thermal-hydraulic calculation. In the paper, NEMs with different orders of basis functions are successfully developed and applied to multi-dimensional steady convection diffusion equation. Numerical results show that NEMs with three or higher order basis functions can track the reference solutions very well and are superior to second order upwind scheme and QUICK scheme. However, the false diffusion and unphysical oscillation behavior are discovered for NEMs. To explain the reasons for the above-mentioned behaviors, the stability, accuracy and numerical diffusion properties of NEM are analyzed by the Fourier analysis, and by comparing with exact solutions of difference and differential equation. The theoretical analysis results show that the accuracy of NEM increases with the expansion order. However, the stability and numerical diffusion properties depend not only on the order of basis functions but also on the parity of

  20. Stability, accuracy and numerical diffusion analysis of nodal expansion method for steady convection diffusion equation

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Xiafeng, E-mail: zhou-xf11@mails.tsinghua.edu.cn; Guo, Jiong, E-mail: guojiong12@tsinghua.edu.cn; Li, Fu, E-mail: lifu@tsinghua.edu.cn

    2015-12-15

    Highlights: • NEMs are innovatively applied to solve convection diffusion equation. • Stability, accuracy and numerical diffusion for NEM are analyzed for the first time. • Stability and numerical diffusion depend on the NEM expansion order and its parity. • NEMs have higher accuracy than both second order upwind and QUICK scheme. • NEMs with different expansion orders are integrated into a unified discrete form. - Abstract: The traditional finite difference method or finite volume method (FDM or FVM) is used for HTGR thermal-hydraulic calculation at present. However, both FDM and FVM require the fine mesh sizes to achieve the desired precision and thus result in a limited efficiency. Therefore, a more efficient and accurate numerical method needs to be developed. Nodal expansion method (NEM) can achieve high accuracy even on the coarse meshes in the reactor physics analysis so that the number of spatial meshes and computational cost can be largely decreased. Because of higher efficiency and accuracy, NEM can be innovatively applied to thermal-hydraulic calculation. In the paper, NEMs with different orders of basis functions are successfully developed and applied to multi-dimensional steady convection diffusion equation. Numerical results show that NEMs with three or higher order basis functions can track the reference solutions very well and are superior to second order upwind scheme and QUICK scheme. However, the false diffusion and unphysical oscillation behavior are discovered for NEMs. To explain the reasons for the above-mentioned behaviors, the stability, accuracy and numerical diffusion properties of NEM are analyzed by the Fourier analysis, and by comparing with exact solutions of difference and differential equation. The theoretical analysis results show that the accuracy of NEM increases with the expansion order. However, the stability and numerical diffusion properties depend not only on the order of basis functions but also on the parity of