WorldWideScience

Sample records for panther neutronics code

  1. Development of an integrated thermal-hydraulics capability incorporating RELAP5 and PANTHER neutronics code

    Energy Technology Data Exchange (ETDEWEB)

    Page, R.; Jones, J.R.

    1997-07-01

    Ensuring that safety analysis needs are met in the future is likely to lead to the development of new codes and the further development of existing codes. It is therefore advantageous to define standards for data interfaces and to develop software interfacing techniques which can readily accommodate changes when they are made. Defining interface standards is beneficial but is necessarily restricted in application if future requirements are not known in detail. Code interfacing methods are of particular relevance with the move towards automatic grid frequency response operation where the integration of plant dynamic, core follow and fault study calculation tools is considered advantageous. This paper describes the background and features of a new code TALINK (Transient Analysis code LINKage program) used to provide a flexible interface to link the RELAP5 thermal hydraulics code with the PANTHER neutron kinetics and the SIBDYM whole plant dynamic modelling codes used by Nuclear Electric. The complete package enables the codes to be executed in parallel and provides an integrated whole plant thermal-hydraulics and neutron kinetics model. In addition the paper discusses the capabilities and pedigree of the component codes used to form the integrated transient analysis package and the details of the calculation of a postulated Sizewell `B` Loss of offsite power fault transient.

  2. Development of an integrated thermal-hydraulics capability incorporating RELAP5 and PANTHER neutronics code

    International Nuclear Information System (INIS)

    Page, R.; Jones, J.R.

    1997-01-01

    Ensuring that safety analysis needs are met in the future is likely to lead to the development of new codes and the further development of existing codes. It is therefore advantageous to define standards for data interfaces and to develop software interfacing techniques which can readily accommodate changes when they are made. Defining interface standards is beneficial but is necessarily restricted in application if future requirements are not known in detail. Code interfacing methods are of particular relevance with the move towards automatic grid frequency response operation where the integration of plant dynamic, core follow and fault study calculation tools is considered advantageous. This paper describes the background and features of a new code TALINK (Transient Analysis code LINKage program) used to provide a flexible interface to link the RELAP5 thermal hydraulics code with the PANTHER neutron kinetics and the SIBDYM whole plant dynamic modelling codes used by Nuclear Electric. The complete package enables the codes to be executed in parallel and provides an integrated whole plant thermal-hydraulics and neutron kinetics model. In addition the paper discusses the capabilities and pedigree of the component codes used to form the integrated transient analysis package and the details of the calculation of a postulated Sizewell 'B' Loss of offsite power fault transient

  3. 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

  4. Application of the coupled Relap5/Panther codes for PWR steam. Line break accident analysis

    International Nuclear Information System (INIS)

    Guisset, J.-P.; Bosso, S.; Charlier, A.; Delhaye, X.; Ergo, O.; Ouliddren, K.; Schneidesch, C.; Zhang, J.

    2001-01-01

    A dynamic coupling between the existing 1-dimensional thermal-hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER is applied via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel thermal-hydraulic analysis code COBRA 3C allows direct evaluation of the Departure from Nucleate Boiling Ratio in parallel with the coupled PANTHER/RELAP5 simulation. The coupled codes are applied to develop a Final Safety Analysis Report (FSAR) accident analysis methodology for the major Steam Line Break (SLB) accident at hot zero power in a typical three-loop pressurised water reactor. In this methodology, the uncertainties related to the plant, core thermal-hydraulic and neutronic parameters are combined in a deterministic bounding approach based on sensitivity studies. The results of coupled thermal-hydraulic and neutronic analysis of SLB are presented and discussed. It is shown that there exists an important margin in the traditional FSAR accident analysis for SLB, which can be attributed by the conservatism's introduced by de-coupling the plant sub-systems. (author)

  5. Application of the coupled Relap5/Panther codes for PWR steam. Line break accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Guisset, J.-P.; Bosso, S.; Charlier, A.; Delhaye, X.; Ergo, O.; Ouliddren, K.; Schneidesch, C.; Zhang, J. [Tractebel Energy Engineering, Brussels (Belgium)

    2001-07-01

    A dynamic coupling between the existing 1-dimensional thermal-hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER is applied via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel thermal-hydraulic analysis code COBRA 3C allows direct evaluation of the Departure from Nucleate Boiling Ratio in parallel with the coupled PANTHER/RELAP5 simulation. The coupled codes are applied to develop a Final Safety Analysis Report (FSAR) accident analysis methodology for the major Steam Line Break (SLB) accident at hot zero power in a typical three-loop pressurised water reactor. In this methodology, the uncertainties related to the plant, core thermal-hydraulic and neutronic parameters are combined in a deterministic bounding approach based on sensitivity studies. The results of coupled thermal-hydraulic and neutronic analysis of SLB are presented and discussed. It is shown that there exists an important margin in the traditional FSAR accident analysis for SLB, which can be attributed by the conservatism's introduced by de-coupling the plant sub-systems. (author)

  6. PANTHER - Polarisation Analysis with Thermal neutron

    International Nuclear Information System (INIS)

    Deen, P.P.; Fennell, T.; Schober, H.; Orecchini, A.; Rols, S.; Andersen, K.H.; Stewart, J.R.

    2011-01-01

    PANTHER will build on the success of IN4, the world's most intense time-of-flight spectrometer. A large position-sensitive detector (PSD) will improve data collection rates significantly, the background will be greatly reduced, and it will incorporate features indispensable for magnetic studies (small angles, polarisation analysis, high magnetic field devices). The new instrument will enable rapid surveys of (Q,ω) space, as well as more detailed studies in fields ranging from magnetism to the structural excitations - phonon densities of states, dispersion of collective modes and molecular vibrations - that govern the behaviour of many important physical and chemical systems. (authors)

  7. 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

  8. The TE coupled RELAP5/PANTHER/COBRA code package and methodology for integrated PWR accident analysis

    International Nuclear Information System (INIS)

    Schneidesch, Christophe R.; Zhang, Jinzhao; Ammirabile, Luca; Dalleur, Jean-Paul

    2006-01-01

    At Tractebel Engineering (TE), a dynamic coupling has been developed between the best estimate thermal hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel core thermal-hydraulic analysis code COBRA 3C has been established for on-line calculation of the Departure from Nucleate Boiling Ratio (DNBR). In addition to the standard RELAP5-PANTHER coupling, the fully dynamic coupling of the RELAP5/PANTHER/COBRA3C-TE code package can be activated for evaluation purposes in which the PANTHER close-channel thermal-hydraulics module is replaced by the COBRA3C-TE with cross flow modelling and extended T-H flow conditions capabilities. The qualification of the RELAP5-PANTHER coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been approved by the Belgian Safety Authorities and is used at TE for analyzing asymmetric PWR accidents with strong core-system interactions. In particular, the TE coupled code package was first used to develop a main steam line break in hot shutdown conditions (SLBHZP) accident analysis methodology based on the TE deterministic bounding approach. This methodology has been reviewed and accepted by the Belgian Safety Authorities for specific applications. Those specific applications are related to the power up-rate and steam generator replacement project of the Doel 2 plant or to the Tihange-3 SLB accident re-analysis. A coupled feedwater line break (FLB) accident analysis methodology is currently being reviewed for application approval. The results of coupled thermal-hydraulic and neutronic analysis of SLB and FLB show that there exist important margins in the traditional final safety analysis report (FSAR) accident analysis. Those margins can be used to increase the operational flexibility of the plants. Moreover, the

  9. The TE coupled RELAP5/PANTHER/COBRA code package and methodology for integrated PWR accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Schneidesch, Christophe R.; Zhang, Jinzhao; Ammirabile, Luca; Dalleur, Jean-Paul [Suez-Tractebel Engineering, Avenue Ariane 7, B-1200 Brussels (Belgium)

    2006-07-01

    At Tractebel Engineering (TE), a dynamic coupling has been developed between the best estimate thermal hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel core thermal-hydraulic analysis code COBRA 3C has been established for on-line calculation of the Departure from Nucleate Boiling Ratio (DNBR). In addition to the standard RELAP5-PANTHER coupling, the fully dynamic coupling of the RELAP5/PANTHER/COBRA3C-TE code package can be activated for evaluation purposes in which the PANTHER close-channel thermal-hydraulics module is replaced by the COBRA3C-TE with cross flow modelling and extended T-H flow conditions capabilities. The qualification of the RELAP5-PANTHER coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been approved by the Belgian Safety Authorities and is used at TE for analyzing asymmetric PWR accidents with strong core-system interactions. In particular, the TE coupled code package was first used to develop a main steam line break in hot shutdown conditions (SLBHZP) accident analysis methodology based on the TE deterministic bounding approach. This methodology has been reviewed and accepted by the Belgian Safety Authorities for specific applications. Those specific applications are related to the power up-rate and steam generator replacement project of the Doel 2 plant or to the Tihange-3 SLB accident re-analysis. A coupled feedwater line break (FLB) accident analysis methodology is currently being reviewed for application approval. The results of coupled thermal-hydraulic and neutronic analysis of SLB and FLB show that there exist important margins in the traditional final safety analysis report (FSAR) accident analysis. Those margins can be used to increase the operational flexibility of the plants. Moreover, the

  10. Simulation of the OECD Main-Steam-Line-Break Benchmark Exercise 3 Using the Coupled RELAP5/PANTHER Codes

    International Nuclear Information System (INIS)

    Schneidesch, Christophe R.; Zhang Jinzhao

    2004-01-01

    The RELAP5 best-estimate thermal-hydraulic system code has been coupled with the PANTHER three-dimensional neutron kinetics code via the TALINK dynamic data exchange control and processing tool. The coupled RELAP5/PANTHER code package has been qualified and will be used at Tractebel Engineering (TE) for analyzing asymmetric pressurized water reactor (PWR) accidents with strong core-system interactions. The Organization for Economic Cooperation and Development/U.S. Nuclear Regulatory Commission PWR main-steam-line-break benchmark problem was analyzed as part of the qualification efforts to demonstrate the capability of the coupled code package of simulating such transients. This paper reports the main results of TE's contribution to the benchmark Exercise 3

  11. Qualification of the coupled RELAP5/PANTHER/COBRA code package for licensing applications

    International Nuclear Information System (INIS)

    Schneidesch, C.R.; Zhang Jinzhao

    2004-01-01

    A coupled thermal hydraulics-neutronics code package has been developed at Tractebel Engineering (TE), in which the best-estimate thermal-hydraulic system code, RELAP5/mod2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via the dynamic data exchange interface, TALINK. The Departure from Nucleate Boiling Ratio (DNBR) is calculated by the sub-channel thermal-hydraulic analysis code COBRA-3C. The package provides the capability to accurately simulate the key physical phenomena in nuclear power plant accidents with strong asymmetric behaviours and system-core interactions. This paper presents the TE coupled code package and focuses on the methodology followed for qualifying it for licensing applications. The qualification of the coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been qualified and will be used at Tractebel Engineering (TE) for analyzing asymmetric PWR accidents with strong core-system interactions

  12. Analysis of the main steam line break accident with loss of offsite power using the fully coupled RELAP5/PANTHER/COBRA code package

    International Nuclear Information System (INIS)

    Ruben Van Parys; Sandrine Bosso; Christophe Schneidesch; Jinzhao Zhang

    2005-01-01

    Full text of publication follows: A coupled thermal hydraulics-neutronics code package (RELAP5/PANTHER/COBRA) has been qualified for accident analysis at Tractebel Engineering. In the TE coupled code package, the best estimate thermal-hydraulic system code, RELAP5/MOD2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via a dynamic data exchange control and processing tool, TALINK. An interface between PANTHER code and the sub-channel thermal-hydraulic analysis code COBRA-IIIC is developed in order to perform online calculation of Departure from Nucleate Boiling Ratio (DNBR). The TE coupled code package has been applied to develop a MSLB accident analysis methodology using the TE deterministic bounding approach. The methodology has been applied for MSLB accident analysis in support of licensing of the power up-rate and steam generator replacement of the Doel 2 plant. The results of coupled thermal-hydraulic and neutronic analysis of SLB show that there exists an important margin in the traditional FSAR MSLB accident analysis. As a specific licensing requirement, the main steam line break accident with loss of offsite power has to be analyzed. In the standard methodology with the coupled RELAP5/PANTHER code, and some corrective methods has to be taken in order to overcome the limitations due to the close-channel T/H model in PANTHER at low flow conditions. The results show that the steam line break accident with loss of offsite power is far less limiting. In order to verify the effect of the cross-flow at low flow conditions, the fully dynamic coupling of RELAP5/PANTHER/COBRA code package is used for reanalysis of this case, in which the PANTHER close-channel T/H model is replaced by the COBRA sub-channel T/H model with crossflow option. It has been demonstrated that, although the consideration of cross-flow in this challenging situation may lead to higher core return to power and slightly lower DNBR than in the standard methodology

  13. Analysis of the main steam line break accident with loss of offsite power using the fully coupled RELAP5/PANTHER/COBRA code package

    Energy Technology Data Exchange (ETDEWEB)

    Ruben Van Parys; Sandrine Bosso; Christophe Schneidesch; Jinzhao Zhang [Nuclear Department, Suez-Tractebel Engineering, avenue Ariane 5, B-1200 Brussels (Belgium)

    2005-07-01

    Full text of publication follows: A coupled thermal hydraulics-neutronics code package (RELAP5/PANTHER/COBRA) has been qualified for accident analysis at Tractebel Engineering. In the TE coupled code package, the best estimate thermal-hydraulic system code, RELAP5/MOD2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via a dynamic data exchange control and processing tool, TALINK. An interface between PANTHER code and the sub-channel thermal-hydraulic analysis code COBRA-IIIC is developed in order to perform online calculation of Departure from Nucleate Boiling Ratio (DNBR). The TE coupled code package has been applied to develop a MSLB accident analysis methodology using the TE deterministic bounding approach. The methodology has been applied for MSLB accident analysis in support of licensing of the power up-rate and steam generator replacement of the Doel 2 plant. The results of coupled thermal-hydraulic and neutronic analysis of SLB show that there exists an important margin in the traditional FSAR MSLB accident analysis. As a specific licensing requirement, the main steam line break accident with loss of offsite power has to be analyzed. In the standard methodology with the coupled RELAP5/PANTHER code, and some corrective methods has to be taken in order to overcome the limitations due to the close-channel T/H model in PANTHER at low flow conditions. The results show that the steam line break accident with loss of offsite power is far less limiting. In order to verify the effect of the cross-flow at low flow conditions, the fully dynamic coupling of RELAP5/PANTHER/COBRA code package is used for reanalysis of this case, in which the PANTHER close-channel T/H model is replaced by the COBRA sub-channel T/H model with crossflow option. It has been demonstrated that, although the consideration of cross-flow in this challenging situation may lead to higher core return to power and slightly lower DNBR than in the standard methodology

  14. Improvements to the transient solution in the PANTHER space-time code

    International Nuclear Information System (INIS)

    Kutt, P.K.; Knight, M.P.

    1993-01-01

    The three dimensional, two-group, nodal diffusion code PANTHER has been developed for the analysis of almost all thermal reactor types [pressurized water reactor (PWR), boiling water reactor, VVER, RBMK, advanced gas-cooled reactor, MAGNOX]. It can perform a comprehensive range of calculations for fuel management, operational support including on-line application, and transient analysis. Transient results for a number of light water reactor (LWR) benchmark problems have been reported previously. This paper outlines some recent developments of the transient solution in PANTHER, showing results for two LWR benchmark problems. Recently, PANTHER results have been accepted as the reference solutions for a Nuclear Energy Agency Committee on Reactor Physics (NEACRP) rod ejection benchmark Unlike previous simplified rod ejection benchmarks, it represents a real PWR with a detailed thermal model and cross sections dependent on boron, fuel temperature, and water density and temperature. This reference solution was computed with fine time steps

  15. Neutronics codes

    International Nuclear Information System (INIS)

    Buckel, G.

    1983-01-01

    The objectives are the development, testing and cultivation of reliable, efficient and user-optimized neutron-physical calculation methods and conformity with users' requirements concerning design of power reactors, planning and analysis of experiments necessary for their protection as well as research on physical key problems. A short outline of available computing programmes for the following objectives is given: - Provision of macroscopic group constants, - Calculation of neutron flux distribution in transport theory and diffusion approximation, - Evaluation of neutron flux-distribution, - Execution of disturbance calculations for the determination reactivity coefficients, and - graphical representation of results. (orig./RW) [de

  16. AGR fuel management using PANTHER

    International Nuclear Information System (INIS)

    Haddock, S.A.; Parks, G.T.

    1995-01-01

    This paper describes recent improvements in the AGR fuel management methodology implemented within PANTHER and the use of the code both for stand-alone calculations and within an automatic optimisation procedure. (author)

  17. The CORSYS neutronics code system

    International Nuclear Information System (INIS)

    Caner, M.; Krumbein, A.D.; Saphier, D.; Shapira, M.

    1994-01-01

    The purpose of this work is to assemble a code package for LWR core physics including coupled neutronics, burnup and thermal hydraulics. The CORSYS system is built around the cell code WIMS (for group microscopic cross section calculations) and 3-dimension diffusion code CITATION (for burnup and fuel management). We are implementing such a system on an IBM RS-6000 workstation. The code was rested with a simplified model of the Zion Unit 2 PWR. (authors). 6 refs., 8 figs., 1 tabs

  18. Pressurised water reactor fuel management using PANTHER

    International Nuclear Information System (INIS)

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

    1996-01-01

    This paper describes the integration of Nuclear Electric's reactor physics code PANTHER with an automatic optimisation procedure designed to search for optimal PWR reload cores and assesses its performance. (Author)

  19. Monte Carlo code for neutron radiography

    International Nuclear Information System (INIS)

    Milczarek, Jacek J.; Trzcinski, Andrzej; El-Ghany El Abd, Abd; Czachor, Andrzej

    2005-01-01

    The concise Monte Carlo code, MSX, for simulation of neutron radiography images of non-uniform objects is presented. The possibility of modeling the images of objects with continuous spatial distribution of specific isotopes is included. The code can be used for assessment of the scattered neutron component in neutron radiograms

  20. Monte Carlo code for neutron radiography

    Energy Technology Data Exchange (ETDEWEB)

    Milczarek, Jacek J. [Institute of Atomic Energy, Swierk, 05-400 Otwock (Poland)]. E-mail: jjmilcz@cyf.gov.pl; Trzcinski, Andrzej [Institute for Nuclear Studies, Swierk, 05-400 Otwock (Poland); El-Ghany El Abd, Abd [Institute of Atomic Energy, Swierk, 05-400 Otwock (Poland); Nuclear Research Center, PC 13759, Cairo (Egypt); Czachor, Andrzej [Institute of Atomic Energy, Swierk, 05-400 Otwock (Poland)

    2005-04-21

    The concise Monte Carlo code, MSX, for simulation of neutron radiography images of non-uniform objects is presented. The possibility of modeling the images of objects with continuous spatial distribution of specific isotopes is included. The code can be used for assessment of the scattered neutron component in neutron radiograms.

  1. Monte Carlo codes use in neutron therapy

    International Nuclear Information System (INIS)

    Paquis, P.; Mokhtari, F.; Karamanoukian, D.; Pignol, J.P.; Cuendet, P.; Iborra, N.

    1998-01-01

    Monte Carlo calculation codes allow to study accurately all the parameters relevant to radiation effects, like the dose deposition or the type of microscopic interactions, through one by one particle transport simulation. These features are very useful for neutron irradiations, from device development up to dosimetry. This paper illustrates some applications of these codes in Neutron Capture Therapy and Neutron Capture Enhancement of fast neutrons irradiations. (authors)

  2. Computer codes for neutron data evaluation

    International Nuclear Information System (INIS)

    Nakagawa, Tsuneo

    1979-01-01

    Data compilation codes such as NESTOR and REPSTOR, and NDES (Neutron Data Evaluation System) are mainly discussed. NDES is a code for neutron data evaluation using a TSS terminal, TEKTRONIX 4014. Users of NDES can perform plotting of data and calculation with nuclear models under conversational mode. (author)

  3. Ultrafast protein structure-based virtual screening with Panther

    Science.gov (United States)

    Niinivehmas, Sanna P.; Salokas, Kari; Lätti, Sakari; Raunio, Hannu; Pentikäinen, Olli T.

    2015-10-01

    Molecular docking is by far the most common method used in protein structure-based virtual screening. This paper presents Panther, a novel ultrafast multipurpose docking tool. In Panther, a simple shape-electrostatic model of the ligand-binding area of the protein is created by utilizing the protein crystal structure. The features of the possible ligands are then compared to the model by using a similarity search algorithm. On average, one ligand can be processed in a few minutes by using classical docking methods, whereas using Panther processing takes Panther protocol can be used in several applications, such as speeding up the early phases of drug discovery projects, reducing the number of failures in the clinical phase of the drug development process, and estimating the environmental toxicity of chemicals. Panther-code is available in our web pages (http://www.jyu.fi/panther) free of charge after registration.

  4. PANTHERMIX. A PANTHER-THERMIX interaction

    International Nuclear Information System (INIS)

    Oppe, J.; Haas, J.B.M. de; Kuijper, J.C.

    1996-05-01

    This report describes the combination of the general purpose modular rector code PANTHER and the pebble bed HTR thermal hydraulics code THERMIX-DIREKT. After the installation of a stand-alone THERMIX-DIREKT at ECN, the conversion programs PATOTH and THTOPA were developed at ECN. This report describes the details of the conversion programs. For PANTHER and THERMIX-DIREKT only those elements important for their interaction via the conversion programs are described here in detail. In appendices an example is given of a total interaction job. (orig.)

  5. Numerical and computational aspects of the coupled three-dimensional core/ plant simulations: organization for economic cooperation and development/ U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-II. 3. Analysis of the OECD TMI-1 Main-Steam- Line-Break Benchmark Accident Using the Coupled RELAP5/PANTHER Codes

    International Nuclear Information System (INIS)

    Schneidesch, C.R.; Guisset, J.P.; Zhang, J.; Bryce, P.; Parkes, M.

    2001-01-01

    The RELAP5 best-estimate thermal-hydraulic system code has been coupled with the PANTHER three-dimensional (3-D) neutron kinetics code via the TALINK dynamic data exchange control and processing tool. The coupled RELAP5/PANTHER code package is being qualified and will be used at British Energy (BE) and Tractebel Energy Engineering (TEE), independently, to analyze pressurized water reactor (PWR) transients where strong core-system interactions occur. The Organization for Economic Cooperation and Development/Nuclear Energy Agency PWR Main-Steam-Line-Break (MSLB) Benchmark problem was performed to demonstrate the capability of the coupled code package to simulate such transients, and this paper reports the BE and TEE contributions. In the first exercise, a point-kinetics (PK) calculation is performed using the RELAP5 code. Two solutions have been derived for the PK case. The first corresponds to scenario, 1 where calculations are carried out using the original (BE) rod worth and where no significant return to power (RTP) occurs. The second corresponds to scenario 2 with arbitrarily reduced rod worth in order to obtain RTP (and was not part of the 'official' results). The results, as illustrated in Fig. 1, show that the thermalhydraulic system response and rod worth are essential in determining the core response. The second exercise consists of a 3-D neutron kinetics transient calculation driven by best-estimate time-dependent core inlet conditions on a 18 T and H zones basis derived from TRAC-PF1/MOD2 (PSU), again analyzing two scenarios of different rod worths. Two sets of PANTHER solutions were submitted for exercise 2. The first solution uses a spatial discretization of one node per assembly and 24 core axial layers for both flux and T and H mesh. The second is characterized by spatial refinement (2 x 2 nodes per assembly, 48 core layers for flux, and T and H calculation), time refinement (half-size time steps), and an increased radial discretization for solution

  6. 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

  7. Fast neutron analysis code SAD1

    International Nuclear Information System (INIS)

    Jung, M.; Ott, C.

    1985-01-01

    A listing and an example of outputs of the M.C. code SAD1 are given here. This code has been used many times to predict responses of fast neutrons in hydrogenic materials (in our case emulsions or plastics) towards the elastic n, p scattering. It can be easily extended to other kinds of such materials and to any kind of incident fast neutron spectrum

  8. The PANTHER User Experience

    Energy Technology Data Exchange (ETDEWEB)

    Coram, Jamie L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Morrow, James D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Perkins, David Nikolaus [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    This document describes the PANTHER R&D Application, a proof-of-concept user interface application developed under the PANTHER Grand Challenge LDRD. The purpose of the application is to explore interaction models for graph analytics, drive algorithmic improvements from an end-user point of view, and support demonstration of PANTHER technologies to potential customers. The R&D Application implements a graph-centric interaction model that exposes analysts to the algorithms contained within the GeoGraphy graph analytics library. Users define geospatial-temporal semantic graph queries by constructing search templates based on nodes, edges, and the constraints among them. Users then analyze the results of the queries using both geo-spatial and temporal visualizations. Development of this application has made user experience an explicit driver for project and algorithmic level decisions that will affect how analysts one day make use of PANTHER technologies.

  9. Comparison of neutron spectrum unfolding codes

    International Nuclear Information System (INIS)

    Zijp, W.

    1979-02-01

    This final report contains a set of four ECN-reports. The first is dealing with the comparison of the neutron spectrum unfolding codes CRYSTAL BALL, RFSP-JUL, SAND II and STAY'SL. The other three present the results of calculations about the influence of statistical weights in CRYSTAL BALL, SAND II and RFSP-JUL

  10. PANTHERMIX (PANTHER-THERMIX). User manual

    International Nuclear Information System (INIS)

    Oppe, J.; De Haas, J.B.M.; Kuijper, J.C.

    1998-06-01

    The PANTHER code calculates steady-state or time-dependent power distribution in a reactor with a given temperature distribution. The THERMIX-DIREKT code calculates temperature and coolant flow distributions, in steady-state or transient mode, in a system with a given power distribution. It is described how to use the combination of the general purpose modular reactor code PANTHER and the HTR thermal hydraulics code THERMIX-DIREKT. An earlier version of PANTHERMIX consisted of THERMIX-DIREKT plus 2 conversion programs. The jobs and scripts to be edited by the user were very complex in their interactions. Therefore this version of PANTHERMIX has been extended with macros that take care of all these interactions, so the interaction parts of the jobs become much less complex. 6 refs

  11. Code system for fast reactor neutronics analysis

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Abe, Junji; Sato, Wakaei.

    1983-04-01

    A code system for analysis of fast reactor neutronics has been developed for the purpose of handy use and error reduction. The JOINT code produces the input data file to be used in the neutronics calculation code and also prepares the cross section library file with an assigned format. The effective cross sections are saved in the PDS file with an unified format. At the present stage, this code system includes the following codes; SLAROM, ESELEM5, EXPANDA-G for the production of effective cross sections and CITATION-FBR, ANISN-JR, TWOTRAN2, PHENIX, 3DB, MORSE, CIPER and SNPERT. In the course of the development, some utility programs and service programs have been additionaly developed. These are used for access of PDS file, edit of the cross sections and graphic display. Included in this report are a description of input data format of the JOINT and other programs, and of the function of each subroutine and utility programs. The usage of PDS file is also explained. In Appendix A, the input formats are described for the revised version of the CIPER code. (author)

  12. Neutron spectrum unfolding using computer code SAIPS

    International Nuclear Information System (INIS)

    Karim, S.

    1999-01-01

    The main objective of this project was to study the neutron energy spectrum at rabbit station-1 in Pakistan Research Reactor (PARR-I). To do so, multiple foils activation method was used to get the saturated activities. The computer code SAIPS was used to unfold the neutron spectra from the measured reaction rates. Of the three built in codes in SAIPS, only SANDI and WINDOWS were used. Contribution of thermal part of the spectra was observed to be higher than the fast one. It was found that the WINDOWS gave smooth spectra while SANDII spectra have violet oscillations in the resonance region. The uncertainties in the WINDOWS results are higher than those of SANDII. The results show reasonable agreement with the published results. (author)

  13. Generic programming for deterministic neutron transport codes

    International Nuclear Information System (INIS)

    Plagne, L.; Poncot, A.

    2005-01-01

    This paper discusses the implementation of neutron transport codes via generic programming techniques. Two different Boltzmann equation approximations have been implemented, namely the Sn and SPn methods. This implementation experiment shows that generic programming allows us to improve maintainability and readability of source codes with no performance penalties compared to classical approaches. In the present implementation, matrices and vectors as well as linear algebra algorithms are treated separately from the rest of source code and gathered in a tool library called 'Generic Linear Algebra Solver System' (GLASS). Such a code architecture, based on a linear algebra library, allows us to separate the three different scientific fields involved in transport codes design: numerical analysis, reactor physics and computer science. Our library handles matrices with optional storage policies and thus applies both to Sn code, where the matrix elements are computed on the fly, and to SPn code where stored matrices are used. Thus, using GLASS allows us to share a large fraction of source code between Sn and SPn implementations. Moreover, the GLASS high level of abstraction allows the writing of numerical algorithms in a form which is very close to their textbook descriptions. Hence the GLASS algorithms collection, disconnected from computer science considerations (e.g. storage policy), is very easy to read, to maintain and to extend. (authors)

  14. Fast-neutron, coded-aperture imager

    Science.gov (United States)

    Woolf, Richard S.; Phlips, Bernard F.; Hutcheson, Anthony L.; Wulf, Eric A.

    2015-06-01

    This work discusses a large-scale, coded-aperture imager for fast neutrons, building off a proof-of concept instrument developed at the U.S. Naval Research Laboratory (NRL). The Space Science Division at the NRL has a heritage of developing large-scale, mobile systems, using coded-aperture imaging, for long-range γ-ray detection and localization. The fast-neutron, coded-aperture imaging instrument, designed for a mobile unit (20 ft. ISO container), consists of a 32-element array of 15 cm×15 cm×15 cm liquid scintillation detectors (EJ-309) mounted behind a 12×12 pseudorandom coded aperture. The elements of the aperture are composed of 15 cm×15 cm×10 cm blocks of high-density polyethylene (HDPE). The arrangement of the aperture elements produces a shadow pattern on the detector array behind the mask. By measuring of the number of neutron counts per masked and unmasked detector, and with knowledge of the mask pattern, a source image can be deconvolved to obtain a 2-d location. The number of neutrons per detector was obtained by processing the fast signal from each PMT in flash digitizing electronics. Digital pulse shape discrimination (PSD) was performed to filter out the fast-neutron signal from the γ background. The prototype instrument was tested at an indoor facility at the NRL with a 1.8-μCi and 13-μCi 252Cf neutron/γ source at three standoff distances of 9, 15 and 26 m (maximum allowed in the facility) over a 15-min integration time. The imaging and detection capabilities of the instrument were tested by moving the source in half- and one-pixel increments across the image plane. We show a representative sample of the results obtained at one-pixel increments for a standoff distance of 9 m. The 1.8-μCi source was not detected at the 26-m standoff. In order to increase the sensitivity of the instrument, we reduced the fastneutron background by shielding the top, sides and back of the detector array with 10-cm-thick HDPE. This shielding configuration led

  15. Fast-neutron, coded-aperture imager

    International Nuclear Information System (INIS)

    Woolf, Richard S.; Phlips, Bernard F.; Hutcheson, Anthony L.; Wulf, Eric A.

    2015-01-01

    This work discusses a large-scale, coded-aperture imager for fast neutrons, building off a proof-of concept instrument developed at the U.S. Naval Research Laboratory (NRL). The Space Science Division at the NRL has a heritage of developing large-scale, mobile systems, using coded-aperture imaging, for long-range γ-ray detection and localization. The fast-neutron, coded-aperture imaging instrument, designed for a mobile unit (20 ft. ISO container), consists of a 32-element array of 15 cm×15 cm×15 cm liquid scintillation detectors (EJ-309) mounted behind a 12×12 pseudorandom coded aperture. The elements of the aperture are composed of 15 cm×15 cm×10 cm blocks of high-density polyethylene (HDPE). The arrangement of the aperture elements produces a shadow pattern on the detector array behind the mask. By measuring of the number of neutron counts per masked and unmasked detector, and with knowledge of the mask pattern, a source image can be deconvolved to obtain a 2-d location. The number of neutrons per detector was obtained by processing the fast signal from each PMT in flash digitizing electronics. Digital pulse shape discrimination (PSD) was performed to filter out the fast-neutron signal from the γ background. The prototype instrument was tested at an indoor facility at the NRL with a 1.8-μCi and 13-μCi 252Cf neutron/γ source at three standoff distances of 9, 15 and 26 m (maximum allowed in the facility) over a 15-min integration time. The imaging and detection capabilities of the instrument were tested by moving the source in half- and one-pixel increments across the image plane. We show a representative sample of the results obtained at one-pixel increments for a standoff distance of 9 m. The 1.8-μCi source was not detected at the 26-m standoff. In order to increase the sensitivity of the instrument, we reduced the fastneutron background by shielding the top, sides and back of the detector array with 10-cm-thick HDPE. This shielding configuration led

  16. Fast-neutron, coded-aperture imager

    Energy Technology Data Exchange (ETDEWEB)

    Woolf, Richard S., E-mail: richard.woolf@nrl.navy.mil; Phlips, Bernard F., E-mail: bernard.phlips@nrl.navy.mil; Hutcheson, Anthony L., E-mail: anthony.hutcheson@nrl.navy.mil; Wulf, Eric A., E-mail: eric.wulf@nrl.navy.mil

    2015-06-01

    This work discusses a large-scale, coded-aperture imager for fast neutrons, building off a proof-of concept instrument developed at the U.S. Naval Research Laboratory (NRL). The Space Science Division at the NRL has a heritage of developing large-scale, mobile systems, using coded-aperture imaging, for long-range γ-ray detection and localization. The fast-neutron, coded-aperture imaging instrument, designed for a mobile unit (20 ft. ISO container), consists of a 32-element array of 15 cm×15 cm×15 cm liquid scintillation detectors (EJ-309) mounted behind a 12×12 pseudorandom coded aperture. The elements of the aperture are composed of 15 cm×15 cm×10 cm blocks of high-density polyethylene (HDPE). The arrangement of the aperture elements produces a shadow pattern on the detector array behind the mask. By measuring of the number of neutron counts per masked and unmasked detector, and with knowledge of the mask pattern, a source image can be deconvolved to obtain a 2-d location. The number of neutrons per detector was obtained by processing the fast signal from each PMT in flash digitizing electronics. Digital pulse shape discrimination (PSD) was performed to filter out the fast-neutron signal from the γ background. The prototype instrument was tested at an indoor facility at the NRL with a 1.8-μCi and 13-μCi 252Cf neutron/γ source at three standoff distances of 9, 15 and 26 m (maximum allowed in the facility) over a 15-min integration time. The imaging and detection capabilities of the instrument were tested by moving the source in half- and one-pixel increments across the image plane. We show a representative sample of the results obtained at one-pixel increments for a standoff distance of 9 m. The 1.8-μCi source was not detected at the 26-m standoff. In order to increase the sensitivity of the instrument, we reduced the fastneutron background by shielding the top, sides and back of the detector array with 10-cm-thick HDPE. This shielding configuration led

  17. A code for leakage neutron spectra through thick shields

    International Nuclear Information System (INIS)

    Nagarajan, P.S.; Sethulakshmi, P.; Raghavendran, C.P.

    1975-01-01

    An exponential transform Monte Carlo code has been developed for deep penetration of neutrons and the results of leakage neutron spectra of this code have been compared with those of a basic Monte Carlo code for small thickness. The development of the code and optimisation of certain transform parameters are discussed and results are presented for a few thick shields of concrete and water in the context of neutron monitoring in the environs of accelerator and reactor shields. (author)

  18. SNAP - a three dimensional neutron diffusion code

    International Nuclear Information System (INIS)

    McCallien, C.W.J.

    1993-02-01

    This report describes a one- two- three-dimensional multi-group diffusion code, SNAP, which is primarily intended for neutron diffusion calculations but can also carry out gamma calculations if the diffusion approximation is accurate enough. It is suitable for fast and thermal reactor core calculations and for shield calculations. SNAP can solve the multi-group neutron diffusion equations using finite difference methods. The one-dimensional slab, cylindrical and spherical geometries and the two-dimensional case are all treated as simple special cases of three-dimensional geometries. Numerous reflective and periodic symmetry options are available and may be used to reduce the number of mesh points necessary to represent the system. Extrapolation lengths can be specified at internal and external boundaries. (Author)

  19. Interface requirements to couple thermal-hydraulic codes to 3D neutronic codes

    Energy Technology Data Exchange (ETDEWEB)

    Langenbuch, S.; Austregesilo, H.; Velkov, K. [GRS, Garching (Germany)] [and others

    1997-07-01

    The present situation of thermalhydraulics codes and 3D neutronics codes is briefly described and general considerations for coupling of these codes are discussed. Two different basic approaches of coupling are identified and their relative advantages and disadvantages are discussed. The implementation of the coupling for 3D neutronics codes in the system ATHLET is presented. Meanwhile, this interface is used for coupling three different 3D neutronics codes.

  20. Interface requirements to couple thermal-hydraulic codes to 3D neutronic codes

    International Nuclear Information System (INIS)

    Langenbuch, S.; Austregesilo, H.; Velkov, K.

    1997-01-01

    The present situation of thermalhydraulics codes and 3D neutronics codes is briefly described and general considerations for coupling of these codes are discussed. Two different basic approaches of coupling are identified and their relative advantages and disadvantages are discussed. The implementation of the coupling for 3D neutronics codes in the system ATHLET is presented. Meanwhile, this interface is used for coupling three different 3D neutronics codes

  1. Loading pattern optimization in hexagonal geometry using PANTHER

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  2. SRAC95; general purpose neutronics code system

    International Nuclear Information System (INIS)

    Okumura, Keisuke; Tsuchihashi, Keichiro; Kaneko, Kunio.

    1996-03-01

    SRAC is a general purpose neutronics code system applicable to core analyses of various types of reactors. Since the publication of JAERI-1302 for the revised SRAC in 1986, a number of additions and modifications have been made for nuclear data libraries and programs. Thus, the new version SRAC95 has been completed. The system consists of six kinds of nuclear data libraries(ENDF/B-IV, -V, -VI, JENDL-2, -3.1, -3.2), five modular codes integrated into SRAC95; collision probability calculation module (PIJ) for 16 types of lattice geometries, Sn transport calculation modules(ANISN, TWOTRAN), diffusion calculation modules(TUD, CITATION) and two optional codes for fuel assembly and core burn-up calculations(newly developed ASMBURN, revised COREBN). In this version, many new functions and data are implemented to support nuclear design studies of advanced reactors, especially for burn-up calculations. SRAC95 is available not only on conventional IBM-compatible computers but also on scalar or vector computers with the UNIX operating system. This report is the SRAC95 users manual which contains general description, contents of revisions, input data requirements, detail information on usage, sample input data and list of available libraries. (author)

  3. SRAC95; general purpose neutronics code system

    Energy Technology Data Exchange (ETDEWEB)

    Okumura, Keisuke; Tsuchihashi, Keichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio

    1996-03-01

    SRAC is a general purpose neutronics code system applicable to core analyses of various types of reactors. Since the publication of JAERI-1302 for the revised SRAC in 1986, a number of additions and modifications have been made for nuclear data libraries and programs. Thus, the new version SRAC95 has been completed. The system consists of six kinds of nuclear data libraries(ENDF/B-IV, -V, -VI, JENDL-2, -3.1, -3.2), five modular codes integrated into SRAC95; collision probability calculation module (PIJ) for 16 types of lattice geometries, Sn transport calculation modules(ANISN, TWOTRAN), diffusion calculation modules(TUD, CITATION) and two optional codes for fuel assembly and core burn-up calculations(newly developed ASMBURN, revised COREBN). In this version, many new functions and data are implemented to support nuclear design studies of advanced reactors, especially for burn-up calculations. SRAC95 is available not only on conventional IBM-compatible computers but also on scalar or vector computers with the UNIX operating system. This report is the SRAC95 users manual which contains general description, contents of revisions, input data requirements, detail information on usage, sample input data and list of available libraries. (author).

  4. Results from the Coded Aperture Neutron Imaging System (CANIS)

    International Nuclear Information System (INIS)

    Brubaker, Erik; Steele, John T.; Brennan, James S.; Hilton, Nathan R.; Marleau, Peter

    2010-01-01

    Because of their penetrating power, energetic neutrons and gamma rays (∼1 MeV) offer the best possibility of detecting highly shielded or distant special nuclear material (SNM). Of these, fast neutrons offer the greatest advantage due to their very low and well understood natural background. We are investigating a new approach to fast-neutron imaging- a coded aperture neutron imaging system (CANIS). Coded aperture neutron imaging should offer a highly efficient solution for improved detection speed, range, and sensitivity. We have demonstrated fast neutron and gamma ray imaging with several different configurations of coded masks patterns and detectors including an 'active' mask that is composed of neutron detectors. Here we describe our prototype detector and present some initial results from laboratory tests and demonstrations.

  5. Results from the coded aperture neutron imaging system

    International Nuclear Information System (INIS)

    Brubaker, Erik; Steele, John T.; Brennan, James S.; Marleau, Peter

    2010-01-01

    Because of their penetrating power, energetic neutrons and gamma rays (∼1 MeV) offer the best possibility of detecting highly shielded or distant special nuclear material (SNM). Of these, fast neutrons offer the greatest advantage due to their very low and well understood natural background. We are investigating a new approach to fast-neutron imaging - a coded aperture neutron imaging system (CANIS). Coded aperture neutron imaging should offer a highly efficient solution for improved detection speed, range, and sensitivity. We have demonstrated fast neutron and gamma ray imaging with several different configurations of coded masks patterns and detectors including an 'active' mask that is composed of neutron detectors. Here we describe our prototype detector and present some initial results from laboratory tests and demonstrations.

  6. Monte Carlo codes use in neutron therapy; Application de codes Monte Carlo en neutrontherapie

    Energy Technology Data Exchange (ETDEWEB)

    Paquis, P.; Mokhtari, F.; Karamanoukian, D. [Hopital Pasteur, 06 - Nice (France); Pignol, J.P. [Hopital du Hasenrain, 68 - Mulhouse (France); Cuendet, P. [CEA Centre d' Etudes de Saclay, 91 - Gif-sur-Yvette (France). Direction des Reacteurs Nucleaires; Fares, G.; Hachem, A. [Faculte des Sciences, 06 - Nice (France); Iborra, N. [Centre Antoine-Lacassagne, 06 - Nice (France)

    1998-04-01

    Monte Carlo calculation codes allow to study accurately all the parameters relevant to radiation effects, like the dose deposition or the type of microscopic interactions, through one by one particle transport simulation. These features are very useful for neutron irradiations, from device development up to dosimetry. This paper illustrates some applications of these codes in Neutron Capture Therapy and Neutron Capture Enhancement of fast neutrons irradiations. (authors)

  7. Scintillator Based Coded-Aperture Imaging for Neutron Detection

    International Nuclear Information System (INIS)

    Hayes, Sean-C.; Gamage, Kelum-A-A.

    2013-06-01

    In this paper we are going to assess the variations of neutron images using a series of Monte Carlo simulations. We are going to study neutron images of the same neutron source with different source locations, using a scintillator based coded-aperture system. The Monte Carlo simulations have been conducted making use of the EJ-426 neutron scintillator detector. This type of detector has a low sensitivity to gamma rays and is therefore of particular use in a system with a source that emits a mixed radiation field. From the use of different source locations, several neutron images have been produced, compared both qualitatively and quantitatively for each case. This allows conclusions to be drawn on how suited the scintillator based coded-aperture neutron imaging system is to detecting various neutron source locations. This type of neutron imaging system can be easily used to identify and locate nuclear materials precisely. (authors)

  8. SRAC2006: A comprehensive neutronics calculation code system

    International Nuclear Information System (INIS)

    Okumura, Keisuke; Kugo, Teruhiko; Kaneko, Kunio; Tsuchihashi, Keichiro

    2007-02-01

    The SRAC is a code system applicable to neutronics analysis of a variety of reactor types. Since the publication of the second version of the users manual (JAERI-1302) in 1986 for the SRAC system, a number of additions and modifications to the functions and the library data have been made to establish a comprehensive neutronics code system. The current system includes major neutron data libraries (JENDL-3.3, JENDL-3.2, ENDF/B-VII, ENDF/B-VI.8, JEFF-3.1, JEF-2.2, etc.), and integrates five elementary codes for neutron transport and diffusion calculation; PIJ based on the collision probability method applicable to 16 kind of lattice models, S N transport codes ANISN(1D) and TWOTRN(2D), diffusion codes TUD(1D) and CITATION(multi-D). The system also includes an auxiliary code COREBN for multi-dimensional core burn-up calculation. (author)

  9. A New Monte Carlo Neutron Transport Code at UNIST

    International Nuclear Information System (INIS)

    Lee, Hyunsuk; Kong, Chidong; Lee, Deokjung

    2014-01-01

    Monte Carlo neutron transport code named MCS is under development at UNIST for the advanced reactor design and research purpose. This MC code can be used for fixed source calculation and criticality calculation. Continuous energy neutron cross section data and multi-group cross section data can be used for the MC calculation. This paper presents the overview of developed MC code and its calculation results. The real time fixed source calculation ability is also tested in this paper. The calculation results show good agreement with commercial code and experiment. A new Monte Carlo neutron transport code is being developed at UNIST. The MC codes are tested with several benchmark problems: ICSBEP, VENUS-2, and Hoogenboom-Martin benchmark. These benchmarks covers pin geometry to 3-dimensional whole core, and results shows good agreement with reference results

  10. Description of the CAREM Reactor Neutronic Calculation Codes

    International Nuclear Information System (INIS)

    Villarino, Eduardo; Hergenreder, Daniel

    2000-01-01

    In this work is described the neutronic calculation line used to design the CAREM reactor.A description of the codes used and the interfaces between the different programs are presented.Both, the normal calculation line and the alternative or verification calculation line are included.The calculation line used to obtain the kinetics parameters (effective delayed-neutron fraction and prompt-neutron lifetime) is also included

  11. Neutron cross section library production code system for continuous energy Monte Carlo code MVP. LICEM

    International Nuclear Information System (INIS)

    Mori, Takamasa; Nakagawa, Masayuki; Kaneko, Kunio.

    1996-05-01

    A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author)

  12. Neutron cross section library production code system for continuous energy Monte Carlo code MVP. LICEM

    Energy Technology Data Exchange (ETDEWEB)

    Mori, Takamasa; Nakagawa, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio

    1996-05-01

    A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author).

  13. A neutron spectrum unfolding code based on iterative procedures

    International Nuclear Information System (INIS)

    Ortiz R, J. M.; Vega C, H. R.

    2012-10-01

    In this work, the version 3.0 of the neutron spectrum unfolding code called Neutron Spectrometry and Dosimetry from Universidad Autonoma de Zacatecas (NSDUAZ), is presented. This code was designed in a graphical interface under the LabVIEW programming environment and it is based on the iterative SPUNIT iterative algorithm, using as entrance data, only the rate counts obtained with 7 Bonner spheres based on a 6 Lil(Eu) neutron detector. The main features of the code are: it is intuitive and friendly to the user; it has a programming routine which automatically selects the initial guess spectrum by using a set of neutron spectra compiled by the International Atomic Energy Agency. Besides the neutron spectrum, this code calculates the total flux, the mean energy, H(10), h(10), 15 dosimetric quantities for radiation protection porpoises and 7 survey meter responses, in four energy grids, based on the International Atomic Energy Agency compilation. This code generates a full report in html format with all relevant information. In this work, the neutron spectrum of a 241 AmBe neutron source on air, located at 150 cm from detector, is unfolded. (Author)

  14. CONDOR: neutronic code for fuel elements calculation with rods

    International Nuclear Information System (INIS)

    Villarino, E.A.

    1990-01-01

    CONDOR neutronic code is used for the calculation of fuel elements formed by fuel rods. The method employed to obtain the neutronic flux is that of collision probabilities in a multigroup scheme on two-dimensional geometry. This code utilizes new calculation algorithms and normalization of such collision probabilities. Burn-up calculations can be made before the alternative of applying variational methods for response flux calculations or those corresponding to collision normalization. (Author) [es

  15. Unfolding code for neutron spectrometry based on neural nets technology

    International Nuclear Information System (INIS)

    Ortiz R, J. M.; Vega C, H. R.

    2012-10-01

    The most delicate part of neutron spectrometry, is the unfolding process. The derivation of the spectral information is not simple because the unknown is not given directly as a result of the measurements. The drawbacks associated with traditional unfolding procedures have motivated the need of complementary approaches. Novel methods based on Artificial Neural Networks have been widely investigated. In this work, a neutron spectrum unfolding code based on neural nets technology is presented. This unfolding code called Neutron Spectrometry and Dosimetry by means of Artificial Neural Networks was designed in a graphical interface under LabVIEW programming environment. The core of the code is an embedded neural network architecture, previously optimized by the R obust Design of Artificial Neural Networks Methodology . The main features of the code are: is easy to use, friendly and intuitive to the user. This code was designed for a Bonner Sphere System based on a 6 Lil(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. The main feature of the code is that as entrance data, only seven rate counts measurement with a Bonner spheres spectrometer are required for simultaneously unfold the 60 energy bins of the neutron spectrum and to calculate 15 dosimetric quantities, for radiation protection porpoises. This code generates a full report in html format with all relevant information. (Author)

  16. Unfolding code for neutron spectrometry based on neural nets technology

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz R, J. M.; Vega C, H. R., E-mail: morvymm@yahoo.com.mx [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica, Apdo. Postal 336, 98000 Zacatecas (Mexico)

    2012-10-15

    The most delicate part of neutron spectrometry, is the unfolding process. The derivation of the spectral information is not simple because the unknown is not given directly as a result of the measurements. The drawbacks associated with traditional unfolding procedures have motivated the need of complementary approaches. Novel methods based on Artificial Neural Networks have been widely investigated. In this work, a neutron spectrum unfolding code based on neural nets technology is presented. This unfolding code called Neutron Spectrometry and Dosimetry by means of Artificial Neural Networks was designed in a graphical interface under LabVIEW programming environment. The core of the code is an embedded neural network architecture, previously optimized by the {sup R}obust Design of Artificial Neural Networks Methodology{sup .} The main features of the code are: is easy to use, friendly and intuitive to the user. This code was designed for a Bonner Sphere System based on a {sup 6}Lil(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. The main feature of the code is that as entrance data, only seven rate counts measurement with a Bonner spheres spectrometer are required for simultaneously unfold the 60 energy bins of the neutron spectrum and to calculate 15 dosimetric quantities, for radiation protection porpoises. This code generates a full report in html format with all relevant information. (Author)

  17. 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

  18. Calibration of neutron yield activation measurements at JET using MCNP and furnace neutron transport codes

    International Nuclear Information System (INIS)

    Pillon, M.; Martone, M.; Verschuur, K.A.; Jarvis, O.N.; Kaellne, J.

    1989-01-01

    Neutron transport calculations have been performed using fluence ray tracing (FURNACE code) and Monte Carlo particle trajectory sampling methods (MCNP code) in order to determine the neutron fluence and energy distributions at different locations in the JET tokamak. These calculations were used to calibrate the activation measurements used in the determination of the absolute fusion neutron yields from the JET plasma. We present here the neutron activation response coefficients calculated for three different materials. Comparison of the MCNP and FURNACE results helps identify the sources of error in these neutron transport calculations. The accuracy of these calculations was tested by comparing the total 2.5 MeV neutron yields derived from the activation measurements with those obtained with calibrated fission chambers; agreement at the ±15% level was demonstrate. (orig.)

  19. An Auto sequence Code to Integrate a Neutron Unfolding Code with thePC-MCA Accuspec

    International Nuclear Information System (INIS)

    Darsono

    2000-01-01

    In a neutron spectrometry using proton recoil method, the neutronunfolding code is needed to unfold the measured proton spectrum to become theneutron spectrum. The process of the unfolding neutron in the existingneutron spectrometry which was successfully installed last year was doneseparately. This manuscript reports that the auto sequence code to integratethe neutron unfolding code UNFSPEC.EXE with the software facility of thePC-MCA Accuspec has been made and run successfully so that the new neutronspectrometry become compact. The auto sequence code was written based on therules in application program facility of PC-MCA Accuspec and then it wascompiled using AC-EXE. Result of the test of the auto sequence code showedthat for binning width 20, 30, and 40 giving a little different spectrumshape. The binning width around 30 gives a better spectrum in mean of givingsmall error compared to the others. (author)

  20. Calculation of neutron spectra produced in neutron generator target: Code testing.

    Science.gov (United States)

    Gaganov, V V

    2018-03-01

    DT-neutron spectra calculated using the SRIANG code was benchmarked against the results obtained by widely used Monte Carlo codes: PROFIL, SHORIN, TARGET, ENEA-JSI, MCUNED, DDT and NEUSDESC. The comparison of the spectra obtained by different codes confirmed the correctness of SRIANG calculations. The cross-checking of the compared spectra revealed some systematic features and possible errors of analysed codes. Copyright © 2017 Elsevier Ltd. All rights reserved.

  1. Using MCNP code for neutron and photon skyshine analysis

    Energy Technology Data Exchange (ETDEWEB)

    Zharkov, V.P.; Dikareva, O.F.; Kartashev, I.A.; Kiselev, A.N.; Netecha, M.E. [Research and Development Inst. of Power Engineering, Moscow (Russian Federation); Nomura, Y.; Tsubosaka, A. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

    2000-03-01

    The MCNP Monte-Carlo code was used for the investigation of the sensitivity of neutron and neutron-induced secondary photon dose rate, total and thermal neutron fluxes and space-energy distributions to energy and angular distribution of radiation source, to thickness and composition of the ground, air density (including it changing with height), humidities of air and ground, thermalization effects, detector's dimension and its disposal above the ground level. The calculations were performed with the assumption that the source or released radiation into the atmosphere can be treated as a point source and the source containment structure has a negligible perturbation on the skyshine radiation field. (author)

  2. CALTRANS: A parallel, deterministic, 3D neutronics code

    Energy Technology Data Exchange (ETDEWEB)

    Carson, L.; Ferguson, J.; Rogers, J.

    1994-04-01

    Our efforts to parallelize the deterministic solution of the neutron transport equation has culminated in a new neutronics code CALTRANS, which has full 3D capability. In this article, we describe the layout and algorithms of CALTRANS and present performance measurements of the code on a variety of platforms. Explicit implementation of the parallel algorithms of CALTRANS using both the function calls of the Parallel Virtual Machine software package (PVM 3.2) and the Meiko CS-2 tagged message passing library (based on the Intel NX/2 interface) are provided in appendices.

  3. Neutron spallation source and the Dubna cascade code

    CERN Document Server

    Kumar, V; Goel, U; Barashenkov, V S

    2003-01-01

    Neutron multiplicity per incident proton, n/p, in collision of high energy proton beam with voluminous Pb and W targets has been estimated from the Dubna cascade code and compared with the available experimental data for the purpose of benchmarking of the code. Contributions of various atomic and nuclear processes for heat production and isotopic yield of secondary nuclei are also estimated to assess the heat and radioactivity conditions of the targets. Results obtained from the code show excellent agreement with the experimental data at beam energy, E < 1.2 GeV and differ maximum up to 25% at higher energy. (author)

  4. 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)

  5. A guide to the AUS modular neutronics code system

    International Nuclear Information System (INIS)

    Robinson, G.S.

    1987-04-01

    A general description is given of the AUS modular neutronics code system, which may be used for calculations of a very wide range of fission reactors, fusion blankets and other neutron applications. The present system has cross-section libraries derived from ENDF/B-IV and includes modules which provide for lattice calculations, one-dimensional transport calculations, and one, two, and three-dimensional diffusion calculations, burnup calculations and the flexible editing of results. Details of all system aspects of AUS are provided but the major individual modules are only outlined. Sufficient information is given to enable other modules to be added to the system

  6. Landscape Analysis of Adult Florida Panther Habitat.

    Directory of Open Access Journals (Sweden)

    Robert A Frakes

    Full Text Available Historically occurring throughout the southeastern United States, the Florida panther is now restricted to less than 5% of its historic range in one breeding population located in southern Florida. Using radio-telemetry data from 87 prime-aged (≥3 years old adult panthers (35 males and 52 females during the period 2004 through 2013 (28,720 radio-locations, we analyzed the characteristics of the occupied area and used those attributes in a random forest model to develop a predictive distribution map for resident breeding panthers in southern Florida. Using 10-fold cross validation, the model was 87.5 % accurate in predicting presence or absence of panthers in the 16,678 km2 study area. Analysis of variable importance indicated that the amount of forests and forest edge, hydrology, and human population density were the most important factors determining presence or absence of panthers. Sensitivity analysis showed that the presence of human populations, roads, and agriculture (other than pasture had strong negative effects on the probability of panther presence. Forest cover and forest edge had strong positive effects. The median model-predicted probability of presence for panther home ranges was 0.81 (0.82 for females and 0.74 for males. The model identified 5579 km2 of suitable breeding habitat remaining in southern Florida; 1399 km2 (25% of this habitat is in non-protected private ownership. Because there is less panther habitat remaining than previously thought, we recommend that all remaining breeding habitat in south Florida should be maintained, and the current panther range should be expanded into south-central Florida. This model should be useful for evaluating the impacts of future development projects, in prioritizing areas for panther conservation, and in evaluating the potential impacts of sea-level rise and changes in hydrology.

  7. Landscape Analysis of Adult Florida Panther Habitat.

    Science.gov (United States)

    Frakes, Robert A; Belden, Robert C; Wood, Barry E; James, Frederick E

    2015-01-01

    Historically occurring throughout the southeastern United States, the Florida panther is now restricted to less than 5% of its historic range in one breeding population located in southern Florida. Using radio-telemetry data from 87 prime-aged (≥3 years old) adult panthers (35 males and 52 females) during the period 2004 through 2013 (28,720 radio-locations), we analyzed the characteristics of the occupied area and used those attributes in a random forest model to develop a predictive distribution map for resident breeding panthers in southern Florida. Using 10-fold cross validation, the model was 87.5 % accurate in predicting presence or absence of panthers in the 16,678 km2 study area. Analysis of variable importance indicated that the amount of forests and forest edge, hydrology, and human population density were the most important factors determining presence or absence of panthers. Sensitivity analysis showed that the presence of human populations, roads, and agriculture (other than pasture) had strong negative effects on the probability of panther presence. Forest cover and forest edge had strong positive effects. The median model-predicted probability of presence for panther home ranges was 0.81 (0.82 for females and 0.74 for males). The model identified 5579 km2 of suitable breeding habitat remaining in southern Florida; 1399 km2 (25%) of this habitat is in non-protected private ownership. Because there is less panther habitat remaining than previously thought, we recommend that all remaining breeding habitat in south Florida should be maintained, and the current panther range should be expanded into south-central Florida. This model should be useful for evaluating the impacts of future development projects, in prioritizing areas for panther conservation, and in evaluating the potential impacts of sea-level rise and changes in hydrology.

  8. 3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

    Energy Technology Data Exchange (ETDEWEB)

    Langenbuch, S.; Velkov, K. [GRS, Garching (Germany); Lizorkin, M. [Kurchatov-Institute, Moscow (Russian Federation)] [and others

    1997-07-01

    This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

  9. PANTHER solution to the NEA-NSC 3-D PWR core transient benchmark. Uncontrolled withdrawal of control rods at zero power

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C.

    1994-10-01

    This report contains the results of PANTHER calculations for the ``NEA-NSC 3-D PWR Core Transient Benchmark: Uncontrolled Withdrawal of Control Rods at Zero Power``. PANTHER was able to model the benchmark problems without modifications to the code. All the calculations were performed in 3-D. (orig.).

  10. Final PANTHER solution to the NEA-NSC3-DPWR core transient benchmark. Uncontrolled withdrawal of control rods at zero power

    International Nuclear Information System (INIS)

    Kuijper, J.C.

    1996-10-01

    This report contains the final results of PANTHER calculations for the 'NEA-NSC 3-D PWR Core Transient Benchmark: Uncontrolled Withdrawal of Control Rods at Zero Power'. PANTHER was able to model the benchmark problems without modifications to the code. All the calculations were performed in 3-D. (orig.)

  11. PANTHER solution to the NEA-NSC 3-D PWR core transient benchmark. Uncontrolled withdrawal of control rods at zero power

    International Nuclear Information System (INIS)

    Kuijper, J.C.

    1994-10-01

    This report contains the results of PANTHER calculations for the ''NEA-NSC 3-D PWR Core Transient Benchmark: Uncontrolled Withdrawal of Control Rods at Zero Power''. PANTHER was able to model the benchmark problems without modifications to the code. All the calculations were performed in 3-D. (orig.)

  12. Final PANTHER solution to the NEA-NSC3-DPWR core transient benchmark. Uncontrolled withdrawal of control rods at zero power

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C.

    1996-10-01

    This report contains the final results of PANTHER calculations for the `NEA-NSC 3-D PWR Core Transient Benchmark: Uncontrolled Withdrawal of Control Rods at Zero Power`. PANTHER was able to model the benchmark problems without modifications to the code. All the calculations were performed in 3-D. (orig.).

  13. Benchmark of neutron production cross sections with Monte Carlo codes

    Science.gov (United States)

    Tsai, Pi-En; Lai, Bo-Lun; Heilbronn, Lawrence H.; Sheu, Rong-Jiun

    2018-02-01

    Aiming to provide critical information in the fields of heavy ion therapy, radiation shielding in space, and facility design for heavy-ion research accelerators, the physics models in three Monte Carlo simulation codes - PHITS, FLUKA, and MCNP6, were systematically benchmarked with comparisons to fifteen sets of experimental data for neutron production cross sections, which include various combinations of 12C, 20Ne, 40Ar, 84Kr and 132Xe projectiles and natLi, natC, natAl, natCu, and natPb target nuclides at incident energies between 135 MeV/nucleon and 600 MeV/nucleon. For neutron energies above 60% of the specific projectile energy per nucleon, the LAQGMS03.03 in MCNP6, the JQMD/JQMD-2.0 in PHITS, and the RQMD-2.4 in FLUKA all show a better agreement with data in heavy-projectile systems than with light-projectile systems, suggesting that the collective properties of projectile nuclei and nucleon interactions in the nucleus should be considered for light projectiles. For intermediate-energy neutrons whose energies are below the 60% projectile energy per nucleon and above 20 MeV, FLUKA is likely to overestimate the secondary neutron production, while MCNP6 tends towards underestimation. PHITS with JQMD shows a mild tendency for underestimation, but the JQMD-2.0 model with a modified physics description for central collisions generally improves the agreement between data and calculations. For low-energy neutrons (below 20 MeV), which are dominated by the evaporation mechanism, PHITS (which uses GEM linked with JQMD and JQMD-2.0) and FLUKA both tend to overestimate the production cross section, whereas MCNP6 tends to underestimate more systems than to overestimate. For total neutron production cross sections, the trends of the benchmark results over the entire energy range are similar to the trends seen in the dominate energy region. Also, the comparison of GEM coupled with either JQMD or JQMD-2.0 in the PHITS code indicates that the model used to describe the first

  14. General features of the neutronics design code EQUICYCLE

    International Nuclear Information System (INIS)

    Jirlow, K.

    1978-10-01

    The neutronics code EQUICYCLE has been developed and improved over a long period of time. It is expecially adapted to survey type design calculations of large fast power reactors with particular emphasis on the nuclear parameters for a realistic equilibrium fuel cycle. Thus the code is used to evaluate the breeding performance, the power distributions and the uranium and plutonium mass balance for realistic refuelling schemes. In addition reactivity coefficients can be calculated and the influence of burnup could be assessed. The code is two-dimensional and treats the reactor core in R-Z geometry. The basic ideas of the calculating scheme are successive iterative improvement of cross-section sets and flux spectra and use of the mid-cycle flux for burning the fuel according to a specified refuelling scheme. Normally given peak burn-ups and maximum power densities are used as boundary conditions. The code is capable of handling the unconventional, so called heterogeneous cores. (author)

  15. Alignment effects on a neutron imaging system using coded apertures

    International Nuclear Information System (INIS)

    Thfoin, Isabelle; Landoas, Olivier; Caillaud, Tony; Vincent, Maxime; Bourgade, Jean-Luc; Rosse, Bertrand; Disdier, Laurent; Sangster, Thomas C.; Glebov, Vladimir Yu.; Pien, Greg; Armstrong, William

    2010-01-01

    A high resolution neutron imaging system is being developed and tested on the OMEGA laser facility for inertial confinement fusion experiments. This diagnostic uses a coded imaging technique with a penumbral or an annular aperture. The sensitiveness of these techniques to misalignment was pointed out with both experiments and simulations. Results obtained during OMEGA shots are in good agreement with calculations performed with the Monte Carlo code GEANT4. Both techniques are sensitive to the relative position of the source in the field of view. The penumbral imaging technique then demonstrates to be less sensitive to misalignment compared to the ring. These results show the necessity to develop a neutron imaging diagnostic for megajoule class lasers taking into account our alignment capabilities on such facilities.

  16. Thermal neutron self-shielding correction factors for large sample instrumental neutron activation analysis using the MCNP code

    International Nuclear Information System (INIS)

    Tzika, F.; Stamatelatos, I.E.

    2004-01-01

    Thermal neutron self-shielding within large samples was studied using the Monte Carlo neutron transport code MCNP. The code enabled a three-dimensional modeling of the actual source and geometry configuration including reactor core, graphite pile and sample. Neutron flux self-shielding correction factors derived for a set of materials of interest for large sample neutron activation analysis are presented and evaluated. Simulations were experimentally verified by measurements performed using activation foils. The results of this study can be applied in order to determine neutron self-shielding factors of unknown samples from the thermal neutron fluxes measured at the surface of the sample

  17. Vectorization of three-dimensional neutron diffusion code CITATION

    International Nuclear Information System (INIS)

    Harada, Hiroo; Ishiguro, Misako

    1985-01-01

    Three-dimensional multi-group neutron diffusion code CITATION has been widely used for reactor criticality calculations. The code is expected to be run at a high speed by using recent vector supercomputers, when it is appropriately vectorized. In this paper, vectorization methods and their effects are described for the CITATION code. Especially, calculation algorithms suited for vectorization of the inner-outer iterative calculations which spend most of the computing time are discussed. The SLOR method, which is used in the original CITATION code, and the SOR method, which is adopted in the revised code, are vectorized by odd-even mesh ordering. The vectorized CITATION code is executed on the FACOM VP-100 and VP-200 computers, and is found to run over six times faster than the original code for a practical-scale problem. The initial value of the relaxation factor and the number of inner-iterations given as input data are also investigated since the computing time depends on these values. (author)

  18. Testing of a Code for the Calculation of Spectra of Neutrons Produced in a Target of a Neutron Generator

    Science.gov (United States)

    Gaganov, V. V.

    2017-12-01

    The correctness of calculations performed with the SRIANG code for modeling the spectra of DT neutrons is estimated by comparing the obtained spectra to the results of calculations carried out with five different codes based on the Monte Carlo method.

  19. SNAP-3D: a three-dimensional neutron diffusion code

    International Nuclear Information System (INIS)

    McCallien, C.W.J.

    1975-10-01

    A preliminary report is presented describing the data requirements of a one- two- or three-dimensional multi-group diffusion code, SNAP-3D. This code is primarily intended for neutron diffusion calculations but it can also carry out gamma calculations if the diffuse approximation is accurate enough. It is suitable for fast and thermal reactor core calculations and for shield calculations. It is assumed the reader is familiar with the older, two-dimensional code SNAP and can refer to the report [TRG-Report-1990], describing it. The present report concentrates on the enhancements to SNAP that have been made to produce the three-dimensional version, SNAP-3D, and is intended to act a a guide on data preparation until a single, comprehensive report can be published. (author)

  20. CONIFERS: a neutronics code for reactors with channels

    International Nuclear Information System (INIS)

    Davis, R.S.

    1977-04-01

    CONIFERS is a neutronics code for nuclear reactors whose fuel is in channels that are separated from each other by several neutron mean-free-path lengths of moderator. It can treat accurately situations in which the usual homogenized-cell diffusion equation becomes inaccurate, but is more economical than other advanced methods such as response-matrix and source-sink formalisms. CONIFERS uses exact solutions of the neutron diffusion equation within each cell. It allows for the breakdown of this equation near a channel by means of data that almost any cell code can supply. It uses the results of these cell analyses in a reactor equations set that is as readily solvable as the familiar finite-difference equations set. CONIFERS can model almost any configuration of channels and other structures in two or three dimensions. It can use any number of energy groups and any reactivity scales, including scales based on control operations. It is also flexible from a programming point of view, and has convenient input and output provisions. (author)

  1. The MCUCN simulation code for ultracold neutron physics

    Science.gov (United States)

    Zsigmond, G.

    2018-02-01

    Ultracold neutrons (UCN) have very low kinetic energies 0-300 neV, thereby can be stored in specific material or magnetic confinements for many hundreds of seconds. This makes them a very useful tool in probing fundamental symmetries of nature (for instance charge-parity violation by neutron electric dipole moment experiments) and contributing important parameters for the Big Bang nucleosynthesis (neutron lifetime measurements). Improved precision experiments are in construction at new and planned UCN sources around the world. MC simulations play an important role in the optimization of such systems with a large number of parameters, but also in the estimation of systematic effects, in benchmarking of analysis codes, or as part of the analysis. The MCUCN code written at PSI has been extensively used for the optimization of the UCN source optics and in the optimization and analysis of (test) experiments within the nEDM project based at PSI. In this paper we present the main features of MCUCN and interesting benchmark and application examples.

  2. KAMCCO, a reactor physics Monte Carlo neutron transport code

    International Nuclear Information System (INIS)

    Arnecke, G.; Borgwaldt, H.; Brandl, V.; Lalovic, M.

    1976-06-01

    KAMCCO is a 3-dimensional reactor Monte Carlo code for fast neutron physics problems. Two options are available for the solution of 1) the inhomogeneous time-dependent neutron transport equation (census time scheme), and 2) the homogeneous static neutron transport equation (generation cycle scheme). The user defines the desired output, e.g. estimates of reaction rates or neutron flux integrated over specified volumes in phase space and time intervals. Such primary quantities can be arbitrarily combined, also ratios of these quantities can be estimated with their errors. The Monte Carlo techniques are mostly analogue (exceptions: Importance sampling for collision processes, ELP/MELP, Russian roulette and splitting). Estimates are obtained from the collision and track length estimators. Elastic scattering takes into account first order anisotropy in the center of mass system. Inelastic scattering is processed via the evaporation model or via the excitation of discrete levels. For the calculation of cross sections, the energy is treated as a continuous variable. They are computed by a) linear interpolation, b) from optionally Doppler broadened single level Breit-Wigner resonances or c) from probability tables (in the region of statistically distributed resonances). (orig.) [de

  3. Validation of SCALE code package on high performance neutron shields

    International Nuclear Information System (INIS)

    Bace, M.; Jecmenica, R.; Smuc, T.

    1999-01-01

    The shielding ability and other properties of new high performance neutron shielding materials from the KRAFTON series have been recently published. A comparison of the published experimental and MCNP results for the two materials of the KRAFTON series, with our own calculations has been done. Two control modules of the SCALE-4.4 code system have been used, one of them based on one dimensional radiation transport analysis (SAS1) and other based on the three dimensional Monte Carlo method (SAS3). The comparison of the calculated neutron dose equivalent rates shows a good agreement between experimental and calculated results for the KRAFTON-N2 material.. Our results indicate that the N2-M-N2 sandwich type is approximately 10% inferior as neutron shield to the KRAFTON-N2 material. All values of neutron dose equivalent obtained by SAS1 are approximately 25% lower in comparison with the SAS3 results, which indicates proportions of discrepancies introduced by one-dimensional geometry approximation.(author)

  4. Identifying suitable sites for Florida panther reintroduction

    Science.gov (United States)

    Thatcher, Cindy A.; van Manen, Frank T.; Clark, Joseph D.

    2006-01-01

    A major objective of the 1995 Florida Panther (Puma concolor cory) Recovery Plan is the establishment of 2 additional panther populations within the historic range. Our goal was to identify prospective sites for Florida panther reintroduction within the historic range based on quantitative landscape assessments. First, we delineated 86 panther home ranges using telemetry data collected from 1981 to 2001 in south Florida to develop a Mahalanobis distance (D2) habitat model, using 4 anthropogenic variables and 3 landscape variables mapped at a 500-m resolution. From that analysis, we identified 9 potential reintroduction sites of sufficient size to support a panther population. We then developed a similar D2 model at a higher spatial resolution to quantify the area of favorable panther habitat at each site. To address potential for the population to expand, we calculated the amount of favorable habitat adjacent to each prospective reintroduction site within a range of dispersal distances of female panthers. We then added those totals to the contiguous patches to estimate the total amount of effective panther habitat at each site. Finally, we developed an expert-assisted model to rank and incorporate potentially important habitat variables that were not appropriate for our empirical analysis (e.g., area of public lands, livestock density). Anthropogenic factors heavily influenced both the landscape and the expert-assisted models. Of the 9 areas we identified, the Okefenokee National Wildlife Refuge, Ozark National Forest, and Felsenthal National Wildlife Refuge regions had the highest combination of effective habitat area and expert opinion scores. Sensitivity analyses indicated that variability among key model parameters did not affect the high ranking of those sites. Those sites should be considered as starting points for the field evaluation of potential reintroduction sites.

  5. Random mask optimization for fast neutron coded aperture imaging

    Energy Technology Data Exchange (ETDEWEB)

    McMillan, Kyle [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Univ. of California, Los Angeles, CA (United States); Marleau, Peter [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Brubaker, Erik [Sandia National Lab. (SNL-CA), Livermore, CA (United States)

    2015-05-01

    In coded aperture imaging, one of the most important factors determining the quality of reconstructed images is the choice of mask/aperture pattern. In many applications, uniformly redundant arrays (URAs) are widely accepted as the optimal mask pattern. Under ideal conditions, thin and highly opaque masks, URA patterns are mathematically constructed to provide artifact-free reconstruction however, the number of URAs for a chosen number of mask elements is limited and when highly penetrating particles such as fast neutrons and high-energy gamma-rays are being imaged, the optimum is seldom achieved. In this case more robust mask patterns that provide better reconstructed image quality may exist. Through the use of heuristic optimization methods and maximum likelihood expectation maximization (MLEM) image reconstruction, we show that for both point and extended neutron sources a random mask pattern can be optimized to provide better image quality than that of a URA.

  6. MADNIX a code to calculate prompt fission neutron spectra and average prompt neutron multiplicities

    International Nuclear Information System (INIS)

    Merchant, A.C.

    1986-03-01

    A code has been written and tested on the CDC Cyber-170 to calculate the prompt fission neutron spectrum, N(E), as a function of both the fissioning nucleus and its excitation energy. In this note a brief description of the underlying physical principles involved and a detailed explanation of the required input data (together with a sample output for the fission of 235 U induced by 14 MeV neutrons) are presented. Weisskopf's standard nuclear evaporation theory provides the basis for the calculation. Two important refinements are that the distribution of fission-fragment residual nuclear temperature and the cooling of the fragments as neutrons are emitted approximately taken into account, and also the energy dependence of the cross section for the inverse process of compound nucleus formation is included. This approach is then used to calculate the average number of prompt neutrons emitted per fission, v-bar p . At high excitation energies, where fission is still possible after neutron emission, the consequences of the competition between first, second and third chance fission on N(E) and v-bar p are calculated. Excellent agreement with all the examples given in the original work of Madland and Nix is obtained. (author) [pt

  7. Modifications to the Monte Carlo neutronics code MONK

    International Nuclear Information System (INIS)

    Hutton, J.L.

    1979-09-01

    The Monte Carlo neutronics code MONK has been widely used for criticality calculations, and is one of the standard methods for assessing the safety of transport flasks and fuel storage facilities in the UK. Recently, attempts have been made to extend the range of applications of this calculational technique. In particular studies have been carried out using Monte Carlo to analyse reactor physics experiments. In these applications various shortcomings of the standard version MONK5 became apparent. The basic data library was found to be inadequate and additional estimates of parameters (eg power distribution) not normally included in criticality studies were required. These features which required improvement, primarily in the context of using the code for reactor physics calculations, are enumerated. To facilitate the use of the code as a reactor physics calculational tool a series of modifications have been carried out. The code has been modified so that the user can use group data tabulations of the cross sections instead of the present 'point' data values. The code can now interface with a number of reactor physics group data preparation schemes but in particular it can use WIMS-E interfaces as a source of group data. Details of the changes are outlined and a new version of MONK incorporating these modifications has been created. This version is called MONK5W. This paper provides a guide to the use of this version. The data input is described along with other details required to use this code on the Harwell IBM 3033. To aid the user, examples of calculations using the new facilities incorporated in MONK5W are given. (UK)

  8. Improvement of neutron collimator design for thermal neutron radiography using Monte Carlo N-particle transport code version 5

    International Nuclear Information System (INIS)

    Thiagu Supramaniam

    2007-01-01

    The aim of this research was to propose a new neutron collimator design for thermal neutron radiography facility using tangential beam port of PUSPATI TRIGA Mark II reactor, Malaysia Institute of Nuclear Technology Research (MINT). Best geometry and materials for neutron collimator were chosen in order to obtain a uniform beam with maximum thermal neutron flux, high L/ D ratio, high neutron to gamma ratio and low beam divergence with high resolution. Monte Carlo N-particle Transport Code version 5 (MCNP 5) was used to optimize six neutron collimator components such as beam port medium, neutron scatterer, neutron moderator, gamma filter, aperture and collimator wall. The reactor and tangential beam port setup in MCNP5 was plotted according to its actual sizes. A homogeneous reactor core was assumed and population control method of variance reduction technique was applied by using cell importance. The comparison between experimental results and simulated results of the thermal neutron flux measurement of the bare tangential beam port, shows that both graph obtained had similar pattern. This directly suggests the reliability of MCNP5 in order to obtained optimal neutron collimator parameters. The simulated results of the optimal neutron medium, shows that vacuum was the best medium to transport neutrons followed by helium gas and air. The optimized aperture component was boral with 3 cm thickness. The optimal aperture center hole diameter was 2 cm which produces 88 L/ D ratio. Simulation also shows that graphite neutron scatterer improves thermal neutron flux while reducing fast neutron flux. Neutron moderator was used to moderate fast and epithermal neutrons in the beam port. Paraffin wax with 90 cm thick was bound to be the best neutron moderator material which produces the highest thermal neutron flux at the image plane. Cylindrical shape high density polyethylene neutron collimator produces the highest thermal neutron flux at the image plane rather than divergent

  9. HTR core physics and transient analyses by the Panthermix code system

    Energy Technology Data Exchange (ETDEWEB)

    Haas, J.B.M. de; Kuijper, J.C.; Oppe, J. [NRG - Fuels, Actinides and Isotopes group, Petten (Netherlands)

    2005-07-01

    At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes.

  10. HTR core physics and transient analyses by the Panthermix code system

    International Nuclear Information System (INIS)

    Haas, J.B.M. de; Kuijper, J.C.; Oppe, J.

    2005-01-01

    At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes

  11. Light-water-reactor coupled neutronic and thermal-hydraulic codes

    International Nuclear Information System (INIS)

    Diamond, D.J.

    1982-01-01

    An overview is presented of computer codes that model light water reactor cores with coupled neutronics and thermal-hydraulics. This includes codes for transient analysis and codes for steady state analysis which include fuel depletion and fission product buildup. Applications in nuclear design, reactor operations and safety analysis are given and the major codes in use in the USA are identified. The neutronic and thermal-hydraulic methodologies and other code features are outlined for three steady state codes (PDQ7, NODE-P/B and SIMULATE) and four dynamic codes (BNL-TWIGL, MEKIN, RAMONA-3B, RETRAN-02). Speculation as to future trends with such codes is also presented

  12. Development of a neutronic analysis code using data from Monju

    International Nuclear Information System (INIS)

    Rooijen, W.F.G. van; Yamano, N.; Shimazu, Y.

    2015-01-01

    In recent years three major sets of modern evaluated nuclear data have become available: JENDL-4.0, JEFF-3.1.2 and ENDF/B-VII.1. The authors were involved with a research project to establish analysis method for a future commercial-scale LMFBR. This project focused on JENDL-4.0 and conventional Japanese codes. As a cross check, we decided to also apply the fast reactor code ERANOS. This necessitated to produce nuclear data (cross sections, etc) for the ERANOS code system, discussed in this paper. We developed a nuclear data processing system to produce cross sections, probability tables, delayed neutron data, and covariance data from the evaluated nuclear data files for ERANOS. A benchmark calculation on the MZA/MZB benchmark showed very satisfying results. Subsequently, we analyzed the prototype LMFBR Monju with ERANOS and our own sets of nuclear data. The results are very satisfactory. The results from ERANOS indicate that the target accuracies for nuclear data have not been met, although the three sets of evaluated nuclear data all performed very well in our analysis. In the future, the covariance on nuclear data should be reduced to meet the target accuracies on criticality and feedback coefficients. (author)

  13. Virginia Tech Wildlife Professor Helping To Save Florida Panther

    OpenAIRE

    Davis, Lynn

    2003-01-01

    With few Florida panthers now in existence, Mike Vaughan, Virginia Tech professor of wildlife and sciences in the College of Natural Resources, has been appointed to serve on the Florida Panther Scientific Review Team (SRT). Vaughan and other SRT members have made several trips to Naples, Fla., to interview state and federal biologists directly involved with the recovery of the Florida panther.

  14. Four energy group neutron flux distribution in the Syrian miniature neutron source reactor using the WIMSD4 and CITATION code

    International Nuclear Information System (INIS)

    Khattab, K.; Omar, H.; Ghazi, N.

    2009-01-01

    A 3-D (R, θ , Z) neutronic model for the Miniature Neutron Source Reactor (MNSR) was developed earlier to conduct the reactor neutronic analysis. The group constants for all the reactor components were generated using the WIMSD4 code. The reactor excess reactivity and the four group neutron flux distributions were calculated using the CITATION code. This model is used in this paper to calculate the point wise four energy group neutron flux distributions in the MNSR versus the radius, angle and reactor axial directions. Good agreement is noticed between the measured and the calculated thermal neutron flux in the inner and the outer irradiation site with relative difference less than 7% and 5% respectively. (author)

  15. Accuracy estimation for intermediate and low energy neutron transport calculation with Monte Carlo code MCNP

    International Nuclear Information System (INIS)

    Kotegawa, Hiroshi; Sasamoto, Nobuo; Tanaka, Shun-ichi

    1987-02-01

    Both ''measured radioactive inventory due to neutron activation in the shield concrete of JPDR'' and ''measured intermediate and low energy neutron spectra penetrating through a graphite sphere'' are analyzed using a continuous energy model Monte Carlo code MCNP so as to estimate calculational accuracy of the code for neutron transport in thermal and epithermal energy regions. Analyses reveal that MCNP calculates thermal neutron spectra fairly accurately, while it apparently over-estimates epithermal neutron spectra (of approximate 1/E distribution) as compared with the measurements. (author)

  16. Coupled RELAP5/PANTHER/COBRA steam line break accident analysis in support of licensing DOEL 2 power uprate and steam generator replacement

    International Nuclear Information System (INIS)

    Zhang, J.; Bosso, S.; Henno, X.; Ouliddren, K.; Schneidesch, C.R.; Hove, W. van

    2004-01-01

    The nuclear reactor accident analyses using best estimate codes provide a better understanding and more accurate modeling of the key physical phenomena, which allows a more realistic evaluation of the conservatism and margins in the final safety analysis report (FSAR) accident analysis. The use of the best estimate codes and methods is necessary to meet the increasing technical, licensing and regulatory requirements for major plant changes (e.g. steam generator replacement), power uprate, core design optimization (cycle extension), as well as Periodic Safety Review. Since 1992, Tractebel Engineering (TE) has developed and applied a deterministic bounding approach to FASR accident analysis using the best estimate system thermal hydraulic code RELAP5/MOD2.5 and the subchannel thermal hydraulic code COBRA-3C. This approach has been accepted by the Belgian Safety Authorities, and turned out to be cost effective for most of the non-LOCA transient analyses. Since this approach adapts a decoupled modeling of the core responses, the analysis results often involved too large un-quantified conservatisms, due to either simplistic approximations for asymmetric accidents with strong 3D core neutronics - plant thermal hydraulics interactions, or additional penalties introduced from 'incoherent' initial/boundary conditions for separate plant and core analyses. Therefore, an external dynamic coupling between the RELAP5/MOD2.5 code and the 3-D neutronic code PANTHER was implemented since 1997 via the transient analysis code linkage program TALINK. Furthermore, a static linkage between the PANTHER code and the COBRA-3C code was developed for on-line calculation of (Departure from Nucleate Boiling Ratio (DNBR). TE intends to use the coupled code package for licensing non-symmetric FSAR accident analysis. The TE coupled code package has been applied to develop a main steam line break (MSLB) accident analysis methodology [using the TE deterministic bounding approach. The methodology

  17. A novel neutron energy spectrum unfolding code using particle swarm optimization

    International Nuclear Information System (INIS)

    Shahabinejad, H.; Sohrabpour, M.

    2017-01-01

    A novel neutron Spectrum Deconvolution using Particle Swarm Optimization (SDPSO) code has been developed to unfold the neutron spectrum from a pulse height distribution and a response matrix. The Particle Swarm Optimization (PSO) imitates the bird flocks social behavior to solve complex optimization problems. The results of the SDPSO code have been compared with those of the standard spectra and recently published Two-steps Genetic Algorithm Spectrum Unfolding (TGASU) code. The TGASU code have been previously compared with the other codes such as MAXED, GRAVEL, FERDOR and GAMCD and shown to be more accurate than the previous codes. The results of the SDPSO code have been demonstrated to match well with those of the TGASU code for both under determined and over-determined problems. In addition the SDPSO has been shown to be nearly two times faster than the TGASU code. - Highlights: • Introducing a novel method for neutron spectrum unfolding. • Implementation of a particle swarm optimization code for neutron unfolding. • Comparing results of the PSO code with those of recently published TGASU code. • Match results of the PSO code with those of TGASU code. • Greater convergence rate of implemented PSO code than TGASU code.

  18. Generation of neutron cross sections library for the Thermos code of the Fuel management System (FMS)

    International Nuclear Information System (INIS)

    Alonso V, G.; Viais J, J.

    1990-10-01

    There is developed a method to generate the library of neutron cross sections for the Thermos code by means of the database ENDF-B/IV and the NJOY code. The obtained results are compared with the version previous of the library of neutron cross sections which was processed using the version ENDF-B/III. (Author)

  19. Investigation of coupling scheme for neutronic and thermal-hydraulic codes

    International Nuclear Information System (INIS)

    Wang Guoli; Yu Jianfeng; Pen Muzhang; Zhang Yuman.

    1988-01-01

    Recently, a number of coupled neutronics/thermal-hydraulics codes have been used in reaction design and safty analysis, which have been obtained by coupling previous neutronic and thermal-hydraulic codes. The different coupling schemes affect computer time and accuracy of calculation results. Numberical experiments of several different coupling schemes and some heuristic results are described

  20. Development Of The Computer Code For Comparative Neutron Activation Analysis

    International Nuclear Information System (INIS)

    Purwadi, Mohammad Dhandhang

    2001-01-01

    The qualitative and quantitative chemical analysis with Neutron Activation Analysis (NAA) is an importance utilization of a nuclear research reactor, and this should be accelerated and promoted in application and its development to raise the utilization of the reactor. The application of Comparative NAA technique in GA Siwabessy Multi Purpose Reactor (RSG-GAS) needs special (not commercially available yet) soft wares for analyzing the spectrum of multiple elements in the analysis at once. The application carried out using a single spectrum software analyzer, and comparing each result manually. This method really degrades the quality of the analysis significantly. To solve the problem, a computer code was designed and developed for comparative NAA. Spectrum analysis in the code is carried out using a non-linear fitting method. Before the spectrum analyzed, it was passed to the numerical filter which improves the signal to noise ratio to do the deconvolution operation. The software was developed using the G language and named as PASAN-K The testing result of the developed software was benchmark with the IAEA spectrum and well operated with less than 10 % deviation

  1. Applications guide to the RSIC-distributed version of the MCNP code (coupled Monte Carlo neutron-photon Code)

    International Nuclear Information System (INIS)

    Cramer, S.N.

    1985-09-01

    An overview of the RSIC-distributed version of the MCNP code (a soupled Monte Carlo neutron-photon code) is presented. All general features of the code, from machine hardware requirements to theoretical details, are discussed. The current nuclide cross-section and other libraries available in the standard code package are specified, and a realistic example of the flexible geometry input is given. Standard and nonstandard source, estimator, and variance-reduction procedures are outlined. Examples of correct usage and possible misuse of certain code features are presented graphically and in standard output listings. Finally, itemized summaries of sample problems, various MCNP code documentation, and future work are given

  2. A computer code for calculating neutron cross-sections from resonance parameter data

    International Nuclear Information System (INIS)

    Mill, A.J.

    1979-08-01

    A computer code, XSEC, has been written which calculates neutron cross-sections from resonance data. Although the program was originally written in order to identify neutron 'windows' in enriched nuclides, it may be used to evaluate the total neutron cross-section of any medium mass nuclide at intermediate energies. XSEC has proved very useful in identifying suitable nuclides for use as neutron filters at intermediate energies. (author)

  3. Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system

    International Nuclear Information System (INIS)

    Iga, Kiminori; Takada, Hiroshi; Nagao, Tadashi.

    1998-01-01

    In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B 4 C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)

  4. Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system

    Energy Technology Data Exchange (ETDEWEB)

    Iga, Kiminori [Kyushu Univ., Fukuoka (Japan); Takada, Hiroshi; Nagao, Tadashi

    1998-01-01

    In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B{sub 4}C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)

  5. PANTHER version 11: expanded annotation data from Gene Ontology and Reactome pathways, and data analysis tool enhancements.

    Science.gov (United States)

    Mi, Huaiyu; Huang, Xiaosong; Muruganujan, Anushya; Tang, Haiming; Mills, Caitlin; Kang, Diane; Thomas, Paul D

    2017-01-04

    The PANTHER database (Protein ANalysis THrough Evolutionary Relationships, http://pantherdb.org) contains comprehensive information on the evolution and function of protein-coding genes from 104 completely sequenced genomes. PANTHER software tools allow users to classify new protein sequences, and to analyze gene lists obtained from large-scale genomics experiments. In the past year, major improvements include a large expansion of classification information available in PANTHER, as well as significant enhancements to the analysis tools. Protein subfamily functional classifications have more than doubled due to progress of the Gene Ontology Phylogenetic Annotation Project. For human genes (as well as a few other organisms), PANTHER now also supports enrichment analysis using pathway classifications from the Reactome resource. The gene list enrichment tools include a new 'hierarchical view' of results, enabling users to leverage the structure of the classifications/ontologies; the tools also allow users to upload genetic variant data directly, rather than requiring prior conversion to a gene list. The updated coding single-nucleotide polymorphisms (SNP) scoring tool uses an improved algorithm. The hidden Markov model (HMM) search tools now use HMMER3, dramatically reducing search times and improving accuracy of E-value statistics. Finally, the PANTHER Tree-Attribute Viewer has been implemented in JavaScript, with new views for exploring protein sequence evolution. © The Author(s) 2016. Published by Oxford University Press on behalf of Nucleic Acids Research.

  6. TRACG post-test analysis of panthers prototype tests of SBWR passive containment condenser

    International Nuclear Information System (INIS)

    Fitch, J.R.; Billig, P.F.; Abdollahian, D.; Masoni, P.

    1997-01-01

    As part of the validation effort for application of the TRACG code to the Simplified Boiling Water Reactor (SBWR), calculations have been performed for the various test facilities which are part of the SBWR design and technology certification program. These calculations include post-test calculations for tests in the PANTHERS Passive Containment Condenser (PCC) test program. Sixteen tests from the PANTHERS/PCC test matrix were selected for post-test analysis. This set includes three steady-state pure-steam tests, nine steady-state steam-air tests, and four transient tests. The purpose of this paper is to present and discuss the results of the post-test analysis. The author includes a brief description of the PANTHERS/PCC test facility and test matrix, a description of the PANTHERS/PCC post-test TRACG model and the manner in which the various types of tests in the post-test evaluation were simulated, and a presentation of the results of the TRACG simulation

  7. Comparison of DT neutron production codes MCUNED, ENEA-JSI source subroutine and DDT

    Energy Technology Data Exchange (ETDEWEB)

    Čufar, Aljaž, E-mail: aljaz.cufar@ijs.si [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Lengar, Igor; Kodeli, Ivan [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Milocco, Alberto [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Sauvan, Patrick [Departamento de Ingeniería Energética, E.T.S. Ingenieros Industriales, UNED, C/Juan del Rosal 12, 28040 Madrid (Spain); Conroy, Sean [VR Association, Uppsala University, Department of Physics and Astronomy, PO Box 516, SE-75120 Uppsala (Sweden); Snoj, Luka [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia)

    2016-11-01

    Highlights: • Results of three codes capable of simulating the accelerator based DT neutron generators were compared on a simple model where only a thin target made of mixture of titanium and tritium is present. Two typical deuteron beam energies, 100 keV and 250 keV, were used in the comparison. • Comparisons of the angular dependence of the total neutron flux and spectrum as well as the neutron spectrum of all the neutrons emitted from the target show general agreement of the results but also some noticeable differences. • A comparison of figures of merit of the calculations using different codes showed that the computational time necessary to achieve the same statistical uncertainty can vary for more than 30× when different codes for the simulation of the DT neutron generator are used. - Abstract: As the DT fusion reaction produces neutrons with energies significantly higher than in fission reactors, special fusion-relevant benchmark experiments are often performed using DT neutron generators. However, commonly used Monte Carlo particle transport codes such as MCNP or TRIPOLI cannot be directly used to analyze these experiments since they do not have the capabilities to model the production of DT neutrons. Three of the available approaches to model the DT neutron generator source are the MCUNED code, the ENEA-JSI DT source subroutine and the DDT code. The MCUNED code is an extension of the well-established and validated MCNPX Monte Carlo code. The ENEA-JSI source subroutine was originally prepared for the modelling of the FNG experiments using different versions of the MCNP code (−4, −5, −X) and was later extended to allow the modelling of both DT and DD neutron sources. The DDT code prepares the DT source definition file (SDEF card in MCNP) which can then be used in different versions of the MCNP code. In the paper the methods for the simulation of the DT neutron production used in the codes are briefly described and compared for the case of a

  8. NSPEC - A neutron spectrum code for beam-heated fusion plasmas

    International Nuclear Information System (INIS)

    Scheffel, J.

    1983-06-01

    A 3-dimensional computer code is described, which computes neutron spectra due to beam heating of fusion plasmas. Three types of interactions are considered; thermonuclear of plasma-plasma, beam-plasma and beam-beam interactions. Beam deposition is modelled by the NFREYA code. The applied steady state beam distribution as a function of pitch angle and velocity contains the effects of energy diffusion, friction, angular scattering, charge exchange, electric field and source pitch angle distribution. The neutron spectra, generated by Monte-Carlo methods, are computed with respect to given lines of sight. This enables the code to be used for neutron diagnostics. (author)

  9. A novel approach to correct the coded aperture misalignment for fast neutron imaging

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, F. N.; Hu, H. S., E-mail: huasi-hu@mail.xjtu.edu.cn; Wang, D. M.; Jia, J. [School of Energy and Power Engineering, Xi’an Jiaotong University, Xi’an 710049 (China); Zhang, T. K. [Laser Fusion Research Center, CAEP, Mianyang, 621900 Sichuan (China); Jia, Q. G. [Institute of Applied Physics and Computational Mathematics, Beijing 100094 (China)

    2015-12-15

    Aperture alignment is crucial for the diagnosis of neutron imaging because it has significant impact on the coding imaging and the understanding of the neutron source. In our previous studies on the neutron imaging system with coded aperture for large field of view, “residual watermark,” certain extra information that overlies reconstructed image and has nothing to do with the source is discovered if the peak normalization is employed in genetic algorithms (GA) to reconstruct the source image. Some studies on basic properties of residual watermark indicate that the residual watermark can characterize coded aperture and can thus be used to determine the location of coded aperture relative to the system axis. In this paper, we have further analyzed the essential conditions for the existence of residual watermark and the requirements of the reconstruction algorithm for the emergence of residual watermark. A gamma coded imaging experiment has been performed to verify the existence of residual watermark. Based on the residual watermark, a correction method for the aperture misalignment has been studied. A multiple linear regression model of the position of coded aperture axis, the position of residual watermark center, and the gray barycenter of neutron source with twenty training samples has been set up. Using the regression model and verification samples, we have found the position of the coded aperture axis relative to the system axis with an accuracy of approximately 20 μm. Conclusively, a novel approach has been established to correct the coded aperture misalignment for fast neutron coded imaging.

  10. A neutron spectrum unfolding computer code based on artificial neural networks

    International Nuclear Information System (INIS)

    Ortiz-Rodríguez, J.M.; Reyes Alfaro, A.; Reyes Haro, A.; Cervantes Viramontes, J.M.; Vega-Carrillo, H.R.

    2014-01-01

    The Bonner Spheres Spectrometer consists of a thermal neutron sensor placed at the center of a number of moderating polyethylene spheres of different diameters. From the measured readings, information can be derived about the spectrum of the neutron field where measurements were made. Disadvantages of the Bonner system are the weight associated with each sphere and the need to sequentially irradiate the spheres, requiring long exposure periods. Provided a well-established response matrix and adequate irradiation conditions, the most delicate part of neutron spectrometry, is the unfolding process. The derivation of the spectral information is not simple because the unknown is not given directly as a result of the measurements. The drawbacks associated with traditional unfolding procedures have motivated the need of complementary approaches. Novel methods based on Artificial Intelligence, mainly Artificial Neural Networks, have been widely investigated. In this work, a neutron spectrum unfolding code based on neural nets technology is presented. This code is called Neutron Spectrometry and Dosimetry with Artificial Neural networks unfolding code that was designed in a graphical interface. The core of the code is an embedded neural network architecture previously optimized using the robust design of artificial neural networks methodology. The main features of the code are: easy to use, friendly and intuitive to the user. This code was designed for a Bonner Sphere System based on a 6 LiI(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. The main feature of the code is that as entrance data, for unfolding the neutron spectrum, only seven rate counts measured with seven Bonner spheres are required; simultaneously the code calculates 15 dosimetric quantities as well as the total flux for radiation protection purposes. This code generates a full report with all information of the unfolding

  11. Calculation of thermal neutron self-shielding correction factors for aqueous bulk sample prompt gamma neutron activation analysis using the MCNP code

    International Nuclear Information System (INIS)

    Nasrabadi, M.N.; Jalali, M.; Mohammadi, A.

    2007-01-01

    In this work thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing materials is studied using bulk sample prompt gamma neutron activation analysis (BSPGNAA) with the MCNP code. The code was used to perform three dimensional simulations of a neutron source, neutron detector and sample of various material compositions. The MCNP model was validated against experimental measurements of the neutron flux performed using a BF 3 detector. Simulations were performed to predict thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing solutes. In practice, the MCNP calculations are combined with experimental measurements of the relative thermal neutron flux over the sample's surface, with respect to a reference water sample, to derive the thermal neutron self-shielding within the sample. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the average thermal neutron flux within the sample volume is required

  12. 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

  13. A neutron spectrum unfolding computer code based on artificial neural networks

    Science.gov (United States)

    Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

    2014-02-01

    The Bonner Spheres Spectrometer consists of a thermal neutron sensor placed at the center of a number of moderating polyethylene spheres of different diameters. From the measured readings, information can be derived about the spectrum of the neutron field where measurements were made. Disadvantages of the Bonner system are the weight associated with each sphere and the need to sequentially irradiate the spheres, requiring long exposure periods. Provided a well-established response matrix and adequate irradiation conditions, the most delicate part of neutron spectrometry, is the unfolding process. The derivation of the spectral information is not simple because the unknown is not given directly as a result of the measurements. The drawbacks associated with traditional unfolding procedures have motivated the need of complementary approaches. Novel methods based on Artificial Intelligence, mainly Artificial Neural Networks, have been widely investigated. In this work, a neutron spectrum unfolding code based on neural nets technology is presented. This code is called Neutron Spectrometry and Dosimetry with Artificial Neural networks unfolding code that was designed in a graphical interface. The core of the code is an embedded neural network architecture previously optimized using the robust design of artificial neural networks methodology. The main features of the code are: easy to use, friendly and intuitive to the user. This code was designed for a Bonner Sphere System based on a 6LiI(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. The main feature of the code is that as entrance data, for unfolding the neutron spectrum, only seven rate counts measured with seven Bonner spheres are required; simultaneously the code calculates 15 dosimetric quantities as well as the total flux for radiation protection purposes. This code generates a full report with all information of the unfolding in

  14. Analysis of panthers full-scale heat transfer tests with RELAP5

    International Nuclear Information System (INIS)

    Parlatan, Y.; Boyer, B.D.; Jo, J.; Rohatgi, S.

    1996-01-01

    The RELAP5 code is being assessed on the full-scale Passive Containment Cooling System (PCCS) in the Performance ANalysis and Testing of HEat Removal Systems (PANTHERS) facility at Societa Informazioni Termoidrauliche (SIET) in Italy. PANTHERS is a test facility with fall-size prototype beat exchangers for the PCCS in support of the General Electric's (GE) Simplified Boiling Water Reactor (SBWR) program. PANTHERS tests with a low noncondensable gas concentration and with a high noncondensable gas concentration were analyzed with RELAP5. The results showed that beat transfer rate decreases significantly along the PCCS tubes. In the test case with a higher inlet noncondensable gas fraction, the PCCS removed 35% less heat than in the test case with the lower noncondensable gas fraction. The dominant resistance to the overall heat transfer is the condensation beat transfer resistance inside the tubes. This resistance increased by about 5-fold between the inlet and exit of the tube due to the build up of noncondensable gases along the tube. The RELAP5 calculations also predicted that 4% to 5% of the heat removed to the PCCS pool occurs in the inlet steam piping and PCCS upper and lower headers. These piping needs to be modeled for other tests systems. The full-scale PANTHERS predictions are also compared against 1/400 scale GIRAFFE tests. GIRAFFE has 33% larger heat surface area, but its efficiency is only 15% and 23% higher than PANTHERS for the two cases analyzed This was explained by the high heat transfer resistance inside the tubes near the exit

  15. ICPP - a collision probability module for the AUS neutronics code system

    International Nuclear Information System (INIS)

    Robinson, G.S.

    1985-10-01

    The isotropic collision probability program (ICPP) is a module of the AUS neutronics code system which calculates first flight collision probabilities for neutrons in one-dimensional geometries and in clusters of rods. Neutron sources, including scattering, are assumed to be isotropic and to be spatially flat within each mesh interval. The module solves the multigroup collision probability equations for eigenvalue or fixed source problems

  16. Some neutronics and thermal-hydraulics codes for reactor analysis using personal computers

    International Nuclear Information System (INIS)

    Woodruff, W.L.

    1990-01-01

    Some neutronics and thermal-hydraulics codes formerly available only for main frame computers may now be run on personal computers. Brief descriptions of the codes are provided. Running times for some of the codes are compared for an assortment of personal and main frame computers. With some limitations in detail, personal computer versions of the codes can be used to solve many problems of interest in reactor analyses at very modest costs. 11 refs., 4 tabs

  17. Modeling of LVRF critical experiments in ZED-2 using WIMS9A/PANTHER and MCNP5

    International Nuclear Information System (INIS)

    Sissaoui, M.T.; Carlson, P.A.; Lebenhaft, J.R.

    2009-01-01

    The accuracy of WIMS9A/PANTHER and MCNP5 in modeling D 2 O-moderated, and H 2 O-, D 2 O- or air-cooled, doubly heterogeneous lattices of fuel clusters was demonstrated using Low Void Reactivity Fuel (LVRF) substitution experiments in the ZED-2 critical facility. MCNP5 with ENDF/B-VI (Release 5) underpredicted k eff but gave excellent coolant void reactivity (CVR) bias values. WIMS9A/PANTHER with JEF-2.2 overpredicted k eff and underpredicted the CVR bias relative to MCNP5 by 100-200 pcm. Both codes reproduced the measured axial and radial flux shapes accurately

  18. Development of EASYQAD version β: A Visualization Code System for QAD-CGGP-A Gamma and Neutron Shielding Calculation Code

    International Nuclear Information System (INIS)

    Kim, Jae Cheon; Lee, Hwan Soo; Ha, Pham Nhu Viet; Kim, Soon Young; Shin, Chang Ho; Kim, Jong Kyung

    2007-01-01

    EASYQAD had been previously developed by using MATLAB GUI (Graphical User Interface) in order to perform conveniently gamma and neutron shielding calculations at Hanyang University. It had been completed as version α of radiation shielding analysis code. In this study, EASYQAD was upgraded to version β with many additional functions and more user-friendly graphical interfaces. For general users to run it on Windows XP environment without any MATLAB installation, this version was developed into a standalone code system

  19. MAXED, a computer code for the deconvolution of multisphere neutron spectrometer data using the maximum entropy method

    International Nuclear Information System (INIS)

    Reginatto, M.; Goldhagen, P.

    1998-06-01

    The problem of analyzing data from a multisphere neutron spectrometer to infer the energy spectrum of the incident neutrons is discussed. The main features of the code MAXED, a computer program developed to apply the maximum entropy principle to the deconvolution (unfolding) of multisphere neutron spectrometer data, are described, and the use of the code is illustrated with an example. A user's guide for the code MAXED is included in an appendix. The code is available from the authors upon request

  20. ACDOS2: a code for neutron-induced activities and dose rates

    International Nuclear Information System (INIS)

    Ruby, L.; Keney, G.S.; Lagache, J.C.

    1981-10-01

    In order to anticipate problems from the radioactivation of neutral beam sources as a result of testing, a code has been developed which calculates both the radioactivities produced and the dose rates resulting therefrom. The code ACDOS2 requires neutron source strength and spectral distribution as input, or alternately, the source strength can be calculated internally from an input of neutral beam source parameters. A variety of simple geometries can be specified, and up to 12 times of interest following the shutdown of the neutron source. Radiation attenuating and daughter radioactivities are treated accurately. ACDOS2 is also of use for neutron-induced radioactivation problems involving accelerators, fusion reactors, or fission reactors

  1. Neutron secondary-particle production cross sections and their incorporation into Monte-Carlo transport codes

    International Nuclear Information System (INIS)

    Brenner, D.J.; Prael, R.E.; Little, R.C.

    1987-01-01

    Realistic simulations of the passage of fast neutrons through tissue require a large quantity of cross-sectional data. What are needed are differential (in particle type, energy and angle) cross sections. A computer code is described which produces such spectra for neutrons above ∼14 MeV incident on light nuclei such as carbon and oxygen. Comparisons have been made with experimental measurements of double-differential secondary charged-particle production on carbon and oxygen at energies from 27 to 60 MeV; they indicate that the model is adequate in this energy range. In order to utilize fully the results of these calculations, they should be incorporated into a neutron transport code. This requires defining a generalized format for describing charged-particle production, putting the calculated results in this format, interfacing the neutron transport code with these data, and charged-particle transport. The design and development of such a program is described. 13 refs., 3 figs

  2. Benchmarking time-dependent neutron problems with Monte Carlo codes

    International Nuclear Information System (INIS)

    Couet, B.; Loomis, W.A.

    1990-01-01

    Many nuclear logging tools measure the time dependence of a neutron flux in a geological formation to infer important properties of the formation. The complex geometry of the tool and the borehole within the formation does not permit an exact deterministic modelling of the neutron flux behaviour. While this exact simulation is possible with Monte Carlo methods the computation time does not facilitate quick turnaround of results useful for design and diagnostic purposes. Nonetheless a simple model based on the diffusion-decay equation for the flux of neutrons of a single energy group can be useful in this situation. A combination approach where a Monte Carlo calculation benchmarks a deterministic model in terms of the diffusion constants of the neutrons propagating in the media and their flux depletion rates thus offers the possibility of quick calculation with assurance as to accuracy. We exemplify this approach with the Monte Carlo benchmarking of a logging tool problem, showing standoff and bedding response. (author)

  3. PAD: a one-dimensional, coupled neutronic-thermodynamic-hydrodynamic computer code

    International Nuclear Information System (INIS)

    Peterson, D.M.; Stratton, W.R.; McLaughlin, T.P.

    1976-12-01

    Theoretical and numerical foundations, utilization guide, sample problems, and program listing and glossary are given for the PAD computer code which describes dynamic systems with interactive neutronics, thermodynamics, and hydrodynamics in one-dimensional spherical, cylindrical, and planar geometries. The code has been applied to prompt critical excursions in various fissioning systems (solution, metal, LMFBR, etc.) as well as to nonfissioning systems

  4. Coupling of 3D neutronics models with the system code ATHLET

    International Nuclear Information System (INIS)

    Langenbuch, S.; Velkov, K.

    1999-01-01

    The system code ATHLET for plant transient and accident analysis has been coupled with 3D neutronics models, like QUABOX/CUBBOX, for the realistic evaluation of some specific safety problems under discussion. The considerations for the coupling approach and its realization are discussed. The specific features of the coupled code system established are explained and experience from first applications is presented. (author)

  5. General-purpose Monte Carlo codes for neutron and photon transport calculations. MVP version 3

    International Nuclear Information System (INIS)

    Nagaya, Yasunobu

    2017-01-01

    JAEA has developed a general-purpose neutron/photon transport Monte Carlo code MVP. This paper describes the recent development of the MVP code and reviews the basic features and capabilities. In addition, capabilities implemented in Version 3 are also described. (author)

  6. HADES. A computer code for fast neutron cross section from the Optical Model

    International Nuclear Information System (INIS)

    Guasp, J.; Navarro, C.

    1973-01-01

    A FORTRAN V computer code for UNIVAC 1108/6 using a local Optical Model with spin-orbit interaction is described. The code calculates fast neutron cross sections, angular distribution, and Legendre moments for heavy and intermediate spherical nuclei. It allows for the possibility of automatic variation of potential parameters for experimental data fitting. (Author) 55 refs

  7. A comparison of neutron resonance absorption in thermal reactor lattices in the AUS neutronics code system with Monte Carlo calculations

    International Nuclear Information System (INIS)

    Robinson, G.S.

    1985-08-01

    The calculation of resonance shielding by the subgroup method, as incorporated in the MIRANDA module of the AUS neutronics code system, is compared with Monte Carlo calculatons for a number of thermal reactor lattices. For the large range of single rod and rod cluster lattices considered, AUS results for resonance absorption were high by up to two per cent

  8. Test and validation of the iterative code for the neutrons spectrometry and dosimetry: NSDUAZ

    International Nuclear Information System (INIS)

    Reyes H, A.; Ortiz R, J. M.; Reyes A, A.; Castaneda M, R.; Solis S, L. O.; Vega C, H. R.

    2014-08-01

    In this work was realized the test and validation of an iterative code for neutronic spectrometry known as Neutron Spectrometry and Dosimetry of the Universidad Autonoma de Zacatecas (NSDUAZ). This code was designed in a user graph interface, friendly and intuitive in the environment programming of LabVIEW using the iterative algorithm known as SPUNIT. The main characteristics of the program are: the automatic selection of the initial spectrum starting from the neutrons spectra catalog compiled by the International Atomic Energy Agency, the possibility to generate a report in HTML format that shows in graph and numeric way the neutrons flowing and calculates the ambient dose equivalent with base to this. To prove the designed code, the count rates of a spectrometer system of Bonner spheres were used with a detector of 6 LiI(Eu) with 7 polyethylene spheres with diameter of 0, 2, 3, 5, 8, 10 and 12. The count rates measured with two neutron sources: 252 Cf and 239 PuBe were used to validate the code, the obtained results were compared against those obtained using the BUNKIUT code. We find that the reconstructed spectra present an error that is inside the limit reported in the literature that oscillates around 15%. Therefore, it was concluded that the designed code presents similar results to those techniques used at the present time. (Author)

  9. BCG: a code for calculating pointwise neutron spectra and criticality in fast reactor cells

    International Nuclear Information System (INIS)

    Leite, S.B.; Caldeira, A.D.; Garcia, R.D.M.

    1988-02-01

    The BCG code for determining the space and energy neutron flux distribution and criticality of fast reactor cylindrical cells is presented. The code solves the unidimensional neutron transport equation together with interface current relations at each energy in an unionized grid prepared for the cell and at an arbitrary number of spatial zones. While the spatial resolution is user specified, the energy dependence of the flux distribution is resolved according to the degree of variation in the reconstructed total microscopic cross sections of the atomic species in the cell. Results for a defined sample problem illustrate the high resolution and accuracy that can be obtained with the code. (author) [pt

  10. BCG: a computer code for calculating neutron spectra and criticality in cells of fast reactors

    International Nuclear Information System (INIS)

    Leite, S.B.; Caldeira, A.D.; Garcia, R.D.M.

    1988-01-01

    The BCG code for determining the space and energy neutron flux distribution and criticality of fast reactor cylindrical cells is discussed. The code solves the unidimensional neutron transport equation together with interface current relations at each energy point in an unionized energy grid prepared for the cell and at an arbitrary number of spatial zones. While the spatial resolution is user specified, the energy dependence of the flux distribution is resolved according to the degree of variation in the reconstruced total microscopic cross sections of the atomic species in the cell. Results for a simplified fuel cell illustrate the high resolution and accuracy that can be obtained with the code. (author) [pt

  11. Florida panther habitat use response to prescribed fire

    Science.gov (United States)

    Dees, Catherine S.; Clark, Joseph D.; van Manen, Frank T.

    2001-01-01

    The Florida panther (Puma concolor coryi) is one of the most endangered mammals in the world, with only 30-50 adults surviving in and around Florida Panther National Wildlife Refuge and the adjacent Big Cypress National Preserve. Managers at these areas conduct annual prescribed burns in pine (Pinus sp.) as a cost-effective method of managing wildlife habitat. Our objectives were to determine if temporal and spatial relationships existed between prescribed fire an panther use of pine. to accomplish this, we paired fire-event data from the Refuge an the Preserve with panther radiolocations collected between 1989 and 1998, determined the time that had elapsed since burning had occurred in management units associated with the radiolocations, and generated a frequency distribution based on those times. We then generated ant expected frequency distribution, based on random use relative to time since burning. This analysis revealed that panther use of burned pine habitats was greatest during the first year after a management unit was burned. Also, compositional analysis indicated that panthers were more likely to position their home ranges in areas that contained pine. We conclude that prescribed burning is important to panther ecology. We suggest that panthers were attracted to effects of shorter burning intervals on vegetation composition and evaluate the landscape-scale changes that would result. 

  12. Genetic introgression and the survival of Florida panther kittens

    Science.gov (United States)

    Hostetler, Jeffrey A.; Onorato, David P.; Nichols, James D.; Johnson, Warren E.; Roelke, Melody E.; O'Brien, Stephen J.; Jansen, Deborah; Oli, Madan K.

    2010-01-01

    Estimates of survival for the young of a species are critical for population models. These models can often be improved by determining the effects of management actions and population abundance on this demographic parameter. We used multiple sources of data collected during 1982–2008 and a live-recapture dead-recovery modeling framework to estimate and model survival of Florida panther (Puma concolor coryi) kittens (age 0–1 year). Overall, annual survival of Florida panther kittens was 0.323 ± 0.071 (SE), which was lower than estimates used in previous population models. In 1995, female pumas from Texas (P. c. stanleyana) were released into occupied panther range as part of an intentional introgression program to restore genetic variability. We found that kitten survival generally increased with degree of admixture: F1 admixed and backcrossed to Texas kittens survived better than canonical Florida panther and backcrossed to canonical kittens. Average heterozygosity positively influenced kitten and older panther survival, whereas index of panther abundance negatively influenced kitten survival. Our results provide strong evidence for the positive population-level impact of genetic introgression on Florida panthers. Our approach to integrate data from multiple sources was effective at improving robustness as well as precision of estimates of Florida panther kitten survival, and can be useful in estimating vital rates for other elusive species with sparse data.

  13. Current and anticipated uses of thermalhydraulic and neutronic codes at PSI

    Energy Technology Data Exchange (ETDEWEB)

    Aksan, S.N.; Zimmermann, M.A.; Yadigaroglu, G. [Paul Scherrer Institut, Villigen (Switzerland)

    1997-07-01

    The thermalhydraulic and/or neutronic codes in use at PSI mainly provide the capability to perform deterministic safety analysis for Swiss NPPs and also serve as analysis tools for experimental facilities for LWR and ALWR simulations. In relation to these applications, physical model development and improvements, and assessment of the codes are also essential components of the activities. In this paper, a brief overview is provided on the thermalhydraulic and/or neutronic codes used for safety analysis of LWRs, at PSI, and also of some experiences and applications with these codes. Based on these experiences, additional assessment needs are indicated, together with some model improvement needs. The future needs that could be used to specify both the development of a new code and also improvement of available codes are summarized.

  14. Current and anticipated uses of thermalhydraulic and neutronic codes at PSI

    International Nuclear Information System (INIS)

    Aksan, S.N.; Zimmermann, M.A.; Yadigaroglu, G.

    1997-01-01

    The thermalhydraulic and/or neutronic codes in use at PSI mainly provide the capability to perform deterministic safety analysis for Swiss NPPs and also serve as analysis tools for experimental facilities for LWR and ALWR simulations. In relation to these applications, physical model development and improvements, and assessment of the codes are also essential components of the activities. In this paper, a brief overview is provided on the thermalhydraulic and/or neutronic codes used for safety analysis of LWRs, at PSI, and also of some experiences and applications with these codes. Based on these experiences, additional assessment needs are indicated, together with some model improvement needs. The future needs that could be used to specify both the development of a new code and also improvement of available codes are summarized

  15. Study on MPI/OpenMP hybrid parallelism for Monte Carlo neutron transport code

    International Nuclear Information System (INIS)

    Liang Jingang; Xu Qi; Wang Kan; Liu Shiwen

    2013-01-01

    Parallel programming with mixed mode of messages-passing and shared-memory has several advantages when used in Monte Carlo neutron transport code, such as fitting hardware of distributed-shared clusters, economizing memory demand of Monte Carlo transport, improving parallel performance, and so on. MPI/OpenMP hybrid parallelism was implemented based on a one dimension Monte Carlo neutron transport code. Some critical factors affecting the parallel performance were analyzed and solutions were proposed for several problems such as contention access, lock contention and false sharing. After optimization the code was tested finally. It is shown that the hybrid parallel code can reach good performance just as pure MPI parallel program, while it saves a lot of memory usage at the same time. Therefore hybrid parallel is efficient for achieving large-scale parallel of Monte Carlo neutron transport. (authors)

  16. Transient calculation performance of the MASTER code for control rod ejection problem

    International Nuclear Information System (INIS)

    Cho, B. O.; Joo, H. G.; Yoo, Y. J.; Park, S. Y.; Zee, S. Q.

    1999-01-01

    The accuracy and the effectiveness of the solution methods of the MASTER code for reactor transient problems were analyzed with a set of NEACRP PWR control rod ejection benchmark problems. A series of sensitivity study for the effects on the solution by the neutronic solution methods and the neutronic and thermal-hydraulic model parameters were thus investigated. The MASTER results were then compared with the reference PANTHER results. This indicates that the MASTER solution is sufficiently accurate and the computing time is fast enough for nuclear design application

  17. Transient calculation performance of the MASTER code for control rod ejection problem

    Energy Technology Data Exchange (ETDEWEB)

    Cho, B. O.; Joo, H. G.; Yoo, Y. J.; Park, S. Y.; Zee, S. Q. [KAERI, Taejon (Korea, Republic of)

    1999-10-01

    The accuracy and the effectiveness of the solution methods of the MASTER code for reactor transient problems were analyzed with a set of NEACRP PWR control rod ejection benchmark problems. A series of sensitivity study for the effects on the solution by the neutronic solution methods and the neutronic and thermal-hydraulic model parameters were thus investigated. The MASTER results were then compared with the reference PANTHER results. This indicates that the MASTER solution is sufficiently accurate and the computing time is fast enough for nuclear design application.

  18. Shopping Centers as Panther Habitat: Inferring Animal Locations from Models

    Directory of Open Access Journals (Sweden)

    David S. Maehr

    2004-12-01

    Full Text Available A recent model of Florida panther (Puma concolor coryi habitat erred in arbitrarily creating buffers around radio locations collected during daylight hours on the assumption that study animals were only at rest during these times. The buffers generated by this method likely cause an overestimation of the amounts and kinds of habitats that are used by the panther. This, and other errors, could lead to the impression that unfragmented forest cover is unimportant to panther conservation, and could encourage inaccurate characterizations of panther habitat. Previous 24-hour monitoring of activity and activity readings made during routine telemetry flights indicate that high levels of activity occur in the early morning hours. Literature on the behavior of the species does not support the creation of large buffers around telemetry locations to compensate for the lack of nighttime telemetry data. A thorough examination of ongoing studies that use global positioning systems may help calibrate future Florida panther habitat models.

  19. Fast neutron fluence evaluation of the smart reactor pressure vessel by using the GEOSHIELD code

    International Nuclear Information System (INIS)

    Kim, K.Y.; Kim, K.S.; Kim, H.Y.; Lee, C.C.; Zee, S.Q.

    2007-01-01

    In Korea, the design of an advanced integral reactor system called SMART has been developed by KAERI to supply energy for seawater desalination as well as an electricity generation. A fast neutron fluence distribution at the SMART reactor pressure vessel was evaluated to confirm the integrity of the vessel by using the GEOSHIELD code. The GEOSHIELD code was developed by KAERI in order to prepare an input list including a geometry modeling of the DORT code and to process results from the DORT code output list. Results by a GEOSHIELD code processing and by a manual processing of the DORT show a good agreement. (author)

  20. Development of neutron diffuse scattering analysis code by thin film and multilayer film

    International Nuclear Information System (INIS)

    Soyama, Kazuhiko

    2004-01-01

    To research surface structure of thin film and multilayer film by neutron, a neutron diffuse scattering analysis code using DWBA (Distorted-Wave Bron Approximation) principle was developed. Subjects using this code contain the surface and interface properties of solid/solid, solid/liquid, liquid/liquid and gas/liquid, and metal, magnetism and polymer thin film and biomembran. The roughness of surface and interface of substance shows fractal self-similarity and its analytical model is based on DWBA theory by Sinha. The surface and interface properties by diffuse scattering are investigated on the basis of the theoretical model. The calculation values are proved to be agreed with the experimental values. On neutron diffuse scattering by thin film, roughness of surface of thin film, correlation function, neutron propagation by thin film, diffuse scattering by DWBA theory, measurement model, SDIFFF (neutron diffuse scattering analysis program by thin film) and simulation results are explained. On neutron diffuse scattering by multilayer film, roughness of multilayer film, principle of diffuse scattering, measurement method and simulation examples by MDIFF (neutron diffuse scattering analysis program by multilayer film) are explained. (S.Y.)To research surface structure of thin film and multilayer film by neutron, a neutron diffuse scattering analysis code using DWBA (Distorted-Wave Bron Approximation) principle was developed. Subjects using this code contain the surface and interface properties of solid/solid, solid/liquid, liquid/liquid and gas/liquid, and metal, magnetism and polymer thin film and biomembran. The roughness of surface and interface of substance shows fractal self-similarity and its analytical model is based on DWBA theory by Sinha. The surface and interface properties by diffuse scattering are investigated on the basis of the theoretical model. The calculation values are proved to be agreed with the experimental values. On neutron diffuse scattering

  1. A point-kernel shielding code for calculations of neutron and secondary gamma-ray 1cm dose equivalents: PKN

    International Nuclear Information System (INIS)

    Kotegawa, Hiroshi; Tanaka, Shun-ichi

    1991-09-01

    A point-kernel integral technique code, PKN, and the related data library have been developed to calculate neutron and secondary gamma-ray dose equivalents in water, concrete and iron shields for neutron sources in 3-dimensional geometry. The comparison between calculational results of the present code and those of the 1-dimensional transport code ANISN = JR, and the 2-dimensional transport code DOT4.2 showed a sufficient accuracy, and the availability of the PKN code has been confirmed. (author)

  2. Spallation neutron production and the current intra-nuclear cascade and transport codes

    International Nuclear Information System (INIS)

    Filges, D.; Goldenbaum, F.

    2001-01-01

    A recent renascent interest in energetic proton-induced production of neutrons originates largely from the inception of projects for target stations of intense spallation neutron sources, like the planned European Spallation Source (ESS), accelerator-driven nuclear reactors, nuclear waste transmutation, and also from the application for radioactive beams. In the framework of such a neutron production, of major importance is the search for ways for the most efficient conversion of the primary beam energy into neutron production. Although the issue has been quite successfully addressed experimentally by varying the incident proton energy for various target materials and by covering a huge collection of different target geometries --providing an exhaustive matrix of benchmark data-- the ultimate challenge is to increase the predictive power of transport codes currently on the market. To scrutinize these codes, calculations of reaction cross-sections, hadronic interaction lengths, average neutron multiplicities, neutron multiplicity and energy distributions, and the development of hadronic showers are confronted with recent experimental data of the NESSI collaboration. Program packages like HERMES, LCS or MCNPX master the prevision of reaction cross-sections, hadronic interaction lengths, averaged neutron multiplicities and neutron multiplicity distributions in thick and thin targets for a wide spectrum of incident proton energies, geometrical shapes and materials of the target generally within less than 10% deviation, while production cross-section measurements for light charged particles on thin targets point out that appreciable distinctions exist within these models. (orig.)

  3. Spallation neutron production and the current intra-nuclear cascade and transport codes

    Science.gov (United States)

    Filges, D.; Goldenbaum, F.; Enke, M.; Galin, J.; Herbach, C.-M.; Hilscher, D.; Jahnke, U.; Letourneau, A.; Lott, B.; Neef, R.-D.; Nünighoff, K.; Paul, N.; Péghaire, A.; Pienkowski, L.; Schaal, H.; Schröder, U.; Sterzenbach, G.; Tietze, A.; Tishchenko, V.; Toke, J.; Wohlmuther, M.

    A recent renascent interest in energetic proton-induced production of neutrons originates largely from the inception of projects for target stations of intense spallation neutron sources, like the planned European Spallation Source (ESS), accelerator-driven nuclear reactors, nuclear waste transmutation, and also from the application for radioactive beams. In the framework of such a neutron production, of major importance is the search for ways for the most efficient conversion of the primary beam energy into neutron production. Although the issue has been quite successfully addressed experimentally by varying the incident proton energy for various target materials and by covering a huge collection of different target geometries --providing an exhaustive matrix of benchmark data-- the ultimate challenge is to increase the predictive power of transport codes currently on the market. To scrutinize these codes, calculations of reaction cross-sections, hadronic interaction lengths, average neutron multiplicities, neutron multiplicity and energy distributions, and the development of hadronic showers are confronted with recent experimental data of the NESSI collaboration. Program packages like HERMES, LCS or MCNPX master the prevision of reaction cross-sections, hadronic interaction lengths, averaged neutron multiplicities and neutron multiplicity distributions in thick and thin targets for a wide spectrum of incident proton energies, geometrical shapes and materials of the target generally within less than 10% deviation, while production cross-section measurements for light charged particles on thin targets point out that appreciable distinctions exist within these models.

  4. SUNF, Simplified UNF Code, Fast Neutron Calculation by Unified Hauser-Feshbach Theory

    International Nuclear Information System (INIS)

    Zhang Jingshang

    2001-01-01

    1 - Description of program or function: The SUNF code is the simplified version of UNF code and is based on the unified Hauser-Feshbach and exciton model. SUNF code has been developed for calculations of fast neutron data for structural materials with neutron energies below 20 MeV. Besides elastic scattering channel, the code may handle decay sequence up to (n,3n) reaction, including 14 reaction channels. The energy spectra can be obtained and the output form is in the ENDF/B-6 format, but in file 5 form. For the ENDF-B-6 output, the incident energies are divided into two types: only cross section calculation; and those including neutron energy spectra. 2 - Methods: Gaussian integration is used for all numerical integration. 3 - Restrictions on the complexity of the problem: The incident energies of neutrons are from 1 KeV to 20 MeV. There are two parameters in this code: incident neutron energies number 'NEL'; and the number of discrete levels of residual nuclei for the first particle emissions 'NLV'. The users can set the values of NEL and NLV according to the storage size of the computer used. The number of discrete levels of residual nuclei for the multi-particle emissions is not greater than 20

  5. Monte Carlo simulation of a coded-aperture thermal neutron camera

    International Nuclear Information System (INIS)

    Dioszegi, I.; Salwen, C.; Forman, L.

    2011-01-01

    We employed the MCNPX Monte Carlo code to simulate image formation in a coded-aperture thermal-neutron camera. The camera, developed at Brookhaven National Laboratory (BNL), consists of a 20 x 17 cm"2 active area "3He-filled position-sensitive wire chamber in a cadmium enclosure box. The front of the box is a coded-aperture cadmium mask (at present with three different resolutions). We tested the detector experimentally with various arrangements of moderated point-neutron sources. The purpose of using the Monte Carlo modeling was to develop an easily modifiable model of the device to predict the detector's behavior using different mask patterns, and also to generate images of extended-area sources or large numbers (up to ten) of them, that is important for nonproliferation and arms-control verification, but difficult to achieve experimentally. In the model, we utilized the advanced geometry capabilities of the MCNPX code to simulate the coded aperture mask. Furthermore, the code simulated the production of thermal neutrons from fission sources surrounded by a thermalizer. With this code we also determined the thermal-neutron shadow cast by the cadmium mask; the calculations encompassed fast- and epithermal-neutrons penetrating into the detector through the mask. Since the process of signal production in "3He-filled position-sensitive wire chambers is well known, we omitted this part from our modeling. Simplified efficiency values were used for the three (thermal, epithermal, and fast) neutron-energy regions. Electronic noise and the room's background were included as a uniform irradiation component. We processed the experimental- and simulated-images using identical LabVIEW virtual instruments. (author)

  6. BRC neutron evaluations of actinides with the TALYS code

    International Nuclear Information System (INIS)

    Morillon, B.; Romain, P.

    2014-01-01

    We briefly report here part of the list of problems to overcome in order to build evaluations as predictive as possible for simulation of criticality benchmarks. Dispersive potential, large coupling scheme and neutron inelastic scattering are the most crucial points of this list of problems. Different tools to distinguish differences between evaluations are also presented. (authors)

  7. TMCC: a transient three-dimensional neutron transport code by the direct simulation method - 222

    International Nuclear Information System (INIS)

    Shen, H.; Li, Z.; Wang, K.; Yu, G.

    2010-01-01

    A direct simulation method (DSM) is applied to solve the transient three-dimensional neutron transport problems. DSM is based on the Monte Carlo method, and can be considered as an application of the Monte Carlo method in the specific type of problems. In this work, the transient neutronics problem is solved by simulating the dynamic behaviors of neutrons and precursors of delayed neutrons during the transient process. DSM gets rid of various approximations which are always necessary to other methods, so it is precise and flexible in the requirement of geometric configurations, material compositions and energy spectrum. In this paper, the theory of DSM is introduced first, and the numerical results obtained with the new transient analysis code, named TMCC (Transient Monte Carlo Code), are presented. (authors)

  8. Neutron shielding point kernel integral calculation code for personal computer: PKN-pc

    International Nuclear Information System (INIS)

    Kotegawa, Hiroshi; Sakamoto, Yukio; Nakane, Yoshihiro; Tomita, Ken-ichi; Kurosawa, Naohiro.

    1994-07-01

    A personal computer version of PKN code, PKN-pc, has been developed to calculate neutron and secondary gamma-ray 1cm depth dose equivalents in water, ordinary concrete and iron for neutron source. Characteristics of PKN code are, to able to calculate dose equivalents in multi-layer three-dimensional system, which are described with two-dimensional surface, for monoenergetic neutron source from 0.01 to 14.9 MeV, 252 Cf fission and 241 Am-Be neutron source quick and easily. In addition to these features, the PKN-pc is possible to process interactive input and to get graphical system configuration and graphical results easily. (author)

  9. PANTHER: A Library of Protein Families and Subfamilies Indexed by Function

    OpenAIRE

    Thomas, Paul D.; Campbell, Michael J.; Kejariwal, Anish; Mi, Huaiyu; Karlak, Brian; Daverman, Robin; Diemer, Karen; Muruganujan, Anushya; Narechania, Apurva

    2003-01-01

    In the genomic era, one of the fundamental goals is to characterize the function of proteins on a large scale. We describe a method, PANTHER, for relating protein sequence relationships to function relationships in a robust and accurate way. PANTHER is composed of two main components: the PANTHER library (PANTHER/LIB) and the PANTHER index (PANTHER/X). PANTHER/LIB is a collection of “books,” each representing a protein family as a multiple sequence alignment, a Hidden Markov Model (HMM)...

  10. TDTORT: Time-Dependent, 3-D, Discrete Ordinates, Neutron Transport Code System with Delayed Neutrons

    International Nuclear Information System (INIS)

    2002-01-01

    1 - Description of program or function: TDTORT solves the time-dependent, three-dimensional neutron transport equation with explicit representation of delayed neutrons to estimate the fission yield from fissionable material transients. This release includes a modified version of TORT from the C00650MFMWS01 DOORS3.1 code package plus the time-dependent TDTORT code. GIP is also included for cross-section preparation. TORT calculates the flux or fluence of particles due to particles incident upon the system from extraneous sources or generated internally as a result of interaction with the system in two- or three-dimensional geometric systems. The principle application is to the deep-penetration transport of neutrons and photons. Reactor eigenvalue problems can also be solved. Numerous printed edits of the results are available, and results can be transferred to output files for subsequent analysis. TDTORT reads ANISN-format cross-section libraries, which are not included in the package. Users may choose from several available in RSICC's data library collection which can be identified by the keyword 'ANISN FORMAT'. 2 - Methods:The time-dependent spatial flux is expressed as a product of a space-, energy-, and angle-dependent shape function, which is usually slowly varying in time and a purely time-dependent amplitude function. The shape equation is solved for the shape using TORT; and the result is used to calculate the point kinetics parameters (e.g., reactivity) by using their inner product definitions, which are then used to solve the time-dependent amplitude and precursor equations. The amplitude function is calculated by solving the kinetics equations using the LSODE solver. When a new shape calculation is needed, the flux is calculated using the newly computed amplitude function. The Boltzmann transport equation is solved using the method of discrete ordinates to treat the directional variable and weighted finite-difference methods, in addition to Linear Nodal

  11. Comparison of Neutron Cross-Sections Using IAEA Nuclear Codes ''ABAREX'' and ''SCAT2''

    International Nuclear Information System (INIS)

    Myint Myint Moe; Win Sin; Sein Htoon

    2004-05-01

    Moel calculations can be used to provide nuclear data for applications in science and technology. The energy averaged neutron induced nuclear reaction cross-sections particular for Al-27, Mg-24, Cr-52, Mn-55, Zn-64 and U-238 with neutrons of energy (0.005 to 10 MeV) are calculated using IAEA nuclear codes ''ABAREX'' and ''SCAT2''. The results are compared with those given in ENDF- 3 nuclear data

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

    International Nuclear Information System (INIS)

    Nguyen Kien Cuong; Toru Obara

    2008-01-01

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

  13. Coded moderator approach for fast neutron source detection and localization at standoff

    Energy Technology Data Exchange (ETDEWEB)

    Littell, Jennifer [Department of Nuclear Engineering, University of Tennessee, 305 Pasqua Engineering Building, Knoxville, TN 37996 (United States); Lukosi, Eric, E-mail: elukosi@utk.edu [Department of Nuclear Engineering, University of Tennessee, 305 Pasqua Engineering Building, Knoxville, TN 37996 (United States); Institute for Nuclear Security, University of Tennessee, 1640 Cumberland Avenue, Knoxville, TN 37996 (United States); Hayward, Jason; Milburn, Robert; Rowan, Allen [Department of Nuclear Engineering, University of Tennessee, 305 Pasqua Engineering Building, Knoxville, TN 37996 (United States)

    2015-06-01

    Considering the need for directional sensing at standoff for some security applications and scenarios where a neutron source may be shielded by high Z material that nearly eliminates the source gamma flux, this work focuses on investigating the feasibility of using thermal neutron sensitive boron straw detectors for fast neutron source detection and localization. We utilized MCNPX simulations to demonstrate that, through surrounding the boron straw detectors by a HDPE coded moderator, a source-detector orientation-specific response enables potential 1D source localization in a high neutron detection efficiency design. An initial test algorithm has been developed in order to confirm the viability of this detector system's localization capabilities which resulted in identification of a 1 MeV neutron source with a strength equivalent to 8 kg WGPu at 50 m standoff within ±11°.

  14. Image enhancement using MCNP5 code and MATLAB in neutron radiography.

    Science.gov (United States)

    Tharwat, Montaser; Mohamed, Nader; Mongy, T

    2014-07-01

    This work presents a method that can be used to enhance the neutron radiography (NR) image for objects with high scattering materials like hydrogen, carbon and other light materials. This method used Monte Carlo code, MCNP5, to simulate the NR process and get the flux distribution for each pixel of the image and determines the scattered neutron distribution that caused image blur, and then uses MATLAB to subtract this scattered neutron distribution from the initial image to improve its quality. This work was performed before the commissioning of digital NR system in Jan. 2013. The MATLAB enhancement method is quite a good technique in the case of static based film neutron radiography, while in neutron imaging (NI) technique, image enhancement and quantitative measurement were efficient by using ImageJ software. The enhanced image quality and quantitative measurements were presented in this work. Copyright © 2014 Elsevier Ltd. All rights reserved.

  15. Library of neutron cross sections of the Thermos code

    International Nuclear Information System (INIS)

    Alonso V, G.; Hernandez L, H.

    1991-10-01

    The present work is the complement of the IT.SN/DFR-017 report in which the structure and the generation of the library of the Thermos code is described. In this report the comparison among the values of the cross sections that has the current library of the Thermos code and those generated by means of the ENDF-B/NJOY it is shown. (Author)

  16. Application of Monte Carlo codes to neutron dosimetry

    International Nuclear Information System (INIS)

    Prevo, C.T.

    1982-01-01

    In neutron dosimetry, calculations enable one to predict the response of a proposed dosimeter before effort is expended to design and fabricate the neutron instrument or dosimeter. The nature of these calculations requires the use of computer programs that implement mathematical models representing the transport of radiation through attenuating media. Numerical, and in some cases analytical, solutions of these models can be obtained by one of several calculational techniques. All of these techniques are either approximate solutions to the well-known Boltzmann equation or are based on kernels obtained from solutions to the equation. The Boltzmann equation is a precise mathematical description of neutron behavior in terms of position, energy, direction, and time. The solution of the transport equation represents the average value of the particle flux density. Integral forms of the transport equation are generally regarded as the formal basis for the Monte Carlo method, the results of which can in principle be made to approach the exact solution. This paper focuses on the Monte Carlo technique

  17. 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

  18. Application of neutron/gamma transport codes for the design of explosive detection systems

    International Nuclear Information System (INIS)

    Elias, E.; Shayer, Z.

    1994-01-01

    Applications of neutron and gamma transport codes to the design of nuclear techniques for detecting concealed explosives material are discussed. The methodology of integrating radiation transport computations in the development, optimization and analysis phases of these new technologies is discussed. Transport and Monte Carlo codes are used for proof of concepts, guide the system integration, reduce the extend of experimental program and provide insight into the physical problem involved. The paper concentrates on detection techniques based on thermal and fast neutron interactions in the interrogated object. (authors). 6 refs., 1 tab., 5 figs

  19. Reactor physics computer code development for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs

    International Nuclear Information System (INIS)

    Rastogi, B.P.

    1989-01-01

    This report discusses various reactor physics codes developed for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs. These code packages have been utilized for nuclear design of 500 MWe and new 235 MWe PHWRs. (author)

  20. Evolutionary and functional mitogenomics associated with the genetic restoration of the Florida panther

    Science.gov (United States)

    Ochoa, Alexander; Onorato, David P.; Fitak, Robert R.; Roelke-Parker, Melody; Culver, Melanie

    2017-01-01

    Florida panthers are endangered pumas that currently persist in reduced patches of habitat in South Florida, USA. We performed mitogenome reference-based assemblies for most parental lines of the admixed Florida panthers that resulted from the introduction of female Texas pumas into South Florida in 1995. With the addition of 2 puma mitogenomes, we characterized 174 single nucleotide polymorphisms (SNPs) across 12 individuals. We defined 5 haplotypes (Pco1–Pco5), one of which (Pco1) had a geographic origin exclusive to Costa Rica and Panama and was possibly introduced into the Everglades National Park, Florida, prior to 1995. Haplotype Pco2 was native to Florida. Haplotypes Pco3 and Pco4 were exclusive to Texas, whereas haplotype Pco5 had an undetermined geographic origin. Phylogenetic inference suggests that haplotypes Pco1–Pco4 diverged ~202000 (95% HPDI = 83000–345000) years ago and that haplotypes Pco2–Pco4 diverged ~61000 (95% HPDI = 9000–127000) years ago. These results are congruent with a south-to-north continental expansion and with a recent North American colonization by pumas. Furthermore, pumas may have migrated from Texas to Florida no earlier than ~44000 (95% HPDI = 2000–98000) years ago. Synonymous mutations presented a greater mean substitution rate than other mitochondrial functional regions: nonsynonymous mutations, tRNAs, rRNAs, and control region. Similarly, all protein-coding genes were under predominant negative selection constraints. We directly and indirectly assessed the presence of potential deleterious SNPs in the ND2 and ND5 genes in Florida panthers prior to and as a consequence of the introduction of Texas pumas. Screenings for such variants are recommended in extant Florida panthers.

  1. 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

  2. Shielding calculations for neutron calibration bunker using Monte Carlo code MCNP-4C

    International Nuclear Information System (INIS)

    Suman, H.; Kharita, M. H.; Yousef, S.

    2008-02-01

    In this work, the dose arising from an Am-Be source of 10 8 neutron/sec strength located inside the newly constructed neutron calibration bunker in the National Radiation Metrology Laboratories, was calculated using MCNP-4C code. It was found that the shielding of the neutron calibration bunker is sufficient. As the calculated dose is not expected to exceed in inhabited areas 0.183 μSv/hr, which is 10 times smaller than the regulatory dose constraints. Hence, it can be concluded that the calibration bunker can house - from the external exposure point of view - an Am-Be neutron source of 10 9 neutron/sec strength. It turned out that the neutron dose from the source is few times greater than the photon dose. The sky shine was found to contribute significantly to the total dose. This contribution was estimated to be 60% of the neutron dose and 10% of the photon dose. The systematic uncertainties due to various factors have been assessed and was found to be between 4 and 10% due to concrete density variations; 15% due to the dose estimation method; 4 -10% due to weather variations (temperature and moisture). The calculated dose was highly sensitive to the changes in source spectra. The uncertainty due to the use of two different neutron spectra is about 70%.(author)

  3. Testing the new stochastic neutronic code ANET in simulating safety important parameters

    International Nuclear Information System (INIS)

    Xenofontos, T.; Delipei, G.-K.; Savva, P.; Varvayanni, M.; Maillard, J.; Silva, J.; Catsaros, N.

    2017-01-01

    Highlights: • ANET is a new neutronics stochastic code. • Criticality calculations in both subcritical and critical nuclear systems of conventional design were conducted. • Simulations of thermal, lower epithermal and fast neutron fluence rates were performed. • Axial fission rate distributions in standard and MOX fuel pins were computed. - Abstract: ANET (Advanced Neutronics with Evolution and Thermal hydraulic feedback) is an under development Monte Carlo code for simulating both GEN II/III reactors as well as innovative nuclear reactor designs, based on the high energy physics code GEANT3.21 of CERN. ANET is built through continuous GEANT3.21 applicability amplifications, comprising the simulation of particles’ transport and interaction in low energy along with the accessibility of user-provided libraries and tracking algorithms for energies below 20 MeV, as well as the simulation of elastic and inelastic collision, capture and fission. Successive testing applications performed throughout the ANET development have been utilized to verify the new code capabilities. In this context the ANET reliability in simulating certain reactor parameters important to safety is here examined. More specifically the reactor criticality as well as the neutron fluence and fission rates are benchmarked and validated. The Portuguese Research Reactor (RPI) after its conversion to low enrichment in U-235 and the OECD/NEA VENUS-2 MOX international benchmark were considered appropriate for the present study, the former providing criticality and neutron flux data and the latter reaction rates. Concerning criticality benchmarking, the subcritical, Training Nuclear Reactor of the Aristotle University of Thessaloniki (TNR-AUTh) was also analyzed. The obtained results are compared with experimental data from the critical infrastructures and with computations performed by two different, well established stochastic neutronics codes, i.e. TRIPOLI-4.8 and MCNP5. Satisfactory agreement

  4. Advanced thermal-hydraulic and neutronic codes: current and future applications. Summary and conclusions

    International Nuclear Information System (INIS)

    2001-05-01

    An OECD Workshop on Advanced Thermal-Hydraulic and Neutronic Codes Applications was held from 10 to 13 April 2000, in Barcelona, Spain, sponsored by the Committee on the Safety of Nuclear Installations (CSNI) of the OECD Nuclear Energy Agency (NEA). It was organised in collaboration with the Spanish Nuclear Safety Council (CSN) and hosted by CSN and the Polytechnic University of Catalonia (UPC) in collaboration with the Spanish Electricity Association (UNESA). The objectives of the Workshop were to review the developments since the previous CSNI Workshop held in Annapolis [NEA/CSNI/ R(97)4; NUREG/CP-0159], to analyse the present status of maturity and remnant needs of thermal-hydraulic (TH) and neutronic system codes and methods, and finally to evaluate the role of these tools in the evolving regulatory environment. The Technical Sessions and Discussion Sessions covered the following topics: - Regulatory requirements for Best-Estimate (BE) code assessment; - Application of TH and neutronic codes for current safety issues; - Uncertainty analysis; - Needs for integral plant transient and accident analysis; - Simulators and fast running codes; - Advances in next generation TH and neutronic codes; - Future trends in physical modeling; - Long term plans for development of advanced codes. The focus of the Workshop was on system codes. An incursion was made, however, in the new field of applying Computational Fluid Dynamic (CFD) codes to nuclear safety analysis. As a general conclusion, the Barcelona Workshop can be considered representative of the progress towards the targets marked at Annapolis almost four years ago. The Annapolis Workshop had identified areas where further development and specific improvements were needed, among them: multi-field models, transport of interfacial area, 2D and 3D thermal-hydraulics, 3-D neutronics consistent with level of details of thermal-hydraulics. Recommendations issued at Annapolis included: developing small pilot/test codes for

  5. Proceedings of the workshop on advanced thermal-hydraulic and neutronic codes: current and future applications

    International Nuclear Information System (INIS)

    2001-01-01

    An OECD Workshop on Advanced Thermal-Hydraulic and Neutronic Codes Applications was held from 10 to 13 April 2000, in Barcelona, Spain, sponsored by the Committee on the Safety of Nuclear Installations (CSNI) of the OECD Nuclear Energy Agency (NEA). It was organised in collaboration with the Spanish Nuclear Safety Council (CSN) and hosted by CSN and the Polytechnic University of Catalonia (UPC) in collaboration with the Spanish Electricity Association (UNESA). The objectives of the Workshop were to review the developments since the previous CSNI Workshop held in Annapolis [NEA/CSNI/ R(97)4; NUREG/CP-0159], to analyse the present status of maturity and remnant needs of thermal-hydraulic (TH) and neutronic system codes and methods, and finally to evaluate the role of these tools in the evolving regulatory environment. The Technical Sessions and Discussion Sessions covered the following topics: - Regulatory requirements for Best-Estimate (BE) code assessment; - Application of TH and neutronic codes for current safety issues; - Uncertainty analysis; - Needs for integral plant transient and accident analysis; - Simulators and fast running codes; - Advances in next generation TH and neutronic codes; - Future trends in physical modeling; - Long term plans for development of advanced codes. The focus of the Workshop was on system codes. An incursion was made, however, in the new field of applying Computational Fluid Dynamic (CFD) codes to nuclear safety analysis. As a general conclusion, the Barcelona Workshop can be considered representative of the progress towards the targets marked at Annapolis almost four years ago. The Annapolis Workshop had identified areas where further development and specific improvements were needed, among them: multi-field models, transport of interfacial area, 2D and 3D thermal-hydraulics, 3-D neutronics consistent with level of details of thermal-hydraulics. Recommendations issued at Annapolis included: developing small pilot/test codes for

  6. SHINE-III. Simple code for skyshine dose calculation up to 3 GeV neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Tsukiyama, Toshihisa; Tayama, Ryuichi; Handa, Hiroyuki [Hitachi Engineering Co. Ltd., Ibaraki (Japan)] [and others

    2000-03-01

    Skyshine dose at site boundary is considered as one of the most fundamental issues to get approval of constructing nuclear installations. Skyshine conical beam response functions (CBRF) for high energy neutrons up to 3 GeV are obtained using NMTC-JAERI and MCNP code. This CBRF is fitted to the four parameters equation. Simple code named SHINE-III using this equation with updated data is developed. (author)

  7. Systems guide to MCNP (Monte Carlo Neutron and Photon Transport Code)

    International Nuclear Information System (INIS)

    Kirk, B.L.; West, J.T.

    1984-06-01

    The subject of this report is the implementation of the Los Alamos National Laboratory Monte Carlo Neutron and Photon Transport Code - Version 3 (MCNP) on the different types of computer systems, especially the IBM MVS system. The report supplements the documentation of the RSIC computer code package CCC-200/MCNP. Details of the procedure to follow in executing MCNP on the IBM computers, either in batch mode or interactive mode, are provided

  8. ACDOS2: an improved neutron-induced dose rate code

    International Nuclear Information System (INIS)

    Lagache, J.C.

    1981-06-01

    To calculate the expected dose rate from fusion reactors as a function of geometry, composition, and time after shutdown a computer code, ACDOS2, was written, which utilizes up-to-date libraries of cross-sections and radioisotope decay data. ACDOS2 is in ANSI FORTRAN IV, in order to make it readily adaptable elsewhere

  9. ACDOS2: an improved neutron-induced dose rate code

    Energy Technology Data Exchange (ETDEWEB)

    Lagache, J.C.

    1981-06-01

    To calculate the expected dose rate from fusion reactors as a function of geometry, composition, and time after shutdown a computer code, ACDOS2, was written, which utilizes up-to-date libraries of cross-sections and radioisotope decay data. ACDOS2 is in ANSI FORTRAN IV, in order to make it readily adaptable elsewhere.

  10. A group of neutronics calculations in the MNSR using the MCNP-4C code

    International Nuclear Information System (INIS)

    Khattab, K.; Sulieman, I.

    2009-11-01

    The MCNP-4C code was used to model the 3-D core configuration for the Syrian Miniature Neutron Source Reactor (MNSR). The continuous energy neutron cross sections were evaluated from ENDF/B-VI library to calculate the thermal and fast neutron fluxes in the MNSR inner and outer irradiation sites. The thermal fluxes in the MNSR inner irradiation sites were measured for the first time using the multiple foil activation method. Good agreements were noticed between the calculated and measured results. This model is used as well to calculate neutron flux spectrum in the reactor inner and outer irradiation sites and the reactor thermal power. Three 3-D neutronic models for the Syrian MNSR reactor using the MCNP-4C code were developed also to assess the possibility of fuel conversion from 89.87 % HEU fuel (UAl 4 -Al) to 19.75 % LEU fuel (UO 2 ). This model is used in this paper to calculate the following reactor core physics parameters: clean cold core excess reactivity, calibration of the control rod worth and calculation its shut down margin, calibration of the top beryllium shim plate reflector, axial neutron flux distributions in the inner and outer irradiation sites and the kinetics parameters ( ι p l and β e ff). (authors)

  11. A validation study of the BURNUP and associated options of the MONTE CARLO neutronics code MONK5W

    International Nuclear Information System (INIS)

    Howard, E.A.

    1985-11-01

    This is a report on the validation of the burnup option of the Monte Carlo Neutronics Code MONK5W, together with the associated facilities which allow for control rod movements and power changes. The validation uses reference solutions produced by the Deterministic Neutronics Code LWR-WIMS for a 2D model which represents a whole reactor calculation with control rod movements. (author)

  12. Release of WIMS10: a versatile reactor physics code for thermal and fast systems - 15467

    International Nuclear Information System (INIS)

    Lindley, B.A.; Newton, T.D.; Hosking, J.G.; Smith, P.N.; Powney, D.J.; Tollit, B.; Smith, P.J.

    2015-01-01

    the WIMS code provides a versatile software package for neutronic calculations, which can be applied to all thermal reactor types including mixed moderator systems. It can provide lattice cell and supercell calculations using a range of flux solutions methods to produce the neutronic libraries for use in PANTHER or other whole core analysis codes. With the release of WIMS10, the range of problems which WIMS can solve has been greatly extended. A WIMS/PANTHER calculation route has been developed and validated for part MOX-fuelled PWRs, with calculations showing excellent agreement with 2D core deterministic and Monte Carlo transport solutions. A flexible geometry 3D method of characteristics transport solver, CACTUS3D has been added to the code. CACTUS3D has been benchmarked for a 3D BWR assembly model, and was in good agreement with a direct 172-group solution in the Monte Carlo code MONK. Fast reactor calculations using the ECCO deterministic calculation route have been validated using experimental data from the ZEBRA reactor. Power deposition can be treated through following neutrons and/or photons to their point of interaction. The improved methodology is shown to give more accurate calculation of heat deposition and improve agreement between calculated and measured detector responses for part MOX-fuelled cores. (authors)

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  14. Radiation Shielding Information Center: a source of computer codes and data for fusion neutronics studies

    International Nuclear Information System (INIS)

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

    1980-01-01

    The Radiation Shielding Information Center (RSIC), established in 1962 to collect, package, analyze, and disseminate information, computer codes, and data in the area of radiation transport related to fission, is now being utilized to support fusion neutronics technology. The major activities include: (1) answering technical inquiries on radiation transport problems, (2) collecting, packaging, testing, and disseminating computing technology and data libraries, and (3) reviewing literature and operating a computer-based information retrieval system containing material pertinent to radiation transport analysis. The computer codes emphasize methods for solving the Boltzmann equation such as the discrete ordinates and Monte Carlo techniques, both of which are widely used in fusion neutronics. The data packages include multigroup coupled neutron-gamma-ray cross sections and kerma coefficients, other nuclear data, and radiation transport benchmark problem results

  15. Application of an integrated PC-based neutronics code system to criticality safety

    International Nuclear Information System (INIS)

    Briggs, J.B.; Nigg, D.W.

    1991-01-01

    An integrated system of neutronics and radiation transport software suitable for operation in an IBM PC-class environment has been under development at the Idaho National Engineering Laboratory (INEL) for the past four years. Four modules within the system are particularly useful for criticality safety applications. Using the neutronics portion of the integrated code system, effective neutron multiplication values (k eff values) have been calculated for a variety of benchmark critical experiments for metal systems (Plutonium and Uranium), Aqueous Systems (Plutonium and Uranium) and LWR fuel rod arrays. A description of the codes and methods used in the analysis and the results of the benchmark critical experiments are presented in this paper. In general, excellent agreement was found between calculated and experimental results. (Author)

  16. Simulation code for the interaction of 14 MeV neutrons on cells

    Energy Technology Data Exchange (ETDEWEB)

    Nenot, M.L.; Alard, J.P.; Dionet, C.; Arnold, J.; Tchirkov, A.; Meunier, H.; Bodez, V.; Rapp, M.; Verrelle, P

    2002-07-01

    The structure of the survival curve of melanoma cells irradiated by 14 MeV neutrons displays unusual features at very low dose rate where a marked increase in cell killings at 0.05 Gy is followed by a plateau for survival from 0.1 to 0.32 Gy. In parallel a simulation code was constructed for the interaction of 14 MeV neutrons with cellular cultures. The code describes the interaction of the neutrons with the atomic nuclei of the cellular medium and of the external medium (flask culture and culture medium), and is used to compute the deposited energy into the cell volume. It was found that the large energy transfer events associated with heavy charged recoil can occur and that a large part of the energy deposition events are due to recoil protons emitted from the external medium. It is suggested that such events could partially explain the experimental results. (author)

  17. The neutrons flux density calculations by Monte Carlo code for the double heterogeneity fuel

    International Nuclear Information System (INIS)

    Gurevich, M.I.; Brizgalov, V.I.

    1994-01-01

    This document provides the calculation technique for the fuel elements which consists of the one substance as a matrix and the other substance as the corn embedded in it. This technique can be used in the neutron flux density calculation by the universal Monte Carlo code. The estimation of accuracy is presented too. (authors). 6 refs., 1 fig

  18. The neutron transport code DTF-Traca users manual and input data

    Energy Technology Data Exchange (ETDEWEB)

    Ahnert, C

    1979-07-01

    This is a users manual of the neutron transport code DTF-TRACA, which is a version of the original DTF-IV with some modifications made at JEN. A detailed input data descriptions is given. The new options developed at JEN are included too. (Author) 18 refs.

  19. Code ACTIVE for calculation of the transmutation, induced activity and decay heat in neutron irradiation

    International Nuclear Information System (INIS)

    Ioki, Kimihiro; Harada, Yuhei; Asami, Naoto.

    1976-03-01

    The computer code ACTIVE has been prepared for calculation of the transmutation rate, induced activity and decay heat. Calculations are carried out with activation chain and spatial distribution of neutron energy spectrum. The spatial distribution of secondary gamma-ray source due to the unstable nuclides is also obtainable. Special attension is paid to the short life decays. (auth.)

  20. Verification of RRC Ki code package for neutronic calculations of WWER core with GD

    International Nuclear Information System (INIS)

    Aleshin, S.S.; Bolshagin, S.N.; Lazarenko, A.P.; Markov, A.V.; Pavlov, V.I.; Pavlovitchev, A.M.; Sidorenko, V.D.; Tsvetkov, V.M.

    2001-01-01

    The report presented is concerned with verification results of TVS-M/PERMAK-A/BIPR-7A code package for WWERs neutronic calculation as applied to calculation of systems containing U-GD pins. The verification is based on corresponded benchmark calculations, data critical experiments and on operation data obtained WWER units with Gd. The comparison results are discussed (Authors)

  1. The neutron transport code DTF-Traca users manual and input data

    International Nuclear Information System (INIS)

    Ahnert, C.

    1979-01-01

    This is a users manual of the neutron transport code DTF-TRACA, which is a version of the original DTF-IV with some modifications made at JEN. A detailed input data descriptions is given. The new options developed at JEN are included too. (Author) 18 refs

  2. Resolution of the neutron transport equation by massively parallel computer in the Cronos code

    International Nuclear Information System (INIS)

    Zardini, D.M.

    1996-01-01

    The feasibility of neutron transport problems parallel resolution by CRONOS code's SN module is here studied. In this report we give the first data about the parallel resolution by angular variable decomposition of the transport equation. Problems about parallel resolution by spatial variable decomposition and memory stage limits are also explained here. (author)

  3. The neutron transport code DTF-TRACA. User's manual and input data

    International Nuclear Information System (INIS)

    Anhert, C.

    1979-01-01

    A user's manual of the neutron transport code DTF-TRACA, which is a version of the original DTF-IV with some modifications made at JEN. A detailed input data description is given. The new options developped at JEN are included too. (author)

  4. Simulations of the neutronic REP behaviour using the codes DRAGON/DONJON

    International Nuclear Information System (INIS)

    Le Mer, J.

    2007-01-01

    Neutron flux calculation is necessary to understand how a nuclear reactor works. This flux is derived from the transport equation on the whole core. Because of its really complex structure and the angular dependence of the transport equation, it is impossible to compute the flux directly and several neutronic calculation codes must be used to solve the equation for different discretizations which require different modelisations. This chain of successive models, known as a calculation scheme, compute the neutron flux of a reactor from its geometry, its isotopic compositions and a cross-section library. Pressurised light Water Reactor (PWR) are the most common nuclear reactor used today. It is necessary for each neutronic code to be validated for this type of reactor. The goal of this work is to create a complete calculation scheme which can be applied to the evolution of the core of a pressurised light water nuclear reactor using the lattice code DRAGON and the reactor code DONJON. Each step of this scheme will be validated by comparisons with other codes or with experimental results. The unit cell calculation will be computed for a benchmark submitted by R. Mosteller. The assembly calculations will be used to compare the results given by DRAGON, APOLLO2 and MCNP for an assembly used by EDF for code testing. The core calculations will show that the codes DRAGON and DONJON can produce accurate macroscopic results for a real core. Those studies will be used to show the effects of many factors on the flux distribution including the cross section library, the number of energy groups, spatial discretization of the unit cell, the tracking model, the self-shielding of the resonant isotopes or the burnup steps. (author)

  5. MCT: a Monte Carlo code for time-dependent neutron thermalization problems

    International Nuclear Information System (INIS)

    Cupini, E.; Simonini, R.

    1974-01-01

    In the Monte Carlo simulation of pulse source experiments, the neutron energy spectrum, spatial distribution and total density may be required for a long time after the pulse. If the assemblies are very small, as often occurs in the cases of interest, sophisticated Monte Carlo techniques must be applied which force neutrons to remain in the system during the time interval investigated. In the MCT code a splitting technique has been applied to neutrons exceeding assigned target times, and we have found that this technique compares very favorably with more usual ones, such as the expected leakage probability, giving large gains in computational time and variance. As an example, satisfactory asymptotic thermal spectra with a neutron attenuation of 10 -5 were quickly obtained. (U.S.)

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

    International Nuclear Information System (INIS)

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

    1976-06-01

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

  7. Image enhancement using MCNP5 code and MATLAB in neutron radiography

    International Nuclear Information System (INIS)

    Tharwat, Montaser; Mohamed, Nader; Mongy, T.

    2014-01-01

    This work presents a method that can be used to enhance the neutron radiography (NR) image for objects with high scattering materials like hydrogen, carbon and other light materials. This method used Monte Carlo code, MCNP5, to simulate the NR process and get the flux distribution for each pixel of the image and determines the scattered neutron distribution that caused image blur, and then uses MATLAB to subtract this scattered neutron distribution from the initial image to improve its quality. This work was performed before the commissioning of digital NR system in Jan. 2013. The MATLAB enhancement method is quite a good technique in the case of static based film neutron radiography, while in neutron imaging (NI) technique, image enhancement and quantitative measurement were efficient by using ImageJ software. The enhanced image quality and quantitative measurements were presented in this work. - Highlights: • This work is applicable for static based film neutron radiography and digital neutron imaging. • MATLAB is a useful tool for imaging enhancement in radiographic film. • Advanced imaging processing is available in the ETRR-2 for imaging processing and data extraction. • The digital imaging system is suitable for complex shapes and sizes, while MATLAB technique is suitable for simple shapes and sizes. • Quantitative measurements are available

  8. An accuracy estimation on neutron penetration calculation through concrete shield with PALLAS codes using bunched component nuclides of concrete

    International Nuclear Information System (INIS)

    Sasamoto, Nobuo; Kotegawa, Hiroshi

    1984-11-01

    In order to improve computational efficiency of PALLAS code, an accuracy is estimated on the neutron penetration calculation through a concrete shield, using bunched component nuclides of concrete. The calculated fast neutron flux is observed to depend weakly on how the nuclides are bunched. Contrary to this, the calculated thermal neutron fluxes are strongly dependent on the manner of bunching, mainly due to the fact that iron cross section has exceptionally large negative sensitivity to thermal neutron flux. (author)

  9. ABAREX -- A neutron spherical optical-statistical-model code -- A user`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Smith, A.B. [ed.; Lawson, R.D.

    1998-06-01

    The contemporary version of the neutron spherical optical-statistical-model code ABAREX is summarized with the objective of providing detailed operational guidance for the user. The physical concepts involved are very briefly outlined. The code is described in some detail and a number of explicit examples are given. With this document one should very quickly become fluent with the use of ABAREX. While the code has operated on a number of computing systems, this version is specifically tailored for the VAX/VMS work station and/or the IBM-compatible personal computer.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  11. Coupled neutronic and thermal-hydraulic code benchmark activities at the International Nuclear Safety Center

    International Nuclear Information System (INIS)

    Podlazov, L. N.

    1998-01-01

    Two realistic benchmark problems are defined and used to assess the performance of coupled thermal-hydraulic and neutronic codes used in simulating dynamic processes in VVER-1000 and RBMK reactor systems. One of the problems simulates a design basis accident involving the ejection of three control and protection system rods from a VVER-1000 reactor. The other is based on a postulated rod withdrawal from an operating RBMK reactor. Preliminary results calculated by various codes are compared. While these results show significant differences, the intercomparisons performed so far provide a basis for further evaluation of code limitations and modeling assumptions

  12. Development of a CAD-based neutron transport code with the method of characteristics

    International Nuclear Information System (INIS)

    Chen Zhenping; Wang Dianxi; He Tao; Wang Guozhong; Zheng Huaqing

    2012-01-01

    The main problem determining whether the method of characteristics (MOC) can be used in complicated and highly heterogeneous geometry is how to combine an effective geometry processing method with MOC. In this study, a new idea making use of MCAM, which is a Mutlti-Calculation Automatic Modeling for Neutronics and Radiation Transport program developed by FDS Team, for geometry description and ray tracing of particle transport was brought forward to solve the geometry problem mentioned above. Based on the theory and approach as the foregoing statement, a two dimensional neutron transport code was developed which had been integrated into VisualBUS, developed by FDS Team. Several benchmarks were used to verify the validity of the code and the numerical results were coincident with the reference values very well, which indicated the accuracy and feasibility of the method and the MOC code. (authors)

  13. The use of the SRIM code for calculation of radiation damage induced by neutrons

    Science.gov (United States)

    Mohammadi, A.; Hamidi, S.; Asadabad, Mohsen Asadi

    2017-12-01

    Materials subjected to neutron irradiation will being evolve to structural changes by the displacement cascades initiated by nuclear reaction. This study discusses a methodology to compute primary knock-on atoms or PKAs information that lead to radiation damage. A program AMTRACK has been developed for assessing of the PKAs information. This software determines the specifications of recoil atoms (using PTRAC card of MCNPX code) and also the kinematics of interactions. The deterministic method was used for verification of the results of (MCNPX+AMTRACK). The SRIM (formely TRIM) code is capable to compute neutron radiation damage. The PKAs information was extracted by AMTRACK program, which can be used as an input of SRIM codes for systematic analysis of primary radiation damage. Then the Bushehr Nuclear Power Plant (BNPP) radiation damage on reactor pressure vessel is calculated.

  14. GPU-accelerated 3D neutron diffusion code based on finite difference method

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Q.; Yu, G.; Wang, K. [Dept. of Engineering Physics, Tsinghua Univ. (China)

    2012-07-01

    Finite difference method, as a traditional numerical solution to neutron diffusion equation, although considered simpler and more precise than the coarse mesh nodal methods, has a bottle neck to be widely applied caused by the huge memory and unendurable computation time it requires. In recent years, the concept of General-Purpose computation on GPUs has provided us with a powerful computational engine for scientific research. In this study, a GPU-Accelerated multi-group 3D neutron diffusion code based on finite difference method was developed. First, a clean-sheet neutron diffusion code (3DFD-CPU) was written in C++ on the CPU architecture, and later ported to GPUs under NVIDIA's CUDA platform (3DFD-GPU). The IAEA 3D PWR benchmark problem was calculated in the numerical test, where three different codes, including the original CPU-based sequential code, the HYPRE (High Performance Pre-conditioners)-based diffusion code and CITATION, were used as counterpoints to test the efficiency and accuracy of the GPU-based program. The results demonstrate both high efficiency and adequate accuracy of the GPU implementation for neutron diffusion equation. A speedup factor of about 46 times was obtained, using NVIDIA's Geforce GTX470 GPU card against a 2.50 GHz Intel Quad Q9300 CPU processor. Compared with the HYPRE-based code performing in parallel on an 8-core tower server, the speedup of about 2 still could be observed. More encouragingly, without any mathematical acceleration technology, the GPU implementation ran about 5 times faster than CITATION which was speeded up by using the SOR method and Chebyshev extrapolation technique. (authors)

  15. GPU-accelerated 3D neutron diffusion code based on finite difference method

    International Nuclear Information System (INIS)

    Xu, Q.; Yu, G.; Wang, K.

    2012-01-01

    Finite difference method, as a traditional numerical solution to neutron diffusion equation, although considered simpler and more precise than the coarse mesh nodal methods, has a bottle neck to be widely applied caused by the huge memory and unendurable computation time it requires. In recent years, the concept of General-Purpose computation on GPUs has provided us with a powerful computational engine for scientific research. In this study, a GPU-Accelerated multi-group 3D neutron diffusion code based on finite difference method was developed. First, a clean-sheet neutron diffusion code (3DFD-CPU) was written in C++ on the CPU architecture, and later ported to GPUs under NVIDIA's CUDA platform (3DFD-GPU). The IAEA 3D PWR benchmark problem was calculated in the numerical test, where three different codes, including the original CPU-based sequential code, the HYPRE (High Performance Pre-conditioners)-based diffusion code and CITATION, were used as counterpoints to test the efficiency and accuracy of the GPU-based program. The results demonstrate both high efficiency and adequate accuracy of the GPU implementation for neutron diffusion equation. A speedup factor of about 46 times was obtained, using NVIDIA's Geforce GTX470 GPU card against a 2.50 GHz Intel Quad Q9300 CPU processor. Compared with the HYPRE-based code performing in parallel on an 8-core tower server, the speedup of about 2 still could be observed. More encouragingly, without any mathematical acceleration technology, the GPU implementation ran about 5 times faster than CITATION which was speeded up by using the SOR method and Chebyshev extrapolation technique. (authors)

  16. Proceedings of the OECD/CSNI workshop on transient thermal-hydraulic and neutronic codes requirements

    International Nuclear Information System (INIS)

    Ebert, D.

    1997-07-01

    This is a report on the CSNI Workshop on Transient Thermal-Hydraulic and Neutronic Codes Requirements held at Annapolis, Maryland, USA November 5-8, 1996. This experts' meeting consisted of 140 participants from 21 countries; 65 invited papers were presented. The meeting was divided into five areas: (1) current and prospective plans of thermal hydraulic codes development; (2) current and anticipated uses of thermal-hydraulic codes; (3) advances in modeling of thermal-hydraulic phenomena and associated additional experimental needs; (4) numerical methods in multi-phase flows; and (5) programming language, code architectures and user interfaces. The workshop consensus identified the following important action items to be addressed by the international community in order to maintain and improve the calculational capability: (a) preserve current code expertise and institutional memory, (b) preserve the ability to use the existing investment in plant transient analysis codes, (c) maintain essential experimental capabilities, (d) develop advanced measurement capabilities to support future code validation work, (e) integrate existing analytical capabilities so as to improve performance and reduce operating costs, (f) exploit the proven advances in code architecture, numerics, graphical user interfaces, and modularization in order to improve code performance and scrutibility, and (g) more effectively utilize user experience in modifying and improving the codes

  17. Proceedings of the OECD/CSNI workshop on transient thermal-hydraulic and neutronic codes requirements

    Energy Technology Data Exchange (ETDEWEB)

    Ebert, D.

    1997-07-01

    This is a report on the CSNI Workshop on Transient Thermal-Hydraulic and Neutronic Codes Requirements held at Annapolis, Maryland, USA November 5-8, 1996. This experts` meeting consisted of 140 participants from 21 countries; 65 invited papers were presented. The meeting was divided into five areas: (1) current and prospective plans of thermal hydraulic codes development; (2) current and anticipated uses of thermal-hydraulic codes; (3) advances in modeling of thermal-hydraulic phenomena and associated additional experimental needs; (4) numerical methods in multi-phase flows; and (5) programming language, code architectures and user interfaces. The workshop consensus identified the following important action items to be addressed by the international community in order to maintain and improve the calculational capability: (a) preserve current code expertise and institutional memory, (b) preserve the ability to use the existing investment in plant transient analysis codes, (c) maintain essential experimental capabilities, (d) develop advanced measurement capabilities to support future code validation work, (e) integrate existing analytical capabilities so as to improve performance and reduce operating costs, (f) exploit the proven advances in code architecture, numerics, graphical user interfaces, and modularization in order to improve code performance and scrutibility, and (g) more effectively utilize user experience in modifying and improving the codes.

  18. A neutron spectrum unfolding code based on generalized regression artificial neural networks

    International Nuclear Information System (INIS)

    Ortiz R, J. M.; Martinez B, M. R.; Castaneda M, R.; Solis S, L. O.; Vega C, H. R.

    2015-10-01

    The most delicate part of neutron spectrometry, is the unfolding process. Then derivation of the spectral information is not simple because the unknown is not given directly as result of the measurements. Novel methods based on Artificial Neural Networks have been widely investigated. In prior works, back propagation neural networks (BPNN) have been used to solve the neutron spectrometry problem, however, some drawbacks still exist using this kind of neural nets, as the optimum selection of the network topology and the long training time. Compared to BPNN, is usually much faster to train a generalized regression neural network (GRNN). That is mainly because spread constant is the only parameter used in GRNN. Another feature is that the network will converge to a global minimum. In addition, often are more accurate than BPNN in prediction. These characteristics make GRNN be of great interest in the neutron spectrometry domain. In this work is presented a computational tool based on GRNN, capable to solve the neutron spectrometry problem. This computational code, automates the pre-processing, training and testing stages, the statistical analysis and the post-processing of the information, using 7 Bonner spheres rate counts as only entrance data. The code was designed for a Bonner Spheres System based on a 6 LiI(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. (Author)

  19. A neutron spectrum unfolding code based on generalized regression artificial neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz R, J. M.; Martinez B, M. R.; Castaneda M, R.; Solis S, L. O. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica, Av. Ramon Lopez Velarde 801, Col. Centro, 98000 Zacatecas, Zac. (Mexico); Vega C, H. R., E-mail: morvymm@yahoo.com.mx [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico)

    2015-10-15

    The most delicate part of neutron spectrometry, is the unfolding process. Then derivation of the spectral information is not simple because the unknown is not given directly as result of the measurements. Novel methods based on Artificial Neural Networks have been widely investigated. In prior works, back propagation neural networks (BPNN) have been used to solve the neutron spectrometry problem, however, some drawbacks still exist using this kind of neural nets, as the optimum selection of the network topology and the long training time. Compared to BPNN, is usually much faster to train a generalized regression neural network (GRNN). That is mainly because spread constant is the only parameter used in GRNN. Another feature is that the network will converge to a global minimum. In addition, often are more accurate than BPNN in prediction. These characteristics make GRNN be of great interest in the neutron spectrometry domain. In this work is presented a computational tool based on GRNN, capable to solve the neutron spectrometry problem. This computational code, automates the pre-processing, training and testing stages, the statistical analysis and the post-processing of the information, using 7 Bonner spheres rate counts as only entrance data. The code was designed for a Bonner Spheres System based on a {sup 6}LiI(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. (Author)

  20. Neutronics code VALE for two-dimensional triagonal (hexagonal) and three-dimensional geometries

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.

    1981-08-01

    This report documents the computer code VALE designed to solve multigroup neutronics problems with the diffusion theory approximation to neutron transport for a triagonal arrangement of mesh points on planes in two- and three-dimensional geometry. This code parallels the VENTURE neutronics code in the local computation system, making exposure and fuel management capabilities available. It uses and generates interface data files adopted in the cooperative effort sponsored by Reactor Physics RRT Division of the US DOE. The programming in FORTRAN is straightforward, although data is transferred in blocks between auxiliary storage devices and main core, and direct access schemes are used. The size of problems which can be handled is essentially limited only by cost of calculation since the arrays are variably dimensioned. The memory requirement is held down while data transfer during iteration is increased only as necessary with problem size. There is provision for the more common boundary conditions including the repeating boundary, 180 0 rotational symmetry, and the rotational symmetry conditions for the 30 0 , 60 0 , and 120 0 triangular grids on planes. A variety of types of problems may be solved: the usual neutron flux eignevalue problem, or a direct criticality search on the buckling, on a reciprocal velocity absorber (prompt mode), or on nuclide concentrations. The adjoint problem and fixed source problem may be solved, as well as the dominating higher harmonic, or the importance problem for an arbitrary fixed source

  1. Iterative code for the reconstruction of the neutrons spectrum using the Bonner spheres

    International Nuclear Information System (INIS)

    Reyes H, A.; Ortiz R, J. M.; Vega C, H. R.

    2012-10-01

    The neutrons are the particles more difficult of detecting for their intrinsic nature. The absence of the neutrons charge makes that an interaction exists with the matter in a different way. The term radiation spectrometry can use to describe the measurement of the intensity of a radiation field with regard to the energy. The intensity distribution with relationship to the energy is commonly known as spectrum. A method to know the neutrons spectrum in the radiation fields to those that people are exposed is the use of the known system as spectrometry system of Bonner spheres, being the more used for the purposes of the radiological protection. The current interest in the electrons spectrometry has stimulated the development of several procedures to carry out the reconstruction of the spectra. During the last decades new codes have been developed such as BUNKIUT, Bums, Fruit, UMG, etc., however, these methods still present several inconveniences as the complexity in their use, the necessity of an expert user and a very near initial spectrum to the spectrum that is wanted to obtain. To solve the mentioned problems it was development the program NSDUAZ (Neutron Spectrometry and Dosimetry from Autonomous University of Zacatecas). The objective of the present work is to prove and to validate the code before mentioned making an analysis of likeness and differences and of advantages and disadvantages with relationship to the codes used at the present time. (Author)

  2. Comparison of the thermal neutron scattering treatment in MCNP6 and GEANT4 codes

    Science.gov (United States)

    Tran, H. N.; Marchix, A.; Letourneau, A.; Darpentigny, J.; Menelle, A.; Ott, F.; Schwindling, J.; Chauvin, N.

    2018-06-01

    To ensure the reliability of simulation tools, verification and comparison should be made regularly. This paper describes the work performed in order to compare the neutron transport treatment in MCNP6.1 and GEANT4-10.3 in the thermal energy range. This work focuses on the thermal neutron scattering processes for several potential materials which would be involved in the neutron source designs of Compact Accelerator-based Neutrons Sources (CANS), such as beryllium metal, beryllium oxide, polyethylene, graphite, para-hydrogen, light water, heavy water, aluminium and iron. Both thermal scattering law and free gas model, coming from the evaluated data library ENDF/B-VII, were considered. It was observed that the GEANT4.10.03-patch2 version was not able to account properly the coherent elastic process occurring in crystal lattice. This bug is treated in this work and it should be included in the next release of the code. Cross section sampling and integral tests have been performed for both simulation codes showing a fair agreement between the two codes for most of the materials except for iron and aluminium.

  3. A neutron spectrum unfolding code based on generalized regression artificial neural networks

    International Nuclear Information System (INIS)

    Rosario Martinez-Blanco, Ma. del

    2016-01-01

    The most delicate part of neutron spectrometry, is the unfolding process. The derivation of the spectral information is not simple because the unknown is not given directly as a result of the measurements. Novel methods based on Artificial Neural Networks have been widely investigated. In prior works, back propagation neural networks (BPNN) have been used to solve the neutron spectrometry problem, however, some drawbacks still exist using this kind of neural nets, i.e. the optimum selection of the network topology and the long training time. Compared to BPNN, it's usually much faster to train a generalized regression neural network (GRNN). That's mainly because spread constant is the only parameter used in GRNN. Another feature is that the network will converge to a global minimum, provided that the optimal values of spread has been determined and that the dataset adequately represents the problem space. In addition, GRNN are often more accurate than BPNN in the prediction. These characteristics make GRNNs to be of great interest in the neutron spectrometry domain. This work presents a computational tool based on GRNN capable to solve the neutron spectrometry problem. This computational code, automates the pre-processing, training and testing stages using a k-fold cross validation of 3 folds, the statistical analysis and the post-processing of the information, using 7 Bonner spheres rate counts as only entrance data. The code was designed for a Bonner Spheres System based on a "6LiI(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. - Highlights: • Main drawback of neutron spectrometry with BPNN is network topology optimization. • Compared to BPNN, it’s usually much faster to train a (GRNN). • GRNN are often more accurate than BPNN in the prediction. These characteristics make GRNNs to be of great interest. • This computational code, automates the pre-processing, training

  4. Neutrons Flux Distributions of the Pu-Be Source and its Simulation by the MCNP-4B Code

    Science.gov (United States)

    Faghihi, F.; Mehdizadeh, S.; Hadad, K.

    Neutron Fluence rate of a low intense Pu-Be source is measured by Neutron Activation Analysis (NAA) of 197Au foils. Also, the neutron fluence rate distribution versus energy is calculated using the MCNP-4B code based on ENDF/B-V library. Theoretical simulation as well as our experimental performance are a new experience for Iranians to make reliability with the code for further researches. In our theoretical investigation, an isotropic Pu-Be source with cylindrical volume distribution is simulated and relative neutron fluence rate versus energy is calculated using MCNP-4B code. Variation of the fast and also thermal neutrons fluence rate, which are measured by NAA method and MCNP code, are compared.

  5. New evaluated neutron cross section libraries for the GEANT4 code

    International Nuclear Information System (INIS)

    Mendoza, E.; Cano-Ott, D.; Guerrero, C.; Capote, R.

    2012-04-01

    The so-called High Precision neutron physics model implemented in the GEANT4 simulation package allows simulating the transport of neutrons with energies up to 20 MeV. It relies on the G4NDL cross section libraries, prepared by the GEANT4 collaboration from evaluated cross section files and distributed freely together with the code. Even though the performance of the G4NDL library has been improved over the time, users running complex simulations which involve the transport of neutrons do need more flexibility, in particular when assessing the uncertainties in the simulation results due to the neutron (and hence the nuclear) data library used. For this reason, a software tool has been developed for transforming any evaluated neutron cross section library in the ENDF-6 format into the G4NDL format. Furthermore, eight different releases of ENDF-B, JEFF, JENDL, CENDL and BROND national libraries have been translated into the G4NDL format and are distributed by the IAEA nuclear data service at www-nds.iaea.org/geant4. In this way, GEANT4 users have access to the complete list of standard evaluated neutron data libraries when performing Monte Carlo simulations with GEANT4. Consistency checks and a first validation of the libraries have been made following the methods described in this report. (author)

  6. Benchmarking the MCNP code for Monte Carlo modelling of an in vivo neutron activation analysis system.

    Science.gov (United States)

    Natto, S A; Lewis, D G; Ryde, S J

    1998-01-01

    The Monte Carlo computer code MCNP (version 4A) has been used to develop a personal computer-based model of the Swansea in vivo neutron activation analysis (IVNAA) system. The model included specification of the neutron source (252Cf), collimators, reflectors and shielding. The MCNP model was 'benchmarked' against fast neutron and thermal neutron fluence data obtained experimentally from the IVNAA system. The Swansea system allows two irradiation geometries using 'short' and 'long' collimators, which provide alternative dose rates for IVNAA. The data presented here relate to the short collimator, although results of similar accuracy were obtained using the long collimator. The fast neutron fluence was measured in air at a series of depths inside the collimator. The measurements agreed with the MCNP simulation within the statistical uncertainty (5-10%) of the calculations. The thermal neutron fluence was measured and calculated inside the cuboidal water phantom. The depth of maximum thermal fluence was 3.2 cm (measured) and 3.0 cm (calculated). The width of the 50% thermal fluence level across the phantom at its mid-depth was found to be the same by both MCNP and experiment. This benchmarking exercise has given us a high degree of confidence in MCNP as a tool for the design of IVNAA systems.

  7. TRACE/VALKIN: a neutronics-thermohydraulics coupled code to analyze strong 3D transients

    Energy Technology Data Exchange (ETDEWEB)

    Rafael Miro; Gumersindo Verdu; Ana Maria Sanchez [Chemical and Nuclear Engineering Department. Polytechnic University of Valencia. Cami de Vera s/n. 46022 Valencia (Spain); Damian Ginestar [Applied Mathematics Department. Polytechnic University of Valencia. Cami de Vera s/n. 46022 Valencia (Spain)

    2005-07-01

    Full text of publication follows: A nuclear reactor simulator consists mainly of two different blocks, which solve the models used for the basic physical phenomena taking place in the reactor. In this way, there is a neutronic module which simulates the neutron balance in the reactor core, and a thermal-hydraulics module, which simulates the heat transfer in the fuel, the heat transfer from the fuel to the water, and the different condensation and evaporation processes taking place in the reactor core and in the condenser systems. TRACE is a two-phase, two-fluid thermal-hydraulic reactor systems analysis code. The TRACE acronym stands for TRAC/RELAP Advanced Computational Engine, reflecting its ability to run both RELAP5 and TRAC legacy input models. It includes a three-dimensional kinetics module called PARCS for performing advanced analysis of coupled core thermal-hydraulic/kinetics problems. TRACE-VALKIN code is a new time domain analysis code to study transients in LWR reactors. This code uses the best estimate code TRACE to give account of the heat transfer and thermal-hydraulic processes, and a 3D neutronics module. This module has two options, the MODKIN option that makes use of a modal method based on the assumption that the neutronic flux can be approximately expanded in terms of the dominant lambda modes associated with a static configuration of the reactor core, and the NOKIN option that uses a one-step backward discretization of the neutron diffusion equation. The lambda modes are obtained using the Implicit Restarted Arnoldi approach or the Jacob-Davidson algorithm. To check the performance of the coupled code TRACE-VALKIN against complex 3D neutronic transients, using the cross-sections tables generated with the translator SIMTAB from SIMULATE to TRACE/VALKIN, the Cofrentes NPP SCRAM-61 transient is simulated. Cofrentes NPP is a General Electric BWR-6 design located in Valencia-land (Spain). It is in operation since 1985 and currently in its fifteenth

  8. Panthers and Forests in South Florida: an Ecological Perspective

    Directory of Open Access Journals (Sweden)

    E. Jane Comiskey

    2002-06-01

    Full Text Available The endangered Florida panther (Puma concolor coryi survives in an area of pronounced habitat diversity in southern Florida, occupying extensive home ranges that encompass a mosaic of habitats. Twenty-one years of daytime monitoring via radiotelemetry have provided substantial but incomplete information about panther ecology, mainly because this method fails to capture movement and habitat use between dusk and dawn, when panthers are most active. Broad characterizations of panther habitat suitability have nonetheless been derived from telemetry-based habitat selection studies, focusing narrowly on forests where daytime resting sites are often located. The resulting forest-centered view of panthers attributed their restricted distribution and absence of population growth in the mid-1990s to a scarcity of unfragmented forest for expansion. However, the panther population has doubled since the beginning of genetic restoration in 1995, increasing five-fold in public areas described as unsuitable based on forest criteria. Although the forest-centered view no longer explains panther distribution, it continues to shape management decisions and habitat conservation policies. The assumptions and limitations of this view therefore merit critical examination. We analyze the role of forests in the ecology of the Florida panther. To address the absence of nighttime telemetry data, we use innovative telemetry mapping techniques and incorporate information from field observations indicating habitat use during active hours (e.g., tracks, scats, urine markers, and kill sites. We consider daytime telemetry data in the context of panther home ranges and breeding units. We analyze home range size in relation to the amount of forest within each range, concluding that percent forest cover is a poor predictor of size. We apply fractal analysis techniques to characterize the relative density of forest cover associated with daytime locations and interpret the results in

  9. Parallelization of MCNP Monte Carlo neutron and photon transport code in parallel virtual machine and message passing interface

    International Nuclear Information System (INIS)

    Deng Li; Xie Zhongsheng

    1999-01-01

    The coupled neutron and photon transport Monte Carlo code MCNP (version 3B) has been parallelized in parallel virtual machine (PVM) and message passing interface (MPI) by modifying a previous serial code. The new code has been verified by solving sample problems. The speedup increases linearly with the number of processors and the average efficiency is up to 99% for 12-processor. (author)

  10. Verification of vectorized Monte Carlo code MVP using JRR-4 experiment of fast neutrons penetrating through graphite and water

    International Nuclear Information System (INIS)

    Odano, N.; Miura, T.; Yamaji, A.

    1996-01-01

    Measurement of activation reaction rates was carried out for fast neutrons penetrating through graphite and water from the core of JRR-4 research reactor of JAERI, with paying attention to the energy above 10 MeV. Analysis of the experiment was made using a vectorized continuous energy Monte Carlo code MVP to verify the code. The analysis shows good agreements between the measurement and calculation and the MVP code has been confirmed its validity for the fast neutron transport calculations above 10 MeV in fission neutron field. (author)

  11. Neutron flux distribution inside the cylindrical core of minor excess of reactivity in the IPEN/MB-01 reactor and comparison with citation code and MCNP- 5 code

    Energy Technology Data Exchange (ETDEWEB)

    Aredes, Vitor Ottoni; Bitelli, Ulysses d' Utra; Mura, Luiz Ernesto C.; Santos, Diogo Feliciano dos; Lima, Ana Cecilia de Souza, E-mail: ubitelli@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    This study aims to determine the distribution of thermal neutron flux in the IPEN/MB-01 nuclear reactor core assembled with cylindrical core configuration of minor excess of reactivity with 568 fuel rods (28 fuel rods in diameter). The thermal neutron flux at the positions of irradiation derive from the method of reaction rate using gold foils. The experiment consists in inserting gold activations foils with and without cadmium coverage (cadmium boxes with 0.0502 cm thickness) in several positions throughout the active core. After irradiation, activity induced by nuclear reaction rates over gold foils is assessed by gamma ray spectrometry using a high-purity germanium (HPGe) detector. Experimental results are compared to those derived from calculations performed using a three dimensional CITATION diffusion code and MCNP-5 code and a proper nuclear data library. While calculated neutron flux data shows good agreement with experimental values in regions with little disturbance in the neutron flux, also showing that in the region of the reflectors of neutrons and near the control rods, the diffusion theory is not very precise. The average value of thermal neutron flux obtained experimentally compared to the calculated value by CITATION code and MCNP-5 code respectively show a difference of 1.18% and 0.84% at a nuclear power level of 74.65 ± 3.28 % watts. The average measured value of thermal neutron flux is 4.10 10{sup 8} ± 5.25% n/cm{sup 2}s. (author)

  12. Neutron flux distribution inside the cylindrical core of minor excess of reactivity in the IPEN/MB-01 reactor and comparison with citation code and MCNP- 5 code

    International Nuclear Information System (INIS)

    Aredes, Vitor Ottoni; Bitelli, Ulysses d'Utra; Mura, Luiz Ernesto C.; Santos, Diogo Feliciano dos; Lima, Ana Cecilia de Souza

    2015-01-01

    This study aims to determine the distribution of thermal neutron flux in the IPEN/MB-01 nuclear reactor core assembled with cylindrical core configuration of minor excess of reactivity with 568 fuel rods (28 fuel rods in diameter). The thermal neutron flux at the positions of irradiation derive from the method of reaction rate using gold foils. The experiment consists in inserting gold activations foils with and without cadmium coverage (cadmium boxes with 0.0502 cm thickness) in several positions throughout the active core. After irradiation, activity induced by nuclear reaction rates over gold foils is assessed by gamma ray spectrometry using a high-purity germanium (HPGe) detector. Experimental results are compared to those derived from calculations performed using a three dimensional CITATION diffusion code and MCNP-5 code and a proper nuclear data library. While calculated neutron flux data shows good agreement with experimental values in regions with little disturbance in the neutron flux, also showing that in the region of the reflectors of neutrons and near the control rods, the diffusion theory is not very precise. The average value of thermal neutron flux obtained experimentally compared to the calculated value by CITATION code and MCNP-5 code respectively show a difference of 1.18% and 0.84% at a nuclear power level of 74.65 ± 3.28 % watts. The average measured value of thermal neutron flux is 4.10 10 8 ± 5.25% n/cm 2 s. (author)

  13. Analytical three-dimensional neutron transport benchmarks for verification of nuclear engineering codes. Final report

    International Nuclear Information System (INIS)

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

    1997-01-01

    Because of the requirement of accountability and quality control in the scientific world, a demand for high-quality analytical benchmark calculations has arisen in the neutron transport community. The intent of these benchmarks is to provide a numerical standard to which production neutron transport codes may be compared in order to verify proper operation. The overall investigation as modified in the second year renewal application includes the following three primary tasks. Task 1 on two dimensional neutron transport is divided into (a) single medium searchlight problem (SLP) and (b) two-adjacent half-space SLP. Task 2 on three-dimensional neutron transport covers (a) point source in arbitrary geometry, (b) single medium SLP, and (c) two-adjacent half-space SLP. Task 3 on code verification, includes deterministic and probabilistic codes. The primary aim of the proposed investigation was to provide a suite of comprehensive two- and three-dimensional analytical benchmarks for neutron transport theory applications. This objective has been achieved. The suite of benchmarks in infinite media and the three-dimensional SLP are a relatively comprehensive set of one-group benchmarks for isotropically scattering media. Because of time and resource limitations, the extensions of the benchmarks to include multi-group and anisotropic scattering are not included here. Presently, however, enormous advances in the solution for the planar Green's function in an anisotropically scattering medium have been made and will eventually be implemented in the two- and three-dimensional solutions considered under this grant. Of particular note in this work are the numerical results for the three-dimensional SLP, which have never before been presented. The results presented were made possible only because of the tremendous advances in computing power that have occurred during the past decade

  14. DANDE-a linked code system for core neutronics/depletion analysis

    International Nuclear Information System (INIS)

    LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.

    1986-01-01

    This report describes DANDE-a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the course of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of the reactor fuel under increased burn conditions. The operation of the code system is illustrated in this report by two actual problems

  15. DANDE: a linked code system for core neutronics/depletion analysis

    International Nuclear Information System (INIS)

    LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.

    1986-01-01

    This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the cource of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of reactor fuel under increased burn conditions. The operation of the code system is illustrated in this report by two sample problems. 25 refs

  16. DANDE: a linked code system for core neutronics/depletion analysis

    International Nuclear Information System (INIS)

    LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.

    1985-06-01

    This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the course of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of reactor fuel under increased burn conditions. The operation of the code system is made clear in this report by following a sample problem

  17. Aquelarre. A computer code for fast neutron cross sections from the statistical model

    International Nuclear Information System (INIS)

    Guasp, J.

    1974-01-01

    A Fortran V computer code for Univac 1108/6 using the partial statistical (or compound nucleus) model is described. The code calculates fast neutron cross sections for the (n, n'), (n, p), (n, d) and (n, α reactions and the angular distributions and Legendre moments.for the (n, n) and (n, n') processes in heavy and intermediate spherical nuclei. A local Optical Model with spin-orbit interaction for each level is employed, allowing for the width fluctuation and Moldauer corrections, as well as the inclusion of discrete and continuous levels. (Author) 67 refs

  18. AUS98 - The 1998 version of the AUS modular neutronic code system

    International Nuclear Information System (INIS)

    Robinson, G.S.; Harrington, B.V.

    1998-07-01

    AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energy deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous AUS publications are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM main-frame computers to UNIX workstations This report gives details of all system aspects of AUS and all modules except the POW3D multi-dimensional diffusion module

  19. AUS98 - The 1998 version of the AUS modular neutronic code system

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, G.S.; Harrington, B.V

    1998-07-01

    AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energy deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous AUS publications are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM main-frame computers to UNIX workstations This report gives details of all system aspects of AUS and all modules except the POW3D multi-dimensional diffusion module refs., tabs.

  20. Input data requirements for special processors in the computation system containing the VENTURE neutronics code

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.

    1979-07-01

    User input data requirements are presented for certain special processors in a nuclear reactor computation system. These processors generally read data in formatted form and generate binary interface data files. Some data processing is done to convert from the user oriented form to the interface file forms. The VENTURE diffusion theory neutronics code and other computation modules in this system use the interface data files which are generated

  1. Input data requirements for special processors in the computation system containing the VENTURE neutronics code

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.

    1976-11-01

    This report presents user input data requirements for certain special processors in a nuclear reactor computation system. These processors generally read data in formatted form and generate binary interface data files. Some data processing is done to convert from the user-oriented form to the interface file forms. The VENTURE diffusion theory neutronics code and other computation modules in this system use the interface data files which are generated

  2. NCT-ART - a neutron computer tomography code based on the algebraic reconstruction technique

    International Nuclear Information System (INIS)

    Krueger, A.

    1988-01-01

    A computer code is presented, which calculates two-dimensional cuts of material assemblies from a number of neutron radiographic projections. Mathematically, the reconstruction is performed by an iterative solution of a system of linear equations. If the system is fully determined, clear pictures are obtained. Even for an underdetermined system (low number of projections) reasonable pictures are reconstructed, but then picture artefacts and convergence problems occur increasingly. (orig.) With 37 figs [de

  3. A study of the responses of neutron dose equivalent survey meters with computer codes

    International Nuclear Information System (INIS)

    Sartori, D.E.; Beer, G.P. de

    1983-01-01

    The ANISN and DOT discrete-ordinates radiation transport codes for one and two dimensions have been proved as effective and simple techniques to study the response of dose equivalent neutron detectors. Comparisons between results of an experimental calibration of the Harwell 95/0075 survey meter and calculated results rendered satisfactory agreement, considering the different techniques and sources of error involved. Possible improvements in the methods and designs and causes of error are discussed. (author)

  4. PANTHER version 6: protein sequence and function evolution data with expanded representation of biological pathways

    OpenAIRE

    Mi, Huaiyu; Guo, Nan; Kejariwal, Anish; Thomas, Paul D.

    2006-01-01

    PANTHER is a freely available, comprehensive software system for relating protein sequence evolution to the evolution of specific protein functions and biological roles. Since 2005, there have been three main improvements to PANTHER. First, the sequences used to create evolutionary trees are carefully selected to provide coverage of phylogenetic as well as functional information. Second, PANTHER is now a member of the InterPro Consortium, and the PANTHER hidden markov Models (HMMs) are distri...

  5. Testing and Fielding of the Panther Tank and Lessons for Force XXI

    Science.gov (United States)

    1997-01-01

    decided that the following solution be adopted: the construction of the Tiger Tank , a tank of some 60 tons, which had recently been started would...to minimize the German advantages of the Panther. The Russians learned quickly that charging at the new Panthers (and Tiger tanks as well), and then...vehicle powerful enough to pull a Panther was another Panther or a Tiger tank . Without another tank stopping to retrieve the disabled vehicle, the

  6. A data-based conservation planning tool for Florida panthers

    Science.gov (United States)

    Murrow, Jennifer L.; Thatcher, Cindy A.; Van Manen, Frank T.; Clark, Joseph D.

    2013-01-01

    Habitat loss and fragmentation are the greatest threats to the endangered Florida panther (Puma concolor coryi). We developed a data-based habitat model and user-friendly interface so that land managers can objectively evaluate Florida panther habitat. We used a geographic information system (GIS) and the Mahalanobis distance statistic (D2) to develop a model based on broad-scale landscape characteristics associated with panther home ranges. Variables in our model were Euclidean distance to natural land cover, road density, distance to major roads, human density, amount of natural land cover, amount of semi-natural land cover, amount of permanent or semi-permanent flooded area–open water, and a cost–distance variable. We then developed a Florida Panther Habitat Estimator tool, which automates and replicates the GIS processes used to apply the statistical habitat model. The estimator can be used by persons with moderate GIS skills to quantify effects of land-use changes on panther habitat at local and landscape scales. Example applications of the tool are presented.

  7. Genetic characterization of feline leukemia virus from Florida panthers.

    Science.gov (United States)

    Brown, Meredith A; Cunningham, Mark W; Roca, Alfred L; Troyer, Jennifer L; Johnson, Warren E; O'Brien, Stephen J

    2008-02-01

    From 2002 through 2005, an outbreak of feline leukemia virus (FeLV) occurred in Florida panthers (Puma concolor coryi). Clinical signs included lymphadenopathy, anemia, septicemia, and weight loss; 5 panthers died. Not associated with FeLV outcome were the genetic heritage of the panthers (pure Florida vs. Texas/Florida crosses) and co-infection with feline immunodeficiency virus. Genetic analysis of panther FeLV, designated FeLV-Pco, determined that the outbreak likely came from 1 cross-species transmission from a domestic cat. The FeLV-Pco virus was closely related to the domestic cat exogenous FeLV-A subgroup in lacking recombinant segments derived from endogenous FeLV. FeLV-Pco sequences were most similar to the well-characterized FeLV-945 strain, which is highly virulent and strongly pathogenic in domestic cats because of unique long terminal repeat and envelope sequences. These unique features may also account for the severity of the outbreak after cross-species transmission to the panther.

  8. Simulations of inspiraling and merging double neutron stars using the Spectral Einstein Code

    Science.gov (United States)

    Haas, Roland; Ott, Christian D.; Szilagyi, Bela; Kaplan, Jeffrey D.; Lippuner, Jonas; Scheel, Mark A.; Barkett, Kevin; Muhlberger, Curran D.; Dietrich, Tim; Duez, Matthew D.; Foucart, Francois; Pfeiffer, Harald P.; Kidder, Lawrence E.; Teukolsky, Saul A.

    2016-06-01

    We present results on the inspiral, merger, and postmerger evolution of a neutron star-neutron star (NSNS) system. Our results are obtained using the hybrid pseudospectral-finite volume Spectral Einstein Code (SpEC). To test our numerical methods, we evolve an equal-mass system for ≈22 orbits before merger. This waveform is the longest waveform obtained from fully general-relativistic simulations for NSNSs to date. Such long (and accurate) numerical waveforms are required to further improve semianalytical models used in gravitational wave data analysis, for example, the effective one body models. We discuss in detail the improvements to SpEC's ability to simulate NSNS mergers, in particular mesh refined grids to better resolve the merger and postmerger phases. We provide a set of consistency checks and compare our results to NSNS merger simulations with the independent bam code. We find agreement between them, which increases confidence in results obtained with either code. This work paves the way for future studies using long waveforms and more complex microphysical descriptions of neutron star matter in SpEC.

  9. Review and comparison of effective delayed neutron fraction calculation methods with Monte Carlo codes

    International Nuclear Information System (INIS)

    Bécares, V.; Pérez-Martín, S.; Vázquez-Antolín, M.; Villamarín, D.; Martín-Fuertes, F.; González-Romero, E.M.; Merino, I.

    2014-01-01

    Highlights: • Review of several Monte Carlo effective delayed neutron fraction calculation methods. • These methods have been implemented with the Monte Carlo code MCNPX. • They have been benchmarked against against some critical and subcritical systems. • Several nuclear data libraries have been used. - Abstract: The calculation of the effective delayed neutron fraction, β eff , with Monte Carlo codes is a complex task due to the requirement of properly considering the adjoint weighting of delayed neutrons. Nevertheless, several techniques have been proposed to circumvent this difficulty and obtain accurate Monte Carlo results for β eff without the need of explicitly determining the adjoint flux. In this paper, we make a review of some of these techniques; namely we have analyzed two variants of what we call the k-eigenvalue technique and other techniques based on different interpretations of the physical meaning of the adjoint weighting. To test the validity of all these techniques we have implemented them with the MCNPX code and we have benchmarked them against a range of critical and subcritical systems for which either experimental or deterministic values of β eff are available. Furthermore, several nuclear data libraries have been used in order to assess the impact of the uncertainty in nuclear data in the calculated value of β eff

  10. A comparison of the COG and MCNP codes in computational neutron capture therapy modeling, Part I: boron neutron capture therapy models.

    Science.gov (United States)

    Culbertson, C N; Wangerin, K; Ghandourah, E; Jevremovic, T

    2005-08-01

    The goal of this study was to evaluate the COG Monte Carlo radiation transport code, developed and tested by Lawrence Livermore National Laboratory, for neutron capture therapy related modeling. A boron neutron capture therapy model was analyzed comparing COG calculational results to results from the widely used MCNP4B (Monte Carlo N-Particle) transport code. The approach for computing neutron fluence rate and each dose component relevant in boron neutron capture therapy is described, and calculated values are shown in detail. The differences between the COG and MCNP predictions are qualified and quantified. The differences are generally small and suggest that the COG code can be applied for BNCT research related problems.

  11. Artus code. Calculation of the energy transfer from neutrons to a solid; Programme Artus. Calcul des quantites d'energie cedees par des neutrons a un solide

    Energy Technology Data Exchange (ETDEWEB)

    Barre, B

    1967-07-01

    The described codes have been realized for a particular study concerning the ionization influence of the CO{sub 2} by neutrons on the reaction: CO{sub 2} - graphite. The code hypothesis and formulation are presented and the application to more general physical problems is proposed. (A.L.B.)

  12. Artus code. Calculation of the energy transfer from neutrons to a solid; Programme Artus. Calcul des quantites d'energie cedees par des neutrons a un solide

    Energy Technology Data Exchange (ETDEWEB)

    Barre, B

    1967-07-01

    The described codes have been realized for a particular study concerning the ionization influence of the CO{sub 2} by neutrons on the reaction: CO{sub 2} - graphite. The code hypothesis and formulation are presented and the application to more general physical problems is proposed. (A.L.B.)

  13. Validation of coupled neutronic / thermal-hydraulic codes for VVER reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Mittag, S.; Grundmann, U.; Kliem, S.; Kozmenkov, Y.; Rindelhardt, U.; Rohde, U.; Weiss, F.-P.; Langenbuch, S.; Krzykacz-Hausmann, B.; Schmidt, K.-D.; Vanttola, T.; Haemaelaeinen, A.; Kaloinen, E.; Kereszturi, A.; Hegyi, G.; Panka, I.; Hadek, J.; Strmensky, C.; Darilek, P.; Petkov, P.; Stefanova, S.; Kuchin, A.; Khalimonchuk, V.; Hlbocky, P.; Sico, D.; Danilin, S.; Ionov, V.; Nikonov, S.; Powney, D.

    2004-08-01

    In recent years, the simulation methods for the safety analysis of nuclear power plants have been continuously improved to perform realistic calculations. Therefore in VALCO work package 2 (WP 2), the usual application of coupled neutron-kinetic / thermal-hydraulic codes to VVER has been supplemented by systematic uncertainty and sensitivity analyses. A comprehensive uncertainty analysis has been carried out. The GRS uncertainty and sensitivity method based on the statistical code package SUSA was applied to the two transients studied earlier in SRR-1/95: A load drop of one turbo-generator in Loviisa-1 (VVER-440), and a switch-off of one feed water pump in Balakovo-4 (VVER-1000). The main steps of these analyses and the results obtained by applying different coupled code systems (SMABRE - HEXTRAN, ATHLET - DYN3D, ATHLET - KIKO3D, ATHLET - BIPR-8) are described in this report. The application of this method is only based on variations of input parameter values. No internal code adjustments are needed. An essential result of the analysis using the GRS SUSA methodology is the identification of the input parameters, such as the secondary-circuit pressure, the control-assembly position (as a function of time), and the control-assembly efficiency, that most sensitively affect safety-relevant output parameters, like reactor power, coolant heat-up, and primary pressure. Uncertainty bands for these output parameters have been derived. The variation of potentially uncertain input parameter values as a consequence of uncertain knowledge can activate system actions causing quite different transient evolutions. This gives indications about possible plant conditions that might be reached from the initiating event assuming only small disturbances. In this way, the uncertainty and sensitivity analysis reveals the spectrum of possible transient evolutions. Deviations of SRR-1/95 coupled code calculations from measurements also led to the objective to separate neutron kinetics from

  14. Evaluating the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz-Rodriguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solis Sanches, L. O.; Miranda, R. Castaneda; Cervantes Viramontes, J. M. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica. Av. Ramon Lopez Velarde 801. Col. Centro Zacatecas, Zac (Mexico); Vega-Carrillo, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica. Av. Ramon Lopez Velarde 801. Col. Centro Zacatecas, Zac., Mexico. and Unidad Academica de Estudios Nucleares. C. Cip (Mexico)

    2013-07-03

    In this work the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks is evaluated. The first one code based on traditional iterative procedures and called Neutron spectrometry and dosimetry from the Universidad Autonoma de Zacatecas (NSDUAZ) use the SPUNIT iterative algorithm and was designed to unfold neutron spectrum and calculate 15 dosimetric quantities and 7 IAEA survey meters. The main feature of this code is the automated selection of the initial guess spectrum trough a compendium of neutron spectrum compiled by the IAEA. The second one code known as Neutron spectrometry and dosimetry with artificial neural networks (NDSann) is a code designed using neural nets technology. The artificial intelligence approach of neural net does not solve mathematical equations. By using the knowledge stored at synaptic weights on a neural net properly trained, the code is capable to unfold neutron spectrum and to simultaneously calculate 15 dosimetric quantities, needing as entrance data, only the rate counts measured with a Bonner spheres system. Similarities of both NSDUAZ and NSDann codes are: they follow the same easy and intuitive user's philosophy and were designed in a graphical interface under the LabVIEW programming environment. Both codes unfold the neutron spectrum expressed in 60 energy bins, calculate 15 dosimetric quantities and generate a full report in HTML format. Differences of these codes are: NSDUAZ code was designed using classical iterative approaches and needs an initial guess spectrum in order to initiate the iterative procedure. In NSDUAZ, a programming routine was designed to calculate 7 IAEA instrument survey meters using the fluence-dose conversion coefficients. NSDann code use artificial neural networks for solving the ill-conditioned equation system of neutron spectrometry problem through synaptic weights of a properly trained neural network. Contrary to iterative procedures, in

  15. Evaluating the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks

    Science.gov (United States)

    Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solís Sánches, L. O.; Miranda, R. Castañeda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

    2013-07-01

    In this work the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks is evaluated. The first one code based on traditional iterative procedures and called Neutron spectrometry and dosimetry from the Universidad Autonoma de Zacatecas (NSDUAZ) use the SPUNIT iterative algorithm and was designed to unfold neutron spectrum and calculate 15 dosimetric quantities and 7 IAEA survey meters. The main feature of this code is the automated selection of the initial guess spectrum trough a compendium of neutron spectrum compiled by the IAEA. The second one code known as Neutron spectrometry and dosimetry with artificial neural networks (NDSann) is a code designed using neural nets technology. The artificial intelligence approach of neural net does not solve mathematical equations. By using the knowledge stored at synaptic weights on a neural net properly trained, the code is capable to unfold neutron spectrum and to simultaneously calculate 15 dosimetric quantities, needing as entrance data, only the rate counts measured with a Bonner spheres system. Similarities of both NSDUAZ and NSDann codes are: they follow the same easy and intuitive user's philosophy and were designed in a graphical interface under the LabVIEW programming environment. Both codes unfold the neutron spectrum expressed in 60 energy bins, calculate 15 dosimetric quantities and generate a full report in HTML format. Differences of these codes are: NSDUAZ code was designed using classical iterative approaches and needs an initial guess spectrum in order to initiate the iterative procedure. In NSDUAZ, a programming routine was designed to calculate 7 IAEA instrument survey meters using the fluence-dose conversion coefficients. NSDann code use artificial neural networks for solving the ill-conditioned equation system of neutron spectrometry problem through synaptic weights of a properly trained neural network. Contrary to iterative procedures, in neural

  16. Evaluating the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks

    International Nuclear Information System (INIS)

    Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solís Sánches, L. O.; Miranda, R. Castañeda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

    2013-01-01

    In this work the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks is evaluated. The first one code based on traditional iterative procedures and called Neutron spectrometry and dosimetry from the Universidad Autonoma de Zacatecas (NSDUAZ) use the SPUNIT iterative algorithm and was designed to unfold neutron spectrum and calculate 15 dosimetric quantities and 7 IAEA survey meters. The main feature of this code is the automated selection of the initial guess spectrum trough a compendium of neutron spectrum compiled by the IAEA. The second one code known as Neutron spectrometry and dosimetry with artificial neural networks (NDSann) is a code designed using neural nets technology. The artificial intelligence approach of neural net does not solve mathematical equations. By using the knowledge stored at synaptic weights on a neural net properly trained, the code is capable to unfold neutron spectrum and to simultaneously calculate 15 dosimetric quantities, needing as entrance data, only the rate counts measured with a Bonner spheres system. Similarities of both NSDUAZ and NSDann codes are: they follow the same easy and intuitive user's philosophy and were designed in a graphical interface under the LabVIEW programming environment. Both codes unfold the neutron spectrum expressed in 60 energy bins, calculate 15 dosimetric quantities and generate a full report in HTML format. Differences of these codes are: NSDUAZ code was designed using classical iterative approaches and needs an initial guess spectrum in order to initiate the iterative procedure. In NSDUAZ, a programming routine was designed to calculate 7 IAEA instrument survey meters using the fluence-dose conversion coefficients. NSDann code use artificial neural networks for solving the ill-conditioned equation system of neutron spectrometry problem through synaptic weights of a properly trained neural network. Contrary to iterative procedures, in neural

  17. Test and validation of the iterative code for the neutrons spectrometry and dosimetry: NSDUAZ; Prueba y validacion del codigo iterativo para la espectrometria y dosimetria de neutrones: NSDUAZ

    Energy Technology Data Exchange (ETDEWEB)

    Reyes H, A.; Ortiz R, J. M.; Reyes A, A.; Castaneda M, R.; Solis S, L. O.; Vega C, H. R., E-mail: alfredo_reyesh@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica, Av. Lopez Velarde 801, Col. Centro, 98000 Zacatecas (Mexico)

    2014-08-15

    In this work was realized the test and validation of an iterative code for neutronic spectrometry known as Neutron Spectrometry and Dosimetry of the Universidad Autonoma de Zacatecas (NSDUAZ). This code was designed in a user graph interface, friendly and intuitive in the environment programming of LabVIEW using the iterative algorithm known as SPUNIT. The main characteristics of the program are: the automatic selection of the initial spectrum starting from the neutrons spectra catalog compiled by the International Atomic Energy Agency, the possibility to generate a report in HTML format that shows in graph and numeric way the neutrons flowing and calculates the ambient dose equivalent with base to this. To prove the designed code, the count rates of a spectrometer system of Bonner spheres were used with a detector of {sup 6}LiI(Eu) with 7 polyethylene spheres with diameter of 0, 2, 3, 5, 8, 10 and 12. The count rates measured with two neutron sources: {sup 252}Cf and {sup 239}PuBe were used to validate the code, the obtained results were compared against those obtained using the BUNKIUT code. We find that the reconstructed spectra present an error that is inside the limit reported in the literature that oscillates around 15%. Therefore, it was concluded that the designed code presents similar results to those techniques used at the present time. (Author)

  18. BLINDAGE: A neutron and gamma-ray transport code for shieldings with the removal-diffusion technique coupled with the point-kernel technique

    International Nuclear Information System (INIS)

    Fanaro, L.C.C.B.

    1984-01-01

    It was developed the BLINDAGE computer code for the radiation transport (neutrons and gammas) calculation. The code uses the removal - diffusion method for neutron transport and point-kernel technique with buil-up factors for gamma-rays. The results obtained through BLINDAGE code are compared with those obtained with the ANISN and SABINE computer codes. (Author) [pt

  19. Description of the CAREM Reactor Neutronic Calculation Codes; Descripcion de la Linea de Calculo Neutronica del CAREM

    Energy Technology Data Exchange (ETDEWEB)

    Villarino, Eduardo; Hergenreder, Daniel [Investigacion Aplicada, INVAP, San Carlos de Bariloche (Argentina)

    2000-07-01

    In this work is described the neutronic calculation line used to design the CAREM reactor.A description of the codes used and the interfaces between the different programs are presented.Both, the normal calculation line and the alternative or verification calculation line are included.The calculation line used to obtain the kinetics parameters (effective delayed-neutron fraction and prompt-neutron lifetime) is also included.

  20. Modeling of LVRF Critical Experiments in ZED-2 Using WIMS9A/PANTHER and MCNP5

    International Nuclear Information System (INIS)

    Sissaoui, M.T.; Lebenhaft, J.R; Carlson, P.A.

    2008-01-01

    The accuracy of WIMS9A/PANTHER and MCNP5 in modeling D 2 O-moderated, and H 2 O-, D 2 O- or air-cooled, doubly heterogeneous lattices of fuel clusters was demonstrated using Low Void Reactivity Fuel (LVRF) substitution experiments in the ZED-2 critical facility. MCNP5 with ENDF/B-VI (Release 5) under-predicted k eff but gave excellent coolant void reactivity (CVR) bias values. WIMS9A/PANTHER with JEF-2.2 over-predicted k eff and under-predicted the CVR bias relative to MCNP5 by 100 pcm to 200 pcm. Both codes reproduced the measured axial and radial flux shapes accurately. (authors)

  1. Verification of 3-D generation code package for neutronic calculations of WWERs

    International Nuclear Information System (INIS)

    Sidorenko, V.D.; Aleshin, S.S.; Bolobov, P.A.; Bolshagin, S.N.; Lazarenko, A.P.; Markov, A.V.; Morozov, V.V.; Syslov, A.A.; Tsvetkov, V.M.

    2000-01-01

    Materials on verification of the 3 -d generation code package for WWERs neutronic calculations are presented. The package includes: - spectral code TVS-M; - 2-D fine mesh diffusion code PERMAK-A for 4- or 6-group calculation of WWER core burnup; - 3-D coarse mesh diffusion code BIPR-7A for 2-group calculations of quasi-stationary WWERs regimes. The materials include both TVS-M verification data and verification data on PERMAK-A and BIPR-7A codes using constant libraries generated with TVS-M. All materials are related to the fuel without Gd. TVS-M verification materials include results of comparison both with benchmark calculations obtained by other codes and with experiments carried out at ZR-6 critical facility. PERMAK-A verification materials contain results of comparison with TVS-M calculations and with ZR-6 experiments. BIPR-7A materials include comparison with operation data for Dukovany-2 and Loviisa-1 NPPs (WWER-440) and for Balakovo NPP Unit 4 (WWER-1000). The verification materials demonstrate rather good accuracy of calculations obtained with the use of code package of the 3 -d generation. (Authors)

  2. Sizewell B cycle 5 core design with Framatome ANP's CASCADE-3D and British Energy's PANTHER

    International Nuclear Information System (INIS)

    Attale, F.; Koegl, J.; Knight, M.; Bryce, P.

    2001-01-01

    Sizewell B Cycle 5 is the first cycle, after 4 cycles with BNFL fuel, with a reload consisting of Framatome ANP HTP (high thermal performance) fuel assemblies. The impact of this fuel vendor change on the Nuclear Design area is that, according to British energy's (BE) practice, the Framatome ANP's nuclear design code system CASCADE-3D is used for the majority of the cycle specific safety case calculations. However, other parts of the safety submission (e.g. 3D transient analyses) are made by using the BE code PANTHER. Before using in parallel two different code systems for reload core licensing extensive comparisons of applied methodologies and obtained results were required to ensure an acceptable level of agreement. (orig.)

  3. Summary report for ITER task - D10: Update and implementation of neutron transport and activation codes and processed libraries

    International Nuclear Information System (INIS)

    Attaya, H.

    1995-01-01

    The primary goal of this task is to provide the capabilities in the activation code RACC, to treat pulsed operation modes. In addition, it is required that the code utilizes the same spatial mesh and geometrical models as employed in the one or multidimensional neutron transport codes used in ITER design. This would ensure the use of the same neutron flux generated by those codes to calculate the different activation parameters. It is also required to have the capabilities for generating graphical outputs for the calculated activation parameters

  4. JENDL-4.0 benchmarking for effective delayed neutron fraction with a continuous-energy Monte Carlo code MVP

    International Nuclear Information System (INIS)

    Nagaya, Yasunobu

    2013-01-01

    Benchmark calculations with a continuous-energy Monte Carlo code have been performed for delayed neutron data of JENDL-4.0. JENDL-4.0 gives good prediction for the effective delayed neutron fraction in the present benchmarks but further detailed analysis is required for some cores. (author)

  5. Calculation of the fast neutron flux spectrum in the MNSR inner irradiation site using the WIMSD4 code

    International Nuclear Information System (INIS)

    Khattab, K.

    2005-03-01

    The Miniature Neutron Source Reactor (MNSR) in Syria has five inner irradiation sites in the annulus Beryllium reflectors to analyze the unknown samples using the Neutron Activation Analysis technique and to produce medium and short half life isotopes. The fast neutron flux spectrum has a special importance in the MNSR reactor physics where this spectrum is required to measure the fast neutron flux in the MNSR inner irradiation sites. Hence, calculation of the fast neutron flux spectrum in the MNSR inner irradiation site is conducted in this work using the WIMSD4 code. The energy range is divided in the WIMSD4 to 69 energy groups. The first six energy groups represent the fast neutron ranging from 0.5 to 10 MeV. To calculate the fast neutron flux spectrum in the MNSR inner irradiation site using the WIMSD4 code, the MNSR is modeled as a super unit cell. This cell consists of three regions which are: the homogenized core, annulus Beryllium, and water. The fast neutron spectrum is calculated also using the U 235 fission neutron spectrum approximation. The U 235 fission neutron spectrum agrees very good with the WIMSD4 results when neutron energy exceeds 1 MeV, but it fails when the neutron energy ranges from 0.5 to 1 MeV. The WIMSD4 code is used as well to calculate the microscopic fission cross sections for the U 238 using six energy groups where a unit cell of U 238 is used since the U 238 is usually used to measure the fast neutron flux in the reactor. The macroscopic fission cross sections for the U 238 are calculated first then the microscopic fission cross sections are calculated knowing the U 238 atomic density. (Author)

  6. Calculation of the fast neutron flux spectrum in the MNSR inner irradiation site using the WIMSD4 code

    International Nuclear Information System (INIS)

    Khattab, K.

    2006-01-01

    The Miniature Neutron Source Reactor (MNSR) in Syria has five inner irradiation sites in the annulus Beryllium reflectors to analyze the unknown samples using the Neutron Activation Analysis technique and to produce medium and short half life isotopes. The fast neutron flux spectrum has a special importance in the MNSR reactor physics where this spectrum is required to measure the fast neutron flux in the MNSR inner irradiation sites. Hence, calculation of the fast neutron flux spectrum in the MNSR inner irradiation site is conducted in this work using the WIMSD4 code. The energy range is divided in the WIMSD4 to 69 energy groups. The first six energy groups represent the fast neutron ranging from 0.5 to 10 MeV. To calculate the fast neutron flux spectrum in the MNSR inner irradiation site using the WIMSD4 code, the MNSR is modeled as a super unit cell. This cell consists of three regions which are: the homogenized core, annulus Beryllium, and water. The fast neutron spectrum is calculated also using the U 235 fission neutron spectrum approximation. The U 235 fission neutron spectrum agrees very good with the WIMSD4 results when neutron energy exceeds 1 MeV, but it fails when the neutron energy ranges from 0.5 to 1 MeV. The WIMSD4 code is used as well to calculate the microscopic fission cross sections for the U 238 using six energy groups where a unit cell of U 238 is used since the U 238 is usually used to measure the fast neutron flux in the reactor. The macroscopic fission cross sections for the U 238 are calculated first then the microscopic fission cross sections are calculated knowing the U 238 atomic density. (Author)

  7. Angular resolution study of a combined gamma-neutron coded aperture imager for standoff detection

    International Nuclear Information System (INIS)

    Ayaz-Maierhafer, Birsen; Hayward, Jason P.; Ziock, Klaus P.; Blackston, Matthew A.; Fabris, Lorenzo

    2013-01-01

    Nuclear threat source observables at standoff distances of tens of meters from mCi class sources include both gamma-rays and neutrons. This work uses simulations to investigate the effects of the angular resolution of a mobile gamma-ray and neutron coded aperture imaging system upon orphan source detection significance and specificity. The design requires maintaining high sensitivity and specificity while keeping the system size as compact as possible to reduce weight, footprint, and cost. A mixture of inorganic and organic scintillators was considered in the detector plane for high sensitivity to both gamma-rays and fast neutrons. For gamma-rays (100 to 2500 keV) and fission spectrum neutrons, angular resolutions of 1–9° and radiation angles of incidence appropriate for mobile search were evaluated. Detection significance for gamma-rays considers those events that contribute to the photopeak of the image pixel corresponding the orphan source location. For detection of fission spectrum neutrons, energy depositions above a set pulse shape discrimination threshold were tallied. The results show that the expected detection significance for the system at an angular resolution of 1° is significantly lower compared to its detection significance an angular resolution of ∼3–4°. An angular resolution of ∼3–4° is recommended both for better detection significance and improved false alarm rate, considering that finer angular resolution does not result in improved background rejection when the coded aperture method is used. Instead, over-pixelating the search space may result in an unacceptably high false alarm rate

  8. EPRI-LATTICE: a multigroup neutron transport code for light water reactor lattice physics calculations

    International Nuclear Information System (INIS)

    Jones, D.B.

    1986-01-01

    EPRI-LATTICE is a multigroup neutron transport computer code for the analysis of light water reactor fuel assemblies. It can solve the two-dimensional neutron transport problem by two distinct methods: (a) the method of collision probabilities and (b) the method of discrete ordinates. The code was developed by S. Levy Inc. as an account of work sponsored by the Electric Power Research Institute (EPRI). The collision probabilities calculation in EPRI-LATTICE (L-CP) is based on the same methodology that exists in the lattice codes CPM-2 and EPRI-CPM. Certain extensions have been made to the data representations of the CPM programs to improve the overall accuracy of the calculation. The important extensions include unique representations of scattering matrices and fission fractions (chi) for each composition in the problem. A new capability specifically developed for the EPRI-LATTICE code is a discrete ordinates methodology. The discrete ordinates calculation in EPRI-LATTICE (L-SN) is based on the discrete S/sub n/ methodology that exists in the TWODANT program. In contrast to TWODANT, which utilizes synthetic diffusion acceleration and supports multiple geometries, only the transport equations are solved by L-SN and only the data representations for the two-dimensional geometry are treated

  9. Application of the three-dimensional transport code to analysis of the neutron streaming experiment

    International Nuclear Information System (INIS)

    Chatani, K.; Slater, C.O.

    1990-01-01

    The neutron streaming through an experimental mock-up of a Clinch River Breeder Reactor (CRBR) prototypic coolant pipe chaseway was recalculated with a three-dimensional discrete ordinates code. The experiment was conducted at the Tower Shielding Facility at Oak Ridge National Laboratory in 1976 and 1977. The measurement of the neutron flux, using Bonner ball detectors, indicated nine orders of attenuation in the empty pipeway, which contained two 90-deg bends and was surrounded by concrete walls. The measurement data were originally analyzed using the DOT3.5 two-dimensional discrete ordinates radiation transport code. However, the results did not agree with measurement data at the bend because of the difficulties in modeling the three-dimensional configurations using two-dimensional methods. The two-dimensional calculations used a three-step procedure in which each of the three legs making the two 90-deg bends was a separate calculation. The experiment was recently analyzed with the TORT three-dimensional discrete ordinates radiation transport code, not only to compare the calculational results with the experimental results, but also to compare with results obtained from analyses in Japan using DOT3.5, MORSE, and ENSEMBLE, which is a three-dimensional discrete ordinates radiation transport code developed in Japan

  10. NADAC and MERGE: computer codes for processing neutron activation analysis data

    International Nuclear Information System (INIS)

    Heft, R.E.; Martin, W.E.

    1977-01-01

    Absolute disintegration rates of specific radioactive products induced by neutron irradition of a sample are determined by spectrometric analysis of gamma-ray emissions. Nuclide identification and quantification is carried out by a complex computer code GAMANAL (described elsewhere). The output of GAMANAL is processed by NADAC, a computer code that converts the data on observed distintegration rates to data on the elemental composition of the original sample. Computations by NADAC are on an absolute basis in that stored nuclear parameters are used rather than the difference between the observed disintegration rate and the rate obtained by concurrent irradiation of elemental standards. The NADAC code provides for the computation of complex cases including those involving interrupted irradiations, parent and daughter decay situations where the daughter may also be produced independently, nuclides with very short half-lives compared to counting interval, and those involving interference by competing neutron-induced reactions. The NADAC output consists of a printed report, which summarizes analytical results, and a card-image file, which can be used as input to another computer code MERGE. The purpose of MERGE is to combine the results of multiple analyses and produce a single final answer, based on all available information, for each element found

  11. MCNP: a general Monte Carlo code for neutron and photon transport

    International Nuclear Information System (INIS)

    1979-11-01

    The general-purpose Monte Carlo code MCNP ca be used for neutron, photon, or coupled neutron-photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces (elliptical tori). Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation are accounted for. Thermal neutrons are described by both the free-gas and S(α,β) models. For photons, the code takes account of incoherent and coherent scattering, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. MCNP includes an elaborate, interactive plotting capability that allows the user to view his input geometry to help check for setup errors. Standard features which are available to improve computational efficiency include geometry splitting and Russian roulette, weight cutoff with Russian roulette, correlated sampling, analog capture or capture by weight reduction, the exponential transformation, energy splitting, forced collisions in designated cells, flux estimates at point or ring detectors, deterministically transporting pseudo-particles to designated regions, track-length estimators, source biasing, and several parameter cutoffs. Extensive summary information is provided to help the user better understand the physics and Monte Carlo simulation of his problem. The standard, user-defined output of MCNP includes two-way current as a function of direction across any set of surfaces or surface segments in the problem. Flux across any set of surfaces or surface segments is available. 58 figures, 28 tables

  12. NULIF: neutron spectrum generator, few-group constant calculator, and fuel depletion code

    International Nuclear Information System (INIS)

    Wittkopf, W.A.; Tilford, J.M.; Andrews, J.B. II; Kirschner, G.; Hassan, N.M.; Colpo, P.N.

    1977-02-01

    The NULIF code generates a microgroup neutron spectrum and calculates spectrum-weighted few-group parameters for use in a spatial diffusion code. A wide variety of fuel cells, non-fuel cells, and fuel lattices, typical of PWR (or BWR) lattices, are treated. A fuel depletion routine and change card capability allow a broad range of problems to be studied. Coefficient variation with fuel burnup, fuel temperature change, moderator temperature change, soluble boron concentration change, burnable poison variation, and control rod insertion are readily obtained. Heterogeneous effects, including resonance shielding and thermal flux depressions, are treated. Coefficients are obtained for one thermal group and up to three epithermal groups. A special output routine writes the few-group coefficient data in specified format on an output tape for automated fitting in the PDQ07-HARMONY system of spatial diffusion-depletion codes

  13. 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

  14. NEWSPEC: A computer code to unfold neutron spectra from Bonner sphere data

    International Nuclear Information System (INIS)

    Lemley, E.C.; West, L.

    1996-01-01

    A new computer code, NEWSPEC, is in development at the University of Arkansas. The NEWSPEC code allows a user to unfold, fold, rebin, display, and manipulate neutron spectra as applied to Bonner sphere measurements. The SPUNIT unfolding algorithm, a new rebinning algorithm, and the graphical capabilities of Microsoft (MS) Windows and MS Excel are utilized to perform these operations. The computer platform for NEWSPEC is a personal computer (PC) running MS Windows 3.x or Win95, while the code is written in MS Visual Basic (VB) and MS VB for Applications (VBA) under Excel. One of the most useful attributes of the NEWSPEC software is the link to Excel allowing additional manipulation of program output or creation of program input

  15. 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

  16. 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.

  17. Neutron dose rate analysis on HTGR-10 reactor using Monte Carlo code

    Science.gov (United States)

    Suwoto; Adrial, H.; Hamzah, A.; Zuhair; Bakhri, S.; Sunaryo, G. R.

    2018-02-01

    The HTGR-10 reactor is cylinder-shaped core fuelled with kernel TRISO coated fuel particles in the spherical pebble with helium cooling system. The outlet helium gas coolant temperature outputted from the reactor core is designed to 700 °C. One advantage HTGR type reactor is capable of co-generation, as an addition to generating electricity, the reactor was designed to produce heat at high temperature can be used for other processes. The spherical fuel pebble contains 8335 TRISO UO2 kernel coated particles with enrichment of 10% and 17% are dispersed in a graphite matrix. The main purpose of this study was to analysis the distribution of neutron dose rates generated from HTGR-10 reactors. The calculation and analysis result of neutron dose rate in the HTGR-10 reactor core was performed using Monte Carlo MCNP5v1.6 code. The problems of double heterogeneity in kernel fuel coated particles TRISO and spherical fuel pebble in the HTGR-10 core are modelled well with MCNP5v1.6 code. The neutron flux to dose conversion factors taken from the International Commission on Radiological Protection (ICRP-74) was used to determine the dose rate that passes through the active core, reflectors, core barrel, reactor pressure vessel (RPV) and a biological shield. The calculated results of neutron dose rate with MCNP5v1.6 code using a conversion factor of ICRP-74 (2009) for radiation workers in the radial direction on the outside of the RPV (radial position = 220 cm from the center of the patio HTGR-10) provides the respective value of 9.22E-4 μSv/h and 9.58E-4 μSv/h for enrichment 10% and 17%, respectively. The calculated values of neutron dose rates are compliant with BAPETEN Chairman’s Regulation Number 4 Year 2013 on Radiation Protection and Safety in Nuclear Energy Utilization which sets the limit value for the average effective dose for radiation workers 20 mSv/year or 10μSv/h. Thus the protection and safety for radiation workers to be safe from the radiation source has

  18. Primary Displacement Damage Calculation Induced by Neutron and Ion Using Binary Collision Approximation Techniques (Marlowe Code)

    International Nuclear Information System (INIS)

    Mota, F.; Ortiz, C. J.; Vila, R.

    2012-01-01

    Irradiation Experimental Area of TechnoFusion will emulate the extreme irradiation fusion conditions in materials by means of three ion accelerators: one used for self-implanting heavy ions (Fe, Si, C,...) to emulate the displacement damage induced by fusion neutrons and the other two for light ions (H and He) to emulate the transmutation induced by fusion neutrons. This Laboratory will play an essential role in the selection of functional materials for DEMO reactor since it will allow reproducing the effects of neutron radiation on fusion materials. Ion irradiation produces little or no residual radioactivity, allowing handling of samples without the need for special precautions. Currently, two different methods are used to calculate the primary displacement damage by neutron irradiation or by ion irradiation. On one hand, the displacement damage doses induced by neutrons are calculated considering the NRT model based on the electronic screening theory of Linhard. This methodology is commonly used since 1975. On the other hand, for experimental research community the SRIM code is commonly used to calculate the primary displacement damage dose induced by ion irradiation. Therefore, both methodologies of primary displacement damage calculation have nothing in common. However, if we want to design ion irradiation experiments capable to emulate the neutron fusion effect in materials, it is necessary to develop comparable methodologies of damage calculation for both kinds of radiation. It would allow us to define better the ion irradiation parameters (Ion, current, Ion energy, dose, etc) required to emulate a specific neutron irradiation environment. Therefore, our main objective was to find the way to calculate the primary displacement damage induced by neutron irradiation and by ion irradiation starting from the same point, that is, the PKA spectrum. In order to emulate the neutron irradiation that would prevail under fusion conditions, two approaches are contemplated: a) on

  19. Investigation of the energy correlations of spallation neutrons by the MCNPX code

    International Nuclear Information System (INIS)

    Szieberth, Mate; Radocz, Gabor

    2011-01-01

    Earlier works have suggested that the energy correlations in a spallation source may influence the neutron noise measurements in an ADS. For the calculation of this effect not only the generally known and used one-particle spectrum is needed but also the so-called two particle spectrum, which describes also the energy correlations. Since measured data are not available for the energy distribution of the neutrons from a single spallation event the physical models of the MCNPX code have been used to investigate the effect. The calculational model has been successfully validated with measurements of the number distribution of spallation neutrons. The simulated one- and two-particle energy distributions and spectra proved that the energy correlations exist and have an important effect in low multiplicity spallation events and in thin targets. On the other hand for thick targets this effect appears negligible and the factorization of the two-particle spectrum seems an acceptable approximation. Further investigations are in hand to quantify the actual effect of the energy correlations on the neutron noise measurements. (author)

  20. The development of the Nuclear Electric core performance and fault transient analysis code package in support of Sizewell B

    International Nuclear Information System (INIS)

    Hall, P.; Hutt, P.

    1994-01-01

    This paper describes Nuclear Electric's (NE) development of an integrated code package in support of all its reactors including Sizewell B, designed for the provision of fuel management design, core performance studies, operational support and fault transient analysis. The package uses the NE general purpose three-dimensional transient reactor physics code PANTHER with cross-sections derived in the PWR case from the LWRWIMS LWR lattice neutronics code. The package also includes ENIGMA a generic fuel performance code and for PWR application VIPRE-01 a subchannel thermal hydraulics code, RELAP5 the system thermal hydraulics transient code and SCORPIO an on-line surveillance system. The paper describes the capabilities and validation of the elements of this package for PWR, how they are coupled within the package and the way in which they are being applied for Sizewell B to on-line surveillance and fault transient analysis. (Author)

  1. Development of SCINFUL-CG code to calculate response functions of scintillators in various shapes used for neutron measurement

    Energy Technology Data Exchange (ETDEWEB)

    Endo, Akira; Kim, Eunjoo; Yamaguchi, Yasuhiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-10-01

    A Monte Carlo code SCINFUL has been utilized for calculating response functions of organic scintillators for high-energy neutron spectroscopy. However, the applicability of SCINFUL is limited to the calculations for cylindrical NE213 and NE110 scintillators. In the present study, SCINFUL-CG was developed by introducing a geometry specifying function and high-energy neutron cross section data into SCINFUL. The geometry package MARS-CG, the extended version of the CG (Combinatorial Geometry), was programmed into SCINFUL-CG to express various geometries of detectors. Neutron spectra in the regions specified by the CG can be evaluated by the track length estimator. The cross section data of silicon, oxygen and aluminum for neutron transport calculation were incorporated up to 100 MeV using the data of LA150 library. Validity of SCINFUL-CG was examined by comparing calculated results with those by SCINFUL and MCNP and experimental data measured using high-energy neutron fields. SCINFUL-CG can be used for the calculations of the response functions and neutron spectra in the organic scintillators in various shapes. The computer code will be applicable to the designs of high-energy neutron spectrometers and neutron monitors using the organic scintillators. The present report describes the new features of SCINFUL-CG and explains how to use the code. (author)

  2. EBR-II Static Neutronic Calculations by PHISICS / MCNP6 codes

    Energy Technology Data Exchange (ETDEWEB)

    Paolo Balestra; Carlo Parisi; Andrea Alfonsi

    2016-02-01

    The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) on the Shutdown Heat Removal Tests (SHRT) performed in the '80s at the Experimental fast Breeder Reactor EBR-II, USA. The scope of the CRP is to improve and validate the simulation tools for the study and the design of the liquid metal cooled fast reactors. Moreover, training of the next generation of fast reactor analysts is being also considered the other scope of the CRP. In this framework, a static neutronic model was developed, using state-of-the art neutron transport codes like SCALE/PHISICS (deterministic solution) and MCNP6 (stochastic solution). Comparison between both solutions is briefly illustrated in this summary.

  3. A new philosophy for calibrating oil well logging tools based on neutron transport codes

    International Nuclear Information System (INIS)

    Butler, J.; Clayton, C.G.

    1984-01-01

    The current practice of calibrating neutron borehole logging probes is limited by an inability to match calibration conditions to those which pertain in an operational situation. In addition, test boreholes are expensive to construct and, when natural materials are used, rely on an exact correspondence in composition and in structure between the materials of the test facility and representative samples which may not be valid. Now that neutron tansport codes have been developed to a point at which they are able to cope with realistic, complex situations an alternative approach to calibration can be considered. The basis of this philosophy is the construction of a limited number of calibration facilities which are composed of artificial rocks of controlled but variable porosity and accurately known nuclear characteristics

  4. Contribution to the validation of the Apollo code library for thermal neutron reactors

    International Nuclear Information System (INIS)

    Tellier, H.; Van der Gucht, C.; Vanuxeem, J.

    1988-03-01

    The neutron nuclear data which are needed by reactor physicists to perform core calculation are brought together in the evaluated files. The files are processed to provide multigroup cross sections. The accuracy of the core calculations depends on the initial data which are sometimes not accurate enough. Therefore the reactor physicists carry out integral experiments. We show in this paper, how the use of these integral experiments and the application of the tendency research method can improve the accuracy of the neutron data. This technique was applied to the validation of the Apollo code library. For this purpose 60 buckling measurements (34 for uranium fuel multiplying media and 26 for plutonium fuel multiplying media) and 42 spent fuel analysis were used. Small modifications of the initial data are proposed. The final values are compared which recent recommended values of microscopic data and the agreement is good [fr

  5. Development of Unified Code for Environmental Research by Neutron Activation Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Seung Yeon; Kim, Young Sik; Lee, Sang Mi; Chung, Sang Uk; Lee, Kyu Sung; Kang, Sang Hun; Cheon, Ki Hong [Yonsei University, Seoul (Korea, Republic of)

    1997-07-01

    Three codes were developed to improve accuracy and precision of neutron activation analysis with the adoption of IAEA`s recommended `GANAAS` program which has the better peak identification and efficiency calibration algorithm than the currently using commercial program. Quantitative analytical ability of trace element was improved with the codes such that the number of detectable elements including environmentally important elements was increased. Small and over lapped peaks can be detected more efficiently with the good peak shape calibration(energy dependence on peak height, peak base width and FWHM). Several efficiency functions were added to determine the detector efficiency more accurately which was the main source of error in neutron activation analysis. Errors caused by nuclear data themselves were reduced with the introduction of ko method. New graphical program called `POWER NAA` was developed for the recent personal computer environment, Window 95, and for the data compatibility. It also reduced the error caused by operator`s mistake with the easy and comfortable operation of the code. 11 refs., 3 tabs., 9 figs. (author)

  6. ARCADIAR - A New Generation of Coupled Neutronics / Core Thermal- Hydraulics Code System at AREVA NP

    International Nuclear Information System (INIS)

    Curca-Tivig, Florin; Merk, Stephan; Pautz, Andreas; Thareau, Sebastien

    2007-01-01

    Anticipating future needs of our customers and willing to concentrate synergies and competences existing in the company for the benefit of our customers, AREVA NP decided in 2002 to develop the next generation of coupled neutronics/ core thermal-hydraulic (TH) code systems for fuel assembly and core design calculations for both, PWR and BWR applications. The global CONVERGENCE project was born: after a feasibility study of one year (2002) and a conceptual phase of another year (2003), development was started at the beginning of 2004. The present paper introduces the CONVERGENCE project, presents the main feature of the new code system ARCADIA R and concludes on customer benefits. ARCADIA R is designed to meet AREVA NP market and customers' requirements worldwide. Besides state-of-the-art physical modeling, numerical performance and industrial functionality, the ARCADIA R system is featuring state-of-the-art software engineering. The new code system will bring a series of benefits for our customers: e.g. improved accuracy for heterogeneous cores (MOX/ UOX, Gd...), better description of nuclide chains, and access to local neutronics/ thermal-hydraulics and possibly thermal-mechanical information (3D pin by pin full core modeling). ARCADIA is a registered trademark of AREVA NP. (authors)

  7. A Benchmarking Study of High Energy Carbon Ion Induced Neutron Using Several Monte Carlo Codes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Oh, J. H.; Jung, N. S.; Lee, H. S. [Pohang Accelerator Laboratory, Pohang (Korea, Republic of); Shin, Y. S.; Kwon, D. Y.; Kim, Y. M. [Catholic Univ., Gyeongsan (Korea, Republic of); Oranj, L. Mokhtari [POSTECH, Pohang (Korea, Republic of)

    2014-10-15

    In this study, the benchmarking study was done for the representative particle interaction of the heavy ion accelerator, especially carbon-induced reaction. The secondary neutron is an important particle in the shielding analysis to define the source term and penetration ability of radiation fields. The performance of each Monte Carlo codes were verified for selected codes: MCNPX 2.7, PHITS 2.64 and FLUKA 2011.2b.6. For this benchmarking study, the experimental data of Kurosawa et al. in the SINBAD database of NEA was applied. The calculated results of the differential neutron yield produced from several materials irradiated by high energy carbon beam reproduced the experimental data well in small uncertainty. But the MCNPX results showed large discrepancy with experimental data, especially at the forward angle. The calculated results were lower a little than the experimental and it was clear in the cases of lower incident carbon energy, thinner target and forward angle. As expected, the influence of different model was found clearly at forward direction. In the shielding analysis, these characteristics of each Monte Carlo codes should be considered and utilized to determine the safety margin of a shield thickness.

  8. A Benchmarking Study of High Energy Carbon Ion Induced Neutron Using Several Monte Carlo Codes

    International Nuclear Information System (INIS)

    Kim, D. H.; Oh, J. H.; Jung, N. S.; Lee, H. S.; Shin, Y. S.; Kwon, D. Y.; Kim, Y. M.; Oranj, L. Mokhtari

    2014-01-01

    In this study, the benchmarking study was done for the representative particle interaction of the heavy ion accelerator, especially carbon-induced reaction. The secondary neutron is an important particle in the shielding analysis to define the source term and penetration ability of radiation fields. The performance of each Monte Carlo codes were verified for selected codes: MCNPX 2.7, PHITS 2.64 and FLUKA 2011.2b.6. For this benchmarking study, the experimental data of Kurosawa et al. in the SINBAD database of NEA was applied. The calculated results of the differential neutron yield produced from several materials irradiated by high energy carbon beam reproduced the experimental data well in small uncertainty. But the MCNPX results showed large discrepancy with experimental data, especially at the forward angle. The calculated results were lower a little than the experimental and it was clear in the cases of lower incident carbon energy, thinner target and forward angle. As expected, the influence of different model was found clearly at forward direction. In the shielding analysis, these characteristics of each Monte Carlo codes should be considered and utilized to determine the safety margin of a shield thickness

  9. APPLE-3: improvement of APPLE for neutron and gamma-ray flux, spectrum and reaction rate plotting code, and of its code manual

    International Nuclear Information System (INIS)

    Kawasaki, Hiromitu; Maki, Koichi; Seki, Yasushi.

    1991-03-01

    A code APPLE was produced in 1976 for calculating and plotting tritium breeding ratio and tritium production rate distributions. That code was improved as 'APPLE-2' in 1982, to calculate and plot not only tritium breeding ratio but also distributions of neutron and gamma-ray fluxes, their spectra, nuclear heating rates and other reaction rates, and dose rate distributions during operation and after shutdown in 1982. The code APPLE-2 can calculate and plot these nuclear properties derived from neutron and gamma-ray fluxes by ANISN (one dimensional transport code), DOT3.5 (two dimensional transport code) and MORSE (three dimensional Monte Carlo code). We revised the code APPLE-2 as 'APPLE-3' by adding many functions to the APPLE-2 code in accordance with users' requirements proposed in recent progress of fusion reaction nuclear design. With minor modification of APPLE-2, a number of inconsistencies have been found between the code manual and the input data in the code. In the present report, the new functions added to APPLE-2 and improved users' manual are explained. (author)

  10. Generation of the library of neutron cross sections for the Record code of the Fuel Management System (FMS)

    International Nuclear Information System (INIS)

    Alonso V, G.; Hernandez L, H.

    1991-11-01

    On the basis of the library structure of the RECORD code a method to generate the neutron cross sections by means of the ENDF-B/IV database and the NJOY code has been developed. The obtained cross sections are compared with those of the current library which was processed using the ENDF-B/III version. (Author)

  11. MCNP: a general Monte Carlo code for neutron and photon transport

    International Nuclear Information System (INIS)

    1978-07-01

    The general-purpose Monte Carlo code MCNP can be used for neutron, photon, or coupled neutron--photon transport. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces (elliptical tori). Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation (such as ENDF/B-IV) are accounted for. For photons, the code takes account of incoherent and coherent scattering, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. Standard optional variance reduction schemes include geometry splitting and Russian roulette, the exponential transformation, energy splitting, forced collisions in designated cells, flux estimates at point detectors, track-length estimators, and source biasing. The standard output of MCNP includes two-way current as a function of energy, time, and angle with the normal, across any subset of bounding surfaces in the problem. Fluxes across any set of bounding surfaces are available as a function of time and energy. Similarly, the flux at designated points and the average flux in a cell (track length per unit volume) are standard tallies. Reactions such as fissions or absorptions may be obtained in a subset of geometric cells. The heating tallies give the energy deposition per starting particle. In addition, particles may be flagged when they cross specified surfaces or enter designated cells, and the contributions of these flagged particles to certain of the tallies are listed separately. All quantities printed out have their relative errors listed also. 11 figures, 27 tables

  12. Surficial geology of Panther Lake Quadrangle, Oswego County, New York

    Science.gov (United States)

    Miller, Todd S.

    1981-01-01

    The location and extent of eight kinds of surficial deposits in Panther Lake quadrangle, Oswego County, N.Y., are mapped on a 7.5-minute U.S. Geological Survey topographic map. The map was compiled to indicate the lithology and potential for groundwater development at any specific location. (USGS)

  13. Methods and codes for neutronic calculations of the MARIA research reactor

    International Nuclear Information System (INIS)

    Andrzejewski, K.; Kulikowska, T.; Bretscher, M.M.; Hanan, N.A.; Matos, J.E.

    1998-01-01

    The core of the MARIA high flux multipurpose research reactor is highly heterogeneous. It consists of beryllium blocks arranged in 6x8 matrix, tubular fuel assemblies, control rods and irradiation channels. The reflector is also heterogeneous and consists of graphite blocks clad with aluminium. Its structure is perturbed by the experimental beam tubes. This paper presents methods and codes used to calculate the MARIA reactor neutronics characteristics and experience gained thus far at IAE and ANL. At ANL the methods of MARIA calculations were developed in connection with RERTR program. At IAE the package of programs was developed to help its operator in optimization of fuel utilization. (author)

  14. Analysis of neutron data in the resonance region via the computer code SAMMY

    International Nuclear Information System (INIS)

    Larson, N.M.

    1985-01-01

    Procedures for analysis of resonance neutron cross-section data have been implemented in a state-of-the-art computer code SAMMY, developed at the Oak Ridge Electron Linear Accelerator (ORELA) at Oak Ridge National Laboratory. A unique feature of SAMMY is the use of Bayes' equations to determine ''best'' values of parameters, which permits sequential analysis of data sets (or subsets) while giving the same results as would be given by a simultaneous analysis. Another important feature is the inclusion of data-reduction parameters in the fitting procedure. Other features of SAMMY are also described

  15. CHAR and BURNMAC - burnup modules of the AUS neutronics code system

    International Nuclear Information System (INIS)

    Robinson, G.S.

    1986-03-01

    In the AUS neutronics code system, the burnup module CHAR solves the nuclide depletion equations by an analytic technique in a number of spatial zones. CHAR is usually used as one component of a lattice burnup calculation but contains features which also make it suitable for some global burnup calculations. BURNMAC is a simple accounting module based on the assumption that cross sections for a rector zone depend only on irradiation. BURNMAC is used as one component of a global calculation in which burnup is achieved by interpolation in the cross sections produced from a previous lattice calculation

  16. APPLE-2: an improved version of APPLE code for plotting neutron and gamma ray spectra and reaction rates

    International Nuclear Information System (INIS)

    Kawasaki, Hiromitsu; Seki, Yasushi.

    1982-07-01

    A computer code APPLE-2 which plots the spatial distribution of energy spectra of multi-group neutron and/or gamma ray fluxes, and reaction rates has been developed. This code is an improved version of the previously developed APPLE code and has the following features: (1) It plots energy spectra of neutron and/or gamma ray fluxes calculated by ANISN, DOT and MORSE. (2) It calculates and plots the spatial distribution of neutron and gamma ray fluxes and various types of reaction rates such as nuclear heating rates, operational dose rates, displacement damage rates. (3) Input data specification is greatly simplified by the use of standard, response libraries and by close coupling with radiation transport calculation codes. (4) Plotting outputs are given in camera ready form. (author)

  17. BWR transient analysis using neutronic / thermal hydraulic coupled codes including uncertainty quantification

    International Nuclear Information System (INIS)

    Hartmann, C.; Sanchez, V.; Tietsch, W.; Stieglitz, R.

    2012-01-01

    The KIT is involved in the development and qualification of best estimate methodologies for BWR transient analysis in cooperation with industrial partners. The goal is to establish the most advanced thermal hydraulic system codes coupled with 3D reactor dynamic codes to be able to perform a more realistic evaluation of the BWR behavior under accidental conditions. For this purpose a computational chain based on the lattice code (SCALE6/GenPMAXS), the coupled neutronic/thermal hydraulic code (TRACE/PARCS) as well as a Monte Carlo based uncertainty and sensitivity package (SUSA) has been established and applied to different kind of transients of a Boiling Water Reactor (BWR). This paper will describe the multidimensional models of the plant elaborated for TRACE and PARCS to perform the investigations mentioned before. For the uncertainty quantification of the coupled code TRACE/PARCS and specifically to take into account the influence of the kinetics parameters in such studies, the PARCS code has been extended to facilitate the change of model parameters in such a way that the SUSA package can be used in connection with TRACE/PARCS for the U and S studies. This approach will be presented in detail. The results obtained for a rod drop transient with TRACE/PARCS using the SUSA-methodology showed clearly the importance of some kinetic parameters on the transient progression demonstrating that the coupling of a best-estimate coupled codes with uncertainty and sensitivity tools is very promising and of great importance for the safety assessment of nuclear reactors. (authors)

  18. PASC-1, Petten AMPX-II/SCALE-3 Code System for Reactor Neutronics Calculation

    International Nuclear Information System (INIS)

    Yaoqing, W.; Oppe, J.; Haas, J.B.M. de; Gruppelaar, H.; Slobben, J.

    1995-01-01

    1 - Description of program or function: The Petten AMPX-II/SCALE-3 Code System PASC-1 is a reactor neutronics calculation programme system consisting of well known IBM-oriented codes, that have been translated into FORTRAN-77, for calculations on a CDC-CYBER computer. Thus, the portability of these codes has been increased. In this system, some AMPX-II and SCALE-3 modules, the one-dimensional transport code ANISN and the 1 to 3-dimensional diffusion code CITATION are linked together on the CDC-CYBER/855 computer. The new cell code XSDRNPM-S and the old XSDRN code are included in the system. Starting from an AMPX fine group library up to CITATION, calculations can be performed for each individual module. Existing AMPX master interface format libraries, such as CSRL-IV, JEF-1, IRI and SCALE-45, and the old XSDRN-formatted libraries such as the COBB library can be used for the calculations. The code system contains the following modules and codes at present: AIM, AJAX, MALOCS, NITAWL-S, REVERT-I, ICE-2, CONVERT, JUAN, OCTAGN, XSDRNPM-S, XSDRN, ANISN and CITATION. The system will be extended with other SCALE modules and transport codes. 2 - Method of solution: The PASC-1 system is based on AMPX-II/SCALE-3 modules. Except for some SCALE-3 modules taken from the SCALIAS package, the original AMPX-II modules were IBM versions written in FORTRAN IV. These modules have been translated into CDC FORTRAN V. In order to test these modules and link them with some codes, some of the sample problem calculations have been performed for the whole PASC-1 system. During these calculations, some FORTRAN-77 errors were found in MALOCS, REVERT, CONVERT and some subroutines of SUBLIB (FORTRAN-77 subroutine library). These errors have been corrected. Because many corrections were made for the REVERT module, it is renamed as REVERT-I (improved version of REVERT). After these corrections, the whole system is running on a CDC-CYBER Computer (NOS-BE operating system). 3 - Restrictions on the

  19. Correlated sampling added to the specific purpose Monte Carlo code McPNL for neutron lifetime log responses

    International Nuclear Information System (INIS)

    Mickael, M.; Verghese, K.; Gardner, R.P.

    1989-01-01

    The specific purpose neutron lifetime oil well logging simulation code, McPNL, has been rewritten for greater user-friendliness and faster execution. Correlated sampling has been added to the code to enable studies of relative changes in the tool response caused by environmental changes. The absolute responses calculated by the code have been benchmarked against laboratory test pit data. The relative responses from correlated sampling are not directly benchmarked, but they are validated using experimental and theoretical results

  20. PEGASUS: a preequilibrium and multi-step evaporation code for neutron cross section calculation

    Energy Technology Data Exchange (ETDEWEB)

    Nakagawa, Tsuneo; Sugi, Teruo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iijima, Shungo; Nishigori, Takeo

    1999-06-01

    The computer code PEGASUS was developed to calculate neutron-induced reaction cross sections on the basis of the closed form exciton model preequilibrium theory and the multi-step evaporation theory. The cross sections and emitted particle spectra are calculated for the compound elastic scattering, (n,{gamma}), (n,n`), (n,p), (n,{alpha}), (n,d), (n,t), (n,{sup 3}He), (n,2n), (n,n`p), (n,n`{alpha}), (n,n`d), (n,n`t), (n,2p) and (n,3n) reactions. The double differential cross sections of emitted particles are also calculated. The calculated results are written on a magnetic disk in the ENDF format. Parameter files and/or systematics formulas are provided for level densities, mass excess, radiation widths and inverse cross sections so that the input data to the code are made minimum. (author)

  1. Project ''REMAD'', Neutron spectrometry. Code ''SOHO - phase II''. First part ''FORMULATION''

    International Nuclear Information System (INIS)

    Zaborowski, H.L.

    1982-08-01

    ''SOHO'' is a new neutron spectrometric code with large potential applications besides the BONNER spheres system for which he has been initially developped. This code makes use of a new and original iteractive procedure for the search of approximate solutions from the Fredholm integral Equation of the First kind whose Resolution Function is experimentally and/or mathematically defined (e.g. the Log-Normal Hypothesis for the BONNER Spheres). Upon discretization of the Fredholm integral Equation we get systems of non-exact homogeneous linear equations: Q X = e whose approximate solutions are given when: Q X → 0. We show that the iterative procedure converges absolutely leading to the existence of the approximate solutions, independently of the kind of initialization used. For the applications to health physics of the BONNER sphere ''SOHO'' has been programmed on a HP - 41 CV calculator [fr

  2. A portable, parallel, object-oriented Monte Carlo neutron transport code in C++

    International Nuclear Information System (INIS)

    Lee, S.R.; Cummings, J.C.; Nolen, S.D.

    1997-01-01

    We have developed a multi-group Monte Carlo neutron transport code using C++ and the Parallel Object-Oriented Methods and Applications (POOMA) class library. This transport code, called MC++, currently computes k and α-eigenvalues and is portable to and runs parallel on a wide variety of platforms, including MPPs, clustered SMPs, and individual workstations. It contains appropriate classes and abstractions for particle transport and, through the use of POOMA, for portable parallelism. Current capabilities of MC++ are discussed, along with physics and performance results on a variety of hardware, including all Accelerated Strategic Computing Initiative (ASCI) hardware. Current parallel performance indicates the ability to compute α-eigenvalues in seconds to minutes rather than hours to days. Future plans and the implementation of a general transport physics framework are also discussed

  3. 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.

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

    International Nuclear Information System (INIS)

    Sun Wei; Wang Kan; Ni Dongyang; Li Qing

    2013-01-01

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

  5. New Standard Evaluated Neutron Cross Section Libraries for the GEANT4 Code and First Verification

    CERN Document Server

    Mendoza, Emilio; Koi, Tatsumi; Guerrero, Carlos

    2014-01-01

    The Monte Carlo simulation of the interaction of neutrons with matter relies on evaluated nuclear data libraries and models. The evaluated libraries are compilations of measured physical parameters (such as cross sections) combined with predictions of nuclear model calculations which have been adjusted to reproduce the experimental data. The results obtained from the simulations depend largely on the accuracy of the underlying nuclear data used, and thus it is important to have access to the nuclear data libraries available, either of general use or compiled for specific applications, and to perform exhaustive validations which cover the wide scope of application of the simulation code. In this paper we describe the work performed in order to extend the capabilities of the GEANT4 toolkit for the simulation of the interaction of neutrons with matter at neutron energies up to 20 MeV and a first verification of the results obtained. Such a work is of relevance for applications as diverse as the simulation of a n...

  6. Model-Based Least Squares Reconstruction of Coded Source Neutron Radiographs: Integrating the ORNL HFIR CG1D Source Model

    Energy Technology Data Exchange (ETDEWEB)

    Santos-Villalobos, Hector J [ORNL; Gregor, Jens [University of Tennessee, Knoxville (UTK); Bingham, Philip R [ORNL

    2014-01-01

    At the present, neutron sources cannot be fabricated small and powerful enough in order to achieve high resolution radiography while maintaining an adequate flux. One solution is to employ computational imaging techniques such as a Magnified Coded Source Imaging (CSI) system. A coded-mask is placed between the neutron source and the object. The system resolution is increased by reducing the size of the mask holes and the flux is increased by increasing the size of the coded-mask and/or the number of holes. One limitation of such system is that the resolution of current state-of-the-art scintillator-based detectors caps around 50um. To overcome this challenge, the coded-mask and object are magnified by making the distance from the coded-mask to the object much smaller than the distance from object to detector. In previous work, we have shown via synthetic experiments that our least squares method outperforms other methods in image quality and reconstruction precision because of the modeling of the CSI system components. However, the validation experiments were limited to simplistic neutron sources. In this work, we aim to model the flux distribution of a real neutron source and incorporate such a model in our least squares computational system. We provide a full description of the methodology used to characterize the neutron source and validate the method with synthetic experiments.

  7. Development of EASYQAD version β. A visualization code system for gamma and neutron shielding calculations

    International Nuclear Information System (INIS)

    Kim, Jae Cheon; Kim, Soon Young; Lee, Hwan Soo; Ha, Pham Nhu Viet; Kim, Jong Kyung

    2008-01-01

    EASYQAD version β was developed by MATLAB GUI (Graphical User Interface) as a visualization code system based on QAD-CGGP-A point-kernel code for convenient shielding calculations of gammas and neutrons. It consists of four graphic interface modules including GEOMETRY, INPUT, OUTPUT, and SHIELD. These modules were compiled in C++ programming language by using the MATLAB Compiler Toolbox to form a stand-along code system that can be run on the Windows XP operating system without MATLAB installation. In addition, EASYQAD version β has user-friendly graphical interfaces and, additionally, many useful functions in comparison with QAD- CGGP-A such as common material library, line and grid detectors, and multi-group energy calculations so as to increase its applicability in the field of radiation shielding analysis. It is a powerful tool for non-experts to analyze easily the shielding problems without special training. Therefore, EASYOAD version β is expected to contribute effectively to the development of radiation shielding analysis by providing users in medical and industrial fields with an efficient radiation shielding code. (author)

  8. Discrete-ordinates electron transport calculations using standard neutron transport codes

    International Nuclear Information System (INIS)

    Morel, J.E.

    1979-01-01

    The primary purpose of this work was to develop a method for using standard neutron transport codes to perform electron transport calculations. The method is to develop approximate electron cross sections which are sufficiently well-behaved to be treated with standard S/sub n/ methods, but which nonetheless yield flux solutions which are very similar to the exact solutions. The main advantage of this approach is that, once the approximate cross sections are constructed, their multigroup Legendre expansion coefficients can be calculated and input to any standard S/sub n/ code. Discrete-ordinates calculations were performed to determine the accuracy of the flux solutions for problems corresponding to 1.0-MeV electrons incident upon slabs of aluminum and gold. All S/sub n/ calculations were compared with similar calculations performed with an electron Monte Carlo code, considered to be exact. In all cases, the discrete-ordinates solutions for integral flux quantities (i.e., scalar flux, energy deposition profiles, etc.) are generally in agreement with the Monte Carlo solutions to within approximately 5% or less. The central conclusion is that integral electron flux quantities can be efficiently and accurately calculated using standard S/sub n/ codes in conjunction with approximate cross sections. Furthermore, if group structures and approximate cross section construction are optimized, accurate differential flux energy spectra may also be obtainable without having to use an inordinately large number of energy groups. 1 figure

  9. Studies on the molten salt reactor. Code development and neutronics analysis of MSRE-type design

    International Nuclear Information System (INIS)

    Zhuang Kun; Cao Liangzhi; Zheng Youqi; Wu Hongchun

    2015-01-01

    The molten salt reactor is characterized by its use of the fluid-fuel, which serves both as a fuel and as a coolant simultaneously. The position of delayed neutron precursors continuously changes both in the core and in the external loop due to the fuel circulation, and the fission products are extracted by an online fuel reprocessing unit, which all lead to the modeling methods for the conventional reactors using solid fuel not applicable. This study establishes suitable calculation models for the neutronics analysis of the molten salt reactor and develops a new code named MOREL based on the three-dimensional diffusion steady and transient calculations. Some numerical tests are chosen to verify the code and the numerical results indicate that MOREL can be used for the analysis of the molten salt reactor. After verification, it is applied to analyze the characteristics of a typical molten salt reactor, including the steady characteristics, the influence of fuel circulation on the kinetic behaviors. Besides, the influence of online fuel reprocessing simulation is also examined. The results show that inherent safety is the character of the molten salt reactor from the aspect of reactivity feedback and the fuel circulation has great influence on the kinetic characteristics of molten salt reactor. (author)

  10. Feasibility of the integration of CRONOS, a 3-D neutronics code, into real-time simulators

    International Nuclear Information System (INIS)

    Ragusa, J.C.

    2001-01-01

    In its effort to contribute to nuclear power plant safety, CEA proposes the integration of an engineering grade 3-D neutronics code into a real-time plant analyser. This paper describes the capabilities of the neutronics code CRONOS to achieve a fast running performance. First, we will present current core models in simulators and explain their drawbacks. Secondly, the mean features of CRONOS's spatial-kinetics methods will be reviewed. We will then present an optimum core representation with respect to mesh size, choice of finite elements (FE) basis and execution time, for accurate results as well as the multi 1-D thermal-hydraulics (T/H) model developed to take into account 3-D effects in updating the cross-sections. A Main Steam Line Break (MSLB) End-of-Life (EOL) Hot-Zero-Power (HZP) accident will be used as an example, before we conclude with the perspectives of integrating CRONOS's 3-D core model into real-time simulators. (author)

  11. New Three-Dimensional Neutron Transport Calculation Capability in STREAM Code

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, Youqi [Xi' an Jiaotong University, Xi' an (China); Choi, Sooyoung; Lee, Deokjung [UNIST, Ulsan (Korea, Republic of)

    2016-10-15

    The method of characteristics (MOC) is one of the best choices for its powerful capability in the geometry modeling. To reduce the large computational burden in 3D MOC, the 2D/1D schemes were proposed and have achieved great success in the past 10 years. However, such methods have some instability problems during the iterations when the neutron leakage for axial direction is large. Therefore, full 3D MOC methods were developed. A lot of efforts have been devoted to reduce the computational costs. However, it still requires too much memory storage and computational time for the practical modeling of a commercial size reactor core. Recently, a new approach for the 3D MOC calculation without transverse integration has been implemented in the STREAM code. In this approach, the angular flux is expressed as a basis function expansion form of only axial variable z. A new approach based on the axial expansion and 2D MOC sweeping to solve the 3D neutron transport equation is implemented in the STREAM code. This approach avoids using the transverse integration in the traditional 2D/1D scheme of MOC calculation. By converting the 3D equation into the 2D form of angular flux expansion coefficients, it also avoids the complex 3D ray tracing. Current numerical tests using two benchmarks show good accuracy of the new method.

  12. RSAP - A Code for Display of Neutron Cross Section Data and SAMMY Fit Results

    International Nuclear Information System (INIS)

    Sayer, R.O.

    2001-01-01

    RSAP is a computer code for display of neutron cross section data and selected SAMMY output. SAMMY is a multilevel R-matrix code for fitting neutron time-of-flight cross-section data using Bayes' method. RSAP, which runs on the Digital Unix Alpha platform, reads ORELA Data Files (ODF) created by SAMMY and uses graphics routines from the PLPLOT package. In addition, RSAP can read data and/or computed values from ASCII files with a format specified by the user. Plot output may be displayed in an X window, sent to a postscript file (rsap.ps), or sent to a color postscript file (rsap.psc). Thirteen plot types are supported, allowing the user to display cross section data, transmission data, errors, theory, Bayes fits, and residuals in various combinations. In this document the designations theory and Bayes refer to the initial and final theoretical cross sections, respectively, as evaluated by SAMMY. Special plot types include Bayes/Data, Theory--Data, and Bayes--Data. Output from two SAMMY runs may be compared by plotting the ratios Theory2/Theory1 and Bayes2/Bayes1 or by plotting the differences (Theory2-Theory1) and (Bayes2-Bayes1)

  13. Feasibility of the integration of CRONOS, a 3-D neutronics code, into real-time simulators

    Energy Technology Data Exchange (ETDEWEB)

    Ragusa, J.C. [CEA Saclay, Dept. de Mecanique et de Technologie, 91 - Gif-sur-Yvette (France)

    2001-07-01

    In its effort to contribute to nuclear power plant safety, CEA proposes the integration of an engineering grade 3-D neutronics code into a real-time plant analyser. This paper describes the capabilities of the neutronics code CRONOS to achieve a fast running performance. First, we will present current core models in simulators and explain their drawbacks. Secondly, the mean features of CRONOS's spatial-kinetics methods will be reviewed. We will then present an optimum core representation with respect to mesh size, choice of finite elements (FE) basis and execution time, for accurate results as well as the multi 1-D thermal-hydraulics (T/H) model developed to take into account 3-D effects in updating the cross-sections. A Main Steam Line Break (MSLB) End-of-Life (EOL) Hot-Zero-Power (HZP) accident will be used as an example, before we conclude with the perspectives of integrating CRONOS's 3-D core model into real-time simulators. (author)

  14. Fast resolution of the neutron diffusion equation through public domain Ode codes

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, V.M.; Vidal, V.; Garayoa, J. [Universidad Politecnica de Valencia, Departamento de Sistemas Informaticos, Valencia (Spain); Verdu, G. [Universidad Politecnica de Valencia, Departamento de Ingenieria Quimica y Nuclear, Valencia (Spain); Gomez, R. [I.E.S. de Tavernes Blanques, Valencia (Spain)

    2003-07-01

    The time-dependent neutron diffusion equation is a partial differential equation with source terms. The resolution method usually includes discretizing the spatial domain, obtaining a large system of linear, stiff ordinary differential equations (ODEs), whose resolution is computationally very expensive. Some standard techniques use a fixed time step to solve the ODE system. This can result in errors (if the time step is too large) or in long computing times (if the time step is too little). To speed up the resolution method, two well-known public domain codes have been selected: DASPK and FCVODE that are powerful codes for the resolution of large systems of stiff ODEs. These codes can estimate the error after each time step, and, depending on this estimation can decide which is the new time step and, possibly, which is the integration method to be used in the next step. With these mechanisms, it is possible to keep the overall error below the chosen tolerances, and, when the system behaves smoothly, to take large time steps increasing the execution speed. In this paper we address the use of the public domain codes DASPK and FCVODE for the resolution of the time-dependent neutron diffusion equation. The efficiency of these codes depends largely on the preconditioning of the big systems of linear equations that must be solved. Several pre-conditioners have been programmed and tested; it was found that the multigrid method is the best of the pre-conditioners tested. Also, it has been found that DASPK has performed better than FCVODE, being more robust for our problem.We can conclude that the use of specialized codes for solving large systems of ODEs can reduce drastically the computational work needed for the solution; and combining them with appropriate pre-conditioners, the reduction can be still more important. It has other crucial advantages, since it allows the user to specify the allowed error, which cannot be done in fixed step implementations; this, of course

  15. NSDann2BS, a neutron spectrum unfolding code based on neural networks technology and two bonner spheres

    Science.gov (United States)

    Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solís Sánches, L. O.; Miranda, R. Castañeda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

    2013-07-01

    In this work a neutron spectrum unfolding code, based on artificial intelligence technology is presented. The code called "Neutron Spectrometry and Dosimetry with Artificial Neural Networks and two Bonner spheres", (NSDann2BS), was designed in a graphical user interface under the LabVIEW programming environment. The main features of this code are to use an embedded artificial neural network architecture optimized with the "Robust design of artificial neural networks methodology" and to use two Bonner spheres as the only piece of information. In order to build the code here presented, once the net topology was optimized and properly trained, knowledge stored at synaptic weights was extracted and using a graphical framework build on the LabVIEW programming environment, the NSDann2BS code was designed. This code is friendly, intuitive and easy to use for the end user. The code is freely available upon request to authors. To demonstrate the use of the neural net embedded in the NSDann2BS code, the rate counts of 252Cf, 241AmBe and 239PuBe neutron sources measured with a Bonner spheres system.

  16. NSDann2BS, a neutron spectrum unfolding code based on neural networks technology and two bonner spheres

    International Nuclear Information System (INIS)

    Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solís Sánches, L. O.; Miranda, R. Castañeda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

    2013-01-01

    In this work a neutron spectrum unfolding code, based on artificial intelligence technology is presented. The code called ''Neutron Spectrometry and Dosimetry with Artificial Neural Networks and two Bonner spheres'', (NSDann2BS), was designed in a graphical user interface under the LabVIEW programming environment. The main features of this code are to use an embedded artificial neural network architecture optimized with the ''Robust design of artificial neural networks methodology'' and to use two Bonner spheres as the only piece of information. In order to build the code here presented, once the net topology was optimized and properly trained, knowledge stored at synaptic weights was extracted and using a graphical framework build on the LabVIEW programming environment, the NSDann2BS code was designed. This code is friendly, intuitive and easy to use for the end user. The code is freely available upon request to authors. To demonstrate the use of the neural net embedded in the NSDann2BS code, the rate counts of 252 Cf, 241 AmBe and 239 PuBe neutron sources measured with a Bonner spheres system

  17. NSDann2BS, a neutron spectrum unfolding code based on neural networks technology and two bonner spheres

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz-Rodriguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solis Sanches, L. O.; Miranda, R. Castaneda; Cervantes Viramontes, J. M. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica. Av. Ramon Lopez Velarde 801. Col. Centro Zacatecas, Zac (Mexico); Vega-Carrillo, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica. Av. Ramon Lopez Velarde 801. Col. Centro Zacatecas, Zac., Mexico. and Unidad Academica de Estudios Nucleares. C. Cip (Mexico)

    2013-07-03

    In this work a neutron spectrum unfolding code, based on artificial intelligence technology is presented. The code called ''Neutron Spectrometry and Dosimetry with Artificial Neural Networks and two Bonner spheres'', (NSDann2BS), was designed in a graphical user interface under the LabVIEW programming environment. The main features of this code are to use an embedded artificial neural network architecture optimized with the ''Robust design of artificial neural networks methodology'' and to use two Bonner spheres as the only piece of information. In order to build the code here presented, once the net topology was optimized and properly trained, knowledge stored at synaptic weights was extracted and using a graphical framework build on the LabVIEW programming environment, the NSDann2BS code was designed. This code is friendly, intuitive and easy to use for the end user. The code is freely available upon request to authors. To demonstrate the use of the neural net embedded in the NSDann2BS code, the rate counts of {sup 252}Cf, {sup 241}AmBe and {sup 239}PuBe neutron sources measured with a Bonner spheres system.

  18. ORLIB: a computer code that produces one-energy group, time- and spatially-averaged neutron cross sections

    International Nuclear Information System (INIS)

    Blink, J.A.; Dye, R.E.; Kimlinger, J.R.

    1981-12-01

    Calculation of neutron activation of proposed fusion reactors requires a library of neutron-activation cross sections. One such library is ACTL, which is being updated and expanded by Howerton. If the energy-dependent neutron flux is also known as a function of location and time, the buildup and decay of activation products can be calculated. In practice, hand calculation is impractical without energy-averaged cross sections because of the large number of energy groups. A widely used activation computer code, ORIGEN2, also requires energy-averaged cross sections. Accordingly, we wrote the ORLIB code to collapse the ACTL library, using the flux as a weighting function. The ORLIB code runs on the LLNL Cray computer network. We have also modified ORIGEN2 to accept the expanded activation libraries produced by ORLIB

  19. The KINA neutronic module of the LEGO code for steady-state and transient PWR plant simulations

    International Nuclear Information System (INIS)

    Nicolopoulos, D.; Pollacchini, L.; Vimercati, G.; Spelta, S.

    1989-01-01

    The Automation Research Center (CRA) of ENEl has implemented some models for analyzing both incidental and operational transients in PWR power plants. For such models an axial neutron kinetics module characterized by high computational efficency with adequate results accuracy was called for. CISE has been entrusted with the task of implementing such a module named KINA and based on IQS (Improved Quasi Static) method, to be included in the library of LEGO modular code used by CRA to set up PWR power models. Moreover, The KINA module has been adapted to the neutron constants computing model developed by the EdF-SEPTEN, which has been using and improving the LEGO code for a long time in cooperation with ENEL-CRA. In this paper, after some remarks on the LEGO code, a general description of KINA neutronic module is given. The resylts of a preliminary validation activity of KINA for an EdF 1300 MWe PWR plant are also presented

  20. Calculations of the thermal and fast neutron fluxes in the Syrian miniature neutron source reactor using the MCNP-4C code.

    Science.gov (United States)

    Khattab, K; Sulieman, I

    2009-04-01

    The MCNP-4C code, based on the probabilistic approach, was used to model the 3D configuration of the core of the Syrian miniature neutron source reactor (MNSR). The continuous energy neutron cross sections from the ENDF/B-VI library were used to calculate the thermal and fast neutron fluxes in the inner and outer irradiation sites of MNSR. The thermal fluxes in the MNSR inner irradiation sites were also measured experimentally by the multiple foil activation method ((197)Au (n, gamma) (198)Au and (59)Co (n, gamma) (60)Co). The foils were irradiated simultaneously in each of the five MNSR inner irradiation sites to measure the thermal neutron flux and the epithermal index in each site. The calculated and measured results agree well.

  1. Investigation of some possible changes in Am-Be neutron source configuration in order to increase the thermal neutron flux using Monte Carlo code

    Science.gov (United States)

    Basiri, H.; Tavakoli-Anbaran, H.

    2018-01-01

    Am-Be neutrons source is based on (α, n) reaction and generates neutrons in the energy range of 0-11 MeV. Since the thermal neutrons are widely used in different fields, in this work, we investigate how to improve the source configuration in order to increase the thermal flux. These suggested changes include a spherical moderator instead of common cylindrical geometry, a reflector layer and an appropriate materials selection in order to achieve the maximum thermal flux. All calculations were done by using MCNP1 Monte Carlo code. Our final results indicated that a spherical paraffin moderator, a layer of beryllium as a reflector can efficiently increase the thermal neutron flux of Am-Be source.

  2. Contributions at the Tripoli Monte Carlo code qualifying on critical experiences and at neutronic interaction study of fissile units

    International Nuclear Information System (INIS)

    Nouri, A.

    1994-01-01

    Criticality studies in nuclear fuel cycle are based on Monte Carlo method. These codes use multigroup cross sections which can verify by experimental configurations or by use of reference codes such Tripoli 2. In this Tripoli 2 code nuclear data are errors attached and asked for experimental studies with critical experiences. This is one of the aim of this thesis. To calculate the keff of interacted fissile units we have used the multigroup Monte Carlo code Moret with convergence problems. A new estimator of reactions rates permit to better approximate the neutrons exchange between units and a new importance function has been tested. 2 annexes

  3. Study of the microstructure of neutron irradiated beryllium for the validation of the ANFIBE code

    International Nuclear Information System (INIS)

    Rabaglino, E.; Ferrero, C.; Reimann, J.; Ronchi, C.; Schulenberg, T.

    2002-01-01

    The behaviour of beryllium under fast neutron irradiation is a key issue of the helium cooled pebble bed tritium breeding blanket, due to the production of large quantities of helium and of a non-negligible amount of tritium. To optimise the design, a reliable prediction of swelling due to helium bubbles and of tritium inventory during normal and off-normal operation of a fusion power reactor is needed. The ANFIBE code (ANalysis of Fusion Irradiated BEryllium) is being developed to meet this need. The code has to be applied in a range of irradiation conditions where no experimental data are available, therefore a detailed gas kinetics model, and a specific and particularly careful validation strategy are needed. The validation procedure of the first version of the code was based on macroscopic data of swelling and tritium release. This approach is, however, incomplete, since a verification of the microscopic behaviour of the gas in the metal is necessary to obtain a reliable description of swelling. This paper discusses a general strategy for a thorough validation of the gas kinetics models in ANFIBE. The microstructure characterisation of weakly irradiated beryllium pebbles, with different visual examination techniques, is then presented as an example of the application of this strategy. In particular, the advantage of developing 3D techniques, such as X-ray microtomography, is demonstrated

  4. Neutronic / thermal-hydraulic coupling with the code system Trace / Parcs

    International Nuclear Information System (INIS)

    Mejia S, D. M.; Del Valle G, E.

    2015-09-01

    The developed models for Parcs and Trace codes corresponding for the cycle 15 of the Unit 1 of the Laguna Verde nuclear power plant are described. The first focused to the neutronic simulation and the second to thermal hydraulics. The model developed for Parcs consists of a core of 444 fuel assemblies wrapped in a radial reflective layer and two layers, a superior and another inferior, of axial reflector. The core consists of 27 total axial planes. The model for Trace includes the vessel and its internal components as well as various safety systems. The coupling between the two codes is through two maps that allow its intercommunication. Both codes are used in coupled form performing a dynamic simulation that allows obtaining acceptably a stable state from which is carried out the closure of all the main steam isolation valves (MSIVs) followed by the performance of safety relief valves (SRVs) and ECCS. The results for the power and reactivities introduced by the moderator density, the fuel temperature and total temperature are shown. Data are also provided like: the behavior of the pressure in the steam dome, the water level in the downcomer, the flow through the MSIVs and SRVs. The results are explained for the power, the pressure in the steam dome and the water level in the downcomer which show agreement with the actions of the MSIVs, SRVs and ECCS. (Author)

  5. Effects of neutron data libraries and criticality codes on IAEA criticality benchmark problems

    International Nuclear Information System (INIS)

    Sarker, Md.M.; Takano, Makoto; Masukawa, Fumihiro; Naito, Yoshitaka

    1993-10-01

    In order to compare the effects of neutron data libraries and criticality codes to thermal reactors (LWR), the IAEA criticality benchmark calculations have been performed. The experiments selected in this study include TRX-1 and TRX-2 with a simple geometric configuration. Reactor lattice calculation codes WIMS-D/4, MCNP-4, JACS (MGCL, KENO), and SRAC were used in the present calculations. The TRX cores were analyzed by WIMS-D/4 using WIMS original library and also by MCNP-4, JACS (MGCL, KENO), and SRAC using the libraries generated from JENDL-3 and ENDF/B-IV nuclear data files. An intercomparison work for the above mentioned code systems and cross section libraries was performed by analyzing the LWR benchmark experiments TRX-1 and TRX-2. The TRX cores were also analyzed for supercritical and subcritical conditions and these results were compared. In the case of critical condition, the results were in good agreement. But for the supercritical and subcritical conditions, the difference of the results obtained by using the different cross section libraries become larger than for the critical condition. (author)

  6. A comparison of the COG and MCNP codes in computational neutron capture therapy modeling, Part II: gadolinium neutron capture therapy models and therapeutic effects.

    Science.gov (United States)

    Wangerin, K; Culbertson, C N; Jevremovic, T

    2005-08-01

    The goal of this study was to evaluate the COG Monte Carlo radiation transport code, developed and tested by Lawrence Livermore National Laboratory, for gadolinium neutron capture therapy (GdNCT) related modeling. The validity of COG NCT model has been established for this model, and here the calculation was extended to analyze the effect of various gadolinium concentrations on dose distribution and cell-kill effect of the GdNCT modality and to determine the optimum therapeutic conditions for treating brain cancers. The computational results were compared with the widely used MCNP code. The differences between the COG and MCNP predictions were generally small and suggest that the COG code can be applied to similar research problems in NCT. Results for this study also showed that a concentration of 100 ppm gadolinium in the tumor was most beneficial when using an epithermal neutron beam.

  7. TART 2000: A Coupled Neutron-Photon, 3-D, Combinatorial Geometry, Time Dependent, Monte Carlo Transport Code

    International Nuclear Information System (INIS)

    Cullen, D.E

    2000-01-01

    TART2000 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input Preparation, running Monte Carlo calculations, and analysis of output results. TART2000 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART2000 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART2000 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART2000 and its data files

  8. TART 2000 A Coupled Neutron-Photon, 3-D, Combinatorial Geometry, Time Dependent, Monte Carlo Transport Code

    CERN Document Server

    Cullen, D

    2000-01-01

    TART2000 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input Preparation, running Monte Carlo calculations, and analysis of output results. TART2000 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART2000 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART2000 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART2000 and its data files.

  9. Panthere V2: Multipurpose Simulation Software for 3D Dose Rate Calculations

    Science.gov (United States)

    Penessot, Gaël; Bavoil, Éléonore; Wertz, Laurent; Malouch, Fadhel; Visonneau, Thierry; Dubost, Julien

    2017-09-01

    PANTHERE is a multipurpose radiation protection software developed by EDF to calculate gamma dose rates in complex 3D environments. PANTHERE takes a key role in the EDF ALARA process, enabling to predict dose rates and to organize and optimize operations in high radiation environments. PANTHERE is also used for nuclear waste characterization, transport of nuclear materials, etc. It is used in most of the EDF engineering units and their design service providers and industrial partners.

  10. Toward a Panther-centered View of the Forests of South Florida

    OpenAIRE

    Andrew J. Kerkhoff; Bruce T. Milne; David S. Maehr

    2000-01-01

    Anthropogenic habitat degradation and loss is the single largest threat to the endangered Florida panther, Puma concolor coryi. Conservation of the subspecies must be undertaken on the scale of the entire landscape. Thus, a view of the forested landscape of South Florida must be developed that is meaningful with reference to the panther. We approach this problem by analyzing the spatial interactions of panthers and forests at multiple scales. We apply tools derived from fractal geometry to th...

  11. ERANOS 2.0, Modular code and data system for fast reactor neutronics analyses

    International Nuclear Information System (INIS)

    2008-01-01

    1 - Description of program or function: The European Reactor Analysis Optimized calculation System, ERANOS, has been developed and validated with the aim of providing a suitable basis for reliable neutronic calculations of current as well as advanced fast reactor cores. It consists of data libraries, deterministic codes and calculation procedures which have been developed within the European Collaboration on Fast Reactors over the past 20 years or so, in order to answer the needs of both industrial and R and D organisations. The whole system counts roughly 250 functions and 3000 subroutines totalling 450000 lines of FORTRAN-77 and ESOPE instructions. ERANOS is written using the ALOS software which requires only standard FORTRAN compilers and includes advanced programming features. A modular structure was adopted for easier evolution and incorporation of new functionalities. Blocks of data (SETs) can be created or used by the modules themselves or by the user via the LU control language. Programming, and dynamic memory allocation, are performed by means of the ESOPE language. External temporary storage and permanent storage capabilities are provided by the GEMAT and ARCHIVE functions, respectively. ESOPE, LU, GEMAT and ARCHIVE are all part of the ALOS software. This modular structure allows different modules to be linked together in procedures corresponding to recommended calculation routes ranging from fast-running and moderately-accurate 'routine' procedures to slow-running but highly-accurate 'reference' procedures. The main contents of the ERANOS-2.0 package are: nuclear data libraries (multigroup cross-sections from the JEF-2.2 evaluated nuclear data file, and other specific data files), a cell and lattice code (ECCO), reactor flux solvers (diffusion, Sn transport, nodal variational transport), a burn-up module, various processing modules (material and neutron balance, breeding gains,...), tools related to perturbation theory and sensitivity analysis, core

  12. The PANTHER database of protein families, subfamilies, functions and pathways

    OpenAIRE

    Mi, Huaiyu; Lazareva-Ulitsky, Betty; Loo, Rozina; Kejariwal, Anish; Vandergriff, Jody; Rabkin, Steven; Guo, Nan; Muruganujan, Anushya; Doremieux, Olivier; Campbell, Michael J.; Kitano, Hiroaki; Thomas, Paul D.

    2004-01-01

    PANTHER is a large collection of protein families that have been subdivided into functionally related subfamilies, using human expertise. These subfamilies model the divergence of specific functions within protein families, allowing more accurate association with function (ontology terms and pathways), as well as inference of amino acids important for functional specificity. Hidden Markov models (HMMs) are built for each family and subfamily for classifying additional protein sequences. The l...

  13. Calculation of the effective delayed neutron fraction by TRIPOLI-4 code for IPEN/MB-01 research reactor

    International Nuclear Information System (INIS)

    Lee, Y.K.; Hugot, F.X.

    2011-01-01

    The effective delayed neutron fraction βeff is an important reactor physics parameter. Its calculation within the multi-group deterministic transport code can be performed with the aid of adjoint flux weighted integrations. However, in continuous energy Monte Carlo transport code, the adjoint weighted βeff calculation becomes complicated due to the backward treatment of the anisotropy scattering. In TRIPOLI-4 continuous energy Monte Carlo code, the βeff calculation was performed by a two-run method, one run with delayed neutrons and second with only the contribution from prompt fission neutrons. To improve the uncertainty of the βeff two-run calculation for the experimental reactors, two simple and fast one-run methods to estimate the βeff in the continuous energy simulation have been implemented into the TRIPOLI-4 code. First approach is an improved one of the Bretscher's prompt method and second one based on the proposal of Nauchi and Kameyama. In these one-run methods, the prompt and the delayed neutrons are first tagged. Their tracking and statistics are separated performed. The new βeff calculations have been optimized in the power iteration cycles so as to estimate the production of prompt and delayed neutrons from the prompt and delayed neutrons of previous generation. To validate the new βeff calculation by TRIPOLI-4, several benchmarks including fast and thermal systems have been considered. In this paper the recent measurements of βeff in the research reactor IPEN/MB-01 have been benchmarked. The basic components of the βeff and the Keff have been also calculated so as to understand the influences of the cross sections and the delayed neutron yields on the reactor reactivity calculations. Three nuclear data libraries, ENDF/BVI.r4, ENDF/B-VII.0, and JEFF-3.1 were taken into account in this study. (author)

  14. Implementation of Neutronics Analysis Code using the Features of Object Oriented Programming via Fortran90/95

    Energy Technology Data Exchange (ETDEWEB)

    Han, Tae Young; Cho, Beom Jin [KEPCO Nuclear Fuel, Daejeon (Korea, Republic of)

    2011-05-15

    The object-oriented programming (OOP) concept was radically established after 1990s and successfully involved in Fortran 90/95. The features of OOP are such as the information hiding, encapsulation, modularity and inheritance, which lead to producing code that satisfy three R's: reusability, reliability and readability. The major OOP concepts, however, except Module are not mainly used in neutronics analysis codes even though the code was written by Fortran 90/95. In this work, we show that the OOP concept can be employed to develop the neutronics analysis code, ASTRA1D (Advanced Static and Transient Reactor Analyzer for 1-Dimension), via Fortran90/95 and those can be more efficient and reasonable programming methods

  15. Performance, Accuracy and Efficiency Evaluation of a Three-Dimensional Whole-Core Neutron Transport Code AGENT

    International Nuclear Information System (INIS)

    Jevremovic, Tatjana; Hursin, Mathieu; Satvat, Nader; Hopkins, John; Xiao, Shanjie; Gert, Godfree

    2006-01-01

    The AGENT (Arbitrary Geometry Neutron Transport) an open-architecture reactor modeling tool is deterministic neutron transport code for two or three-dimensional heterogeneous neutronic design and analysis of the whole reactor cores regardless of geometry types and material configurations. The AGENT neutron transport methodology is applicable to all generations of nuclear power and research reactors. It combines three theories: (1) the theory of R-functions used to generate real three-dimensional whole-cores of square, hexagonal or triangular cross sections, (2) the planar method of characteristics used to solve isotropic neutron transport in non-homogenized 2D) reactor slices, and (3) the one-dimensional diffusion theory used to couple the planar and axial neutron tracks through the transverse leakage and angular mesh-wise flux values. The R-function-geometrical module allows a sequential building of the layers of geometry and automatic sub-meshing based on the network of domain functions. The simplicity of geometry description and selection of parameters for accurate treatment of neutron propagation is achieved through the Boolean algebraic hierarchically organized simple primitives into complex domains (both being represented with corresponding domain functions). The accuracy is comparable to Monte Carlo codes and is obtained by following neutron propagation through real geometrical domains that does not require homogenization or simplifications. The efficiency is maintained through a set of acceleration techniques introduced at all important calculation levels. The flux solution incorporates power iteration with two different acceleration techniques: Coarse Mesh Re-balancing (CMR) and Coarse Mesh Finite Difference (CMFD). The stand-alone originally developed graphical user interface of the AGENT code design environment allows the user to view and verify input data by displaying the geometry and material distribution. The user can also view the output data such

  16. Application of static fuel management codes for determination of the neutron noise using the adiabatic approximation

    International Nuclear Information System (INIS)

    Garis, N.S.; Karlsson, J.K.H.; Pazsit, I.

    2000-01-01

    The neutron noise, induced by a rod manoeuvring experiment in a pressurized water reactor, has been calculated by the incore fuel management code SIMULATE. The space- and frequency-dependent noise in the thermal group was calculated through the adiabatic approximation in three dimensions and two-group theory, with the spatial resolution of the nodal model underlying the SIMULATE algorithm. The calculated spatial noise profiles were interpreted on physical terms. They were also compared with model calculations in a 2-D one-group model, where various approximations as well as the full space-dependent response could be calculated. The adiabatic results obtained with SIMULATE can be regarded as reliable for sub-plateau frequencies (below 0.1 Hz). (orig.) [de

  17. Properties of SiC semiconductor detector of fast neutrons investigated using MCNPX code

    International Nuclear Information System (INIS)

    Sedlakova, K.; Sagatova, A.; Necas, V.; Zatko, B.

    2013-01-01

    The potential of silicon carbide (SiC) for use in semiconductor nuclear radiation detectors has been long recognized. The wide bandgap of SiC (3.25 eV for 4H-SiC polytype) compared to that for more conventionally used semiconductors, such as silicon (1.12 eV) and germanium (0.67 eV), makes SiC an attractive semiconductor for use in high dose rate and high ionization nuclear environments. The present work focused on the simulation of particle transport in SiC detectors of fast neutrons using statistical analysis of Monte Carlo radiation transport code MCNPX. Its possibilities in detector design and optimization are presented.(authors)

  18. MC++: A parallel, portable, Monte Carlo neutron transport code in C++

    International Nuclear Information System (INIS)

    Lee, S.R.; Cummings, J.C.; Nolen, S.D.

    1997-01-01

    MC++ is an implicit multi-group Monte Carlo neutron transport code written in C++ and based on the Parallel Object-Oriented Methods and Applications (POOMA) class library. MC++ runs in parallel on and is portable to a wide variety of platforms, including MPPs, SMPs, and clusters of UNIX workstations. MC++ is being developed to provide transport capabilities to the Accelerated Strategic Computing Initiative (ASCI). It is also intended to form the basis of the first transport physics framework (TPF), which is a C++ class library containing appropriate abstractions, objects, and methods for the particle transport problem. The transport problem is briefly described, as well as the current status and algorithms in MC++ for solving the transport equation. The alpha version of the POOMA class library is also discussed, along with the implementation of the transport solution algorithms using POOMA. Finally, a simple test problem is defined and performance and physics results from this problem are discussed on a variety of platforms

  19. PERIGEE computer codes for reactor simulation in 3 dimensions, using 1 or 2 neutron velocity groups

    International Nuclear Information System (INIS)

    Olson, A.P.

    1964-02-01

    PERIGEE is a code written in SNAP for the G-20 computer. It solves the one- or two-group neutron diffusion equations by finite-difference methods on a three-dimensional, uniform mesh having a common spacing in the two directions normal to the fuel channels. The positions of mesh points along a fuel channel, relative to points in adjacent channels, may correspond to either NPD or CANDU fuel bundle positions. The extrapolated flux boundary may be specified in sufficient detail to represent a tapered or stepped circumferential reflector, a variable axial length and, for a reactor with axis horizontal, a variable moderator level and a variable plane bottom surface equivalent to the CANDU dump structure. The neutron flux may be normalized to give a specified power output from the hottest fuel bundle or hottest channel, or to give a total thermal power limited by the turbine and generator. Reactor operation may be simulated in finite time steps, taking into account any fuel shifts, any changes in moderator level and the change in nuclear properties of the fuel with increasing irradiation. The appropriate properties are obtained by interpolation from tables supplied for as many as 8 types of fuel bundle. The mean fuel exit burnup can be calculated at equilibrium for a reactor in which the exit burnups for two zones may be adjusted to give radial power flattening and the fuelling schedules may be designed to give axial power flattening in one or both zones. (author)

  20. Performance assessment of new neutron cross section libraries using MCNP code and some critical benchmarks

    International Nuclear Information System (INIS)

    Bakkari, B El; Bardouni, T El.; Erradi, L.; Chakir, E.; Meroun, O.; Azahra, M.; Boukhal, H.; Khoukhi, T El.; Htet, A.

    2007-01-01

    Full text: New releases of nuclear data files made available during the few recent years. The reference MCNP5 code (1) for Monte Carlo calculations is usually distributed with only one standard nuclear data library for neutron interactions based on ENDF/B-VI. The main goal of this work is to process new neutron cross sections libraries in ACE continuous format for MCNP code based on the most recent data files recently made available for the scientific community : ENDF/B-VII.b2, ENDF/B-VI (release 8), JEFF3.0, JEFF-3.1, JENDL-3.3 and JEF2.2. In our data treatment, we used the modular NJOY system (release 99.9) (2) in conjunction with its most recent upadates. Assessment of the processed point wise cross sections libraries performances was made by means of some criticality prediction and analysis of other integral parameters for a set of reactor benchmarks. Almost all the analyzed benchmarks were taken from the international handbook of Evaluated criticality safety benchmarks experiments from OECD (3). Some revised benchmarks were taken from references (4,5). These benchmarks use Pu-239 or U-235 as the main fissionable materiel in different forms, different enrichments and cover various geometries. Monte Carlo calculations were performed in 3D with maximum details of benchmark description and the S(α,β) cross section treatment was adopted in all thermal cases. The resulting one standard deviation confidence interval for the eigenvalue is typically +/-13% to +/-20 pcm [fr

  1. 'ACTIV' - a package of codes for charged particle and neutron activation analysis

    International Nuclear Information System (INIS)

    Cincu, Em.; Alexandreanu, B.; Manu, V.; Moisa, V.

    1997-01-01

    The 'ACTIV' Program is an advanced software package dedicated to applications of the thermal neutron and charged particle activation (NAA and CPA) induced reactions. The program is designed to run on personal computers compatible IBM PC-Models XT/AT, 286 or more advanced, operating under DOS version 5.0 or later, on systems with minimum 5 MB of hard disk memory. The package consists of 6 software modules and a Nuclear Data Base comprising physical, nuclear reaction and decay data for: thermal neutron, proton, deuteron and α-particle induced reactions on 15 selected metallic elements; the nuclear reaction data corresponds to the energy range (5-100) MeV. In the first version - ACTIV 1.0 - the set of input data concerns: the sample type, irradiation and measurement conditions, the γ-ray spectrum identification code, selected detection efficiency calibration curve, selected radionuclides, selected standardization method for elemental analysis, version of results. At present, the 'ACTIV' package comprises 6 soft modules for processing the experimental data, which ensure computation of the quantities: radionuclide activities, activation yield data (case of CPA) and elemental concentration by relative and absolute standardization methods. Recently, the software designed to processing complex γ-ray spectra was acquired and installed on our PC 486 (8 MB RAM, 100 MHz). The next step in developing the 'ACTIV' program envisages improving the existing computing codes, completing the data libraries, incorporating a new soft for the direct use of the 'Quantum TM MCA' data, developing modules dedicated to uncertainty computation and optimization of the activation experiments

  2. VENTURE: a code block for solving multigroup neutronics problems applying the finite-difference diffusion-theory approximation to neutron transport, version II

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.

    1977-11-01

    The report documents the computer code block VENTURE designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P 1 ) in up to three-dimensional geometry. It uses and generates interface data files adopted in the cooperative effort sponsored by the Reactor Physics Branch of the Division of Reactor Research and Development of the Energy Research and Development Administration. Several different data handling procedures have been incorporated to provide considerable flexibility; it is possible to solve a wide variety of problems on a variety of computer configurations relatively efficiently

  3. VENTURE: a code block for solving multigroup neutronics problems applying the finite-difference diffusion-theory approximation to neutron transport, version II. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.

    1977-11-01

    The report documents the computer code block VENTURE designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P/sub 1/) in up to three-dimensional geometry. It uses and generates interface data files adopted in the cooperative effort sponsored by the Reactor Physics Branch of the Division of Reactor Research and Development of the Energy Research and Development Administration. Several different data handling procedures have been incorporated to provide considerable flexibility; it is possible to solve a wide variety of problems on a variety of computer configurations relatively efficiently.

  4. VENTURE: a code block for solving multigroup neutronics problems applying the finite-difference diffusion-theory approximation to neutron transport

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.

    1975-10-01

    The computer code block VENTURE, designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P 1 ) in up to three-dimensional geometry is described. A variety of types of problems may be solved: the usual eigenvalue problem, a direct criticality search on the buckling, on a reciprocal velocity absorber (prompt mode), or on nuclide concentrations, or an indirect criticality search on nuclide concentrations, or on dimensions. First-order perturbation analysis capability is available at the macroscopic cross section level

  5. 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.

  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. MCTP, a code for the thermo-mechanical analysis of a fuel rod of BWR type reactors (Neutron part)

    International Nuclear Information System (INIS)

    Hernandez L, H.; Ortiz V, J.

    2003-01-01

    In the National Institute of Nuclear Research of Mexico a code for the thermo-mechanical analysis of the fuel rods of the BWR type reactors of the Nucleo electric Central of Laguna Verde is developed. The code solves the diffusion equation in cylindrical coordinates with several energy groups. The code, likewise, calculates the temperature distribution and power distribution in those fuel rods. The code is denominated Multi groups With Temperatures and Power (MCTP). In the code, the energy with which the fission neutrons are emitted it is divided in six groups. They are also considered the produced perturbations by the changes in the temperatures of the materials that constitute the fuel rods, the content of fission products, the uranium consumption and in its case the gadolinium, as well as the plutonium production. In this work there are present preliminary results obtained with the code, using data of operation of the Nucleo electric Central of Laguna Verde. (Author)

  8. The bidimensional neutron transport code TWOTRAN-GG. Users manual and input data TWOTRAN-TRACA version

    International Nuclear Information System (INIS)

    Ahnert, C.; Aragones, J. M.

    1981-01-01

    This Is a users manual of the neutron transport code TWOTRAN-TRACA, which is a version of the original TWOTRAN-GG from the Los Alamos Laboratory, with some modifications made at JEN. A detailed input data description is given as well as the new modifications developed at JEN. (Author) 8 refs

  9. The bidimensional neutron transport code Twotran-GG. User's manual and input data. Twotran-Traca version

    International Nuclear Information System (INIS)

    Ahnert, C.; Aragones, J.M.

    1981-01-01

    A user's manual of the neutron transport code Twotran-Traca is presented; it is a version of the original Twotran-GG from the Los Alamos Laboratory, with some modifications made at J.E.N., Spain. A detailed input data description is given as well as the new modifications developped at J.E.N. (author) [es

  10. Measured performances on vectorization and multitasking with a Monte Carlo code for neutron transport problems

    International Nuclear Information System (INIS)

    Chauvet, Y.

    1985-01-01

    This paper summarized two improvements of a real production code by using vectorization and multitasking techniques. After a short description of Monte Carlo algorithms employed in neutron transport problems, the authors briefly describe the work done in order to get a vector code. Vectorization principles are presented and measured performances on the CRAY 1S, CYBER 205 and CRAY X-MP compared in terms of vector lengths. The second part of this work is an adaptation to multitasking on the CRAY X-MP using exclusively standard multitasking tools available with FORTRAN under the COS 1.13 system. Two examples are presented. The goal of the first one is to measure the overhead inherent to multitasking when tasks become too small and to define a granularity threshold, that is to say a minimum size for a task. With the second example they propose a method that is very X-MP oriented in order to get the best speedup factor on such a computer. In conclusion they prove that Monte Carlo algorithms are very well suited to future vector and parallel computers

  11. The TORT three-dimensional discrete ordinates neutron/photon transport code (TORT version 3)

    Energy Technology Data Exchange (ETDEWEB)

    Rhoades, W.A.; Simpson, D.B.

    1997-10-01

    TORT calculates the flux or fluence of neutrons and/or photons throughout three-dimensional systems due to particles incident upon the system`s external boundaries, due to fixed internal sources, or due to sources generated by interaction with the system materials. The transport process is represented by the Boltzman transport equation. The method of discrete ordinates is used to treat the directional variable, and a multigroup formulation treats the energy dependence. Anisotropic scattering is treated using a Legendre expansion. Various methods are used to treat spatial dependence, including nodal and characteristic procedures that have been especially adapted to resist numerical distortion. A method of body overlay assists in material zone specification, or the specification can be generated by an external code supplied by the user. Several special features are designed to concentrate machine resources where they are most needed. The directional quadrature and Legendre expansion can vary with energy group. A discontinuous mesh capability has been shown to reduce the size of large problems by a factor of roughly three in some cases. The emphasis in this code is a robust, adaptable application of time-tested methods, together with a few well-tested extensions.

  12. Resonance self-shielding methodology of new neutron transport code STREAM

    International Nuclear Information System (INIS)

    Choi, Sooyoung; Lee, Hyunsuk; Lee, Deokjung; Hong, Ser Gi

    2015-01-01

    This paper reports on the development and verification of three new resonance self-shielding methods. The verifications were performed using the new neutron transport code, STREAM. The new methodologies encompass the extension of energy range for resonance treatment, the development of optimum rational approximation, and the application of resonance treatment to isotopes in the cladding region. (1) The extended resonance energy range treatment has been developed to treat the resonances below 4 eV of three resonance isotopes and shows significant improvements in the accuracy of effective cross sections (XSs) in that energy range. (2) The optimum rational approximation can eliminate the geometric limitations of the conventional approach of equivalence theory and can also improve the accuracy of fuel escape probability. (3) The cladding resonance treatment method makes it possible to treat resonances in cladding material which have not been treated explicitly in the conventional methods. These three new methods have been implemented in the new lattice physics code STREAM and the improvement in the accuracy of effective XSs is demonstrated through detailed verification calculations. (author)

  13. Measured performances on vectorization and multitasking with a Monte Carlo code for neutron transport problems

    International Nuclear Information System (INIS)

    Chauvet, Y.

    1985-01-01

    This paper summarized two improvements of a real production code by using vectorization and multitasking techniques. After a short description of Monte Carlo algorithms employed in our neutron transport problems, we briefly describe the work we have done in order to get a vector code. Vectorization principles will be presented and measured performances on the CRAY 1S, CYBER 205 and CRAY X-MP compared in terms of vector lengths. The second part of this work is an adaptation to multitasking on the CRAY X-MP using exclusively standard multitasking tools available with FORTRAN under the COS 1.13 system. Two examples will be presented. The goal of the first one is to measure the overhead inherent to multitasking when tasks become too small and to define a granularity threshold, that is to say a minimum size for a task. With the second example we propose a method that is very X-MP oriented in order to get the best speedup factor on such a computer. In conclusion we prove that Monte Carlo algorithms are very well suited to future vector and parallel computers. (orig.)

  14. Using MCNP-4C code for design of the thermal neutron beam for neutron radiography at the MNSR

    International Nuclear Information System (INIS)

    Shaaban, I.

    2009-11-01

    Studies were carried out for determination of the parameters of a thermal neutron beam at the MNSR reactor (MNSR-30 kW) for neutron radiography in the vertical beam port by using the MCNP-4C (Monte Carlo Neutron - Photon transport). Thermal, epithermal and fast neutron energy ranges were selected as 10 keV respectively. To produce a good neutron beam in terms of intensity and quality, several materials Lead (Pb), Bismuth (Bi), Borated polyethelyene and Alumina Oxide (Al 2 O 3 ) were used as neutron and photon filters. Based on the current design, the L/D of the facility ranges between 125, 110 and 90. The thermal neutron flux at the beam exit is 1.436x10 5 n/cm2 .s ,1.843x10 5 n/cm2 .s and 2.845x10 5 n/cm2 .s respectively, middots with a Cd-ratio of ∼ 2.829, 2.766, 3.191 for the L/D = 125, 110, 90 respectively. The estimated values for gamma doses are 6.705x10 -2 Rem/h and 1.275x10 -1 Rem/h and 2.678x10 -1 Rem/ h with bismuth. The divergent angle of the collimator is 1.348 degree - 2.021 degree. Such neutron beams, if built into the Syrian MNSR reactor, could support the application of NRG in Syria. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-03-01

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

  16. IAEA activities to prepare safety codes and guides for thermal neutron nuclear power plants

    International Nuclear Information System (INIS)

    Iansiti, E.

    1977-01-01

    In accordance with the programme presented to, and endorsed by, the eighteenth General Conference in September 1974, the IAEA is now developing a complete set of safety codes and guides that will represent recommendations for the safety of thermal neutron power plants. The safety codes outline the minimum requirements for achieving this safety, and the safety guides set forth the criteria, procedures and methods to implement the safety codes. The whole programme is directed towards the five areas of Governmental Organization, Siting, Design, Operation, and Quality Assurance. One Scientific Secretary from the Agency Secretariat is responsible for each of these areas and a Co-ordinator takes care of common problems. For the development of each of these documents a working group of a few world experts is first convened which prepare a preliminary draft. This draft is then reviewed by a larger, international Technical Review Committee (one for each of the five areas) and a subsequent review by the Senior Advisory Group - with representatives from 20 states - ensures that the document is well coordinated within the programme. At this stage, it is sent to Member States for comments. The Technical Review Committee concerned is reconvened to integrate these comments into the document, and, after a final review by the Senior Advisory Group, the document is ready for transmission to the Director General of the Agency for endorsement and publication. A preliminary to this procedure is the collation by the Secretariat of large amounts of information submitted by Member States so that the first draft is really based on a very complete knowledge of what is done in each area all over the world. This collation frequently reveals differences in approach which are not random but due, rather, to the local conditions and the types of reactors. These differences must be harmonized in the documents produced without detracting from the effectiveness of the code or guide. The whole

  17. DIANE, a simulation code for the interaction of neutrons with living tissues. Application to low doses of fast neutrons on human tumoral cells

    International Nuclear Information System (INIS)

    Nenot, M.L.

    2003-07-01

    Our work deals with the irradiation of cells and living tissues by 14 MeV neutrons at very low doses (a few 10 -2 Gy). Such experiments require an accurate knowledge of the values of neutron dose rates and fluences at the level of cell cultures. We have performed measurements of fluence rates through an activation method applied to gold and copper foils. The fluence rate is deduced from the gamma rays emitted by the irradiated foils. Neutron doses and dose rates have been measured through varied methods: PIN diodes, ionization tissue equivalent chambers, and Geiger-Mueller counters. We have designed the DIANE code to simulate the impact of energetic neutrons on cells. This code can be used with isolated cells or macroscopic tissues, it takes into account the roles of the ionisation electrons produced by recoil nuclei entering the cell. This point is all the more important since recent works have highlighted the impact of very low energy electrons on DNA. (A.C.)

  18. TITAN: an advanced three-dimensional neutronics/thermal-hydraulics code for light water reactor safety analysis

    International Nuclear Information System (INIS)

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

    1982-01-01

    The initial development of TITAN, a three-dimensional coupled neutronics/thermal-hydraulics code for LWR safety analysis, has been completed. The transient neutronics code QUANDRY has been joined to the two-fluid thermal-hydraulics code THERMIT with the appropriate feedback mechanisms modeled. A detailed steady-state and transient coupling scheme based on the tandem technique was implemented in accordance with the important structural and operational characteristics of QUANDRY and THERMIT. A two channel sample problem formed the basis for steady-state and transient analyses performed with TITAN. TITAN steady-state results were compared with those obtained with MEKIN and showed good agreement. Null transients, simulated turbine trip transients, and a rod withdrawal transient were analyzed with TITAN and reasonable results were obtained

  19. Evaluation of a new neutron energy spectrum unfolding code based on an Adaptive Neuro-Fuzzy Inference System (ANFIS).

    Science.gov (United States)

    Hosseini, Seyed Abolfazl; Esmaili Paeen Afrakoti, Iman

    2018-01-17

    The purpose of the present study was to reconstruct the energy spectrum of a poly-energetic neutron source using an algorithm developed based on an Adaptive Neuro-Fuzzy Inference System (ANFIS). ANFIS is a kind of artificial neural network based on the Takagi-Sugeno fuzzy inference system. The ANFIS algorithm uses the advantages of both fuzzy inference systems and artificial neural networks to improve the effectiveness of algorithms in various applications such as modeling, control and classification. The neutron pulse height distributions used as input data in the training procedure for the ANFIS algorithm were obtained from the simulations performed by MCNPX-ESUT computational code (MCNPX-Energy engineering of Sharif University of Technology). Taking into account the normalization condition of each energy spectrum, 4300 neutron energy spectra were generated randomly. (The value in each bin was generated randomly, and finally a normalization of each generated energy spectrum was performed). The randomly generated neutron energy spectra were considered as output data of the developed ANFIS computational code in the training step. To calculate the neutron energy spectrum using conventional methods, an inverse problem with an approximately singular response matrix (with the determinant of the matrix close to zero) should be solved. The solution of the inverse problem using the conventional methods unfold neutron energy spectrum with low accuracy. Application of the iterative algorithms in the solution of such a problem, or utilizing the intelligent algorithms (in which there is no need to solve the problem), is usually preferred for unfolding of the energy spectrum. Therefore, the main reason for development of intelligent algorithms like ANFIS for unfolding of neutron energy spectra is to avoid solving the inverse problem. In the present study, the unfolded neutron energy spectra of 252Cf and 241Am-9Be neutron sources using the developed computational code were

  20. 75 FR 22551 - Continuation of Hearing on the Department of Justice's Actions Related to the New Black Panther...

    Science.gov (United States)

    2010-04-29

    ... to the New Black Panther Party Litigation and its Enforcement of Section 11(b) of the Voting Rights... New Black Panther Party Litigation and its Enforcement of Section 11(b) of the Voting Rights Act... Department of Justice's actions in the New Black Panther Party Litigation and Enforcement of Section 11(b) of...

  1. A habitat assessment for Florida panther population expansion into central Florida

    Science.gov (United States)

    Thatcher, C.A.; Van Manen, F.T.; Clark, J.D.

    2009-01-01

    One of the goals of the Florida panther (Puma concolor coryi) recovery plan is to expand panther range north of the Caloosahatchee River in central Florida. Our objective was to evaluate the potential of that region to support panthers. We used a geographic information system and the Mahalanobis distance statistic to develop a habitat model based on landscape characteristics associated with panther home ranges. We used cross-validation and an independent telemetry data set to test the habitat model. We also conducted a least-cost path analysis to identify potential habitat linkages and to provide a relative measure of connectivity among habitat patches. Variables in our model were paved road density, major highways, human population density, percentage of the area permanently or semipermanently flooded, and percentage of the area in natural land cover. Our model clearly identified habitat typical of that found within panther home ranges based on model testing with recent telemetry data. We identified 4 potential translocation sites that may support a total of approximately 36 panthers. Although we identified potential habitat linkages, our least-cost path analyses highlighted the extreme isolation of panther habitat in portions of the study area. Human intervention will likely be required if the goal is to establish female panthers north of the Caloosahatchee in the near term.

  2. MCNP-DSP, Monte Carlo Neutron-Particle Transport Code with Digital Signal Processing

    International Nuclear Information System (INIS)

    2002-01-01

    1 - Description of program or function: MCNP-DSP is recommended only for experienced MCNP users working with subcritical measurements. It is a modification of the Los Alamos National Laboratory's Monte Carlo code MCNP4a that is used to simulate a variety of subcritical measurements. The DSP version was developed to simulate frequency analysis measurements, correlation (Rossi-) measurements, pulsed neutron measurements, Feynman variance measurements, and multiplicity measurements. CCC-700/MCNP4C is recommended for general purpose calculations. 2 - Methods:MCNP-DSP performs calculations very similarly to MCNP and uses the same generalized geometry capabilities of MCNP. MCNP-DSP can only be used with the continuous-energy cross-section data. A variety of source and detector options are available. However, unlike standard MCNP, the source and detector options are limited to those described in the manual because these options are specified in the MCNP-DSP extra data file. MCNP-DSP is used to obtain the time-dependent response of detectors that are modeled in the simulation geometry. The detectors represent actual detectors used in measurements. These time-dependent detector responses are used to compute a variety of quantities such as frequency analysis signatures, correlation signatures, multiplicity signatures, etc., between detectors or sources and detectors. Energy ranges are 0-60 MeV for neutrons (data generally only available up to 20 MeV) and 1 keV - 1 GeV for photons and electrons. 3 - Restrictions on the complexity of the problem: None noted

  3. Neutron transport-burnup code MCORGS and its application in fusion fission hybrid blanket conceptual research

    Science.gov (United States)

    Shi, Xue-Ming; Peng, Xian-Jue

    2016-09-01

    Fusion science and technology has made progress in the last decades. However, commercialization of fusion reactors still faces challenges relating to higher fusion energy gain, irradiation-resistant material, and tritium self-sufficiency. Fusion Fission Hybrid Reactors (FFHR) can be introduced to accelerate the early application of fusion energy. Traditionally, FFHRs have been classified as either breeders or transmuters. Both need partition of plutonium from spent fuel, which will pose nuclear proliferation risks. A conceptual design of a Fusion Fission Hybrid Reactor for Energy (FFHR-E), which can make full use of natural uranium with lower nuclear proliferation risk, is presented. The fusion core parameters are similar to those of the International Thermonuclear Experimental Reactor. An alloy of natural uranium and zirconium is adopted in the fission blanket, which is cooled by light water. In order to model blanket burnup problems, a linkage code MCORGS, which couples MCNP4B and ORIGEN-S, is developed and validated through several typical benchmarks. The average blanket energy Multiplication and Tritium Breeding Ratio can be maintained at 10 and 1.15 respectively over tens of years of continuous irradiation. If simple reprocessing without separation of plutonium from uranium is adopted every few years, FFHR-E can achieve better neutronic performance. MCORGS has also been used to analyze the ultra-deep burnup model of Laser Inertial Confinement Fusion Fission Energy (LIFE) from LLNL, and a new blanket design that uses Pb instead of Be as the neutron multiplier is proposed. In addition, MCORGS has been used to simulate the fluid transmuter model of the In-Zinerater from Sandia. A brief comparison of LIFE, In-Zinerater, and FFHR-E will be given.

  4. Improvement of calculation method for temperature coefficient of HTTR by neutronics calculation code based on diffusion theory. Analysis for temperature coefficient by SRAC code system

    International Nuclear Information System (INIS)

    Goto, Minoru; Takamatsu, Kuniyoshi

    2007-03-01

    The HTTR temperature coefficients required for the core dynamics calculations had been calculated from the HTTR core calculation results by the diffusion code with which the corrections had been performed using the core calculation results by the Monte-Carlo code MVP. This calculation method for the temperature coefficients was considered to have some issues to be improved. Then, the calculation method was improved to obtain the temperature coefficients in which the corrections by the Monte-Carlo code were not required. Specifically, from the point of view of neutron spectrum calculated by lattice calculations, the lattice model was revised which had been used for the calculations of the temperature coefficients. The HTTR core calculations were performed by the diffusion code with the group constants which were generated by the lattice calculations with the improved lattice model. The core calculations and the lattice calculations were performed by the SRAC code system. The HTTR core dynamics calculation was performed with the temperature coefficient obtained from the core calculation results. In consequence, the core dynamics calculation result showed good agreement with the experimental data and the valid temperature coefficient could be calculated only by the diffusion code without the corrections by Monte-Carlo code. (author)

  5. Calculation Of Fuel Burnup And Radionuclide Inventory In The Syrian Miniature Neutron Source Reactor Using The GETERA Code

    International Nuclear Information System (INIS)

    Khattab, K.; Dawahra, S.

    2011-01-01

    Calculations of the fuel burnup and radionuclide inventory in the Syrian Miniature Neutron Source Reactor (MNSR) after 10 years (the reactor core expected life) of the reactor operation time are presented in this paper using the GETERA code. The code is used to calculate the fuel group constants and the infinite multiplication factor versus the reactor operating time for 10, 20, and 30 kW operating power levels. The amounts of uranium burnup and plutonium produced in the reactor core, the concentrations and radionuclides of the most important fission product and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core were calculated using the GETERA code as well. It is found that the GETERA code is better than the WIMSD4 code for the fuel burnup calculation in the MNSR reactor since it is newer and has a bigger library of isotopes and more accurate. (author)

  6. Radiation transport and shielding information, computer codes, and nuclear data for use in CTR neutronics research and development

    International Nuclear Information System (INIS)

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

    1976-01-01

    The activities of the Radiation Shielding Information Center (RSIC) of the Oak Ridge National Laboratory are being utilized in support of fusion reactor technology. The major activities of RSIC include the operation of a computer-based information storage and retrieval system, the collection, packaging, and distribution of large computer codes, and the compilation and dissemination of processed and evaluated data libraries, with particular emphasis on neutron and gamma-ray cross-section data. The Center has acquired thirteen years of experience in serving fission reactor, weapons, and accelerator shielding research communities, and the extension of its technical base to fusion reactor research represents a logical progression. RSIC is currently working with fusion reactor researchers and contractors in computer code development to provide tested radiation transport and shielding codes and data library packages. Of significant interest to the CTR community are the 100 energy group neutron and 21 energy group gamma-ray coupled cross-section data package (DLC-37) for neutronics studies, a comprehensive 171 energy group neutron and 36 energy group gamma-ray coupled cross-section data base with retrieval programs, including resonance self-shielding, that are tailored to CTR application, and a data base for the generation of energy-dependent atomic displacement and gas production cross sections and heavy-particle-recoil spectra for estimating radiation damage to CTR structural components

  7. Comparative study of Monte Carlo particle transport code PHITS and nuclear data processing code NJOY for PKA energy spectra and heating number under neutron irradiation

    International Nuclear Information System (INIS)

    Iwamoto, Y.; Ogawa, T.

    2016-01-01

    The modelling of the damage in materials irradiated by neutrons is needed for understanding the mechanism of radiation damage in fission and fusion reactor facilities. The molecular dynamics simulations of damage cascades with full atomic interactions require information about the energy distribution of the Primary Knock on Atoms (PKAs). The most common process to calculate PKA energy spectra under low-energy neutron irradiation is to use the nuclear data processing code NJOY2012. It calculates group-to-group recoil cross section matrices using nuclear data libraries in ENDF data format, which is energy and angular recoil distributions for many reactions. After the NJOY2012 process, SPKA6C is employed to produce PKA energy spectra combining recoil cross section matrices with an incident neutron energy spectrum. However, intercomparison with different processes and nuclear data libraries has not been studied yet. Especially, the higher energy (~5 MeV) of the incident neutrons, compared to fission, leads to many reaction channels, which produces a complex distribution of PKAs in energy and type. Recently, we have developed the event generator mode (EGM) in the Particle and Heavy Ion Transport code System PHITS for neutron incident reactions in the energy region below 20 MeV. The main feature of EGM is to produce PKA with keeping energy and momentum conservation in a reaction. It is used for event-by-event analysis in application fields such as soft error analysis in semiconductors, micro dosimetry in human body, and estimation of Displacement per Atoms (DPA) value in metals and so on. The purpose of this work is to specify differences of PKA spectra and heating number related with kerma between different calculation method using PHITS-EGM and NJOY2012+SPKA6C with different libraries TENDL-2015, ENDF/B-VII.1 and JENDL-4.0 for fusion relevant materials

  8. Performance of a neutron transport code with full phase space decomposition on the Cray Research T3D

    International Nuclear Information System (INIS)

    Dorr, M.R.; Salo, E.M.

    1995-01-01

    We present performance results obtained on a 128-node Cray Research T3D computer by a neutron transport code implementing a standard mtiltigroup, discrete ordinates algorithm on a three-dimensional Cartesian grid. After summarizing the implementation strategy used to obtain a full decomposition of phase space (i.e., simultaneous parallelization of the neutron energy, directional and spatial variables), we investigate the scalability of the fundamental source iteration step with respect to each phase space variable. We also describe enhancements that have enabled performance rates approaching 10 gigaflops on the full 128-node machine

  9. Neutrons and gamma transport in atmosphere by Tripoli-2 code. Energy deposit and electron current time function

    International Nuclear Information System (INIS)

    Vergnaud, T.; Nimal, J.C.; Ulpat, J.P.; Faucheux, G.

    1988-01-01

    The Tripoli-2 computer code has been adapted to calculate, in addition to energy deposit in matter by neutrons (Kerma) the energy deposit by gamma produced in neutronic impacts and the induced recoil electron current. The energy deposit conduces at air ionization, consequently at a conductibility. This knowledge added at that of electron current permit to resolve the Maxwell equations of electromagnetic field. The study is realized for an atmospheric explosion 100 meters high. The calculations of energy deposit and electron current have been conducted as far as 2.5km [fr

  10. 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)

  11. SOURCES-3A: A code for calculating (α, n), spontaneous fission, and delayed neutron sources and spectra

    International Nuclear Information System (INIS)

    Perry, R.T.; Wilson, W.B.; Charlton, W.S.

    1998-04-01

    In many systems, it is imperative to have accurate knowledge of all significant sources of neutrons due to the decay of radionuclides. These sources can include neutrons resulting from the spontaneous fission of actinides, the interaction of actinide decay α-particles in (α,n) reactions with low- or medium-Z nuclides, and/or delayed neutrons from the fission products of actinides. Numerous systems exist in which these neutron sources could be important. These include, but are not limited to, clean and spent nuclear fuel (UO 2 , ThO 2 , MOX, etc.), enrichment plant operations (UF 6 , PuF 4 , etc.), waste tank studies, waste products in borosilicate glass or glass-ceramic mixtures, and weapons-grade plutonium in storage containers. SOURCES-3A is a computer code that determines neutron production rates and spectra from (α,n) reactions, spontaneous fission, and delayed neutron emission due to the decay of radionuclides in homogeneous media (i.e., a mixture of α-emitting source material and low-Z target material) and in interface problems (i.e., a slab of α-emitting source material in contact with a slab of low-Z target material). The code is also capable of calculating the neutron production rates due to (α,n) reactions induced by a monoenergetic beam of α-particles incident on a slab of target material. Spontaneous fission spectra are calculated with evaluated half-life, spontaneous fission branching, and Watt spectrum parameters for 43 actinides. The (α,n) spectra are calculated using an assumed isotropic angular distribution in the center-of-mass system with a library of 89 nuclide decay α-particle spectra, 24 sets of measured and/or evaluated (α,n) cross sections and product nuclide level branching fractions, and functional α-particle stopping cross sections for Z < 106. The delayed neutron spectra are taken from an evaluated library of 105 precursors. The code outputs the magnitude and spectra of the resultant neutron source. It also provides an

  12. A new three-tier architecture design for multi-sphere neutron spectrometer with the FLUKA code

    Science.gov (United States)

    Huang, Hong; Yang, Jian-Bo; Tuo, Xian-Guo; Liu, Zhi; Wang, Qi-Biao; Wang, Xu

    2016-07-01

    The current commercially, available Bonner sphere neutron spectrometer (BSS) has high sensitivity to neutrons below 20 MeV, which causes it to be poorly placed to measure neutrons ranging from a few MeV to 100 MeV. The paper added moderator layers and the auxiliary material layer upon 3He proportional counters with FLUKA code, with a view to improve. The results showed that the responsive peaks to neutrons below 20 MeV gradually shift to higher energy region and decrease slightly with the increasing moderator thickness. On the contrary, the response for neutrons above 20 MeV was always very low until we embed auxiliary materials such as copper (Cu), lead (Pb), tungsten (W) into moderator layers. This paper chose the most suitable auxiliary material Pb to design a three-tier architecture multi-sphere neutron spectrometer (NBSS). Through calculating and comparing, the NBSS was advantageous in terms of response for 5-100 MeV and the highest response was 35.2 times the response of polyethylene (PE) ball with the same PE thickness.

  13. Applying the universal neutron transport codes to the calculation of well-logging probe response at different rock porosities

    International Nuclear Information System (INIS)

    Bogacz, J.; Loskiewicz, J.; Zazula, J.M.

    1991-01-01

    The use of universal neutron transport codes in order to calculate the parameters of well-logging probes presents a new approach first tried in U.S.A. and UK in the eighties. This paper deals with first such an attempt in Poland. The work is based on the use of MORSE code developed in Oak Ridge National Laboratory in U.S.A.. Using CG MORSE code we calculated neutron detector response when surrounded with sandstone of porosities 19% and 38%. During the work it come out that it was necessary to investigate different methods of estimation of the neutron flux. The stochastic estimation method as used currently in the original MORSE code (next collision approximation) can not be used because of slow convergence of its variance. Using the analog type of estimation (calculation of the sum of track lengths inside detector) we obtained results of acceptable variance (∼ 20%) for source-detector spacing smaller than 40 cm. The influence of porosity on detector response is correctly described for detector positioned at 27 cm from the source. At the moment the variances are quite large. (author). 33 refs, 8 figs, 8 tabs

  14. ACDOS1: a computer code to calculate dose rates from neutron activation of neutral beamlines and other fusion-reactor components

    International Nuclear Information System (INIS)

    Keney, G.S.

    1981-08-01

    A computer code has been written to calculate neutron induced activation of neutral-beam injector components and the corresponding dose rates as a function of geometry, component composition, and time after shutdown. The code, ACDOS1, was written in FORTRAN IV to calculate both activity and dose rates for up to 30 target nuclides and 50 neutron groups. Sufficient versatility has also been incorporated into the code to make it applicable to a variety of general activation problems due to neutrons of energy less than 20 MeV

  15. Development of additional module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, K.N.; Bogomazov, D.N.; Poliakov, N. [Moscow Power Engineering Institute (Technical University), Moscow (Russian Federation)

    2007-07-01

    The new special module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation is worked out. The Russian computer code Rainbow has been selected for joint use with a developed module. This code system provides the possibility of EFOCP (Eigen Frequencies of Oscillations of the Coolant Pressure) calculations in any coolant acoustical elements of primary circuits of NPP. EFOCP values have been calculated for transient and for stationary operating. The calculated results for nominal operating were compared with results of measured EFOCP. For example, this comparison was provided for the system: 'pressurizer + surge line' of a WWER-1000 reactor. The calculated result 0.58 Hz practically coincides with the result of measurement (0.6 Hz). The EFOCP variations in transients are also shown. The presented results are intended to be useful for NPP vibration-acoustical certification. There are no serious difficulties for using this module with other computer codes.

  16. Development of additional module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation

    International Nuclear Information System (INIS)

    Proskuryakov, K.N.; Bogomazov, D.N.; Poliakov, N.

    2007-01-01

    The new special module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation is worked out. The Russian computer code Rainbow has been selected for joint use with a developed module. This code system provides the possibility of EFOCP (Eigen Frequencies of Oscillations of the Coolant Pressure) calculations in any coolant acoustical elements of primary circuits of NPP. EFOCP values have been calculated for transient and for stationary operating. The calculated results for nominal operating were compared with results of measured EFOCP. For example, this comparison was provided for the system: 'pressurizer + surge line' of a WWER-1000 reactor. The calculated result 0.58 Hz practically coincides with the result of measurement (0.6 Hz). The EFOCP variations in transients are also shown. The presented results are intended to be useful for NPP vibration-acoustical certification. There are no serious difficulties for using this module with other computer codes

  17. Use of CITATION code for flux calculation in neutron activation analysis with voluminous sample using an Am-Be source

    International Nuclear Information System (INIS)

    Khelifi, R.; Idiri, Z.; Bode, P.

    2002-01-01

    The CITATION code based on neutron diffusion theory was used for flux calculations inside voluminous samples in prompt gamma activation analysis with an isotopic neutron source (Am-Be). The code uses specific parameters related to the energy spectrum source and irradiation system materials (shielding, reflector). The flux distribution (thermal and fast) was calculated in the three-dimensional geometry for the system: air, polyethylene and water cuboidal sample (50x50x50 cm). Thermal flux was calculated in a series of points inside the sample. The results agreed reasonably well with observed values. The maximum thermal flux was observed at a distance of 3.2 cm while CITATION gave 3.7 cm. Beyond a depth of 7.2 cm, the thermal flux to fast flux ratio increases up to twice and allows us to optimise the detection system position in the scope of in-situ PGAA

  18. Neutron Fluence Evaluation of Reactor Internal Structure Using 3D Transport Calculation Code, RAPTOR-M3G

    International Nuclear Information System (INIS)

    Maeng, YoungJae; Lim, MiJoung; Kim, KyungSik; Cho, YoungKi; Yoo, ChoonSung; Kim, ByoungChul

    2015-01-01

    Age-related degradation mechanisms are including the irradiation-assisted stress corrosion cracking(IASCC), void swelling, stress relaxation, fatigue, and etc. A lot of Baffle Former Bolts(BFBs) was installed at the former plate ends between baffle and barrel structure. These would undergo severe experiences, which are high temperature and pressure, bypass water flow and neutron exposure and have some radioactive limitation in inspecting their integrity. The objectives of this paper is to evaluate fast neutron fluence(n/cm 2 , E>1.0MeV) for PWR internals using 3D transport calculation code, RAPTOR-M3G, and to figure out a strategy to manage the effects of aging in PWR internals. One of age-related degradation mechanisms, IASCC, which is affected by fast neutron exposure rate, has been currently issued for PWR internals and has 2 x 10 21 (n/cm 2 ) of the threshold value by MRP-175. Because a lot of BFBs was installed around the internal components, closer inspections are required. As part of an aging management for Kori unit 2, 3D transport calculation code, RAPTOR-M3G, was applied for determining fast neutron fluence at baffle, barrel and former plates regions. As a result, the fast neutron fluence exceeds the screening or threshold values of IASCC in all of baffle, barrel and former plate region. And the most significant region is the baffle because it is located closest to the core. In addition, some regions including former plate tend to be more damaged because of less moderate ability than water. In conclusion, Ice's has been progressed for PWR internals of Kori unit 2. Several regions of internal components were damaged by fast neutron exposure and increase in size as time goes by

  19. Comparison of neutron diffusion theory codes in two and three space dimensions using a sodium cooled fast reactor benchmark

    International Nuclear Information System (INIS)

    Butland, A.T.D.; Putney, J.; Sweet, D.W.

    1980-04-01

    This report describes work performed to compare two UK neutron diffusion theory codes, TIGAR and SNAP, with published results for eight other codes available abroad. Both mesh edge and mesh centred finite difference diffusion theory codes as well as one axial synthesis code are included in the comparison and a range of iteration procedures are used by them. Comparison is made of calculations for a model of the sodium cooled fast reactor SNR-300 in both triangular and rectangular geometry and for a range of spatial meshes, enabling extrapolations to infinite mesh to be made. Calculated values of the effective multiplication constant, keff, for all the codes, agree very well when extrapolated to infinite mesh, indicating that no significant errors arising from the finite difference approximation but independent of mesh spacing are present in the calculations. The variation of keff with mesh area is found to be linear for the small meshes considered here, with the gradients for the mesh centred and mesh edged codes being of opposite sign. The results obtained using the mesh centred codes TIGAR, SNAP and CITATION agree closely with one another for all the meshes considered; the mesh edge codes agree less closely. (author)

  20. The Panther Mountain circular structure, a possible buried meteorite crater

    Science.gov (United States)

    Isachsen, Y. W.; Wright, S. F.; Revetta, F. A.; Duneen, R. J.

    Panther Mountain, located near Phoenicia, New York, is part of the Catskill Mountains, which form the eastern end of the Allegheny Plateau in New York. It is a circular mass defined physiographically by an anomalous circular drainage pattern produced by Esopus Creek and its tributary Woodland Creek. The circular valley that rings the mountain is fracture-controlled; where bedrock is exposed, it shows a joint density 5 to 10 times greater than that on either side of the valley. Where obscured by alluvial valley fill, the bedrock's low seismic velocity suggests that this anomalous fracturing is continuous in the bedrock underlying the rim valley. North-south and east-west gravity and magnetic profiles were made across the structure. Terrane-corrected, residual gravity profiles show an 18-mgal negative anomaly, and very steep gradients indicate a near-surface source. Several possible explanations of the gravity data were modeled. We conclude that the Panther Mountain circular structure is probably a buried meteorite crater that formed contemporaneously with marine or fluvial sedimentation during Silurian or Devonian time. An examination of drill core and cuttings in the region is underway to search for ejecta deposits and possible seismic and tsunami effects in the sedimentary section. Success would result in both dating the impact and furnishing a chronostratigraphic marker horizon.

  1. Party Political Panthers: Hegemonic Tamil Politics and the Dalit Challenge

    Directory of Open Access Journals (Sweden)

    Hugo Gorringe

    2011-12-01

    Full Text Available The Viduthalai Ciruthaigal Katchi (VCK, Liberation Panther Party has successfully transformed from the largest Dalit movement in Tamil Nadu into a recognised political organisation. Social movement theorists like Gamson (1990 view political recognition and engagement as one of the main aims and successes of social mobilisation. Despite the obvious achievements of the VCK, however, activists and commentators express disappointment or disillusionment with its performance. The Panthers clearly reject the caste hierarchy, but they increasingly adopt hegemonic forms of politics which can undermine their aims. This paper, thus, engages with the questions of movement institutionalisation by tracing the political trajectory of the VCK and charting its resistance to and compliance with Dravidian hegemony. It argues that institutionalisation needs to be understood within particular socio-political contexts and notes how the hegemony of Dravidian politics partly explains the disjuncture between activist and political perceptions. It portrays how the dominant political parties have set the template for what it means to ‘do’ politics in Tamil Nadu which serves as both an opportunity and a constraint for potential challengers.

  2. 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)

  3. 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)

  4. Neutronics analysis of the International Thermonuclear Experimental Reactor (ITER) MCNP ''Benchmark CAD Model'' with the ATTILA discrete ordinance code

    International Nuclear Information System (INIS)

    Youssef, M.Z.; Feder, R.; Davis, I.

    2007-01-01

    The ITER IT has adopted the newly developed FEM, 3-D, and CAD-based Discrete Ordinates code, ATTILA for the neutronics studies contingent on its success in predicting key neutronics parameters and nuclear field according to the stringent QA requirements set forth by the Management and Quality Program (MQP). ATTILA has the advantage of providing a full flux and response functions mapping everywhere in one run where components subjected to excessive radiation level and strong streaming paths can be identified. The ITER neutronics community had agreed to use a standard CAD model of ITER (40 degree sector, denoted ''Benchmark CAD Model'') to compare results for several responses selected for calculation benchmarking purposes to test the efficiency and accuracy of the CAD-MCNP approach developed by each party. Since ATTILA seems to lend itself as a powerful design tool with minimal turnaround time, it was decided to benchmark this model with ATTILA as well and compare the results to those obtained with the CAD MCNP calculations. In this paper we report such comparison for five responses, namely: (1) Neutron wall load on the surface of the 18 shield blanket module (SBM), (2) Neutron flux and nuclear heating rate in the divertor cassette, (3) nuclear heating rate in the winding pack of the inner leg of the TF coil, (4) Radial flux profile across dummy port plug and shield plug placed in the equatorial port, and (5) Flux at seven point locations situated behind the equatorial port plug. (orig.)

  5. The coupling of the Star-Cd software to a whole-core neutron transport code Decart for PWR applications

    International Nuclear Information System (INIS)

    Thomas, J.W.; Lee, H.C.; Downar, T.J.; Sofu, T.; Weber, D.P.; Joo, H.G.; Cho, J.Y.

    2003-01-01

    As part of a U.S.- Korea collaborative U.S. Department of Energy INERI project, a comprehensive high-fidelity reactor-core modeling capability is being developed for detailed analysis of existing and advanced PWR reactor designs. An essential element of the project has been the development of an interface between the computational fluid dynamics (CFD) module, STAR-CD, and the neutronics module, DeCART. Since the computational mesh for CFD and neutronics calculations are generally different, the capability to average and decompose data on these different meshes has been an important part of code coupling activities. An averaging process has been developed to extract neutronics zone temperatures in the fuel and coolant and to generate appropriate multi group cross sections and densities. Similar procedures have also been established to map the power distribution from the neutronics zones to the mesh structure used in the CFD module. Since MPI is used as the parallel model in STAR-CD and conflicts arise during initiation of a second level of MPI, the interface developed here is based on using TCP/IP protocol sockets to establish communication between the CFD and neutronics modules. Preliminary coupled calculations have been performed for PWR fuel assembly size problems and converged solutions have been achieved for a series of steady-state problems ranging from a single pin to a 1/8 model of a 17 x 17 PWR fuel assembly. (authors)

  6. Neutronic analysis for core conversion (HEU–LEU of the low power research reactor using the MCNP4C code

    Directory of Open Access Journals (Sweden)

    Aldawahra Saadou

    2015-06-01

    Full Text Available Comparative studies for conversion of the fuel from HEU to LEU in the miniature neutron source reactor (MNSR have been performed using the MCNP4C code. The HEU fuel (UAl4-Al, 90% enriched with Al clad and LEU (UO2 12.6% enriched with zircaloy-4 alloy clad cores have been analyzed in this study. The existing HEU core of MNSR was analyzed to validate the neutronic model of reactor, while the LEU core was studied to prove the possibility of fuel conversion of the existing HEU core. The proposed LEU core contained the same number of fuel pins as the HEU core. All other structure materials and dimensions of HEU and LEU cores were the same except the increase in the radius of control rod material from 0.195 to 0.205 cm and keeping the outer diameter of the control rod unchanged in the LEU core. The effective multiplication factor (keff, excess reactivity (ρex, control rod worth (CRW, shutdown margin (SDM, safety reactivity factor (SRF, delayed neutron fraction (βeff and the neutron fluxes in the irradiation tubes for the existing and the potential LEU fuel were investigated. The results showed that the safety parameters and the neutron fluxes in the irradiation tubes of the LEU fuels were in good agreements with the HEU results. Therefore, the LEU fuel was validated to be a suitable choice for fuel conversion of the MNSR in the future.

  7. A three dimensional calculation of neutron streaming through ITER tokamak pumping ducts with the Monte Carlo code TRIPOLI-2

    International Nuclear Information System (INIS)

    Bresard, I.; Diop, C.M.; Giancarli, L.; Gervaise, F.

    1991-01-01

    In the frame of the ITER tokamak project, the streaming of neutrons through pumping ducts up to the properly so called pumping system is studied. The gas evacuation device of the ITER plasma consists of a set of vacuum pumps which are located in a room which is outside the main machine building. These pumps receive the exhaust gas through several pumping ducts with a cross section of about four square meters and a length of about ten meters. Although insensitive to the magnetic field, the 14 MeV neutrons from plasma D-T thermonuclear reactions can penetrate in the divertor and reach the room pumping device by propagation through the bent ducts. Different components of this system, such as the bellows, turbomolecular pumps, etc., are irradiated and that raises radiation problems. In this study we determine, by using 3D Monte Carlo transport code TRIPOLI-2, neutron fluxes, dose rates and heatings due to neutrons which have streamed out the plasma through the bent ducts, at several points of the pumping room. Results show the neutron flux attenuation reachs a factor 10 -5 from plasma chamber to the pumping hall; the neutron heatings are estimated to 1.9x10 -3 W/cm 3 in bellow stainless steel at duct entrance, and 8x10 -7 W/cm 3 in the turbopumping stainless steel structure, inside pumping hall. The neutron fluxes obtained will be used to compute gamma source produced by radiative, inelastic process and gamma rays from formed activation products. Then, the knowledge of gamma source will allow to compute gamma dose rate and heating. The dose rates and heatings obtained will contribute to the definition of the ITER pumping system technical options and to establish pumping hall access conditions, also. (orig.)

  8. ARTEMIS: The core simulator of AREVA NP's next generation coupled neutronics/thermal-hydraulics code system ARCADIAR

    International Nuclear Information System (INIS)

    Hobson, Greg; Merk, Stephan; Bolloni, Hans-Wilhelm; Breith, Karl-Albert; Curca-Tivig, Florin; Van Geemert, Rene; Heinecke, Jochen; Hartmann, Bettina; Porsch, Dieter; Tiles, Viatcheslav; Dall'Osso, Aldo; Pothet, Baptiste

    2008-01-01

    AREVA NP has developed a next-generation coupled neutronics/thermal-hydraulics code system, ARCADIA R , to fulfil customer's current demands and even anticipate their future demands in terms of accuracy and performance. The new code system will be implemented world-wide and will replace several code systems currently used in various global regions. An extensive phase of verification and validation of the new code system is currently in progress. One of the principal components of this new system is the core simulator, ARTEMIS. Besides the stand-alone tests on the individual computational modules, integrated tests on the overall code are being performed in order to check for non-regression as well as for verification of the code. Several benchmark problems have been successfully calculated. Full-core depletion cycles of different plant types from AREVA's French, American and German regions (e.g. N4 and KONVOI types) have been performed with ARTEMIS (using APOLLO2-A cross sections) and compared directly with current production codes, e.g. with SCIENCE and CASCADE-3D, and additionally with measurements. (authors)

  9. 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

  10. A Conservation-Based Approach to Compensation for Livestock Depredation: The Florida Panther Case Study.

    Science.gov (United States)

    Jacobs, Caitlin E; Main, Martin B

    2015-01-01

    Calf (Bos taurus) depredation by the federally endangered Florida panther (Puma concolor coryi) on ranches in southwest Florida is an important issue because ranches represent mixed landscapes that provide habitat critical to panther recovery. The objectives of this study were to (1) quantify calf depredation by panthers on two ranches in southwest Florida, and (2) develop a habitat suitability model to evaluate the quality of panther hunting habitat on ranchlands, assess whether the model could predict predation risk to calves, and discuss its potential to be incorporated into an incentive-based compensation program. We ear-tagged 409 calves with VHF transmitters on two ranches during 2011-2013 to document calf mortality. We developed a model to evaluate the quality of panther hunting habitat on private lands in southwest Florida using environmental variables obtained from the Florida Natural Areas Inventory (FNAI) Cooperative Landcover Database and nocturnal GPS locations of panthers provided by the Florida Fish and Wildlife Conservation Commission (FWC). We then tested whether the model could predict the location of calf depredation sites. Tagged calf loss to panthers varied between the two ranches (0.5%/yr to 5.3%/yr) and may have been influenced by the amount of panther hunting habitat on each ranch as the ranch that experienced higher depredation rates contained a significantly higher probability of panther presence. Depredation sites of tagged calves had a significantly greater probability of panther presence than depredation sites of untagged calves that were found by ranchers in open pastures. This suggests that there may be more calves killed in high risk environments than are being found and reported by ranchers and that panthers can hunt effectively in open environments. It also suggests that the model may provide a means for evaluating the quality of panther hunting habitat and the corresponding risk of depredation to livestock across the landscape. We

  11. A Conservation-Based Approach to Compensation for Livestock Depredation: The Florida Panther Case Study.

    Directory of Open Access Journals (Sweden)

    Caitlin E Jacobs

    Full Text Available Calf (Bos taurus depredation by the federally endangered Florida panther (Puma concolor coryi on ranches in southwest Florida is an important issue because ranches represent mixed landscapes that provide habitat critical to panther recovery. The objectives of this study were to (1 quantify calf depredation by panthers on two ranches in southwest Florida, and (2 develop a habitat suitability model to evaluate the quality of panther hunting habitat on ranchlands, assess whether the model could predict predation risk to calves, and discuss its potential to be incorporated into an incentive-based compensation program. We ear-tagged 409 calves with VHF transmitters on two ranches during 2011-2013 to document calf mortality. We developed a model to evaluate the quality of panther hunting habitat on private lands in southwest Florida using environmental variables obtained from the Florida Natural Areas Inventory (FNAI Cooperative Landcover Database and nocturnal GPS locations of panthers provided by the Florida Fish and Wildlife Conservation Commission (FWC. We then tested whether the model could predict the location of calf depredation sites. Tagged calf loss to panthers varied between the two ranches (0.5%/yr to 5.3%/yr and may have been influenced by the amount of panther hunting habitat on each ranch as the ranch that experienced higher depredation rates contained a significantly higher probability of panther presence. Depredation sites of tagged calves had a significantly greater probability of panther presence than depredation sites of untagged calves that were found by ranchers in open pastures. This suggests that there may be more calves killed in high risk environments than are being found and reported by ranchers and that panthers can hunt effectively in open environments. It also suggests that the model may provide a means for evaluating the quality of panther hunting habitat and the corresponding risk of depredation to livestock across the

  12. Coupling calculation of CFD-ACE computational fluid dynamics code and DeCART whole-core neutron transport code for development of numerical reactor

    International Nuclear Information System (INIS)

    Shin, Chang Hwan; Seo, Kyong Won; Chun, Tae Hyun; Kim, Kang Seog

    2005-03-01

    Code coupling activities have so far focused on coupling the neutronics modules with the CFD module. An interface module for the CFD-ACE/DeCART coupling was established as an alternative to the original STAR-CD/DeCART interface. The interface module for DeCART/CFD-ACE was validated by single-pin model. The optimized CFD mesh was decided through the calculation of multi-pin model. It was important to consider turbulent mixing of subchannels for calculation of fuel temperature. For the parallel calculation, the optimized decompose process was necessary to reduce the calculation costs and setting of the iteration and convergence criterion for each code was important, too

  13. Coupling calculation of CFD-ACE computational fluid dynamics code and DeCART whole-core neutron transport code for development of numerical reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Chang Hwan; Seo, Kyong Won; Chun, Tae Hyun; Kim, Kang Seog

    2005-03-15

    Code coupling activities have so far focused on coupling the neutronics modules with the CFD module. An interface module for the CFD-ACE/DeCART coupling was established as an alternative to the original STAR-CD/DeCART interface. The interface module for DeCART/CFD-ACE was validated by single-pin model. The optimized CFD mesh was decided through the calculation of multi-pin model. It was important to consider turbulent mixing of subchannels for calculation of fuel temperature. For the parallel calculation, the optimized decompose process was necessary to reduce the calculation costs and setting of the iteration and convergence criterion for each code was important, too.

  14. WIMS/PANTHER analysis of UO2/MOX cores using embedded super-cells

    International Nuclear Information System (INIS)

    Knight, M.; Bryce, P.; Hall, S.

    2012-01-01

    This paper describes a method of analysing PWR UO 2 MOX cores with WIMS/PANTHER. Embedded super-cells, run within the reactor code, are used to correct the standard methodology of using 2-group smeared data from single assembly lattice calculations. In many other codes the weakness of this standard approach has been improved for MOX by imposing a more realistic environment in the lattice code, or by improving the sophistication of the reactor code. In this approach an intermediate set of calculations is introduced, leaving both lattice and reactor calculations broadly unchanged. The essence of the approach is that the whole core is broken down into a set of 'embedded' super-cells, each extending over just four quarter assemblies, with zero leakage imposed at the assembly mid-lines. Each supercell is solved twice, first with a detailed multi-group pin-by-pin solution, and then with the standard single assembly approach. Correction factors are defined by comparing the two solutions, and these can be applied in whole core calculations. The restriction that all such calculations are modelled with zero leakage means that they are independent of each other and of the core-wide flux shape. This allows parallel pre-calculation for the entire cycle once the loading pattern has been determined, in much the same way that single assembly lattice calculations can be pre-calculated once the range of fuel types is known. Comparisons against a whole core pin-by-pin reference demonstrates that the embedding process does not introduce a significant error, even after burnup and refuelling. Comparisons against a WIMS reference demonstrate that a pin-by-pin multi-group diffusion solution is capable of capturing the main interface effects. This therefore defines a practical approach for achieving results close to lattice code accuracy, but broadly at the cost of a standard reactor calculation. (authors)

  15. WIMS/PANTHER analysis of UO{sub 2}/MOX cores using embedded super-cells

    Energy Technology Data Exchange (ETDEWEB)

    Knight, M.; Bryce, P. [EDF Energy, Barnett Way, Barnwood, Gloucester (United Kingdom); Hall, S. [Advanced Modelling and Computation Group, Imperial College, London (United Kingdom)

    2012-07-01

    This paper describes a method of analysing PWR UO{sub 2}MOX cores with WIMS/PANTHER. Embedded super-cells, run within the reactor code, are used to correct the standard methodology of using 2-group smeared data from single assembly lattice calculations. In many other codes the weakness of this standard approach has been improved for MOX by imposing a more realistic environment in the lattice code, or by improving the sophistication of the reactor code. In this approach an intermediate set of calculations is introduced, leaving both lattice and reactor calculations broadly unchanged. The essence of the approach is that the whole core is broken down into a set of 'embedded' super-cells, each extending over just four quarter assemblies, with zero leakage imposed at the assembly mid-lines. Each supercell is solved twice, first with a detailed multi-group pin-by-pin solution, and then with the standard single assembly approach. Correction factors are defined by comparing the two solutions, and these can be applied in whole core calculations. The restriction that all such calculations are modelled with zero leakage means that they are independent of each other and of the core-wide flux shape. This allows parallel pre-calculation for the entire cycle once the loading pattern has been determined, in much the same way that single assembly lattice calculations can be pre-calculated once the range of fuel types is known. Comparisons against a whole core pin-by-pin reference demonstrates that the embedding process does not introduce a significant error, even after burnup and refuelling. Comparisons against a WIMS reference demonstrate that a pin-by-pin multi-group diffusion solution is capable of capturing the main interface effects. This therefore defines a practical approach for achieving results close to lattice code accuracy, but broadly at the cost of a standard reactor calculation. (authors)

  16. Library of neutron cross sections of the Thermos code; Biblioteca de secciones eficaces de neutrones del codigo Thermos

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-10-15

    The present work is the complement of the IT.SN/DFR-017 report in which the structure and the generation of the library of the Thermos code is described. In this report the comparison among the values of the cross sections that has the current library of the Thermos code and those generated by means of the ENDF-B/NJOY it is shown. (Author)

  17. Neutronic calculations with transport and diffusion computer codes for light water moderated critical with UO2 enriched at 4,75% as fuel

    International Nuclear Information System (INIS)

    Sabundjian, G.; Nakata, H.

    1983-02-01

    The neutronic calculational procedure in a 4,75% w/O enriched UO 2 fueled light water moderated critical assembly was tested, using the transport codes and diffusin code available at the Instituto de Pesquisas Energeticas e Nucleares. The results of the tested codes, LEOPARD, CITHAMMER, LASER, GELS and CITATION, were found to be satisfatory and only a slight advantage is presented by CITHAMMER code. (Author) [pt

  18. First Results for Fluid Dynamics, Neutronics and Fission Product Behaviour in HTR applying the HTR Code Package (HCP) Prototype

    International Nuclear Information System (INIS)

    Allelein, H.-J.; Kasselmann, S.; Xhonneux, A.; Lambertz, D.

    2014-01-01

    To simulate the different aspects of High Temperature Reactor (HTR) cores, a variety of specialized computer codes have been developed at Forschungszentrum Jülich (IEK-6) and Aachen University (LRST) in the last decades. In order to preserve knowledge, to overcome present limitations and to make these codes applicable to modern computer clusters, these individual programs are being integrated into a consistent code package. The so-called HTR code package (HCP) couples the related and recently applied physics models in a highly integrated manner and therefore allows to simulate phenomena with higher precision in space and time while at the same time applying state-of-the-art programming techniques and standards. This paper provides an overview of the status of the HCP and reports about first benchmark results for an HCP prototype which couples the fluid dynamics and time dependent neutronics code MGT-3D, the burn up code TNT and the fission product release code STACY. Due to the coupling of MGT-3D and TNT, a first step towards a new reactor operation and accident simulation code was made, where nuclide concentrations calculated by TNT are fed back into a new spectrum code of the HCP. Selected operation scenarios of the HTR-Module 200 concept plant and the HTTR were chosen to be simulated with the HCP prototype. The fission product release during normal operation conditions will be calculated with STACY based on a core status derived from SERPENT and MGT–3D. Comparisons will be shown against data generated by the legacy codes VSOP99/11, NAKURE and FRESCO-II. (author)

  19. Radiation transport and shielding information, computer codes, and nuclear data for use in CTR neutronics research and development

    International Nuclear Information System (INIS)

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

    1976-01-01

    The activities of the Radiation Shielding Information Center (RSIC) of the Oak Ridge National Laboratory are being utilized in support of fusion reactor technology. The major activities of RSIC include the operation of a computer-based information storage and retrieval system, the collection, packaging, and distribution of large computer codes, and the compilation and dissemination of processed and evaluated data libraries, with particular emphasis on neutron and gamma-ray cross-section data. The Center has acquired thirteen years of experience in serving fission reactor, weapons, and accelerator shielding research communities, and the extension of its technical base to fusion reactor research represents a logical progression. RSIC is currently working with fusion reactor researchers and contractors in computer code development to provide tested radiation transport and shielding codes and data library packages. Of significant interest to the CTR community are the 100 energy group neutron and 21 energy group gamma-ray coupled cross-section data package (DLC-37) for neutronics studies, a comprehensive 171 energy group neutron and 36 energy group gamma-ray coupled cross-section data base with retrieval programs, including resonance self-shielding, that are tailored to CTR application, and a data base for the generation of energy-dependent atomic displacement and gas production cross sections and heavy-particle-recoil spectra for estimating radiation damage to CTR structural components. Since 1964, the Center has been involved in the international exchange of information, encouraged and supported by both government and interagency agreements; and to achieve an equally viable and successful program in fusion research, the reciprocal exchange of CTR data and computing technology is encouraged and welcomed

  20. 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)

  1. 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)

  2. Neutronics/Thermo-fluid Coupled Analysis of PMR-200 Equilibrium Cycle by CAPP/GAMMA+ Code System

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyun Chul; Tak, Nam-il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The equilibrium core was obtained by performing CAPP stand-alone multi-cycle depletion calculation with critical rod position search. In this work, a code system for coupled neutronics and thermo-fluids simulation was developed using CAPP and GAMMA+ codes. A server program, INTCA, controls the two codes for coupled calculations and performs the mapping between the variables of the two codes based on the nodalization of the two codes. In order to extend the knowledge about the coupled behavior of a prismatic VHTR, the CAPP/GAMMA+ code system was applied to steady state performance analysis of PMR-200. The coupled calculation was carried out for the equilibrium core of PMR-200 from BOC to EOC. The peak fuel temperature was predicted to be 1372 .deg. C near MOC. However, the cycle-average fuel temperature was calculated as 1230 .deg. C, which is slightly below the design target of 1250 .deg. C. In addition, significant impact of the bypass flow on the central reflector temperature was found. Without bypass flow, the temperature of the active core region was slightly decreased while the temperature of the central and side reflector region was increased much. The both changes in the temperature increase the multiplication factor and the total change of the multiplication factor was more than 300 pcm. On the other hand, the effect of the bypass flow on the power density profile was not significant.

  3. First results for fluid dynamics, neutronics and fission product behavior in HTR applying the HTR code package (HCP) prototype

    Energy Technology Data Exchange (ETDEWEB)

    Allelein, H.-J., E-mail: h.j.allelein@fz-juelich.de [Forschungszentrum Jülich, 52425 Jülich (Germany); Institute for Reactor Safety and Reactor Technology, RWTH Aachen University, 52064 Aachen (Germany); Kasselmann, S.; Xhonneux, A.; Tantillo, F.; Trabadela, A.; Lambertz, D. [Forschungszentrum Jülich, 52425 Jülich (Germany)

    2016-09-15

    To simulate the different aspects of High Temperature Reactor (HTR) cores, a variety of specialized computer codes have been developed at Forschungszentrum Jülich (IEK-6) and Aachen University (LRST) in the last decades. In order to preserve knowledge, to overcome present limitations and to make these codes applicable to modern computer clusters, these individual programs are being integrated into a consistent code package. The so-called HTR code package (HCP) couples the related and recently applied physics models in a highly integrated manner and therefore allows to simulate phenomena with higher precision in space and time while at the same time applying state-of-the-art programming techniques and standards. This paper provides an overview of the status of the HCP and reports about first benchmark results for an HCP prototype which couples the fluid dynamics and time dependent neutronics code MGT-3D, the burn up code TNT and the fission product release code STACY. Due to the coupling of MGT-3D and TNT, a first step towards a new reactor operation and accident simulation code was made, where nuclide concentrations calculated by TNT lead to new cross sections, which are fed back into MGT-3D. Selected operation scenarios of the HTR-Module 200 concept plant and the HTTR were chosen to be simulated with the HCP prototype. The fission product release during normal operation conditions will be calculated with STACY based on a core status derived from SERPENT and MGT-3D. Comparisons will be shown against data generated by SERPENT and the legacy codes VSOP99/11, NAKURE and FRESCO-II.

  4. The application of the PARCS neutronics code to the Atucha-I and Atucha-II NPPs

    Energy Technology Data Exchange (ETDEWEB)

    Ward, Andrew; Collins, Ben; Xu, Yunlin; Downar, Thomas [Purdue University, West Lafayette, IN (United States); Madariaga, Marcelo [Autoridad Nuclear Regulatoria, Buenos Aires (Argentina)

    2008-07-01

    In order to analyze Central Nuclear Atucha II (CNA-II) with coupled RELAP5/PARCS, extensive benchmarking of the neutronics codes HELIOS and PARCS was completed. This benchmarking was performed using a range of test problems designed in collaboration with NA-SA. HELIOS has been previously used to model Candu systems, but the results were validated for this case as well. The validation of both HELIOS and PARCS was performed primarily by comparisons to MCNP results for the same problems. Though originally designed to model light water systems, the capability of the PARCS was validated for predicting the performance of a Pressurized Heavy Water Reactor. The other noteworthy issue was the control rods. Because the insertion of the rods is oblique, a special routine was added to PARCS to treat this effect. Lattice level and Core level calculations were compared to the corresponding NA-SA codes WIMS and PUMA. In all cases there was good agreement in the results which provided confidence that the neutronics methods and the core neutronics modelling would not be a significant source of error in coupled RELAP5/PARCS calculations. (authors)

  5. DIANE, a simulation code for the interaction of neutrons with living tissues. Application to low doses of fast neutrons on human tumoral cells; DIANE, un code de simulation de l'interaction des neutrons avec la matiere vivante. Applications aux faibles doses de neutrons rapides sur des cellules tumorales humaines

    Energy Technology Data Exchange (ETDEWEB)

    Nenot, M.L

    2003-07-15

    Our work deals with the irradiation of cells and living tissues by 14 MeV neutrons at very low doses (a few 10{sup -2} Gy). Such experiments require an accurate knowledge of the values of neutron dose rates and fluences at the level of cell cultures. We have performed measurements of fluence rates through an activation method applied to gold and copper foils. The fluence rate is deduced from the gamma rays emitted by the irradiated foils. Neutron doses and dose rates have been measured through varied methods: PIN diodes, ionization tissue equivalent chambers, and Geiger-Mueller counters. We have designed the DIANE code to simulate the impact of energetic neutrons on cells. This code can be used with isolated cells or macroscopic tissues, it takes into account the roles of the ionisation electrons produced by recoil nuclei entering the cell. This point is all the more important since recent works have highlighted the impact of very low energy electrons on DNA. (A.C.)

  6. 2007 Bureau of Land Management (BLM) Lidar: Panther Creek Watershed, Yamhill County

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The dataset represents LiDAR elevations acquired during a leaf-off and a leaf-on vegetative condition for the Upper Panther Creek Watershed in the Yamhill County...

  7. Reliable estimation of neutron flux in BWR reactor vessel using the tort code (2) application to neutron and gamma flux estimation

    Energy Technology Data Exchange (ETDEWEB)

    Kurosawa, M. [Toshiba Corp., Yokohama (Japan); Tsukiyama, T.; Hayashi, K. [Hitachi Engineering Co. Ltd., Hitachi-shi (Japan)

    2001-07-01

    A neutron and gamma flux distribution around the core of BWR commercial plant in Japan was calculated, using a three-dimensional transport code, TORT in DOORS32 code system. In the external of the core, the bottom of the model was at an elevation of 150 cm below the bottom of active fuel, the top of the model was at an elevation of the top of the shroud head dome and the radial part of the model was to the outside of the reactor pressure vessel. The top guide beams were modeled explicitly to obtain the neutron and gamma flux distribution both in the beams and outside beams. The each control rod guide tube was also modeled with homogeneous region which included the blade wing and poison tubes so that we could obtain the neutron and gamma flux distribution around the each control rod guide tube. The calculation model mentioned above needed very large memory size which exceeded a few decade giga-bytes. As the using the splicing/coupling method had uncertainly at the splicing/coupling boundary, in this work the calculation was performed without this splicing/coupling method. On the other hand, radioactivity data were measured for a few pieces of the top guide beam, shroud and in-core monitor guide tube in the same plant which was analyzed in the above calculation. So the calculation results were able to be compared with those measured data as benchmarking and at the end of this task, the C/M values at these measured points were obtained and calculation model using TORT was evaluated. (authors)

  8. Calculation of neutron shielding for a real loaded C-30 cask by code DORT

    International Nuclear Information System (INIS)

    Lacina, J.

    1999-01-01

    Measured neutron dose rates of real loaded C-30 casks for WWER spent fuel assemblies are compared with calculated values in the frame of benchmark calculation task. The part of this benchmark task concerning neutron shielding was calculated. Neutron sources values were taken from data presented by V. Chrapciak during the eighth symposium Atomic Energy Research, Bystrice pod Perstejnem in 1998 and the data about cask from the article of the same author from the Atomic Energy Research working group E meeting at Stolpen in 1998. (Author)

  9. Basic design of the HANARO cold neutron source using MCNP code

    International Nuclear Information System (INIS)

    Yu, Yeong Jin; Lee, Kye Hong; Kim, Young Jin; Hwang, Dong Gil

    2005-01-01

    The design of the Cold Neutron Source (CNS) for the HANARO research reactor is on progress. The CNS produces neutrons in the low energy range less than 5meV using liquid hydrogen at around 21.6 K as the moderator. The primary goal for the CNS design is to maximize the cold neutron flux with wavelengths of around 2 ∼ 12 A and to minimize the nuclear heat load. In this paper, the basic design of the HANARO CNS is described

  10. 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

  11. Computer calculation of neutron cross sections with Hauser-Feshbach code STAPRE incorporating the hybrid pre-compound emission model

    International Nuclear Information System (INIS)

    Ivascu, M.

    1983-10-01

    Computer codes incorporating advanced nuclear models (optical, statistical and pre-equilibrium decay nuclear reaction models) were used to calculate neutron cross sections needed for fusion reactor technology. The elastic and inelastic scattering (n,2n), (n,p), (n,n'p), (n,d) and (n,γ) cross sections for stable molybdenum isotopes Mosup(92,94,95,96,97,98,100) and incident neutron energy from about 100 keV or a threshold to 20 MeV were calculated using the consistent set of input parameters. The hydrogen production cross section which determined the radiation damage in structural materials of fusion reactors can be simply deduced from the presented results. The more elaborated microscopic models of nuclear level density are required for high accuracy calculations

  12. MIRANDA - a module based on multiregion resonance theory for generating cross sections within the AUS neutronics code system

    International Nuclear Information System (INIS)

    Robinson, G.S.

    1985-12-01

    MIRANDA is the cross-section generation module of the AUS neutronics code system used to prepare multigroup cross-section data which are pertinent to a particular study from a general purpose multigroup library of cross sections. Libraries have been prepared from ENDF/B which are suitable for thermal and fast fission reactors and for fusion blanket studies. The libraries include temperature dependent data, resonance cross sections represented by subgroup parameters and may contain photon as well as neutron data. The MIRANDA module includes a multiregion resonance calculation in slab, cylinder or cluster geometry, a homogeneous B L flux solution, and a group condensation facility. This report documents the modifications to an earlier version of MIRANDA and provides a complete user's manual

  13. EDITAR: a module for reaction rate editing and cross-section averaging within the AUS neutronics code system

    International Nuclear Information System (INIS)

    Robinson, G.S.

    1986-03-01

    The EDITAR module of the AUS neutronics code system edits one and two-dimensional flux data pools produced by other AUS modules to form reaction rates for materials and their constituent nuclides, and to average cross sections over space and energy. The module includes a Bsub(L) flux calculation for application to cell leakage. The STATUS data pool of the AUS system is used to enable the 'unsmearing' of fluxes and nuclide editing with minimal user input. The module distinguishes between neutron and photon groups, and printed reaction rates are formed accordingly. Bilinear weighting may be used to obtain material reactivity worths and to average cross sections. Bilinear weighting is at present restricted to diffusion theory leakage estimates made using mesh-average fluxes

  14. Development of a three-dimensional neutron transport code DFEM based on the double finite element method

    International Nuclear Information System (INIS)

    Fujimura, Toichiro

    1996-01-01

    A three-dimensional neutron transport code DFEM has been developed by the double finite element method to analyze reactor cores with complex geometry as large fast reactors. Solution algorithm is based on the double finite element method in which the space and angle finite elements are employed. A reactor core system can be divided into some triangular and/or quadrangular prism elements, and the spatial distribution of neutron flux in each element is approximated with linear basis functions. As for the angular variables, various basis functions are applied, and their characteristics were clarified by comparison. In order to enhance the accuracy, a general method is derived to remedy the truncation errors at reflective boundaries, which are inherent in the conventional FEM. An adaptive acceleration method and the source extrapolation method were applied to accelerate the convergence of the iterations. The code structure is outlined and explanations are given on how to prepare input data. A sample input list is shown for reference. The eigenvalue and flux distribution for real scale fast reactors and the NEA benchmark problems were presented and discussed in comparison with the results of other transport codes. (author)

  15. Generation of the library of neutron cross sections for the Record code of the Fuel Management System (FMS); Generacion de la biblioteca de secciones eficaces de neutrones para el codigo Record del Sistema de Administracion de Combustible (FMS)

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-11-15

    On the basis of the library structure of the RECORD code a method to generate the neutron cross sections by means of the ENDF-B/IV database and the NJOY code has been developed. The obtained cross sections are compared with those of the current library which was processed using the ENDF-B/III version. (Author)

  16. Computational features of the MELT-III neutronics, thermal-hydraulics computer code system

    International Nuclear Information System (INIS)

    Wilburn, N.P.; Waltar, A.E.

    1976-01-01

    A multichannel, thermal-hydraulics, neutronic accident analysis program for simulating fast reactor behavior from a hypothetical accident inception to the start of core disassembly or to reactor shutdown is described

  17. Calculation of neutron streaming through complicated pathes with the MORSE code

    International Nuclear Information System (INIS)

    Fang Bangcheng.

    1987-01-01

    A procedure to solve a concrete engineering problem of neutron leakage with MORSE program is presented. The methodology used in treating some difficult problems and the biasing techniques are useful to the beginners of the MORSE program

  18. Studies on the liquid fluoride thorium reactor: Comparative neutronics analysis of MCNP6 code with SRAC95 reactor analysis code based on FUJI-U3-(0)

    Energy Technology Data Exchange (ETDEWEB)

    Jaradat, S.Q., E-mail: sqjxv3@mst.edu; Alajo, A.B., E-mail: alajoa@mst.edu

    2017-04-01

    Highlights: • The verification for FUJI-U3-(0)—a molten salt reactor—was performed. • The MCNP6 was used to study the reactor physics characteristics for FUJI-U3 type. • The results from the MCNP6 were comparable with the ones obtained from literature. - Abstract: The verification for FUJI-U3-(0)—a molten salt reactor—was performed. The reactor used LiF-BeF2-ThF4-UF4 as the mixed liquid fuel salt, and the core was graphite moderated. The MCNP6 code was used to study the reactor physics characteristics for the FUJI-U3-(0) reactor. Results for reactor physics characteristic of the FUJI-U3-(0) exist in literature, which were used as reference. The reference results were obtained using SRAC95 (a reactor analysis code) coupled with ORIGEN2 (a depletion code). Some modifications were made in the reconstruction of the FUJI-U3-(0) reactor in MCNP due to unavailability of more detailed description of the reactor core. The assumptions resulted in two representative models of the reactor. The results from the MCNP6 models were compared with the reference results obtained from literature. The results were comparable with each other, but with some notable differences. The differences are because of the approximations that were done on the SRAC95 model of the FUJI-U3 to simplify the simulation. Based on the results, it is concluded that MCNP6 code predicts well the overall simulation of neutronics analysis to the previous simulation works using SRAC95 code.

  19. Large-scale gene function analysis with the PANTHER classification system.

    Science.gov (United States)

    Mi, Huaiyu; Muruganujan, Anushya; Casagrande, John T; Thomas, Paul D

    2013-08-01

    The PANTHER (protein annotation through evolutionary relationship) classification system (http://www.pantherdb.org/) is a comprehensive system that combines gene function, ontology, pathways and statistical analysis tools that enable biologists to analyze large-scale, genome-wide data from sequencing, proteomics or gene expression experiments. The system is built with 82 complete genomes organized into gene families and subfamilies, and their evolutionary relationships are captured in phylogenetic trees, multiple sequence alignments and statistical models (hidden Markov models or HMMs). Genes are classified according to their function in several different ways: families and subfamilies are annotated with ontology terms (Gene Ontology (GO) and PANTHER protein class), and sequences are assigned to PANTHER pathways. The PANTHER website includes a suite of tools that enable users to browse and query gene functions, and to analyze large-scale experimental data with a number of statistical tests. It is widely used by bench scientists, bioinformaticians, computer scientists and systems biologists. In the 2013 release of PANTHER (v.8.0), in addition to an update of the data content, we redesigned the website interface to improve both user experience and the system's analytical capability. This protocol provides a detailed description of how to analyze genome-wide experimental data with the PANTHER classification system.

  20. The study on neutron and photon distribution of AP1000 reactor by MCNP code

    International Nuclear Information System (INIS)

    Chen Defeng; Shen Mingqi

    2014-01-01

    The core and reactor structural of AP1000 was modeled by the MCNP calculation program which is based on the Monte Carlo method in this paper, the neutron and photon distribution of AP1000 reactor core was calculated by the conditions of reactor critical. The results show that the AP1000 reactor neutron and photon distribution is in accordance with the critical design of PWR. (authors)

  1. HADES. A computer code for fast neutron cross section from the Optical Model; HADES. Un programa numerico para el calculo de seccciones eficaces neutronicas mediante el modelo optico

    Energy Technology Data Exchange (ETDEWEB)

    Guasp, J; Navarro, C

    1973-07-01

    A FORTRAN V computer code for UNIVAC 1108/6 using a local Optical Model with spin-orbit interaction is described. The code calculates fast neutron cross sections, angular distribution, and Legendre moments for heavy and intermediate spherical nuclei. It allows for the possibility of automatic variation of potential parameters for experimental data fitting. (Author) 55 refs.

  2. Comparative Neutronics Analysis of DIMPLE S06 Criticality Benchmark with Contemporary Reactor Core Analysis Computer Code Systems

    Directory of Open Access Journals (Sweden)

    Wonkyeong Kim

    2015-01-01

    Full Text Available A high-leakage core has been known to be a challenging problem not only for a two-step homogenization approach but also for a direct heterogeneous approach. In this paper the DIMPLE S06 core, which is a small high-leakage core, has been analyzed by a direct heterogeneous modeling approach and by a two-step homogenization modeling approach, using contemporary code systems developed for reactor core analysis. The focus of this work is a comprehensive comparative analysis of the conventional approaches and codes with a small core design, DIMPLE S06 critical experiment. The calculation procedure for the two approaches is explicitly presented in this paper. Comprehensive comparative analysis is performed by neutronics parameters: multiplication factor and assembly power distribution. Comparison of two-group homogenized cross sections from each lattice physics codes shows that the generated transport cross section has significant difference according to the transport approximation to treat anisotropic scattering effect. The necessity of the ADF to correct the discontinuity at the assembly interfaces is clearly presented by the flux distributions and the result of two-step approach. Finally, the two approaches show consistent results for all codes, while the comparison with the reference generated by MCNP shows significant error except for another Monte Carlo code, SERPENT2.

  3. Development of a tri dimensional Monte Carlo code for the study of the microdosimetry in boron neutron capture therapy

    International Nuclear Information System (INIS)

    Santa Cruz, G.A.

    1998-01-01

    Full text: A charged particles transport Monte Carlo code, specially designed for the boron neutron capture therapy microdosimetry study was developed. The code allows the use of real tri dimensional problem geometry, using serial microscopy slides from a biological substrate where the 10 B(n, Alpha) 7 Li, 14 N(n,p) 14 C reactions and events can occur. The spatial distribution of sources ( 10 B, 14 N concentrations), regions of interest (where the energy deposition, linear energy transfer and other parameters will be calculated) and other zones (without boron) are obtained from the images. The code is in the benchmarking stage, using geometrically simple cases and experimental data obtained from microdosimetric spectra from TEPC (Tissue Equivalent Proportional Counters) doped with 10 B. It allows to obtain LET spectra discriminated by event classes, chord-length distributions, dose and frequency mean values and visualizations of the spatial energy deposition. A similar version of the code uses bidimensional images from a tissue sample containing a great number of cellular structures. An equivalence between the microdosimetry of a bidimensional case and a tri dimensional one can be done. If the real distribution of 10 B is known, for example by high resolution alpha-track autoradiography, the code can use this information explicitly. (author) [es

  4. Development of NRESP98 Monte Carlo codes for the calculation of neutron response functions of neutron detectors. Calculation of the response function of spherical BF{sub 3} proportional counter

    Energy Technology Data Exchange (ETDEWEB)

    Hashimoto, M.; Saito, K.; Ando, H. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-05-01

    The method to calculate the response function of spherical BF{sub 3} proportional counter, which is commonly used as neutron dose rate meter and neutron spectrometer with multi moderator system, is developed. As the calculation code for evaluating the response function, the existing code series NRESP, the Monte Carlo code for the calculation of response function of neutron detectors, is selected. However, the application scope of the existing NRESP is restricted, the NRESP98 is tuned as generally applicable code, with expansion of the geometrical condition, the applicable element, etc. The NRESP98 is tested with the response function of the spherical BF{sub 3} proportional counter. Including the effect of the distribution of amplification factor, the detailed evaluation of the charged particle transportation and the effect of the statistical distribution, the result of NRESP98 calculation fit the experience within {+-}10%. (author)

  5. MVP/GMVP II, MC Codes for Neutron and Photon Transport Calc. based on Continuous Energy and Multigroup Methods

    International Nuclear Information System (INIS)

    2005-01-01

    A - Description of program or function: (1) Problems to be solved: MVP/GMVP can solve eigenvalue and fixed-source problems. The multigroup code GMVP can solve forward and adjoint problems for neutron, photon and neutron-photon coupled transport. The continuous-energy code MVP can solve only the forward problems. Both codes can also perform time-dependent calculations. (2) Geometry description: MVP/GMVP employs combinatorial geometry to describe the calculation geometry. It describes spatial regions by the combination of the 3-dimensional objects (BODIes). Currently, the following objects (BODIes) can be used. - BODIes with linear surfaces: half space, parallelepiped, right parallelepiped, wedge, right hexagonal prism; - BODIes with quadratic surface and linear surfaces: cylinder, sphere, truncated right cone, truncated elliptic cone, ellipsoid by rotation, general ellipsoid; - Arbitrary quadratic surface and torus. The rectangular and hexagonal lattice geometry can be used to describe the repeated geometry. Furthermore, the statistical geometry model is available to treat coated fuel particles or pebbles for high temperature reactors. (3) Particle sources: The various forms of energy-, angle-, space- and time-dependent distribution functions can be specified. (4) Cross sections: The ANISN-type PL cross sections or the double-differential cross sections can be used in the multigroup code GMVP. On the other hand, the specific cross section libraries are used in the continuous-energy code MVP. The libraries are generated from the evaluated nuclear data (JENDL-3.3, ENDF/B-VI, JEF-3.0 etc.) by using the LICEM code. The neutron cross sections in the unresolved resonance region are described by the probability table method. The neutron cross sections at arbitrary temperatures are available for MVP by just specifying the temperatures in the input data. (5) Boundary conditions: Vacuum, perfect reflective, isotropic reflective (white), periodic boundary conditions can be

  6. Simulation of thermal-neutron-induced single-event upset using particle and heavy-ion transport code system

    International Nuclear Information System (INIS)

    Arita, Yutaka; Kihara, Yuji; Mitsuhasi, Junichi; Niita, Koji; Takai, Mikio; Ogawa, Izumi; Kishimoto, Tadafumi; Yoshihara, Tsutomu

    2007-01-01

    The simulation of a thermal-neutron-induced single-event upset (SEU) was performed on a 0.4-μm-design-rule 4 Mbit static random access memory (SRAM) using particle and heavy-ion transport code system (PHITS): The SEU rates obtained by the simulation were in very good agreement with the result of experiments. PHITS is a useful tool for simulating SEUs in semiconductor devices. To further improve the accuracy of the simulation, additional methods for tallying the energy deposition are required for PHITS. (author)

  7. The coupled code system DORT-TD/THERMIX and its application to the OECD/NEA/NSC PBMR400 MW coupled neutronics thermal hydraulics transient benchmark

    International Nuclear Information System (INIS)

    Pautz, A.; Tyobeka, B.; Ivanov, K.

    2009-01-01

    In new reactor designs that are still under review such as the Pebble Bed Modular Reactor (PBMR), not much experimental data exists to benchmark newly developed computer codes against. Such a situation requires that nuclear engineers and designers of this novel reactor design must resort to the validation of a newly developed code through a code-to-code benchmarking exercise because there are validated codes that are currently in use to analyze this reactor design, albeit very few of them. There are numerous HTR core physics benchmarks that are currently being pursued by different organizations, for different purposes. One such benchmark exercise is the PBMR-400MW OECD/NEA coupled neutronics/thermal hydraulics transient benchmark. In this paper, a newly developed coupled neutronics thermal hydraulics code system, DORT-TD/THERMIX with both transport and diffusion theory options, is used to simulate both the steady-state as well as several transient scenarios in this benchmark problem. (orig.)

  8. Approach to the calculation of energy deposition in a container of fuel irradiated by the neutronic codes coupling fluid-dynamics

    International Nuclear Information System (INIS)

    Hueso, C.; Aleman, A.; Colomer, C.; Fabbri, M.; Martin, M.; Saellas, J.

    2013-01-01

    In this work identifies a possible area of improvement through the creation of a code of coupling between deposition energy codes which calculate neutron (MCNP), and data from heading into fluid dynamics (ANSYS-Fluent) or codes thermomechanical, called MAFACS (Monte Carlo ANSYS Fluent Automatic Coupling Software), being possible to so summarize the process by shortening the needs of computing time, increasing the precision of the results and therefore improving the design of the components.

  9. Migros-3: a code for the generation of group constants for reactor calculations from neutron nuclear data in KEDAK format

    International Nuclear Information System (INIS)

    Broeders, I.; Krieg, B.

    1977-01-01

    The code MIGROS-3 was developed from MIGROS-2. The main advantage of MIGROS-3 is its compatibility with the new conventions of the latest version of the Karlsruhe nuclear data library, KEDAK-3. Moreover, to some extent refined physical models were used and numerical methods were improved. MIGROS-3 allows the calculation of microscopic group cross sections of the ABBN type from isotopic neutron data given in KEDAK-format. All group constants, necessary for diffusion-, consistent P 1 - and Ssub(N)-calculations can be generated. Anisotropy of elastic scattering can be taken into account up to P 5 . A description of the code and the underlying theory is given. The input and output description, a sample problem and the program lists are provided. (orig.) [de

  10. Parallelization of MCNP 4, a Monte Carlo neutron and photon transport code system, in highly parallel distributed memory type computer

    International Nuclear Information System (INIS)

    Masukawa, Fumihiro; Takano, Makoto; Naito, Yoshitaka; Yamazaki, Takao; Fujisaki, Masahide; Suzuki, Koichiro; Okuda, Motoi.

    1993-11-01

    In order to improve the accuracy and calculating speed of shielding analyses, MCNP 4, a Monte Carlo neutron and photon transport code system, has been parallelized and measured of its efficiency in the highly parallel distributed memory type computer, AP1000. The code has been analyzed statically and dynamically, then the suitable algorithm for parallelization has been determined for the shielding analysis functions of MCNP 4. This includes a strategy where a new history is assigned to the idling processor element dynamically during the execution. Furthermore, to avoid the congestion of communicative processing, the batch concept, processing multi-histories by a unit, has been introduced. By analyzing a sample cask problem with 2,000,000 histories by the AP1000 with 512 processor elements, the 82 % of parallelization efficiency is achieved, and the calculational speed has been estimated to be around 50 times as fast as that of FACOM M-780. (author)

  11. A cat's tale: the impact of genetic restoration on Florida panther population dynamics and persistence.

    Science.gov (United States)

    Hostetler, Jeffrey A; Onorato, David P; Jansen, Deborah; Oli, Madan K

    2013-05-01

    1. Genetic restoration has been suggested as a management tool for mitigating detrimental effects of inbreeding depression in small, inbred populations, but the demographic mechanisms underlying population-level responses to genetic restoration remain poorly understood. 2. We studied the dynamics and persistence of the endangered Florida panther Puma concolor coryi population and evaluated the potential influence of genetic restoration on population growth and persistence parameters. As part of the genetic restoration programme, eight female Texas pumas P. c. stanleyana were released into Florida panther habitat in southern Florida in 1995. 3. The overall asymptotic population growth rate (λ) was 1.04 (5th and 95th percentiles: 0.95-1.14), suggesting an increase in the panther population of approximately 4% per year. Considering the effects of environmental and demographic stochasticities and density-dependence, the probability that the population will fall below 10 panthers within 100 years was 0.072 (0-0.606). 4. Our results suggest that the population would have declined at 5% per year (λ = 0.95; 0.83-1.08) in the absence of genetic restoration. Retrospective life table response experiment analysis revealed that the positive effect of genetic restoration on survival of kittens was primarily responsible for the substantial growth of the panther population that would otherwise have been declining. 5. For comparative purposes, we also estimated probability of quasi-extinction under two scenarios - implementation of genetic restoration and no genetic restoration initiative - using the estimated abundance of panthers in 1995, the year genetic restoration was initiated. Assuming no density-dependence, the probability that the panther population would fall below 10 panthers by 2010 was 0.098 (0.002-0.332) for the restoration scenario and 0.445 (0.032-0.944) for the no restoration scenario, providing further evidence that the panther population would have faced a

  12. RADHEAT-V3, a code system for generating coupled neutron and gamma-ray group constants and analyzing radiation transport

    International Nuclear Information System (INIS)

    Koyama, Kinji; Taji, Yukichi; Miyasaka, Shun-ichi; Minami, Kazuyoshi.

    1977-07-01

    The modular code system RADHEAT is for producing coupled multigroup neutron and gamma-ray cross section sets, analyzing the neutron and gamma-ray transport, and calculating the energy deposition and atomic displacements due to these radiations in a nuclear reactor or shield. The basic neutron cross sections and secondary gamma-ray production data are taken from ENDF/B and POPOP4 libraries respectively. The system (1) generates multigroup neutron cross sections, energy deposition coefficients and atomic displacement factors due to neutron reactions, (2) generates multigroup gamma-ray cross sections and energy transfer coefficients, (3) generates secondary gamma-ray production cross sections, (4) combines these cross sections into the coupled set, (5) outputs and updates the multigroup cross section libraries in convenient formats for other transport codes, (6) analyzes the neutron and gamma-ray transport and calculates the energy deposition and the number density of atomic displacements in a medium, (7) collapses the cross sections to a broad-group structure, by option, using the weighting functions obtained by one-dimensional transport calculation, and (8) plots, by option, multigroup cross sections, and neutron and gamma-ray distributions. Definitions of the input data required in various options of the code system are also given. (auth.)

  13. Comparative study of Monte Carlo particle transport code PHITS and nuclear data processing code NJOY for recoil cross section spectra under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Iwamoto, Yosuke, E-mail: iwamoto.yosuke@jaea.go.jp; Ogawa, Tatsuhiko

    2017-04-01

    Because primary knock-on atoms (PKAs) create point defects and clusters in materials that are irradiated with neutrons, it is important to validate the calculations of recoil cross section spectra that are used to estimate radiation damage in materials. Here, the recoil cross section spectra of fission- and fusion-relevant materials were calculated using the Event Generator Mode (EGM) of the Particle and Heavy Ion Transport code System (PHITS) and also using the data processing code NJOY2012 with the nuclear data libraries TENDL2015, ENDF/BVII.1, and JEFF3.2. The heating number, which is the integral of the recoil cross section spectra, was also calculated using PHITS-EGM and compared with data extracted from the ACE files of TENDL2015, ENDF/BVII.1, and JENDL4.0. In general, only a small difference was found between the PKA spectra of PHITS + TENDL2015 and NJOY + TENDL2015. From analyzing the recoil cross section spectra extracted from the nuclear data libraries using NJOY2012, we found that the recoil cross section spectra were incorrect for {sup 72}Ge, {sup 75}As, {sup 89}Y, and {sup 109}Ag in the ENDF/B-VII.1 library, and for {sup 90}Zr and {sup 55}Mn in the JEFF3.2 library. From analyzing the heating number, we found that the data extracted from the ACE file of TENDL2015 for all nuclides were problematic in the neutron capture region because of incorrect data regarding the emitted gamma energy. However, PHITS + TENDL2015 can calculate PKA spectra and heating numbers correctly.

  14. Analysis of reaction cross-section production in neutron induced fission reactions on uranium isotope using computer code COMPLET.

    Science.gov (United States)

    Asres, Yihunie Hibstie; Mathuthu, Manny; Birhane, Marelgn Derso

    2018-04-22

    This study provides current evidence about cross-section production processes in the theoretical and experimental results of neutron induced reaction of uranium isotope on projectile energy range of 1-100 MeV in order to improve the reliability of nuclear stimulation. In such fission reactions of 235 U within nuclear reactors, much amount of energy would be released as a product that able to satisfy the needs of energy to the world wide without polluting processes as compared to other sources. The main objective of this work is to transform a related knowledge in the neutron-induced fission reactions on 235 U through describing, analyzing and interpreting the theoretical results of the cross sections obtained from computer code COMPLET by comparing with the experimental data obtained from EXFOR. The cross section value of 235 U(n,2n) 234 U, 235 U(n,3n) 233 U, 235 U(n,γ) 236 U, 235 U(n,f) are obtained using computer code COMPLET and the corresponding experimental values were browsed by EXFOR, IAEA. The theoretical results are compared with the experimental data taken from EXFOR Data Bank. Computer code COMPLET has been used for the analysis with the same set of input parameters and the graphs were plotted by the help of spreadsheet & Origin-8 software. The quantification of uncertainties stemming from both experimental data and computer code calculation plays a significant role in the final evaluated results. The calculated results for total cross sections were compared with the experimental data taken from EXFOR in the literature, and good agreement was found between the experimental and theoretical data. This comparison of the calculated data was analyzed and interpreted with tabulation and graphical descriptions, and the results were briefly discussed within the text of this research work. Copyright © 2018 The Authors. Published by Elsevier Ltd.. All rights reserved.

  15. Study of cold neutron sources: Implementation and validation of a complete computation scheme for research reactor using Monte Carlo codes TRIPOLI-4.4 and McStas

    International Nuclear Information System (INIS)

    Campioni, Guillaume; Mounier, Claude

    2006-01-01

    The main goal of the thesis about studies of cold neutrons sources (CNS) in research reactors was to create a complete set of tools to design efficiently CNS. The work raises the problem to run accurate simulations of experimental devices inside reactor reflector valid for parametric studies. On one hand, deterministic codes have reasonable computation times but introduce problems for geometrical description. On the other hand, Monte Carlo codes give the possibility to compute on precise geometry, but need computation times so important that parametric studies are impossible. To decrease this computation time, several developments were made in the Monte Carlo code TRIPOLI-4.4. An uncoupling technique is used to isolate a study zone in the complete reactor geometry. By recording boundary conditions (incoming flux), further simulations can be launched for parametric studies with a computation time reduced by a factor 60 (case of the cold neutron source of the Orphee reactor). The short response time allows to lead parametric studies using Monte Carlo code. Moreover, using biasing methods, the flux can be recorded on the surface of neutrons guides entries (low solid angle) with a further gain of running time. Finally, the implementation of a coupling module between TRIPOLI- 4.4 and the Monte Carlo code McStas for research in condensed matter field gives the possibility to obtain fluxes after transmission through neutrons guides, thus to have the neutron flux received by samples studied by scientists of condensed matter. This set of developments, involving TRIPOLI-4.4 and McStas, represent a complete computation scheme for research reactors: from nuclear core, where neutrons are created, to the exit of neutrons guides, on samples of matter. This complete calculation scheme is tested against ILL4 measurements of flux in cold neutron guides. (authors)

  16. Analysis of neutronics benchmarks for the utilization of mixed oxide fuel in light water reactor using DRAGON code

    International Nuclear Information System (INIS)

    Nithyadevi, Rajan; Thilagam, L.; Karthikeyan, R.; Pal, Usha

    2016-01-01

    Highlights: • Use of advanced computational code – DRAGON-5 using advanced self shielding model USS. • Testing the capability of DRAGON-5 code for the analysis of light water reactor system. • Wide variety of fuels LEU, MOX and spent fuel have been analyzed. • Parameters such as k ∞ , one, few and multi-group macroscopic cross-sections and fluxes were calculated. • Suitability of deterministic methodology employed in DRAGON-5 code is demonstrated for LWR. - Abstract: Advances in reactor physics have led to the development of new computational technologies and upgraded cross-section libraries so as to produce an accurate approximation to the true solution for the problem. Thus it is necessary to revisit the benchmark problems with the advanced computational code system and upgraded cross-section libraries to see how far they are in agreement with the earlier reported values. Present study is one such analysis with the DRAGON code employing advanced self shielding models like USS and 172 energy group ‘JEFF3.1’ cross-section library in DRAGLIB format. Although DRAGON code has already demonstrated its capability for heavy water moderator systems, it is now tested for light water reactor (LWR) and fast reactor systems. As a part of validation of DRAGON for LWR, a VVER computational benchmark titled “Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel-Volume 3” submitted by the Russian Federation has been taken up. Presently, pincell and assembly calculations are carried out considering variation in fuel temperature (both fresh and spent), moderator temperatures and boron content in the moderator. Various parameters such as infinite neutron multiplication (k ∞ ) factor, one group integrated flux, few group homogenized cross-sections (absorption, nu-fission) and reaction rates (absorption, nu-fission) of individual isotopic nuclides are calculated for different reactor states. Comparisons of results are made with the reported Monte Carlo

  17. LASER-R a computer code for reactor cell and burnup calculations in neutron transport theory

    International Nuclear Information System (INIS)

    Cristian, I.; Cirstoiu, B.; Dumitrache, I.; Cepraga, D.

    1976-04-01

    The LASER-R code is an IBM 370/135 version of the Westinghouse code, LASER, based on the THERMOS and MUFT codes developped by Poncelet. It can be used to perform thermal reactor cell calculations and burnup calculations. The cell exhibits 3-4 concentric areas: fuel, cladding, moderator and scattering ring. Besides directions for use, a short description of the physical model, numerical methods and output is presented

  18. 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

  19. Carmen system: a code block for neutronic PWR calculation by diffusion theory with spacedependent feedback effects

    International Nuclear Information System (INIS)

    Ahnert, C.; Aragones, J.M.

    1982-01-01

    The Carmen code (theory and user's manual) is described. This code for assembly and core calculations uses diffusion theory (Citation), with feedback in the cross sections by zone due to the effects of burnup, water density, fuel temperature, Xenon and Samarium. The burnup calculation of a full cycle is solved in only an execution of Carmen, and in a reduced computer time. (auth.)

  20. Analysis of mixed oxide fuel critical experiments with neutronics analysis codes for boiling water reactors

    International Nuclear Information System (INIS)

    Tamitani, Masashi; Maruyama, Hiromi; Ishii, Kazuya; Izutsu, Sadayuki; Yamaguchi, Masao

    2000-01-01

    Critical experiments of UO 2 and full mixed oxide (MOX) fuel cores conducted at the Tank-type Critical Assembly (TCA) were analyzed using BWR design-purpose codes HINES and CERES with ENDF/B files and Monte Carlo fine analysis codes VMONT and MVP with the JENDL-3.2 library. The averaged values of the multiplication factors calculated with HINES/CERES, VMONT and MVP agreed with those of experiments within 0.3%Δk. The values by the design-purpose codes showed a small difference of 0.1%Δk between UO 2 and MOX cores. Monte Carlo code results showed that the JENDL-3.2 library had a tendency to overestimate the multiplication factors of UO 2 cores by about 0.3%Δk compared with those values of MOX cores. The root mean square errors of calculated power distributions were less than 1% for HINES/CERES and VMONT. These results showed that (1) the accuracy of these codes when applied to full MOX cores was almost the same as their accuracy for UO 2 cores, which confirmed the accuracy of present core design codes for full MOX cores; and (2) the accuracy of the 190-energy-group Monte Carlo calculation code VMONT was almost the same as that of the continuous-energy Monte Carlo calculation code MVP. (author)

  1. Self characterization of a coded aperture array for neutron source imaging

    Energy Technology Data Exchange (ETDEWEB)

    Volegov, P. L., E-mail: volegov@lanl.gov; Danly, C. R.; Guler, N.; Merrill, F. E.; Wilde, C. H. [Los Alamos National Laboratory, Los Alamos, New Mexico 87544 (United States); Fittinghoff, D. N. [Livermore National Laboratory, Livermore, California 94550 (United States)

    2014-12-15

    The neutron imaging system at the National Ignition Facility (NIF) is an important diagnostic tool for measuring the two-dimensional size and shape of the neutrons produced in the burning deuterium-tritium plasma during the stagnation stage of inertial confinement fusion implosions. Since the neutron source is small (∼100 μm) and neutrons are deeply penetrating (>3 cm) in all materials, the apertures used to achieve the desired 10-μm resolution are 20-cm long, triangular tapers machined in gold foils. These gold foils are stacked to form an array of 20 apertures for pinhole imaging and three apertures for penumbral imaging. These apertures must be precisely aligned to accurately place the field of view of each aperture at the design location, or the location of the field of view for each aperture must be measured. In this paper we present a new technique that has been developed for the measurement and characterization of the precise location of each aperture in the array. We present the detailed algorithms used for this characterization and the results of reconstructed sources from inertial confinement fusion implosion experiments at NIF.

  2. Investigation of the response of a neutron moisture meter using a multigroup, two-dimensional diffusion theory code

    International Nuclear Information System (INIS)

    Ritchie, A.I.M.; Wilson, D.J.

    1984-12-01

    A multigroup diffusion code has been used to predict the count rate from a neutron moisture meter for a range of values of soil water content ω, thermal neutron absorption cross section Ssub(a) (defined as Σsub(a)/rho) of the soil matrix and soil matrix density rho. Two dimensions adequately approximated the geometry of the source, detector and soil surrounding the detector. Seven energy groups, the data for which were condensed from 128 group data set over the neutron energy spectrum appropriate to the soil-water mixture under study, proved adequate to describe neutron slowing-down and diffusion. The soil-water mixture was an SiO 2 →water mixture, with the absorption cross section of SiO 2 increased to cover the range of Σsub(a) required. The response to changes in matrix density is, in general, linear but the response to changes in water content is not linear over the range of parameter values investigated. Tabular results are presented which allow interpolation of the response for a particular ω, Ssub(a) and rho. It is shown that R(ω, Ssub(a), rho) rho M(Ssub(a)) + C(ω) is a crude representation of the response over a very limited range of variation of ω, and Ssub(a). As the response is a slowly varying function of rho, Ssub(a) and ω, a polynomial fit will provide a better estimate of the response for values of rho, Ssub(a) and ω not tabulated

  3. Application of modified REFIT code for J-PARC/MLF to evaluation of neutron capture cross section on 155,157Gd

    Science.gov (United States)

    Mizuyama, Kazuhito; Iwamoto, Nobuyuki; Iwamoto, Osamu; Hasemi, Hiroyuki; Kino, Koichi; Kimura, Atsushi; Kiyanagi, Yoshiaki

    2017-09-01

    In order to analyze the experimental data measured by the Accurate Neutron-Nucleus Reaction Measurement Instrument (ANNRI) at the Japan Proton Accelerator Research Complex/Materials and Life Science Experimental Facility (J-PARC/MLF), it is necessary to take into account the double-bunch structure of the neutron pulse and the energy resolution function for the operational condition of the J-PARC/MLF. The modified REFIT code has been developed to treat the double-bunch neutron pulse and the energy resolution function for J-PARC/MLF. In this study, we applied the modified REFIT code to analyze the new data of the neutron capture cross section of 155Gd and 157Gd recently measured by ANNRI in the J-PARC/MLF, and obtained the resonance parameters of two Gd isotopes. We discussed the differences between the our obtained results and the other libraries.

  4. The study of time-dependent neutronics parameters of the 2MW TRIGA Mark II Moroccan research reactor using BUCAL1 computer code

    International Nuclear Information System (INIS)

    Bakkari, B. El; Nacir, B.; El Younoussi, C.; Boulaich, Y.; Riyach, I.; Otmani, S.; Marcih, I.; Elbadri, H.; El Bardouni, T; Merroun, O.; Boukhal, H.; Zoubair, M.; Htet, A.; Chakir, M.

    2010-01-01

    The 2-MW TRIGA MARK II research reactor at Centre National de l'Energie, des Sciences et des Techniques Nucleaires (CNESTEN) achieved initial criticality on May 2, 2007 with 71 fuel elements. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower and training and production of radioisotopes for their use in agriculture, industry and medicine. This work aims to study the time-dependent neutronics parameters of the TRIGA reactor for elaborating and planning of an in-core fuel management strategy to maximize the utilization of the TRIGA fluxes, using a new elaborated burnup computer code called 'BUCAL1'. The code can be used to aid in analysis, prediction, and optimization of fuel burnup performance in a nuclear reactor. It was developed to incorporate the neutron absorption tally/reaction information generated directly by MCNP5 code in the calculation of fissioned or neutron-transmuted isotopes for multi-fueled regions. The use of Monte Carlo method and punctual cross section data characterizing the MCNP code allows an accurate simulation of neutron life cycle in the reactor, and the integration of data on the entire energy spectrum, thus a more accurate estimation of results than deterministic code can do. Also, for the purpose of this study, a full-model of the TRIGA reactor was developed using the MCNP5 code. The validation of the MCNP model of the TRIGA reactor was made by benchmarking the reactivity experiments. (author)

  5. Analysis of the OECD/NRC BWR Turbine Trip Transient Benchmark with the Coupled Thermal-Hydraulics and Neutronics Code TRAC-M/PARCS

    International Nuclear Information System (INIS)

    Lee, Deokjung; Downar, Thomas J.; Ulses, Anthony; Akdeniz, Bedirhan; Ivanov, Kostadin N.

    2004-01-01

    An analysis of the Peach Bottom Unit 2 Turbine Trip 2 (TT2) experiment has been performed using the U.S. Nuclear Regulatory Commission coupled thermal-hydraulics and neutronics code TRAC-M/PARCS. The objective of the analysis was to assess the performance of TRAC-M/PARCS on a BWR transient with significance in two-phase flow and spatial variations of the neutron flux. TRAC-M/PARCS results are found to be in good agreement with measured plant data for both steady-state and transient phases of the benchmark. Additional analyses of four fictitious extreme scenarios are performed to provide a basis for code-to-code comparisons and comprehensive testing of the thermal-hydraulics/neutronics coupling. The obtained results of sensitivity studies on the effect of direct moderator heating on transient simulation indicate the importance of this modeling aspect

  6. MCNP: a general Monte Carlo code for neutron and photon transport. Version 3A. Revision 2

    International Nuclear Information System (INIS)

    Briesmeister, J.F.

    1986-09-01

    This manual is a practical guide for the use of our general-purpose Monte Carlo code MCNP. The first chapter is a primer for the novice user. The second chapter describes the mathematics, data, physics, and Monte Carlo simulation found in MCNP. This discussion is not meant to be exhaustive - details of the particular techniques and of the Monte Carlo method itself will have to be found elsewhere. The third chapter shows the user how to prepare input for the code. The fourth chapter contains several examples, and the fifth chapter explains the output. The appendices show how to use MCNP on particular computer systems at the Los Alamos National Laboratory and also give details about some of the code internals that those who wish to modify the code may find useful. 57 refs

  7. Modelling of the non-stationary thermal neutron transport in hydrogenous media using the MCNP code

    International Nuclear Information System (INIS)

    Wiacek, Urszula

    2006-01-01

    The work is aimed to compare results of the Monte Carlo simulations of pulsed neutron experiments with results of real experiments. The simulations have been performed for homogenous and two-zone system. In the first case the cylinder of Plexiglas has been used. In two-zone systems, Plexiglas has been used as an outer moderator and aqueous solutions of H 3 BO 3 or KCl of known concentrations have been used as the inner samples. The simulations have been performed (1) using for neutron scattering in Plexiglas a standard library for H in polyethylene (a commonly suggested way) and (2) using a modification of this library. The modification gives very good agreement between simulations and experimental results

  8. A user-friendly, graphical interface for the Monte Carlo neutron optics code MCLIB

    International Nuclear Information System (INIS)

    Thelliez, T.; Daemen, L.; Hjelm, R.P.; Seeger, P.A.

    1995-01-01

    The authors describe a prototype of a new user interface for the Monte Carlo neutron optics simulation program MCLIB. At this point in its development the interface allows the user to define an instrument as a set of predefined instrument elements. The user can specify the intrinsic parameters of each element, its position and orientation. The interface then writes output to the MCLIB package and starts the simulation. The present prototype is an early development stage of a comprehensive Monte Carlo simulations package that will serve as a tool for the design, optimization and assessment of performance of new neutron scattering instruments. It will be an important tool for understanding the efficacy of new source designs in meeting the needs of these instruments

  9. 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.

  10. PLUTON: Three-group neutronic code for burnup analysis of isotope generation and depletion in highly irradiated LWR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Lemehov, Sergei E; Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-08-01

    PLUTON is a three-group neutronic code analyzing, as functions of time and burnup, the change of radial profiles, together with average values, of power density, burnup, concentration of trans-uranium elements, plutonium buildup, depletion of fissile elements, and fission product generation in water reactor fuel rod with standard UO{sub 2}, UO{sub 2}-Gd{sub 2}O{sub 3}, inhomogeneous MOX, and UO{sub 2}-ThO{sub 2}. The PLUTON code, which has been designed to be run on Windows PC, has adopted a theoretical shape function of neutron attenuation in pellet, which enables users to perform a very fast and accurate calculation easily. The present code includes the irradiation conditions of the Halden Reactor which gives verification data for the code. The total list of trans-uranium elements included in the calculations consists of {sub 92}U{sup 233-239}, {sub 93}Np{sup 237-239}, {sub 94}Pu{sup 238-243}, {sub 95}Am{sup 241-244} (including isomers), and {sub 96}Cm{sup 242-245}. Poisoning fission products are represented by {sub 54}Xe{sup 131,133,135}, {sub 48}Cd{sup 113}, {sub 62}Sm{sup 149,151,152}, {sub 64}Gd{sup 154-160}, {sub 63}Eu{sup 153,155}, {sub 36}Kr{sup 83,85}, {sub 42}Mo{sup 95}, {sub 43}Tc{sup 99}, {sub 45}Rh{sup 103}, {sub 47}Ag{sup 109}, {sub 53}I{sup 127,129,131}, {sub 55}Cs{sup 133}, {sub 57}La{sup 139}, {sub 59}Pr{sup 141}, {sub 60}Nd{sup 143-150}, {sub 61}Pm{sup 147}. Fission gases and volatiles included in the code are {sub 36}Kr{sup 83-86}, {sub 54}Xe{sup 129-136}, {sub 52}Te{sup 125-130}, {sub 53}I{sup 127-131}, {sub 55}Cs{sup 133-137}, and {sub 56}Ba{sup 135-140}. Verification has been performed up to 83 GWd/tU, and a satisfactory agreement has been obtained. (author)

  11. SURGICAL MANAGEMENT OF APPENDICULAR LONG-BONE FRACTURES IN FREE-RANGING FLORIDA PANTHERS ( PUMA CONCOLOR CORYI): SIX CASES (2000-2014).

    Science.gov (United States)

    Au Yong, Jo Anne; Lewis, Daniel D; Citino, Scott B; Cunningham, Mark W; Cross, Alan R; Farese, James P; Pablo, Luisito S

    2018-03-01

    The clinical outcomes of six free-ranging Florida panthers ( Puma concolor coryi) that underwent surgical stabilization of appendicular long-bone fractures (three femoral fractures, one tibial and one tibial and fibular fracture and two radial and ulnar fractures) were evaluated. These panthers presented to the University of Florida from 2000-2014. Estimated age of the panthers ranged from 0.5 to 4.5 yr, and weights ranged from 22 to 65 kg. Causes of injuries were vehicular collision ( n = 4) and capture related ( n = 2). All panthers underwent open reduction and fracture stabilization. Fixation failure necessitated three subsequent surgeries in one panther. Five panthers survived the immediate postoperative period, and all of these panthers' fractures obtained radiographic union (range, 8-36 [mean, 22] wk). The five surviving panthers underwent convalescence for 7-14 mo at White Oak Conservation Center before being released back into the wild; however, one panther was killed when hit by a car 3 days after release. The remaining four panthers were tracked for up to 106 mo in the wild and successfully integrated back into the native population. Surgical stabilization of appendicular long-bone fractures in free-ranging Florida panthers can be successful, but must take into account the stress that a large, undomesticated felid will place on the stabilized limb during convalescence as well as the difficulties involved in rehabilitating a wild panther in captivity.

  12. 75 FR 1751 - Hearing on the Department of Justice's Actions Related to the New Black Panther Party Litigation...

    Science.gov (United States)

    2010-01-13

    ... COMMISSION ON CIVIL RIGHTS Hearing on the Department of Justice's Actions Related to the New Black Panther Party Litigation and Its Enforcement of Section 11(b) of the Voting Rights Act AGENCY: United... Panther Party Litigation and enforcement of Section 11(b) of the Voting Rights Act. The Commission is...

  13. 75 FR 13076 - Hearing on the Department of Justice's Actions Related to the New Black Panther Party Litigation...

    Science.gov (United States)

    2010-03-18

    ... COMMISSION ON CIVIL RIGHTS Hearing on the Department of Justice's Actions Related to the New Black Panther Party Litigation and Its Enforcement of Section 11(b) of the Voting Rights Act AGENCY: United... Department of Justice's actions in the New Black Panther Party Litigation and enforcement of Section 11(b) of...

  14. 75 FR 7441 - Hearing on the Department of Justice's Actions Related to the New Black Panther Party Litigation...

    Science.gov (United States)

    2010-02-19

    ... COMMISSION ON CIVIL RIGHTS Hearing on the Department of Justice's Actions Related to the New Black Panther Party Litigation and Its Enforcement of Section 11(b) of the Voting Rights Act AGENCY: United... Panther Party Litigation and enforcement of Section 11(b) of the Voting Rights Act. The Commission is...

  15. 77 FR 55817 - Panther Creek Power Operating, LLC; Supplemental Notice that Initial Market-Based Rate Filing...

    Science.gov (United States)

    2012-09-11

    ... DEPARTMENT OF ENERGY Federal Energy Regulatory Commission [Docket No. ER12-2570-000] Panther Creek Power Operating, LLC; Supplemental Notice that Initial Market-Based Rate Filing Includes Request for... Panther Creek Power Operating, LLC's application for market-based rate authority, with an accompanying...

  16. 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.

  17. Parallelization of the MAAP-A code neutronics/thermal hydraulics coupling

    International Nuclear Information System (INIS)

    Froehle, P.H.; Wei, T.Y.C.; Weber, D.P.; Henry, R.E.

    1998-01-01

    A major new feature, one-dimensional space-time kinetics, has been added to a developmental version of the MAAP code through the introduction of the DIF3D-K module. This code is referred to as MAAP-A. To reduce the overall job time required, a capability has been provided to run the MAAP-A code in parallel. The parallel version of MAAP-A utilizes two machines running in parallel, with the DIF3D-K module executing on one machine and the rest of the MAAP-A code executing on the other machine. Timing results obtained during the development of the capability indicate that reductions in time of 30--40% are possible. The parallel version can be run on two SPARC 20 (SUN OS 5.5) workstations connected through the ethernet. MPI (Message Passing Interface standard) needs to be implemented on the machines. If necessary the parallel version can also be run on only one machine. The results obtained running in this one-machine mode identically match the results obtained from the serial version of the code

  18. Sizewell B cycle 5 core design with Framatome ANP's CASCADE-3D and British Energy's PANTHER

    Energy Technology Data Exchange (ETDEWEB)

    Attale, F.; Koegl, J. [Framatome ANP GmbH, Nuclear Fuel Cycle, Erlangen (Germany); Knight, M.; Bryce, P. [British Energy, Nuclear Technology Branch, Gloucester (United Kingdom)

    2001-07-01

    Sizewell B Cycle 5 is the first cycle, after 4 cycles with BNFL fuel, with a reload consisting of Framatome ANP HTP (high thermal performance) fuel assemblies. The impact of this fuel vendor change on the Nuclear Design area is that, according to British energy's (BE) practice, the Framatome ANP's nuclear design code system CASCADE-3D is used for the majority of the cycle specific safety case calculations. However, other parts of the safety submission (e.g. 3D transient analyses) are made by using the BE code PANTHER. Before using in parallel two different code systems for reload core licensing extensive comparisons of applied methodologies and obtained results were required to ensure an acceptable level of agreement. (orig.)

  19. Installation of Monte Carlo neutron and photon transport code system MCNP4

    International Nuclear Information System (INIS)

    Takano, Makoto; Sasaki, Mikio; Kaneko, Toshiyuki; Yamazaki, Takao.

    1993-03-01

    The continuous energy Monte Carlo code MCNP-4 including its graphic functions has been installed on the Sun-4 sparc-2 work station with minor corrections. In order to validate the installed MCNP-4 code, 25 sample problems have been executed on the work station and these results have been compared with the original ones. And, the most of the graphic functions have been demonstrated by using 3 sample problems. Further, additional 14 nuclides have been included to the continuous cross section library edited from JENDL-3. (author)

  20. Advanced Neutron Source Dynamic Model (ANSDM) code description and user guide

    International Nuclear Information System (INIS)

    March-Leuba, J.

    1995-08-01

    A mathematical model is designed that simulates the dynamic behavior of the Advanced Neutron Source (ANS) reactor. Its main objective is to model important characteristics of the ANS systems as they are being designed, updated, and employed; its primary design goal, to aid in the development of safety and control features. During the simulations the model is also found to aid in making design decisions for thermal-hydraulic systems. Model components, empirical correlations, and model parameters are discussed; sample procedures are also given. Modifications are cited, and significant development and application efforts are noted focusing on examination of instrumentation required during and after accidents to ensure adequate monitoring during transient conditions

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-05-16

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

  3. Algorithmic developments and qualification of the ERANOS nuclear code and data system for the characterization of fast neutron reactors

    International Nuclear Information System (INIS)

    Rimpault, G.

    2003-09-01

    In this report, the author discusses the algorithmic and methodological developments in the field of nuclear reactor physics, and more particularly the developments of the ERALIB1/ERANOS nuclear code and data system for the calculation of core critical mass and power of sodium-cooled fast neutron reactors (Phenix and Super Phenix), and of the CAPRA 4/94 core. After a brief recall of nuclear data and methods used to determine critical masses and powers, the author discusses the interpretation of start-up experiments performed on Super-Phenix. The methodology used to characterize the uncertainties of these parameters is then applied to the calculation of the Super-Phenix critical mass and power distribution. He presents the approach chosen to define the validity domain of the ERANOS form

  4. Neutronic and thermal-hydraulic calculations for the AP-1000 NPP with the MCNP6 and SERPENT codes

    International Nuclear Information System (INIS)

    Stefani, Giovanni Laranjo; Maiorino, Jose R.; Santos, Thiago A.

    2015-01-01

    The AP-1000 is an evolutionary PWR reactor designed as an evolution of the AP-600 project. The reactor is already pre-licensed by NRC, and is considered to have achieved high standards of safety, possible short construction time and good economic competitiveness. The core is a 17x17 typical assembly using Zirlo as cladding, 3 different enrichment regions, and is controlled by boron, control banks, and burnable poison. The expected fuel final burnup is 62 MWD/ton U and a cycle of 18 months. In this paper we present results for neutronic and thermal-hydraulic calculations for the AP-1000. We use the MCNP6 and SERPENT codes to calculate the first cycle of operation. The calculated parameters are K eff at BOL and EOL and its variation with burnup and neutron flux, and reactivity coefficients. The production of transuranic elements such as Pu-239 and Pu-241, and burning fuel are calculated over time. In the work a complete reactor was burned for 450 days with no control elements, boron or burnable poison were considered, these results were compared with data provided by the Westinghouse. The results are compared with those reported in the literature. A simple thermal hydraulic analysis allows verification of thermal limits such as fuel and cladding temperatures, and MDNB. (author)

  5. Neutron flux investigation on certain alternative fluids in a hybrid system by using MCNPX Monte Carlo transport code

    Energy Technology Data Exchange (ETDEWEB)

    Guenay, Mehtap [Inoenue Univ., Malatya (Turkey). Physics Dept.

    2014-04-15

    In this study, the molten salt-heavy metal mixtures 93-85 % Li{sub 20}Sn{sub 80} + 5 % SFG-PuO{sub 2} and 2-10 % UO{sub 2}, 93-85 % Li{sub 20}Sn{sub 80} + 5 % SFG-PuO{sub 2} and 2-10 % NpO{sub 2}, 93-85 % Li{sub 20}Sn{sub 80} + 5 % SFG-PuO{sub 2} and 2-10 % UCO were used as fluids. The fluids were used in the liquid first wall, blanket and shield zones of the designed hybrid reactor system. Four centimeter thick 9Cr2WVTa ferritic steel was used as the structural material. In this study, the effect of mixture components on the neutron flux was investigated in a designed fusion-fission hybrid reactor system. The neutron flux was investigated according to the mixture components, radial flux distribution and energy spectrum in the designed system. Three-dimensional analyses were performed using the most recent MCNPX-2.7.0 Monte Carlo radiation transport code and the ENDF/B-VII.0 nuclear data library. (orig.)

  6. Neutronic and thermal-hydraulic calculations for the AP-1000 NPP with the MCNP6 and SERPENT codes

    Energy Technology Data Exchange (ETDEWEB)

    Stefani, Giovanni Laranjo; Maiorino, Jose R.; Santos, Thiago A., E-mail: giovanni.laranjo@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br, E-mail: thiago.santos@ufabc.edu.br [Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais; Rossi, Pedro R., E-mail: pedro.russorossi@gmail.com [FERMIUM - Tecnologia Nuclear, Sao Paulo, SP (Brazil)

    2015-07-01

    The AP-1000 is an evolutionary PWR reactor designed as an evolution of the AP-600 project. The reactor is already pre-licensed by NRC, and is considered to have achieved high standards of safety, possible short construction time and good economic competitiveness. The core is a 17x17 typical assembly using Zirlo as cladding, 3 different enrichment regions, and is controlled by boron, control banks, and burnable poison. The expected fuel final burnup is 62 MWD/ton U and a cycle of 18 months. In this paper we present results for neutronic and thermal-hydraulic calculations for the AP-1000. We use the MCNP6 and SERPENT codes to calculate the first cycle of operation. The calculated parameters are K{sub eff} at BOL and EOL and its variation with burnup and neutron flux, and reactivity coefficients. The production of transuranic elements such as Pu-239 and Pu-241, and burning fuel are calculated over time. In the work a complete reactor was burned for 450 days with no control elements, boron or burnable poison were considered, these results were compared with data provided by the Westinghouse. The results are compared with those reported in the literature. A simple thermal hydraulic analysis allows verification of thermal limits such as fuel and cladding temperatures, and MDNB. (author)

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

    International Nuclear Information System (INIS)

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

    1976-07-01

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

  8. 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.)

  9. Evaluation of wrapper tube temperatures of fast neutron reactors using the TRANSCOEUR-2 code

    Energy Technology Data Exchange (ETDEWEB)

    Valentin, B.; Brun P. [CEA/DRN/DEC/SECA/LHC CEN, St Paul Lez Durance (France); Chaigne, G. [FRAMATOME/NOVATOME, Lyon (France)

    1995-09-01

    This paper deals with the thermal loading estimation of wrapper tubes using the TRANSCOEUR-2 code. This estimation requires a knowledge of two temperature fields: the first involves the peripheral sub-channel temperatures of each sub-assembly calculated by the design code CADET, and the second, outside the sub-assemblies, is the inter-wrapper flow temperature field calculated by the thermal-hydraulic code TRIO-VF with boundary conditions taken from CADET. Theoretical models of the three codes are presented as well as the first TRANSCOEUR-2 wrapper tube temperature calculation performed on the European Fast Reactor (EFR) Core Design 6/91 (CD 6/91) under nominal power conditions. The results show a temperature variation of 115{degrees}C between the bottom of the lower blanket and the top of the upper blanket fuel sub-assemblies in the center of the core and 95{degrees}C at the core periphery. The wrapper tube temperatures are higher in the center than in the external core.

  10. Development of a nuclear data uncertainties propagation code on the residual power in fast neutron reactors

    International Nuclear Information System (INIS)

    Benoit, J.-C.

    2012-01-01

    This PhD study is in the field of nuclear energy, the back end of nuclear fuel cycle and uncertainty calculations. The CEA must design the prototype ASTRID, a sodium cooled fast reactor (SFR) and one of the selected concepts of the Generation IV forum, for which the calculation of the value and the uncertainty of the decay heat have a significant impact. In this study is developed a code of propagation of uncertainties of nuclear data on the decay heat in SFR. The process took place in three stages. The first step has limited the number of parameters involved in the calculation of the decay heat. For this, an experiment on decay heat on the reactor PHENIX (PUIREX 2008) was studied to validate experimentally the DARWIN package for SFR and quantify the source terms of the decay heat. The second step was aimed to develop a code of propagation of uncertainties: CyRUS (Cycle Reactor Uncertainty and Sensitivity). A deterministic propagation method was chosen because calculations are fast and reliable. Assumptions of linearity and normality have been validated theoretically. The code has also been successfully compared with a stochastic code on the example of the thermal burst fission curve of 235 U. The last part was an application of the code on several experiments: decay heat of a reactor, isotopic composition of a fuel pin and the burst fission curve of 235 U. The code has demonstrated the possibility of feedback on nuclear data impacting the uncertainty of this problem. Two main results were highlighted. Firstly, the simplifying assumptions of deterministic codes are compatible with a precise calculation of the uncertainty of the decay heat. Secondly, the developed method is intrusive and allows feedback on nuclear data from experiments on the back end of nuclear fuel cycle. In particular, this study showed how important it is to measure precisely independent fission yields along with their covariance matrices in order to improve the accuracy of the calculation of

  11. Vectorization and multitasking with a Monte-Carlo code for neutron transport problems

    International Nuclear Information System (INIS)

    Chauvet, Y.

    1985-04-01

    This paper summarizes two improvements of a Monte Carlo code by resorting to vectorization and multitasking techniques. After a short presentation of the physical problem to solve and a description of the main difficulties to produce an efficient coding, this paper introduces the vectorization principles employed and briefly describes how the vectorized algorithm works. Next, measured performances on CRAY 1S, CYBER 205 and CRAY X-MP are compared. The second part of this paper is devoted to multitasking technique. Starting from the standard multitasking tools available with FORTRAN on CRAY X-MP/4, a multitasked algorithm and its measured speed-ups are presented. In conclusion we prove that vector and parallel computers are a great opportunity for such Monte Carlo algorithms

  12. Randomly dispersed particle fuel model in the PSG Monte Carlo neutron transport code

    International Nuclear Information System (INIS)

    Leppaenen, J.

    2007-01-01

    High-temperature gas-cooled reactor fuels are composed of thousands of microscopic fuel particles, randomly dispersed in a graphite matrix. The modelling of such geometry is complicated, especially using continuous-energy Monte Carlo codes, which are unable to apply any deterministic corrections in the calculation. This paper presents the geometry routine developed for modelling randomly dispersed particle fuels using the PSG Monte Carlo reactor physics code. The model is based on the delta-tracking method, and it takes into account the spatial self-shielding effects and the random dispersion of the fuel particles. The calculation routine is validated by comparing the results to reference MCNP4C calculations using uranium and plutonium based fuels. (authors)

  13. PWR neutron ex-vessel detection calculations using three-dimensional codes

    International Nuclear Information System (INIS)

    Dekens, O.; Lefebvre, J.C.; Rohart, M.; Chiron, M.

    1997-01-01

    During the accident of TM12, the signal delivered by source detectors was exceptionally high. This phenomenon was found out to be due to the water inventory in the primary system. Thus, in their research activity, Electricite de France (EdF) and Commissariat a l'Energie Atomique (CEA) have jointly launched a programme, whose aim was to determine to what extent the response of ex-vessel neutron detectors are representative of reactor water level (or sources positions) in a French 900 MWe PWR. In this framework, both partners developed the methods needed for each step of the calculation chain. Finally, a simulation of a LOCA indicates that the loss of coolant can be detected by existing monitoring system, and could be more efficiently found by changing the position of the source range detectors. (authors)

  14. Neutronic and thermal-hydraulic coupling using Milonga and OpenFOAM codes: an approach using free software

    International Nuclear Information System (INIS)

    Silva, Vitor Vasconcelos Araújo

    2016-01-01

    The development of a fine mesh coupled neutronics/thermal-hydraulics framework mainly using open source software is presented. The contributions proposed go in two different directions: one, is the focus on the open software development, a concept widely spread in many fields of knowledge but rarely explored in the nuclear engineering field; the second, is the use of operating system shared memory as a fast and reliable storage area to couple the computational fluid dynamics (CFD) software OpenFOAM to the free and flexible reactor core analysis code Milonga. This concept was applied to simulate the behavior of the TRIGA Mark 1 IPR-R1 reactor fuel pin in steady-state mode. The macroscopic cross-sections for the model, a set of two-group cross-sections data, were generated using WIMSD-5B code. The results show that this innovative coupled system gives consistent results, encouraging system further development and its use for complex nuclear systems. (author)

  15. Evaluation of the fluence to dose conversion coefficients for high energy neutrons using a voxel phantom coupled with the GEANT4 code

    CERN Document Server

    Paganini, S

    2005-01-01

    Crews working on present-day jet aircraft are a large occupationally exposed group with a relatively high average effective dose from Galactic cosmic radiation. Crews of future high-speed commercial flying at higher altitudes would be even more exposed. To help reduce the significant uncertainties in calculations of such exposures, the male adult voxels phantom MAX, developed in the Nuclear Energy Department of Pernambuco Federal University in Brazil, has been coupled with the Monte Carlo simulation code GEANT4. This toolkit, distributed and upgraded from the international scientific community of CERN/Switzerland, simulates thermal to ultrahigh energy neutrons transport and interactions in the matter. The high energy neutrons are pointed as the component that contribute about 70% of the neutron effective dose that represent the 35% to 60% total dose at aircraft altitude. In this research calculations of conversion coefficients from fluence to effective dose are performed for neutrons of energies from 100 MeV ...

  16. Method of neutronic calculations for a spherical cell equivalent to cylindrical one for using computer codes in light water reactors in the fluidized bed nuclear reactor

    International Nuclear Information System (INIS)

    Borges, V.; Sefidvash, F.; Rastogi, E.P.; Huria, H.C.; Krishnani, P.D.

    1989-01-01

    In order to use the existing light water reactor cell calculation codes for fluidized bed nuclear reactor having spherical fuel cells, an equivalence method has been developed. This method is shown to be adequate in calculation of the Dancoff factor. This method also was applicable in LEOPARD code and the results obtained in calculation of K ∞ was compared with the obtained using the DTF IV code, the results showed that the method is adequate for the calculations neutronics of the fluidized bed nuclear reactor. (author) [pt

  17. 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

  18. Analysis of PBMR transients using a coupled neutron transport/thermal-hydraulics code DORT-TD/thermix

    International Nuclear Information System (INIS)

    Tyobeka, B.; Ivanov, K.; Pautz, A.

    2007-01-01

    In the advent of increased demand for safety and economics of nuclear power plants, nuclear engineers and designers are called upon to develop advanced computation tools. In these developments, space-time effects in the dynamics of nuclear reactors must be considered within the framework of a full 3-dimensional treatment of both neutron kinetics and thermal hydraulics. In a recent effort at the Pennsylvania State University, a time-dependent version of the discrete ordinates transport code DORT, DORT-TD was coupled to a 2-dimensional core thermal hydraulics code THERMIX-DIREKT. In the coupling process, a feedback model was developed to account for the feedback effects and was implemented into DORT-TD. During the calculation process for each spatial node of the DORT-TD core model, feedback parameters representative of this node are passed to the feedback module. Using these values, cross section tables are then interpolated for the appropriate macroscopic cross section values. The updated macroscopic cross sections are passed back to DORT-TD to perform transport core calculations, and the power distribution is transferred to THERMIX-DIREKT to obtain the relevant thermal-hydraulics data in turn, and this calculation loop continues. In this paper, DORT-TD/THERMIX is used to simulate transients of interest in the PBMR (Pebble Bed Modular Reactor) safety using established benchmark problems: load change from 100% to 40% power and fast control rod ejection (PBMR-268 benchmark problem). The results obtained are compared with those obtained using the diffusion-based module of the code. The results are only preliminary and so far show that diffusion theory is not such a bad approximation for PBMR for the prediction of integral parameters

  19. Assessment of neutron transport codes for application to CANDU fuel lattices analysis

    International Nuclear Information System (INIS)

    Roh, Gyu Hong; Choi, Hang Bok

    1999-08-01

    In order to assess the applicability of WIMS-AECL and HELIOS code to the CANDU fuel lattice analysis, the physics calculations has been carried out for the standard CANDU fuel and DUPIC fuel lattices, and the results were compared with those of Monte Carlo code MCNP-4B. In this study, in order to consider the full isotopic composition and the temperature effect, new MCNP libraries have been generated from ENDF/B-VI release 3 and validated for typical benchmark problems. The TRX-1,2,BAPL-1,2,3 pin -cell lattices and KENO criticality safety benchmark calculations have been performed for the new MCNP libraries, and the results have shown that the new MCNP library has sufficient accuracy to be used for physics calculation. Then, the lattice codes have been benchmarked by the MCNP code for the major physics parameters such as the burnup reactivity, void reactivity, relative pin power and Doppler coefficient, etc. for the standard CANDU fuel and DUPIC fuel lattices. For the standard CANDU fuel lattice, it was found that the results of WIMS-AECL calculations are consistent with those of MCNP. For the DUPIC fuel lattice, however, the results of WIMS-AECL calculations with ENDF/B-V library have shown that the discrepancy from the results of MCNP calculations increases when the fuel burnup is relatively high. The burnup reactivities of WIMS-ACEL calculations with ENDF/B-VI library have shown excellent agreements with those of MCNP calculation for both the standard CANDU and DUPIC fuel lattices. However, the Doppler coefficient have relatively large discrepancies compared with MCNP calculations, and the difference increases as the fuel burns. On the other hand, the results of HELIOS calculation are consistent with those of MCNP even though the discrepancy is slightly larger compared with the case of the standard CANDU fuel lattice. this study has shown that the WIMS-AECL products reliable results for the natural uranium fuel. However, it is recommended that the WIMS

  20. Morphology of the dorsal lingual papillae in the newborn panther and Asian black bear.

    Science.gov (United States)

    Emura, S; Hayakawa, D; Chen, H; Shoumura, S

    2001-12-01

    The dorsal lingual surfaces of a newborn panther (Panthera pardus) and two newborn asian black bears (Selenarctos thibetanus) were examined by scanning electron microscopy (SEM). The tongues of the panther and asian black bear were about 40 mm in length and about 20 mm in width. Filiform, fungiform and vallate papillae were found. The filiform papillae were distributed over the entire dorsal surface of the tongue. In the panther, the filiform papillae on margin of the lingual apex were divided into two shapes which were horny or club-shaped papillae. The filiform papillae on the midportion were larger than those on the lateral region in size. The fungiform papillae also were divided into two shapes which were hemispherical or club-shaped papillae. In the asian black bear, the filiform papillae on the margin of the lingual apex were larger than those on margin of the panther tongue. The vallate papillae in the animals of two species were located on both sides of the posterior end of the lingual body. Each papilla was surrounded by a groove and crescent pad.

  1. Assessing effects of changing land use practices on sediment loads in Panther Creek, north coastal California

    Science.gov (United States)

    Mary Ann Madej; Greg Bundros; Randy Klein

    2012-01-01

    Revisions to the California Forest Practice Rules since 1974 were intended to increase protection of water quality in streams draining timber harvest areas. The effects of improved timber harvesting methods and road designs on sediment loading are assessed for the Panther Creek basin, a 15.4 km2 watershed in Humboldt County, north coastal...

  2. U.S. EPA, Pesticide Product Label, RED PANTHER SUPER JUCE, 06/02/1971

    Science.gov (United States)

    2011-04-21

    ... " 1 ' .... 1. \\ ... 1 , ,.: r', .: • i":;c '\\ I, )1: !~'1' .\\~~I\\1.\\:.: 1" :'!( II·T "i: \\\\. 'II'll P'HI' "\\ l"Il:, I'! 'HI" 1:-:1'.', 'II~ \\\\.I! I-:HI: , J [ •• '. PANTHER JUCE . , ...

  3. If Animals Could Talk: Bald Eagle, Bear, Florida Panther, Gopher Tortoise, Indigo Snake, Manatee, Otter, Raccoon.

    Science.gov (United States)

    Pinellas County District School Board, Clearwater, FL.

    In this series of booklets, eight Florida animals describe their appearance, habitats, food, behavior, and relationships with humans. Each entry is written for elementary students from the animal's point of view and includes a bibliography. Contained are the life stories of the bald eagle, black bear, Florida panther, gopher tortoise, Eastern…

  4. Essential Mac OS X panther server administration integrating Mac OS X server into heterogeneous networks

    CERN Document Server

    Bartosh, Michael

    2004-01-01

    If you've ever wondered how to safely manipulate Mac OS X Panther Server's many underlying configuration files or needed to explain AFP permission mapping--this book's for you. From the command line to Apple's graphical tools, the book provides insight into this powerful server software. Topics covered include installation, deployment, server management, web application services, data gathering, and more

  5. TRIDENT: a two-dimensional, multigroup, triangular mesh discrete ordinates, explicit neutron transport code

    International Nuclear Information System (INIS)

    Seed, T.J.; Miller, W.F. Jr.; Brinkley, F.W. Jr.

    1977-03-01

    TRIDENT solves the two-dimensional-multigroup-transport equations in rectangular (x-y) and cylindrical (r-z) geometries using a regular triangular mesh. Regular and adjoint, inhomogeneous and homogeneous (k/sub eff/ and eigenvalue searches) problems subject to vacuum, reflective, white, or source boundary conditions are solved. General anisotropic scattering is allowed and anisotropic-distributed sources are permitted. The discrete-ordinates approximation is used for the neutron directional variables. An option is included to append a fictitious source to the discrete-ordinates equations that is defined such that spherical-harmonics solutions (in x-y geometry) or spherical-harmonics-like solutions (in r-z geometry) are obtained. A spatial-finite-element method is used in which the angular flux is expressed as a linear polynomial in each triangle that is discontinous at triangle boundaries. Unusual Features of the program: Provision is made for creation of standard interface output files for S/sub N/ constants, angle-integrated (scalar) fluxes, and angular fluxes. Standard interface input files for S/sub N/ constants, inhomogeneous sources, cross sections, and the scalar flux may be read. Flexible edit options as well as a dump and restart capability are provided

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

    International Nuclear Information System (INIS)

    Bilanovic, Z.; McCracken, D.R.

    1994-12-01

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

  7. Analysis of a Neutronic Computational Model for the Core of Material Testing Reactor MTR by Using SQUID Code

    International Nuclear Information System (INIS)

    Al-Taweel, M.H.

    2015-01-01

    It is a conventional practice in the design of nuclear reactor to introduce calculation of hot points to determine spatial variation for energy generated and then determine power distribution.The study had been carried out for core of a reactor type (MTR) by the neutronic code SQUID. In this study, we replace the reflector of the reactor by H 2 O instead of D 2 O as originally the reactor designed.From the study we conclude that the reactor can operates safely, to make sure of that we calculate the multiplication factor where their values ranged from (1.0854) when all control rods are up to (1.001)when three control rods are up.Also the values of hot points were calculated and compared with French documents results with D 2 O as a reflector where the difference is (0.19%), and with light water as reflector instead of heavy water was calculated.For different cases according to control rod position , the values of hot point ranged between (0.46) to (1.64) in case all control rods are up also the values of the average power distributed on different fuel cells were calculated in case of light water as reflector firstly with three control rods are down and the maximum value (2.13*10 -2 Μw).Secondly in case offour control rods are down, the maximum value (1.925*10 -2 Μw) we notice almost coincidence between the neutron flux distribution through the core of reactor and in different positions of control rods

  8. Research on application of system of neutron, thermohydraulic and safety analysis codes in order to simulation of the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Pham Van Lam; Le Vinh Vinh; Huynh Ton Nghiem

    2004-01-01

    Requirements of neutron, thermohydraulic and safety analysis calculation are very important because of issuing new version of SAR for DNRR, research on construction of new research reactor and nuclear power plant. Research on application of system of neutron, thermohydraulic and safety analysis codes in order to simulation of the Dalat Nuclear Research Reactor has been done in the frame work of research theme in the year 2002-2003. The purposes of the research are maintaining safety operation of the DNRR and enhancement of man power and calculation and safety analysis tool potential. (author)

  9. A new modelling of the multigroup scattering cross section in deterministic codes for neutron transport

    International Nuclear Information System (INIS)

    Calloo, A.A.

    2012-01-01

    In reactor physics, calculation schemes with deterministic codes are validated with respect to a reference Monte Carlo code. The remaining biases are attributed to the approximations and models induced by the multigroup theory (self-shielding models and expansion of the scattering law using Legendre polynomials) to represent physical phenomena (resonant absorption and scattering anisotropy respectively). This work focuses on the relevance of a polynomial expansion to model the scattering law. Since the outset of reactor physics, the latter has been expanded on a truncated Legendre polynomial basis. However, the transfer cross sections are highly anisotropic, with non-zero values for a very small range of the cosine of the scattering angle. Besides, the finer the energy mesh and the lighter the scattering nucleus, the more exacerbated is the peaked shape of this cross section. As such, the Legendre expansion is less suited to represent the scattering law. Furthermore, this model induces negative values which are non-physical. In this work, various scattering laws are briefly described and the limitations of the existing model are pointed out. Hence, piecewise-constant functions have been used to represent the multigroup scattering cross section. This representation requires a different model for the diffusion source. The discrete ordinates method which is widely employed to solve the transport equation has been adapted. Thus, the finite volume method for angular discretization has been developed and implemented in Paris environment which hosts the S n solver, Snatch. The angular finite volume method has been compared to the collocation method with Legendre moments to ensure its proper performance. Moreover, unlike the latter, this method is adapted for both the Legendre moments and the piecewise-constant functions representations of the scattering cross section. This hybrid-source method has been validated for different cases: fuel cell in infinite lattice

  10. GANTRAS - a system of codes for the solution of the multigroup transport equation with a rigorous treatment of anisotropic neutron scattering

    International Nuclear Information System (INIS)

    Schwenk-Ferrero, A.

    1986-11-01

    GANTRAS is a system of codes for neutron transport calculations in which the anisotropy of elastic and inelastic (including (n,n'x)-reactions) scattering is fully taken into account. This is achieved by employing a rigorous method, so-called I * -method, to represent the scattering term of the transport equation and with the use of double-differential cross-sections for the description of the emission of secondary neutrons. The I * -method was incorporated into the conventional transport code ONETRAN. The ONETRAN subroutines were modified for the new purpose. An implementation of the updated version ANTRA1 was accomplished for plane and spherical geometry. ANTRA1 was included in GANTRAS and linked to another modules which prepare angle-dependent transfer matrices. The GANTRAS code consists of three modules: 1. The CROMIX code which calculates the macroscopic transfer matrices for mixtures on the base of microscopic nuclide-dependent data. 2. The ATP code which generates discretized angular transfer probabilities (i.e. discretizes the I * -function). 3. The ANTRA1 code to perform S N transport calculations in one-dimensional plane and spherical geometries. This structure of GANTRAS allows to accommodate the system to various transport problems. (orig.) [de

  11. ANITA-2000 activation code package - updating of the decay data libraries and validation on the experimental data of the 14 MeV Frascati Neutron Generator

    Directory of Open Access Journals (Sweden)

    Frisoni Manuela

    2016-01-01

    Full Text Available ANITA-2000 is a code package for the activation characterization of materials exposed to neutron irradiation released by ENEA to OECD-NEADB and ORNL-RSICC. The main component of the package is the activation code ANITA-4M that computes the radioactive inventory of a material exposed to neutron irradiation. The code requires the decay data library (file fl1 containing the quantities describing the decay properties of the unstable nuclides and the library (file fl2 containing the gamma ray spectra emitted by the radioactive nuclei. The fl1 and fl2 files of the ANITA-2000 code package, originally based on the evaluated nuclear data library FENDL/D-2.0, were recently updated on the basis of the JEFF-3.1.1 Radioactive Decay Data Library. This paper presents the results of the validation of the new fl1 decay data library through the comparison of the ANITA-4M calculated values with the measured electron and photon decay heats and activities of fusion material samples irradiated at the 14 MeV Frascati Neutron Generator (FNG of the NEA-Frascati Research Centre. Twelve material samples were considered, namely: Mo, Cu, Hf, Mg, Ni, Cd, Sn, Re, Ti, W, Ag and Al. The ratios between calculated and experimental values (C/E are shown and discussed in this paper.

  12. Application of multi-thread computing and domain decomposition to the 3-D neutronics Fem code Cronos

    International Nuclear Information System (INIS)

    Ragusa, J.C.

    2003-01-01

    The purpose of this paper is to present the parallelization of the flux solver and the isotopic depletion module of the code, either using Message Passing Interface (MPI) or OpenMP. Thread parallelism using OpenMP was used to parallelize the mixed dual FEM (finite element method) flux solver MINOS. Investigations regarding the opportunity of mixing parallelism paradigms will be discussed. The isotopic depletion module was parallelized using domain decomposition and MPI. An attempt at using OpenMP was unsuccessful and will be explained. This paper is organized as follows: the first section recalls the different types of parallelism. The mixed dual flux solver and its parallelization are then presented. In the third section, we describe the isotopic depletion solver and its parallelization; and finally conclude with some future perspectives. Parallel applications are mandatory for fine mesh 3-dimensional transport and simplified transport multigroup calculations. The MINOS solver of the FEM neutronics code CRONOS2 was parallelized using the directive based standard OpenMP. An efficiency of 80% (resp. 60%) was achieved with 2 (resp. 4) threads. Parallelization of the isotopic depletion solver was obtained using domain decomposition principles and MPI. Efficiencies greater than 90% were reached. These parallel implementations were tested on a shared memory symmetric multiprocessor (SMP) cluster machine. The OpenMP implementation in the solver MINOS is only the first step towards fully using the SMPs cluster potential with a mixed mode parallelism. Mixed mode parallelism can be achieved by combining message passing interface between clusters with OpenMP implicit parallelism within a cluster

  13. A new open-source code for spherically symmetric stellar collapse to neutron stars and black holes

    International Nuclear Information System (INIS)

    O'Connor, Evan; Ott, Christian D

    2010-01-01

    We present the new open-source spherically symmetric general-relativistic (GR) hydrodynamics code GR1D. It is based on the Eulerian formulation of GR hydrodynamics (GRHD) put forth by Romero-Ibanez-Gourgoulhon and employs radial-gauge, polar-slicing coordinates in which the 3+1 equations simplify substantially. We discretize the GRHD equations with a finite-volume scheme, employing piecewise-parabolic reconstruction and an approximate Riemann solver. GR1D is intended for the simulation of stellar collapse to neutron stars and black holes and will also serve as a testbed for modeling technology to be incorporated in multi-D GR codes. Its GRHD part is coupled to various finite-temperature microphysical equations of state in tabulated form that we make available with GR1D. An approximate deleptonization scheme for the collapse phase and a neutrino-leakage/heating scheme for the postbounce epoch are included and described. We also derive the equations for effective rotation in 1D and implement them in GR1D. We present an array of standard test calculations and also show how simple analytic equations of state in combination with presupernova models from stellar evolutionary calculations can be used to study qualitative aspects of black hole formation in failing rotating core-collapse supernovae. In addition, we present a simulation with microphysical equations of state and neutrino leakage/heating of a failing core-collapse supernova and black hole formation in a presupernova model of a 40 M o-dot zero-age main-sequence star. We find good agreement on the time of black hole formation (within 20%) and last stable protoneutron star mass (within 10%) with predictions from simulations with full Boltzmann neutrino radiation hydrodynamics.

  14. A new open-source code for spherically symmetric stellar collapse to neutron stars and black holes

    Energy Technology Data Exchange (ETDEWEB)

    O' Connor, Evan; Ott, Christian D, E-mail: evanoc@tapir.caltech.ed, E-mail: cott@tapir.caltech.ed [TAPIR, Mail Code 350-17, California Institute of Technology, Pasadena, CA 91125 (United States)

    2010-06-07

    We present the new open-source spherically symmetric general-relativistic (GR) hydrodynamics code GR1D. It is based on the Eulerian formulation of GR hydrodynamics (GRHD) put forth by Romero-Ibanez-Gourgoulhon and employs radial-gauge, polar-slicing coordinates in which the 3+1 equations simplify substantially. We discretize the GRHD equations with a finite-volume scheme, employing piecewise-parabolic reconstruction and an approximate Riemann solver. GR1D is intended for the simulation of stellar collapse to neutron stars and black holes and will also serve as a testbed for modeling technology to be incorporated in multi-D GR codes. Its GRHD part is coupled to various finite-temperature microphysical equations of state in tabulated form that we make available with GR1D. An approximate deleptonization scheme for the collapse phase and a neutrino-leakage/heating scheme for the postbounce epoch are included and described. We also derive the equations for effective rotation in 1D and implement them in GR1D. We present an array of standard test calculations and also show how simple analytic equations of state in combination with presupernova models from stellar evolutionary calculations can be used to study qualitative aspects of black hole formation in failing rotating core-collapse supernovae. In addition, we present a simulation with microphysical equations of state and neutrino leakage/heating of a failing core-collapse supernova and black hole formation in a presupernova model of a 40 M{sub o-dot} zero-age main-sequence star. We find good agreement on the time of black hole formation (within 20%) and last stable protoneutron star mass (within 10%) with predictions from simulations with full Boltzmann neutrino radiation hydrodynamics.

  15. Application of multi-thread computing and domain decomposition to the 3-D neutronics Fem code Cronos

    Energy Technology Data Exchange (ETDEWEB)

    Ragusa, J.C. [CEA Saclay, Direction de l' Energie Nucleaire, Service d' Etudes des Reacteurs et de Modelisations Avancees (DEN/SERMA), 91 - Gif sur Yvette (France)

    2003-07-01

    The purpose of this paper is to present the parallelization of the flux solver and the isotopic depletion module of the code, either using Message Passing Interface (MPI) or OpenMP. Thread parallelism using OpenMP was used to parallelize the mixed dual FEM (finite element method) flux solver MINOS. Investigations regarding the opportunity of mixing parallelism paradigms will be discussed. The isotopic depletion module was parallelized using domain decomposition and MPI. An attempt at using OpenMP was unsuccessful and will be explained. This paper is organized as follows: the first section recalls the different types of parallelism. The mixed dual flux solver and its parallelization are then presented. In the third section, we describe the isotopic depletion solver and its parallelization; and finally conclude with some future perspectives. Parallel applications are mandatory for fine mesh 3-dimensional transport and simplified transport multigroup calculations. The MINOS solver of the FEM neutronics code CRONOS2 was parallelized using the directive based standard OpenMP. An efficiency of 80% (resp. 60%) was achieved with 2 (resp. 4) threads. Parallelization of the isotopic depletion solver was obtained using domain decomposition principles and MPI. Efficiencies greater than 90% were reached. These parallel implementations were tested on a shared memory symmetric multiprocessor (SMP) cluster machine. The OpenMP implementation in the solver MINOS is only the first step towards fully using the SMPs cluster potential with a mixed mode parallelism. Mixed mode parallelism can be achieved by combining message passing interface between clusters with OpenMP implicit parallelism within a cluster.

  16. High performance 3D neutron transport on peta scale and hybrid architectures within APOLLO3 code

    International Nuclear Information System (INIS)

    Jamelot, E.; Dubois, J.; Lautard, J-J.; Calvin, C.; Baudron, A-M.

    2011-01-01

    APOLLO3 code is a common project of CEA, AREVA and EDF for the development of a new generation system for core physics analysis. We present here the parallelization of two deterministic transport solvers of APOLLO3: MINOS, a simplified 3D transport solver on structured Cartesian and hexagonal grids, and MINARET, a transport solver based on triangular meshes on 2D and prismatic ones in 3D. We used two different techniques to accelerate MINOS: a domain decomposition method, combined with an accelerated algorithm using GPU. The domain decomposition is based on the Schwarz iterative algorithm, with Robin boundary conditions to exchange information. The Robin parameters influence the convergence and we detail how we optimized the choice of these parameters. MINARET parallelization is based on angular directions calculation using explicit message passing. Fine grain parallelization is also available for each angular direction using shared memory multithreaded acceleration. Many performance results are presented on massively parallel architectures using more than 103 cores and on hybrid architectures using some tens of GPUs. This work contributes to the HPC development in reactor physics at the CEA Nuclear Energy Division. (author)

  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. Development of a neutron transport code many-group two-dimensional heterogeneous calculations by the method of characteristics

    International Nuclear Information System (INIS)

    Petkov, P.T.

    2000-01-01

    The method of characteristics (MOC) is gaining increased popularity in the reactor physics community all over the world because it gives a new degree of freedom in nuclear reactor analysis. The MARIKO code solves the neutron transport equation by the MOC in two-dimensional real geometry. The domain of solution can be a rectangle or right hexagon with periodic boundary conditions on the outer boundary. Any reasonable symmetry inside the domain can be fully accounted for. The geometry is described in three levels-macro-cells, cells, and regions. The macro-cells and cells can be any polygon. The outer boundary of a region can be any combination of straight line and circular arc segments. Any level of embedded regions is allowed. Procedures for automatic geometry description of hexagonal fuel assemblies and reflector macro-cells have been developed. The initial ray tracing procedure is performed for the full rectangular or hexagonal domain, but only azimuthal angles in the smallest symmetry interval are tracked. (Authors)

  19. Computer simulation of displacement cascade structures in D-T neutron-irradiated Au, Ag, Cu, Ni and Al with the MARLOWE code

    International Nuclear Information System (INIS)

    Watanabe, N.; Nishiguchi, R.; Shimomura, Y.

    1991-01-01

    Spatial distribution of point defects in displacement damage cascades at the early stage of their formation was simulated with the MARLOWE code for primary knock-on atoms which is relevant to D-T neutron irradiation. Calculations were carried out for Au, Ag, Cu, Ni and Al. Computer-simulated results were analyzed with complement of TEM observations of D-T neutron-irradiated metals at low temperature. The spatial configuration of displacement cascades, the size of small vacancy aggregates and the size of displacement damage cascade were examined. Results suggest that most of vacancy clusters which were formed in damage cascades may be as small as below 20 vacancies. The remarkable difference in defect yield of cascade damage in Ni and Cu is due to interstitial cluster formation and main contribution of cascade energy overlapping observed in cryotransfer TEM of D-T neutron-irradiated Au is due to ejected interstitials from cascade cores. (orig.)

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

    International Nuclear Information System (INIS)

    Hursin, Mathieu; Xiao Shanjie; Jevremovic, Tatjana

    2006-01-01

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

  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. Correlated Production and Analog Transport of Fission Neutrons and Photons using Fission Models FREYA, FIFRELIN and the Monte Carlo Code TRIPOLI-4® .

    Science.gov (United States)

    Verbeke, Jérôme M.; Petit, Odile; Chebboubi, Abdelhazize; Litaize, Olivier

    2018-01-01

    Fission modeling in general-purpose Monte Carlo transport codes often relies on average nuclear data provided by international evaluation libraries. As such, only average fission multiplicities are available and correlations between fission neutrons and photons are missing. Whereas uncorrelated fission physics is usually sufficient for standard reactor core and radiation shielding calculations, correlated fission secondaries are required for specialized nuclear instrumentation and detector modeling. For coincidence counting detector optimization for instance, precise simulation of fission neutrons and photons that remain correlated in time from birth to detection is essential. New developments were recently integrated into the Monte Carlo transport code TRIPOLI-4 to model fission physics more precisely, the purpose being to access event-by-event fission events from two different fission models: FREYA and FIFRELIN. TRIPOLI-4 simulations can now be performed, either by connecting via an API to the LLNL fission library including FREYA, or by reading external fission event data files produced by FIFRELIN beforehand. These new capabilities enable us to easily compare results from Monte Carlo transport calculations using the two fission models in a nuclear instrumentation application. In the first part of this paper, broad underlying principles of the two fission models are recalled. We then present experimental measurements of neutron angular correlations for 252Cf(sf) and 240Pu(sf). The correlations were measured for several neutron kinetic energy thresholds. In the latter part of the paper, simulation results are compared to experimental data. Spontaneous fissions in 252Cf and 240Pu are modeled by FREYA or FIFRELIN. Emitted neutrons and photons are subsequently transported to an array of scintillators by TRIPOLI-4 in analog mode to preserve their correlations. Angular correlations between fission neutrons obtained independently from these TRIPOLI-4 simulations, using

  3. Development of neutronics and thermal hydraulics coupled code – SAC-RIT for plate type fuel and its application to reactivity initiated transient analysis

    International Nuclear Information System (INIS)

    Singh, Tej; Kumar, Jainendra; Mazumdar, Tanay; Raina, V.K.

    2013-01-01

    Highlights: • A point reactor kinetics code coupled with thermal hydraulics of plate type fuel is developed. • This code is applicable for two phase flow of coolant. • Safety analysis of IAEA benchmark reactor core is carried out. • Results agree well with the results available in literature. - Abstract: A point reactor kinetics code SAC-RIT, acronym of Safety Analysis Code for Reactivity Initiated Transient, coupled with thermal hydraulics of two phase coolant flow for plate type fuel, is developed to calculate reactivity initiated transient analysis of nuclear research and test reactors. Point kinetics equations are solved by fourth order Runge Kutta method. Reactivity feedback effect is included into the code. Solution of kinetics equations gives neutronic power and it is then fed into a thermal hydraulic code where mass, momentum and thermal energy conservation equations are solved by explicit finite difference method to find out fuel, clad and coolant temperatures during transients. In this code, all possible flow regimes including laminar flow, transient flow and turbulent flow have been covered. Various heat transfer coefficients suitable for single liquid, sub-cooled boiling, saturation boiling, film boiling and single vapor phases are incorporated in the thermal hydraulic code

  4. Dosimetry boron neutron capture therapy in liver cancer (hepatocellular carcinoma) by means of MCNP-code with neutron source from thermal column

    International Nuclear Information System (INIS)

    Irhas; Andang Widi Harto; Yohannes Sardjono

    2014-01-01

    Boron Neutron Capture Therapy (BNCT) using physics principle when B 10 (Boron-10) irradiated by low energy neutron (thermal neutron). Boron and thermal neutron reaction produced B 11m (Boron-11m) (t 1/2 =10 -2 s). B 11m decay emitted alpha, Li 7 (Lithium-7) particle and gamma ray. Irradiated time needed to ensure cancer dose enough. Liver cancer was primary malignant who located in liver (Hepatocellular carcinoma). Malignant in liver were different to metastatic from Breast, Colon Cancer, and the other. This condition was Metastatic Liver Cancer. Monte Carlo method used by Monte Carlo N-Particle (MCNP) Software. Probabilistic approach used for probability of interaction occurred and record refers to characteristic of particle and material. In this case, thermal neutron produced by model of Collimated Thermal Column Kartini Research Nuclear Reactor, Yogyakarta. Modelling organ and source used liver organ that contain of cancer tissue and research reactor. Variation of boron concentration was 20, 25, 30, 35, 40, 45, and 47 µg/g cancers. Output of MCNP calculation were neutron scattering dose, gamma ray dose and neutron flux from reactor. Neutron flux used to calculate alpha, proton and gamma ray dose from interaction of tissue material and thermal neutron. Variation of boron concentration result dose rate to every variation were 0,059; 0,072; 0,084; 0,098; 0.108; 0,12; 0,125 Gy/sec. Irradiation time who need to every concentration were 841,5 see (14 min 1 sec); 696,07 sec(11 min 36 sec); 593.11 sec (9 min 53 sec); 461,35 sec (8 min 30 sec); 461,238 sec (7 min 41 sec); 414,23 sec (6 min 54 sec); 398,38 sec (6 min 38 sec). Irradiating time could shortly when boron concentration more high. (author)

  5. Assessment of gamma irradiation heating and damage in miniature neutron source reactor vessel using computational methods and SRIM - TRIM code

    International Nuclear Information System (INIS)

    Appiah-Ofori, F. F.

    2014-07-01

    The Effects of Gamma Radiation Heating and Irradiation Damage in the reactor vessel of Ghana Research Reactor 1, Miniature Neutron Source Reactor were assessed using Implicit Control Volume Finite Difference Numerical Computation and validated by SRIM - TRIM Code. It was assumed that 5.0 MeV of gamma rays from the reactor core generate heat which interact and absorbed completely by the interior surface of the MNSR vessel which affects it performance due to the induced displacement damage. This displacement damage is as result of lattice defects being created which impair the vessel through formation of point defect clusters such as vacancies and interstitiaIs which can result in dislocation loops and networks, voids and bubbles and causing changes in the layers in the thickness of the vessel. The microscopic defects produced in the vessel due to γ - radiation damage are referred to as radiation damage while the defects thus produced modify the macroscopic properties of the vessel which are also known as the radiation effects. These radiation damage effects are of major concern for materials used in nuclear energy production. In this study, the overall objective was to assess the effects of gamma radiation heating and damage in GHARR - I MNSR vessel by a well-developed Mathematical model, Analytical and Numerical solutions, simulating the radiation damage in the vessel. SRIM - TRIM Code was used as a computational tool to determine the displacement per atom (dpa) associated with radiation damage while implicit Control Volume Finite Difference Method was used to determine the temperature profile within the vessel due to γ - radiation heating respectively. The methodology adopted in assessing γ - radiation heating in the vessel involved development of the One-Dimensional Steady State Fourier Heat Conduction Equation with Volumetric Heat Generation both analytical and implicit Control Volume Finite Difference Method approach to determine the maximum temperature and

  6. Auto MOC-A 2D neutron transport code for arbitrary geometry based on the method of characteristics and customization of AutoCAD

    International Nuclear Information System (INIS)

    Chen Qichang; Wu Hongchun; Cao Liangzhi

    2008-01-01

    A new 2D neutron transport code AutoMOC for arbitrary geometry has been developed. This code is based on the method of characteristics (MOCs) and the customization of AutoCAD. The MOC solves the neutron transport equation along characteristic lines. It is independent of the geometric shape of boundaries and regions. So theoretically, this method can be used to solve the neutron transport equation in highly complex geometries. However, it is important to describe the geometry and calculate intersection points of each characteristic line with every boundary and region in advance. In complex geometries, due to the complications of treating the arbitrary domain, the selection of geometric shapes and efficiency of ray tracing are generally limited. The geometry treatment through the customization of AutoCAD, a widely used computer-aided design software package, is given in this paper. Thanks to the powerful capability of AutoCAD, the description of arbitrary geometry becomes quite convenient. Moreover, with the language Visual Basic for Applications (VBAs), AutoCAD can be customized to carry out the ray tracing procedure with a high flexibility in geometry. The numerical results show that AutoMOC can solve 2D neutron transport problems in a complex geometry accurately and effectively

  7. Neutron transport by TRIPOLI-2 code in the lower part of a PWR pit and in the pit-access cell

    International Nuclear Information System (INIS)

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

    1984-09-01

    The neutrons, which exit the reactor vessel, leak between the reactor vessel and the primary concrete shield and provide high dose rate in the lower part of the reactor pit. In this part of the reactor the biological shield is reduced at the level of the pit-access cell door and of the ventilation duct. Consequently there is a pin-point increase of dose rate at reactor place where access must be free during power operation. Studies concern 900 MWe and 1300 MWe plants; calculation results are compared with dose rate measurements which have been done at several interesting points of reactor. Neutron transport is studied in two steps: - first Monte Carlo calculation carried out by the TRIPOLI-2 system gives the current of neutrons entering into the pit-access cell; - a second calculation is about neutron transport in this cell; it is also performed with the TRIPOLI-2 code which uses a very realistic model for cell geometry. Then the door efficiency is evaluated by a SN one dimension code

  8. NE213/BC501A scintillator−lightguide assembly response to 241Am−Be neutrons: An MCNPX−PHOTRACK hybrid code simulation

    International Nuclear Information System (INIS)

    Tajik, M.; Ghal-Eh, N.; Etaati, G.R.; Afarideh, H.

    2014-01-01

    The response of an NE213 (or its BICRON equivalent, BC501A) scintillator attached to different sizes of polished/painted lightguides when exposed to 241 Am–Be neutrons has been simulated. This kind of simulation basically needs both particle and light transports: the transport of neutrons and neutron-induced charged particles such as alphas, protons, carbon nuclei and so on has been undertaken using MCNPX whilst the scintillation light transport has been performed with PHOTRACK codes. The comparison between simulated and experimental response functions of NE213 attached to different sizes of polished/painted lightguides and also the influence of length/covering of lightguide on the detection efficiency and uniformity of the scintillator–lightguide assembly response have been studied. - Highlights: • The response of NE213 scintillator with/without lightguides to Am–Be neutrons has been simulated. • The MCNPX–PHOTRACK code has been used for simulation studies in order to model radio-optical properties. • The measured and simulated spectra for an NE213 scintillator exposed to Am–Be source represent a good agreement

  9. Auto MOC-A 2D neutron transport code for arbitrary geometry based on the method of characteristics and customization of AutoCAD

    Energy Technology Data Exchange (ETDEWEB)

    Chen Qichang; Wu Hongchun [School of Nuclear Science and Technology, Xi' an Jiaotong University, Xi' an Shaanxi 710049 (China); Cao Liangzhi [School of Nuclear Science and Technology, Xi' an Jiaotong University, Xi' an Shaanxi 710049 (China)], E-mail: caolz@mail.xjtu.edu.cn

    2008-10-15

    A new 2D neutron transport code AutoMOC for arbitrary geometry has been developed. This code is based on the method of characteristics (MOCs) and the customization of AutoCAD. The MOC solves the neutron transport equation along characteristic lines. It is independent of the geometric shape of boundaries and regions. So theoretically, this method can be used to solve the neutron transport equation in highly complex geometries. However, it is important to describe the geometry and calculate intersection points of each characteristic line with every boundary and region in advance. In complex geometries, due to the complications of treating the arbitrary domain, the selection of geometric shapes and efficiency of ray tracing are generally limited. The geometry treatment through the customization of AutoCAD, a widely used computer-aided design software package, is given in this paper. Thanks to the powerful capability of AutoCAD, the description of arbitrary geometry becomes quite convenient. Moreover, with the language Visual Basic for Applications (VBAs), AutoCAD can be customized to carry out the ray tracing procedure with a high flexibility in geometry. The numerical results show that AutoMOC can solve 2D neutron transport problems in a complex geometry accurately and effectively.

  10. The influence of the number of activation detectors on the Seibersdorf - Milano intercomparison of neutron flux density spectra by WINDOWS code

    International Nuclear Information System (INIS)

    Ertek, C.

    1981-02-01

    This work is a continuation of the work performed within the IAEA programme on standardization of reactor radiation measurements, one of the important objectives of which is the assistance to laboratories in Member States to implement or intercompare the multiple foil activation techniques for different neutron field measurements. The importance of these techniques is well recognized. In CESNEF-FERMI Politecnico di Milano, Italy, they have installed near the core of a water boiler of 50kW, a neutron filter made of B 4 C in order to obtain a neutron flux density spectrum that could be of utility in intercalibration problems connected with irradiation in fast assemblies. Dr. V. Sangiust from CESNEF kindly sent the input guess neutron flux density spectrum and a series of measured reaction rates to be treated by the IAEA Seibersdorf laboratory using the SAND-II and the WINDOWS unfolding codes. The meaningful comparison using partly the same ENDF/B IV cross section data is performed. In the present work we extended the investiga tion using WINDOWS unfolding code for different numbers of activation fo ils or reaction rates

  11. Neutronic study of nuclear reactors. Complete calculation of TRIGA MARKII reactor and calculations of fuel temperature coefficients. (Qualification of WIMS code)

    International Nuclear Information System (INIS)

    Benmansour, L.

    1992-01-01

    The present work shows a group of results, obtained by a neutronic study, concerning the TRIGA MARK II reactor and LIGHT WATER reactors. These studies aim to make cell and diffusion calculations. WIMS D-4 with extended library and DIXY programs are used and tested for those purposes. We also have proceeded to a qualification of WIMS code based on the fuel temperature coefficient calculations. 33 refs.; 23 figs.; 30 tabs. (author)

  12. Evaluation of the Hologic gen-probe PANTHER, APTIMA Combo 2 assay in a tertiary care teaching hospital.

    Science.gov (United States)

    Cheng, Annie; Kirby, James E

    2014-03-01

    To evaluate the performance of the Hologic Gen-Probe (San Diego, CA) PANTHER system. The performance of PANTHER was compared with the Hologic Gen-Probe TIGRIS and/or Roche (Indianapolis, IN) COBAS AMPLICOR systems through testing of patient specimens and the spiked-urine matrix. After discrepant resolution, PANTHER demonstrated a 99.3% (95% confidence interval [CI], 96.0%-99.9%) positive and 100% (98.5%-100.0%) negative agreement for Chlamydia trachomatis (CT) and 100% (96.6%-100.0%) positive and 100% (98.6%-100.0%) negative agreement for Neisseria gonorrhoeae (NG) for all male, female, unsexed, and NG-spiked female urine specimens combined. For other specimen types collectively, the PANTHER demonstrated 100% (95% CI, 90.6%-100.0%) positive and 100% (88.3%-100.0%) negative agreement for CT and 90.9% (62.8%-98.4%) positive and 100% (93.5%-100.0%) negative agreement for NG. Analytical sensitivity of the PANTHER in urine matrix was similar to the TIGRIS system. The PANTHER system provides an excellent new addition to options for detecting CT and NG, is appropriate for testing urine samples, and will facilitate high-throughput testing in the clinical laboratory.

  13. Application of the Synthesis method to the calculations of neutron flow in 3D in the enveloping of a BWR reactor with the DORT code

    International Nuclear Information System (INIS)

    Xolocostli M, J.V.; Gomez T, A.M.; Palacios H, J.C.

    2006-01-01

    The surveillance program of the vessel materials of a BWR reactor requires the determination of the neutron flux in 3D in the core enveloping. To carry out these calculations of the neutron flux, the Regulatory Guide 1.190 of the NRC recommends the use of the following codes: MCNP, TORT and DORT. In the case of using the DORT code, the one which solves the transport equation in discreet coordinates and in two dimensions (xy, rθ, and rz), the regulatory guide in reference, requires to make an approach of the flow in three dimensions by means of the call Synthesis Method. It is denominated like this due to that a flow representation in 3D is achieved 'combining' or 'synthesizing' the calculated flows by DORT in rθ, rz and r. In this work the application of the Synthesis Method it is presented, according to the Regulatory Guide 1.190, to determine the 3D flows in a BWR reactor. To achieve the above mentioned it was implemented the Synthesis Method in a computer program developed in the ININ to which is denominated SYNTHESIS. This program applies the synthesis method, and is 'coupled' with the DORT code to determine by this way the neutronic fluxes in 3D on the enveloping of a BWR reactor. (Author)

  14. Monte Carlo calculation for the development of a BNCT neutron source (1eV-10KeV) using MCNP code.

    Science.gov (United States)

    El Moussaoui, F; El Bardouni, T; Azahra, M; Kamili, A; Boukhal, H

    2008-09-01

    Different materials have been studied in order to produce the epithermal neutron beam between 1eV and 10KeV, which are extensively used to irradiate patients with brain tumors such as GBM. For this purpose, we have studied three different neutrons moderators (H(2)O, D(2)O and BeO) and their combinations, four reflectors (Al(2)O(3), C, Bi, and Pb) and two filters (Cd and Bi). Results of calculation showed that the best obtained assembly configuration corresponds to the combination of the three moderators H(2)O, BeO and D(2)O jointly to Al(2)O(3) reflector and two filter Cd+Bi optimize the spectrum of the epithermal neutron at 72%, and minimize the thermal neutron to 4% and thus it can be used to treat the deep tumor brain. The calculations have been performed by means of the Monte Carlo N (particle code MCNP 5C). Our results strongly encourage further studying of irradiation of the head with epithermal neutron fields.

  15. Condensation and homogenization of cross sections for the deterministic transport codes with Monte Carlo method: Application to the GEN IV fast neutron reactors

    International Nuclear Information System (INIS)

    Cai, Li

    2014-01-01

    In the framework of the Generation IV reactors neutronic research, new core calculation tools are implemented in the code system APOLLO3 for the deterministic part. These calculation methods are based on the discretization concept of nuclear energy data (named multi-group and are generally produced by deterministic codes) and should be validated and qualified with respect to some Monte-Carlo reference calculations. This thesis aims to develop an alternative technique of producing multi-group nuclear properties by a Monte-Carlo code (TRIPOLI-4). At first, after having tested the existing homogenization and condensation functionalities with better precision obtained nowadays, some inconsistencies are revealed. Several new multi-group parameters estimators are developed and validated for TRIPOLI-4 code with the aid of itself, since it has the possibility to use the multi-group constants in a core calculation. Secondly, the scattering anisotropy effect which is necessary for handling neutron leakage case is studied. A correction technique concerning the diagonal line of the first order moment of the scattering matrix is proposed. This is named the IGSC technique and is based on the usage of an approximate current which is introduced by Todorova. An improvement of this IGSC technique is then presented for the geometries which hold an important heterogeneity property. This improvement uses a more accurate current quantity which is the projection on the abscissa X. The later current can represent the real situation better but is limited to 1D geometries. Finally, a B1 leakage model is implemented in the TRIPOLI-4 code for generating multi-group cross sections with a fundamental mode based critical spectrum. This leakage model is analyzed and validated rigorously by the comparison with other codes: Serpent and ECCO, as well as an analytical case.The whole development work introduced in TRIPOLI-4 code allows producing multi-group constants which can then be used in the core

  16. CUTANEOUS SQUAMOUS CELL CARCINOMA IN A PANTHER CHAMELEON (FURCIFER PARDALIS) AND TREATMENT WITH CARBOPLATIN IMPLANTABLE BEADS.

    Science.gov (United States)

    Johnson, James G; Naples, Lisa M; Chu, Caroline; Kinsel, Michael J; Flower, Jennifer E; Van Bonn, William G

    2016-09-01

    A 3-yr-old male panther chameleon (Furcifer pardalis) presented with bilateral raised crusted skin lesions along the lateral body wall that were found to be carcinoma in situ and squamous cell carcinoma. Similar lesions later developed on the caudal body wall and tail. A subcutaneous implantable carboplatin bead was placed in the first squamous cell carcinoma lesion identified. Additional new lesions sampled were also found to be squamous cell carcinomas, and viral polymerase chain reaction was negative for papillomaviruses and herpesviruses. Significant skin loss would have resulted from excision of all the lesions, so treatment with only carboplatin beads was used. No adverse effects were observed. Lesions not excised that were treated with beads decreased in size. This is the first description of cutaneous squamous cell carcinoma and treatment with carboplatin implantable beads in a panther chameleon.

  17. A new open-source pin power reconstruction capability in DRAGON5 and DONJON5 neutronic codes

    Energy Technology Data Exchange (ETDEWEB)

    Chambon, R., E-mail: richard-pierre.chambon@polymtl.ca; Hébert, A., E-mail: alain.hebert@polymtl.ca

    2015-08-15

    In order to better optimize the fuel energy efficiency in PWRs, the burnup distribution has to be known as accurately as possible, ideally in each pin. However, this level of detail is lost when core calculations are performed with homogenized cross-sections. The pin power reconstruction (PPR) method can be used to get back those levels of details as accurately as possible in a small additional computing time frame compared to classical core calculations. Such a de-homogenization technique for core calculations using arbitrarily homogenized fuel assembly geometries was presented originally by Fliscounakis et al. In our work, the same methodology was implemented in the open-source neutronic codes DRAGON5 and DONJON5. The new type of Selengut homogenization, called macro-calculation water gap, also proposed by Fliscounakis et al. was implemented. Some important details on the methodology were emphasized in order to get precise results. Validation tests were performed on 12 configurations of 3×3 clusters where simulations in transport theory and in diffusion theory followed by pin-power reconstruction were compared. The results shows that the pin power reconstruction and the Selengut macro-calculation water gap methods were correctly implemented. The accuracy of the simulations depends on the SPH method and on the homogenization geometry choices. Results show that the heterogeneous homogenization is highly recommended. SPH techniques were investigated with flux-volume and Selengut normalization, but the former leads to inaccurate results. Even though the new Selengut macro-calculation water gap method gives promising results regarding flux continuity at assembly interfaces, the classical Selengut approach is more reliable in terms of maximum and average errors in the whole range of configurations.

  18. PANTHER-PSEP: predicting disease-causing genetic variants using position-specific evolutionary preservation.

    Science.gov (United States)

    Tang, Haiming; Thomas, Paul D

    2016-07-15

    PANTHER-PSEP is a new software tool for predicting non-synonymous genetic variants that may play a causal role in human disease. Several previous variant pathogenicity prediction methods have been proposed that quantify evolutionary conservation among homologous proteins from different organisms. PANTHER-PSEP employs a related but distinct metric based on 'evolutionary preservation': homologous proteins are used to reconstruct the likely sequences of ancestral proteins at nodes in a phylogenetic tree, and the history of each amino acid can be traced back in time from its current state to estimate how long that state has been preserved in its ancestors. Here, we describe the PSEP tool, and assess its performance on standard benchmarks for distinguishing disease-associated from neutral variation in humans. On these benchmarks, PSEP outperforms not only previous tools that utilize evolutionary conservation, but also several highly used tools that include multiple other sources of information as well. For predicting pathogenic human variants, the trace back of course starts with a human 'reference' protein sequence, but the PSEP tool can also be applied to predicting deleterious or pathogenic variants in reference proteins from any of the ∼100 other species in the PANTHER database. PANTHER-PSEP is freely available on the web at http://pantherdb.org/tools/csnpScoreForm.jsp Users can also download the command-line based tool at ftp://ftp.pantherdb.org/cSNP_analysis/PSEP/ CONTACT: pdthomas@usc.edu Supplementary data are available at Bioinformatics online. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please e-mail: journals.permissions@oup.com.

  19. PANTHERE, simulation software of dose flow rates for complex nuclear installations

    International Nuclear Information System (INIS)

    Longeot, M.; Dupont, B.; Coatanea, C.; Schumm, S.; Zweers, M.; Malvagi, F.; Trama, J.C.

    2010-01-01

    The authors present the PANTHERE simulation software developed by EDF-SEPTEN to determine gamma dose flow rate in any point of complex industrial installations. They present the current industrial version (PANTHEREV1) and its different applications, and more particularly an investigation in the field of qualification of hardware under irradiation in case of severe accident. They present the currently under development version (PANTHEREV2) which will be exploited in 2011

  20. Effects of acidity on primary productivity in lakes: phytoplankton. [Lakes Panther, Sagamore, and Woods

    Energy Technology Data Exchange (ETDEWEB)

    Hendrey, G R

    1979-01-01

    Relationships between phytoplankton communities and lake acidity are being studied at Woods Lake (pH ca. 4.9), Sagamore Lake (pH ca. 5.5), and Panther Lake (pH ca. 7.0). Numbers of phytoplankton species observed as of July 31, 1979 are Woods 27, Sagamore 38, and Panther 64, conforming to other observations that species numbers decrease with increasing acidity. Patterns of increasing biomass and productivity found in Woods Lake may be atypical of similar oligotrophic lakes in that they develop rather slowly instead of occuring very close to ice-out. Contributions of netplankton (net > 48 ..mu..m), nannoplankton (48 > nanno > 20 ..mu..m) and ultraplankton (20 > ultra >0.45 ..mu..m) to productivity per m/sup -2/ show that the smaller plankton are relatively more important in the more acid lakes. This pattern could be determined by nutrient availability (lake acidification leading to decreased availability of phosphorus). The amount of /sup 14/C-labelled dissolved photosynthate (/sup 14/C-DOM), as a percent of total productivity, is ordered Woods > Sagamore > Panther. This is consistent with a hypothesis that microbial heterotrophic activity is reduced with increasing acidity, but the smaller phytoplankton may be more leaky at low pH. (ERB)