ORIGNATE: PC input processor for ORIGEN-S

International Nuclear Information System (INIS)

Bowman, S.M.

1992-01-01

ORIGNATE is a personal computer program that serves as a user- friendly interface for the ORIGEN-S isotopic generation and depletion code. It is designed to assist an ORIGEN-S user in preparing an input file for execution of light-water-reactor fuel depletion and decay cases. Output from ORIGNATE is a card-image input file that may be uploaded to a mainframe computer to execute ORIGEN-S in SCALE-4. ORIGNATE features a pulldown menu system that accesses sophisticated data entry screens. The program allows the user to quickly set up an ORIGEN-S input file and perform error checking

An ORIGEN2 update for PCs and mainframes

International Nuclear Information System (INIS)

Ludwig, S.B.

1991-01-01

The ORIGEN2 computer code was developed by Oak Ridge National Laboratory (ORNL) in the late 1970s and made available to users worldwide in 1980 through the Radiation Shielding Information Center (RSIC). The purpose of ORIGEN2 is to calculate the buildup, decay, and processing of radioactive materials. Since 1980, more than 500 users have acquired the ORIGEN2 code from RSIC. ORIGEN2 is often the starting point for many analyses involving the shielding, criticality, safety, performance assessment, accident analysis, risk assessment, and design of reactors, transportation casks, reprocessing plants, and waste disposal facilities. An updated version of the ORIGEN2 code package has been prepared by ORNL. This new version of the ORIGEN2 package (called ORIGEN2, Version 2.1 (8-1-91)) includes revised ORIGEN2 cross-section libraries for standard- and extended-burnup pressurized-water and boiling-water reactors (PWRs and BWRs). These libraries were used in the preparation of Revision 1 of the Characteristic Data Base (CDB). The ORIGEN2 code has also been revised to include numerous maintenance fixes and to consolidate various hardware versions into a single source code capable of execution of both 80386/80486 PCs as well as most mainframe computers; these improvements are part of Version 2.1. 6 refs., 2 tabs

Analysis of Beryllium Having Irradiated at the RSG-GAS Core using ORIGEN2 Code

International Nuclear Information System (INIS)

Jaja Sukmana; Jonnie AK; S-Suwarto; Irwan

2012-01-01

Analysis of activation products generated by irradiated beryllium at the RSG-GAS core has been done using ORIGEN2 code. By assuming that irradiation is 176 days, neutron flux average of 2.30e+14 n/cm 2 s, radioisotopes rose from activated Be are tritium, lithium, beryllium, carbon, magnesium, aluminum, silicon, argon, calcium, scandium, vanadium, chromium, manganese, iron, cobalt, nickel, copper, zinc, silver, and lead. The highest activity after 100 days of irradiation demonstrated by Be-10 (7.99 E-03 Curie), H-3 (2.97 E-03 Curie), Cr-51, Fe-55 and Co-60. Radioactivity generated getting smaller when irradiation time are long. From this analysis it can be conclude that radioactivity was caused by impurities present in Be such as Mn-54, Fe-59, Zn-65, and Li-6. (author)

Originate: PC input processor for origen-S

International Nuclear Information System (INIS)

Bowman, S.M.

1994-01-01

ORIGINATE is a personal computer program developed at Oak Ridge National Laboratory to serve as a user-friendly interface for the ORIGEN-S isotopic generation and depletion code. It is designed to assist an ORIGEN-S user in preparing an input file for execution of light-water-reactor fuel depletion and decay cases. Output from ORIGINATE is a card-image input file that may be uploaded to a mainframe computer to execute ORIGEN-S in SCALE-4. ORIGINATE features a pull down menu system that accesses sophisticated data entry screens. The program allows the user to quickly set up an ORIGEN-S input file and perform error checking. This capability increases productivity and decreases chance of user error. (authors). 6 refs., 3 tabs

ARP: A PC-compatible scheme for generating ORIGEN-S cross section library

International Nuclear Information System (INIS)

Leal, L.C.; Hermann, O.W.; Parks, C.V.

1995-01-01

The SAS2H sequence of the SCALE code system has been widely used for treating problems related to the characterization of nuclear systems for disposal, storage, and shipment. The calculations, in general, consist of determining the isotope compositions of the different materials present in the problem as a function of time, which subsequently enable determination of the heat generation and radiation source terms. In the SAS2H scheme, time-dependent material concentrations are obtained using the ORIGEN-S code based on a point-depletion calculation that utilizes problem-dependent cross-section libraries generated by distinct codes of the SAS2H sequence. In this paper we will be concerned with the methodology utilized in the SAS2H control module to create cross-section libraries for point-depletion calculations with the ORIGEN-S code. A brief description of the SAS2H scheme will be given, and a new capability, the automatic rapid processing (ARP), for generating problem-dependent ORIGEN-S cross-section libraries will be presented. Use of ARP can enable execution of ORIGEN-S on a personal computer with identical accuracy to that obtained with SAS2H

Uncertainties in source term calculations generated by the ORIGEN2 computer code for Hanford Production Reactors

International Nuclear Information System (INIS)

Heeb, C.M.

1991-03-01

The ORIGEN2 computer code is the primary calculational tool for computing isotopic source terms for the Hanford Environmental Dose Reconstruction (HEDR) Project. The ORIGEN2 code computes the amounts of radionuclides that are created or remain in spent nuclear fuel after neutron irradiation and radioactive decay have occurred as a result of nuclear reactor operation. ORIGEN2 was chosen as the primary code for these calculations because it is widely used and accepted by the nuclear industry, both in the United States and the rest of the world. Its comprehensive library of over 1,600 nuclides includes any possible isotope of interest to the HEDR Project. It is important to evaluate the uncertainties expected from use of ORIGEN2 in the HEDR Project because these uncertainties may have a pivotal impact on the final accuracy and credibility of the results of the project. There are three primary sources of uncertainty in an ORIGEN2 calculation: basic nuclear data uncertainty in neutron cross sections, radioactive decay constants, energy per fission, and fission product yields; calculational uncertainty due to input data; and code uncertainties (i.e., numerical approximations, and neutron spectrum-averaged cross-section values from the code library). 15 refs., 5 figs., 5 tabs

Requests from use experience of ORIGEN code. Activity of the working group on evaluation of nuclide generation and depletion

International Nuclear Information System (INIS)

Matsumura, Tetsuo

2005-01-01

A questionnaire survey was carried out through the committee members of the working group on evaluation of nuclide generation and depletion about the demand accuracy of the ORIGEN code which is used widely in various fields of design analysis and evaluation. WG committee asked each organization's ORIGEN user, and obtained the replay from various fields. (author)

The quiescent state of the neutron-star X-ray transient GRS 1747-312 in the globular cluster Terzan 6

Science.gov (United States)

Vats, Smriti; Wijnands, Rudy; Parikh, Aastha S.; Ootes, Laura; Degenaar, Nathalie; Page, Dany

2018-04-01

We studied the transient neutron-star low-mass X-ray binary GRS 1747-312, located in the globular cluster Terzan 6, in its quiescent state after its outburst in August 2004, using an archival XMM-Newton observation. A source was detected in this cluster and its X-ray spectrum can be fitted with the combination of a soft, neutron-star atmosphere model and a hard, power-law model. Both contributed roughly equally to the observed 0.5-10 keV luminosity (˜4.8 × 1033 erg s-1). This type of X-ray spectrum is typically observed for quiescent neutron-star X-ray transients that are perhaps accreting in quiescence at very low rates. Therefore, if this X-ray source is the quiescent counterpart of GRS 1747-312, then this source is also accreting at low levels in-between outbursts. Since source confusion a likely problem in globular clusters, it is quite possible that part, if not all, of the emission we observed is not related to GRS 1747-312, and is instead associated with another source or conglomeration of sources in the cluster. Currently, it is not possible to determine exactly which part of the emission truly originates from GRS1747-312, and a Chandra observation (when no source is in outburst in Terzan 6) is needed to be conclusive. Assuming that the detected emission is due to GRS 1747-312, we discuss the observed results in the context of what is known about other quiescent systems. We also investigated the thermal evolution of the neutron-star in GRS 1747-312, and inferred that GRS 1747-312 can be considered a typical quiescent system under our assumptions.

The quiescent state of the neutron-star X-ray transient GRS 1747-312 in the globular cluster Terzan 6

Science.gov (United States)

Vats, Smriti; Wijnands, Rudy; Parikh, Aastha S.; Ootes, Laura; Degenaar, Nathalie; Page, Dany

2018-06-01

We studied the transient neutron-star low-mass X-ray binary GRS 1747-312, located in the globular cluster Terzan 6, in its quiescent state after its outburst in August 2004, using an archival XMM-Newton observation. A source was detected in this cluster and its X-ray spectrum can be fitted with the combination of a soft, neutron-star atmosphere model and a hard, power-law model. Both contributed roughly equally to the observed 0.5-10 keV luminosity (˜4.8 × 1033 erg s-1). This type of X-ray spectrum is typically observed for quiescent neutron-star X-ray transients that are perhaps accreting in quiescence at very low rates. Therefore, if this X-ray source is the quiescent counterpart of GRS 1747-312, then this source is also accreting at low levels in-between outbursts. Since source confusion is a likely problem in globular clusters, it is quite possible that part, if not all, of the emission we observed is not related to GRS 1747-312, and is instead associated with another source or conglomeration of sources in the cluster. Currently, it is not possible to determine exactly which part of the emission truly originates from GRS 1747-312, and a Chandra observation (when no source is in outburst in Terzan 6) is needed to be conclusive. Assuming that the detected emission is due to GRS 1747-312, we discuss the observed results in the context of what is known about other quiescent systems. We also investigated the thermal evolution of the neutron star in GRS 1747-312, and inferred that GRS 1747-312 can be considered a typical quiescent system under our assumptions.

GB - a preliminary linking code between MCNP4C and Origen2.1 - DEN/UFMG version

International Nuclear Information System (INIS)

Campolina, Daniel; Pereira, Claubia; Veloso, Maria Auxiliadora Fortini; Cavatoni, Andre

2009-01-01

Nowadays it is possible to perform burnup simulation in a detailed 3D geometry and a continuous energy description by the Monte Carlo method. This paper describes an initial project to create and verify a connection code to link Origen2.1 (Oak Ridge National Laboratory) and MCNP4C (Los Alamos National Laboratory). Essentially the code includes point depletion capability to the MCNP code. The incorporation of point depletion capability is explicit and can be summarized by three steps: 1-Monte Carlo determines reaction rates, 2-the reaction rates are used to determine microscopic cross sections for depletion equations, 3-solution of depletion equations (given by Origen2.1) determines number densities for next MCNP step. To evaluate the initial version of the program, we focused on comparing the results with one of the major Monte Carlo burnup codes: MCNPX version 2.6.0. The input files for all codes share the same MCNP geometry, nuclear data library and core thermal power. While simulating 75 time steps at 800 kw of a Heat Pipe Power System model, we have found that the codes generate very similar results. The neutron flux and criticality value of the core agree, especially in the begin of burnup when the influence of fission products are not very considerable. The small difference encountered was probably caused by the difference in the number of isotopes considered in the transport models (89 MCNPX x 25 GB (author)

Development of a friendly interface for ORIGEN 2.1 code using MatLab software

International Nuclear Information System (INIS)

Vieira, Joao Paulo

2011-01-01

In an event of accidental liberation of radioactive material to the environment from a nuclear power plant, decisions must be taken quickly to supply the need of mitigating actions. Thus, it is important a fast, clear and safe access to all information about the source term. This work describes the initiative to develop a graphic interface to output data for ORIGEN 2.1 code, intending a friendly and secure approach with the output data and other important parameters for an analysis in emergency case, known the historic of operation of a nuclear power plant type PWR. By using the software MATLAB it is possible to develop an output routine with graphic presentation to some necessary data for an emergency analysis. The interface output must be able of fix up the ORIGEN conventional tables in graphics. In advance, preliminary results will be presented. (author)

Simulation approaches for the two-phase flow in saline repositories using the code TOUGH2-GRS. Report in the frame of the project ZIESEL. Two-phase flow in a saline repository using the example ERAM; Ansaetze zur Simulation der Zweiphasenstroemung in salinaren Endlagern mit dem Code TOUGH2-GRS. Bericht im Vorhaben ZIESEL. Zweiphasenfluss in einem salinaren Endlager am Beispiel des ERAM

Energy Technology Data Exchange (ETDEWEB)

Navarro, Martin; Fischer, Heidemarie; Seher, Holger; Weyand, Torben

2016-10-15

The simulation approaches for the two-phase flow in saline repositories using the code TOUGH2-GRS cover the following issues: simulation of gravitational flows in horizontal galleries without vertical discretization, homogenization approach for the simulation of the two-phase flow in converging partly backfilled galleries, qualification of the convergence approach implemented by GRS into the code TOUGH2-GRS, discretization effects during replacement of liquid by gas, consequences for the system analyses in the frame of the project ZIESEL.

2008 GRS specialized forum

International Nuclear Information System (INIS)

Butz, H.P.

2008-01-01

After the successful trial run in 2007, the 2008 GRS Specialized Forum held at the Cologne Wolkenburg on April 7 and 8, 2008 was a full success. GRS, Gesellschaft fuer Anlagen- und Reaktorsicherheit, plans to resume in this way its GRS specialized lectures series with a tradition of many years. Approximately 180 participants from public authorities, licensees, the nuclear industry, research and engineering establishments as well as universities showed their interest in current GRS activities. On 2 days, only GRS experts presented fundamental and GRS-specific approaches in scientific analysis, new research findings, and possibilities of advancing scientific methods under the four headings of Reactor Safety Research and Evaluation, Topical Subjects of Reactor Safety, Radiation Protection Activities, and Repository Safety and Environmental Research. The discussions held after each presentation offered a number of critical, but valuable and welcome, contributions from the different perspectives of the discussants. (orig.)

SANDIA-ORIGEN user's manual

International Nuclear Information System (INIS)

Bennett, D.E.

1979-10-01

The SANDIA-ORIGEN code calculates the detailed isotopic composition as a function of time in nuclear reactor fuel irradiation and radioactive decay problems. This code was developed specifically for Control Data Corporation computers from the original Oak Ridge National Laboratory ORIGEN code. The nuclear data file used with the code at Sandia Laboratories contains 1063 isotopes (254 structural materials, 101 actinides, and 708 fission products). SANDIA-ORIGEN is oriented toward simple, easy use and includes NAMELIST input, convenient control of the output, and versatile options for the blending and reprocessing of reactor fuel. System operating instructions and the input decks for numerous sample problems are also presented. 13 references, 14 figures

Status Report on NEAMS PROTEUS/ORIGEN Integration

Energy Technology Data Exchange (ETDEWEB)

Wieselquist, William A [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-02-18

The US Department of Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program has contributed significantly to the development of the PROTEUS neutron transport code at Argonne National Laboratory and to the Oak Ridge Isotope Generation and Depletion Code (ORIGEN) depletion/decay code at Oak Ridge National Laboratory. PROTEUS’s key capability is the efficient and scalable (up to hundreds of thousands of cores) neutron transport solver on general, unstructured, three-dimensional finite-element-type meshes. The scalability and mesh generality enable the transfer of neutron and power distributions to other codes in the NEAMS toolkit for advanced multiphysics analysis. Recently, ORIGEN has received considerable modernization to provide the high-performance depletion/decay capability within the NEAMS toolkit. This work presents a description of the initial integration of ORIGEN in PROTEUS, mainly performed during FY 2015, with minor updates in FY 2016.

X-Ray Transient in Musca (GRS 1121-68 = GS 1124-683)

DEFF Research Database (Denmark)

Lund, Niels; Brandt, Søren; Makino, F.

1991-01-01

Research Institute, Lyngby; and Space Research Institute, Moscow), report: "A new x-ray source in Musca, GRS 1121-68, has been discovered in data obtained by the WATCH detectors on Jan. 9. The spectrum of the source was harder than that of the Crab nebula, and the flux was about twice that from t...... was taken as La Silla, showing no new object in the 2 x 2 deg field brighter than mag 7-8; a search plate will be taken with the Schmidt telescope at La Silla....

Automated procedure for calculating time-dependent sensitivities in ORIGEN2

International Nuclear Information System (INIS)

Worley, B.A.; Wright, R.Q.

1985-10-01

ORIGEN2 is a widely used point-depletion and radioactive-decay computer code for use in simulating nuclear fuel cycles and/or spent fuel characteristics. This paper presents the application of the GRESS procedure to the ORIGEN2 code for performing a sensitivity analysis of a high-level waste disposal problem. The GRESS procedure uses computer calculus and the GRESS precompiler to automate the generation and calculation of gradients in a computer code. The GRESS version of ORIGEN2 is used to calculate the nuclide-dependent sensitivities of the decay heat and radioactivity of 1008 nuclides comprising reprocessed high-level waste to changes in data and input parameters. The sensitivities are calculated in a single execution of the revised code as compared to the conventional method of rerunning the code numerous times. The availability of sensitivity data as an option in ORIGEN2 reveals relationships not easily recognized even with reruns

ZZ CANDULIB-AECL, Burnup-Dependent ORIGEN-S Cross-Section Libraries for Candu Reactor Fuels

International Nuclear Information System (INIS)

2002-01-01

1 - Historical background and information: - 28-element fuel cross-section library: Format: Designed for use with the ORIGEN-S isotope generation and depletion code. Materials: Co, Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, Nb, Mo, Tc, Ru, Rh, Pd, Ag, Cd, In, Sn, Sb, Te, I, Xe, Cs, Ba, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy, Ho, Er, Lu, Ta, W, Re, Au, Th, Pa, U, Np, Pu, Am, Cm. Origin: ENDSF, ENDF/B-IV, -V and -VI Weighting spectrum: determined using WIMS-AECL transport code. - 37-element fuel cross-section library: Format: Designed for use with the ORIGEN-S isotope generation and depletion code. Materials: Co, Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, Nb, Mo, Tc, Ru, Rh, Pd, Ag, Cd, In, Sn, Sb, Te, I, Xe, Cs, Ba, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy, Ho, Er, Lu, Ta, W, Re, Au, Th, Pa, U, Np, Pu, Am, Cm. Origin: ENDSF, ENDF/B-IV, -V and -VI Weighting spectrum: determined using WIMS-AECL transport code. In 1995, updated ORIGEN-S cross-section libraries were created as part of a program to upgrade and standardize the computer codes and nuclear data employed for used fuel characterization. This effort was funded through collaboration between Atomic Energy of Canada Limited and the Canadian Nuclear Power Utilities, under the Candu Owners Group (COG). The updated cross sections were generated using the WIMS-AECL lattice code and ENDF/B-V and -VI based data to provide cross section consistency with reactor physics codes. 2 - Application of the data: The libraries in this data collection are designed for characterising used fuel from Candu pressurized heavy water reactors. Two libraries are provided: one for the standard 28-element fuel bundle design, the other for the 37-element fuel bundle design. The libraries were generated for typical reactor operating conditions. The libraries are designed for use with the ORIGEN-S isotope generation and depletion code. 3 - Source and scope of data: The Candu libraries are updated with cross sections from a variety of different sources. Capture

OREST, LWR Burnup Simulation Using Program HAMMER and ORIGEN

International Nuclear Information System (INIS)

Hesse, Ulrich; Sieberer, Johann

2006-01-01

1 - Description of program or function: In OREST, the 1-dimensional lattice code HAMMER and the isotope generation and depletion code ORIGEN are directly coupled for burnup simulation in light-water reactor fuels (GRS recommended). Additionally heavy water and graphite moderated systems can be calculated. New version differs from the previous version in the following features: An 84-group-library LIB84 for up to 200 isotopes is used to update the 3-group -POISON-XS. LIB84 uses the same energy boundaries as THERMOS and HAMLET in . In this way, high flexibility is achieved in very different reactor models. The coupling factor between THERMOS and HAMLET is now directly transferred from HAMMER to THERES and omits the equation 4 (see page 6 of the manual). Sandwich-reactor fuel reactivity and burnup calculations can be started with NGEOM = 1. Thorium graphite reactivity and burnup calculations can be started with NLIBE = 1. High enriched U-235 heavy water moderated reactivity and burnup calculations can be started. HAMLET libraries in for U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-242, Am-241, Am-243 and Zirconium are updated using resonance parameters. NEA-1324/04: A new version of the module hamme97.f has replaced the old one. 2 - Method of solution: For the user-defined irradiation history, an input data processor generates program loops over small burnup steps for the main codes HAMMER and ORIGEN. The user defined assembly description is transformed to an equivalent HAMMER fuel cell. HAMMER solves the integral neutron transport equation in a four-region cylindrical or sandwiched model with reflecting boundaries and runs with fuel power calculated rod temperatures. ORIGEN runs with HAMMER-calculated cross sections and neutron spectra and calculates isotope concentrations during burnup by solving the buildup-, depletion- and decay-chain equations. An output data processor samples the outputs of the program modules and generates tabular works for the

Development of libraries for ORIGEN2 code based on JENDL-3.2

Energy Technology Data Exchange (ETDEWEB)

Suyama, Kenya; Katakura, Jun-ichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ishikawa, Makoto; Ohkawachi, Yasushi

1998-03-01

The working Group of JNDC `Nuclide Generation Evaluation` has launched a project to make libraries for ORIGEN2 code based on the latest nuclear data library `JENDL-3.2` for current design of LWR and FBR fuels. Many of these libraries are under validation. (author)

Enhancing co-operation between AVN, IRSN and GRS: the junior staff pilot project on the comparative testing of IPA codes

International Nuclear Information System (INIS)

Hoyos, A. de; Keesmann, S.; Smidts, O.

2006-01-01

- Objectives: The project takes place within the framework of the Junior Staff Program of AVN, GRS and IRSN which aims at creating a junior staff network among European TSOs. The objective of this project is to apply integrated performance assessment (IPA) tools used by AVN, IRSN and GRS to two generic and simplified models (Bure site in France and Mol site in Belgium) for disposal systems in argillaceous formations. The comparison of the results from different codes applied to the disposal systems of the two mentioned sites aims at a better understanding of the confinement capabilities of the considered geological formations and of the IPA methodology in general. The incentive is a common understanding of approaches developed by each partner and the improvement of this expertise. More specifically, this pilot project aims at enhancing exchanges of views and mutual experiences in the field of understanding major safety functions. - Tools and Methods: A new code for the assessment of barrier systems in argillaceous formations has only recently been developed at GRS, as in the past such formations played a minor role as a possible hosting environment for a repository in Germany. The project also serves as a test case for this code. The considered disposal systems are defined on the basis of the concepts and data available for Mol and Bure. The program packages used for the performance assessment calculations are: HYDRUS-1D with source term module (AVN), GoldSim (IRSN) and EMOS-modules CLAYPOS and CHET (GRS). While the coupling of HYDRUS-1D with a source term module and the EMOS-modules are FORTRAN77- coded programs specifically developed for the simulation of parts of a barrier system of a final repository, GoldSim is a general purpose simulation environment with an integrated graphical user interface for modelling and data output. Models realized in GoldSim are flexible and can be easily adapted to new requirements. The software also offers an intrinsic

Development of an interface between MCNP and ORIGEN codes for calculations of fuel evolution in nuclear systems. Initial project

International Nuclear Information System (INIS)

Campolina, Daniel de Almeida Magalhaes

2009-01-01

In Many situations of nuclear system study, it is necessary to know the detailed particle flux in a geometry. Deterministic 1-D and 2-D methods aren't suitable to represent some strong 3-D behavior configurations, for example in cores where the neutron flux varies considerably in the space and Monte Carlo analysis are necessary. The majority of Monte Carlo transport calculation codes, performs time static simulations, in terms of fuel isotopic composition. This work is a initial project to incorporate depletion capability to the MCNP code, by means of a connection with ORIGEN2.1 burnup code. The method to develop the program proposed followed the methodology of other programs used to the same purpose. Essentially, MCNP data library are used to generate one group microscopic cross sections that override default ORIGEN libraries. To verify the actual implemented part, comparisons which MCNPX (version 2.6.0) results were made. The neutron flux and criticality value of core agree. The neutron flux and criticality value of the core agree, especially in beginning of burnup when the influence of fission products are not very considerable. The small difference encountered was probably caused by the difference in the number of isotopes considered in the transport models (89 MCNPX x 25 GB). Next step of this work is to adapt MCNP version 4C to work with a memory higher than its standard value (4MB), in order to allow a greater number of isotopes in the transport model. (author)

An ORIGEN-2 update for PCs and mainframes

International Nuclear Information System (INIS)

Ludwig, S.B.

1992-01-01

This paper reports that an updated version of the ORIGEN2 code package has been prepared by Oak Ridge National Laboratory. ORIGEN2 is used extensively by the DOE office of Civilian Radioactive Waste Management (OCRWM) and its contractors to determine the characteristics of spent fuel and high-level radioactive waste due to irradiation, decay, and processing. Included in this update are revised ORIGEN2 cross-section libraries for standard- and extended-burnup PWRs and BWRs. This update also includes improvements to the ORIGEN2 computer code (designated as ORIGEN2, Version 2.1 8-1-91 release). This version of ORIGEN2 provides a single source code that may be executed on both mainframes and 80386 or 80486 PCs, effectively smashing the 640 KB barrier that limited previous PC implementations

Analysis of effects of updated decay and fission yield data on ORIGEN 2 results

International Nuclear Information System (INIS)

Daniel, P.R.

1993-01-01

Work has been performed to improve the accuracy of ORIGEN2 results by updating both the decay library and the fission yield data in the cross-section library. This effort was performed under the auspices of Oak Ridge National Laboratory (ORNL) to ensure that ORIGEN2 uses the most up-to-date data. The impact of the new data was then quantitatively evaluated by solving a set of standard light water reactor (LWR) problems solved with ORIGEN2. The ORIGEN code, developed at ORNL in the late 1960's, is a point depletion code used to determine the composition and characteristics of spent fuel. The results from calculations performed with the code often form the basis for the study and design of reprocessing plants, spent-fuel shipping casks, waste treatment systems, and disposal facilities. The decay data were updated using data from ENDF/B-VI; fission yield data were updated using data from ENDF/B-V. The impact of these new data was then evaluated

The outburst duration and duty cycle of GRS1915+105

Science.gov (United States)

Deegan, Patrick; Combet, Céline; Wynn, Graham A.

2009-12-01

The extraordinarily long outburst of GRS1915+105 makes it one of the most remarkable low-mass X-ray binaries (LMXBs). It has been in a state of constant outburst since its discovery in 1992, an eruption which has persisted ~100 times longer than those of more typical LXMBs. The long orbital period of GRS1915+105 implies that it contains large and massive accretion disc which is able to fuel its extreme outburst. In this paper, we address the longevity of the outburst and quiescence phases of GRS1915+105 using smooth particle hydrodynamics (SPH) simulations of its accretion disc through many outburst cycles. Our model is set in the two-α framework and includes the effects of the thermoviscous instability, tidal torques, irradiation by central X-rays and wind mass loss. We explore the model parameter space and examine the impact of the various ingredients. We predict that the outburst of GRS1915+105 should last a minimum of 20yr and possibly up to ~100yr if X-ray irradiation is very significant. The predicted recurrence times are of the order of 104yr, making the X-ray duty cycle a few 0.1 per cent. Such a low duty cycle may mean that GRS1915+105 is not an anomaly among the more standard LMXBs and that many similar, but quiescent, systems could be present in the Galaxy.

ORIGEN-S data libraries

International Nuclear Information System (INIS)

Ryman, J.C.

1984-01-01

There are five card-image nuclear data libraries: (1) a small light element library for 253 nuclides, (2) a large light element library for 687 nuclides, (3) an actinide library for 101 nuclides, (4) a small fission product library for 461 nuclides, and (5) a large fission product library for 821 nuclides. The data for each nuclide are contained on five card-image records. The first card image contains decay data (half-life, branching fractions, recoverable energy per decay and the fraction of recoverable energy from photons), percent natural abundance, and radioactivity concentration guides. The last four card images contain cross section and (for fission product nuclides) fission yield data for four reactor types (HTGR, LWR, LMFBR, and MSBR), with one card for each reactor type. The card-image nuclear data libraries are the basic libraries for ORIGEN-S. The code can be run using these libraries directly, or it can be run from a binary data library which (prior to any cross section or other nuclear data updating) was created by running the COUPLE code to convert one or more of these card-image libraries

Production and validation of ORIGEN-S libraries from JEF2.2 and EAF3 data

Energy Technology Data Exchange (ETDEWEB)

Hoogenboom, J.E. [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.; Kloosterman, J.L. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands)

1995-12-01

The data libraries for light elements, actinides and fission products of the ORIGEN-S code for depletion and transmutation calculations in the SCALE4.1 computer code system have been updated with respect to cross-section data, radioactive-decay data and fission-product yield data using JEF2.2 as the basic data source and EAF3 as an additional source. This required the fission-product library to be extended with 201 new fission-product nuclides or isomeric states. The effect of the update of different quantities involved is evaluated with a burnup benchmark. When ORIGEN-S is used as a stand-alone code, i.e without regular update of cross sections of the major nuclides due to changes in the neutron spectrum during burnup, the results show appreciable differences in actinide and fission-product densities due to the cross-section update. The effects of decay data and fission-product yield updates are generally small, but with noticeable exceptions. The update of fission and capture reaction energies gives a small but systematic change in actinide and fission-product concentration. (orig.).

ORIGEN-S: scale system module to calculate fuel depletion, actinide transmutation, fission product buildup and decay, and associated radiation source terms

International Nuclear Information System (INIS)

Anon.

1982-01-01

ORIGEN-S computes time-dependent concentrations and source terms of a large number of isotopes, which are simultaneously generated or depleted through neutronic transmutation, fission, radioactive decay, input feet rates and physical or chemical removal rates. The calculations may pertain to fuel irradiation within nuclear reactors, or the storage, management, transportation or subsequent chemical processing of removed fuel elements. The matrix exponential expansion model of the ORIGIN code is unaltered in ORIGEN-S. Essentially all features of ORIGEN were retained, expanded or supplemented within new computations. The primary objective of ORIGEN-S, as requested by the Nuclear Regulatory Commission, is that the calculations may utilize the multi-energy group cross sections from any currently processed standardized ENDF/B data base. This purpose has been implemented through the prior execution of codes within either the SCALE System or the AMPX System, developed at the Oak Ridge National Laboratory. These codes compute flux-weighted cross sections, simulating conditions within any given reactor fuel assembly, and convert the data into a library that can be input to ORIGEN-S. Time-dependent libraries may be produced, reflecting fuel composition variations during irradiation. Presented in the document are: detailed and condensed input instructions, model theory, features available, range of applicability, brief subroutine descriptions, sample input, and I/O requirements. Presently the code is operable on IBM 360/370 computers and may be converted for CDC computers. ORIGEN-S is a functional module in the SCALE System and will be one of the modules invoked in the SAS2 Control Module, presently being developed, or may be applied as a stand alone program. It can be used in nuclear reactor and processing plant design studies, radiation safety analyses, and environmental assessments

Modification to ORIGEN2 for generating N Reactor source terms. Volume 1

International Nuclear Information System (INIS)

Schwarz, R.A.

1997-04-01

This report discusses work that has been done to upgrade the ORIGEN2 code cross sections to be compatible with the WIMS computer code data. Because of the changes in the ORIGEN2 calculations. Details on changes made to the ORIGEN2 computer code and the Radnuc code will be discussed along with additional work that should be done in the future to upgrade both ORIGEN2 and Radnuc. A detailed historical description of how source terms have been generated for N Reactor fuel stored in the K Basins has been generated. The neutron source discussed in this description was generated by the WIMS computer code (Gubbins et al. 1982) because of known shortcomings in the ORIGEN2 (Croff 1980) cross sections. Another document includes a discussion of the ORIGEN2 cross sections

Decay heat rates calculated using ORIGEN-S and CINDER10 with common data libraries

International Nuclear Information System (INIS)

Brady, M.C.; Hermann, O.W.; Beard, C.A.; Bohnhoff, W.J.; England, T.R.

1991-01-01

A set of two benchmark problems were proposed as part of an international comparison of decay heat codes. Problem specifications included explicit fission-yield, decay and capture data libraries to be used in the calculations. This paper describes the results obtained using these common data to perform the benchmark calculations with two popular depletion codes, ORIGEN-S and CINDER10. Short descriptions of the methods used by each of these codes are also presented. Results from other contributors to the international comparison are discussed briefly. This comparison of decay heat codes using common data libraries demonstrates that discrepant results in calculated decay heat rates are the result of differences in the nuclear data input to the codes and not the method of solution. 15 refs., 2 figs., 8 tabs

GRS annual report 1992/93

International Nuclear Information System (INIS)

Knoll, R.

1994-01-01

In the course of Europe growing together, the GRS has further intensified its international cooperation, focusing its work on the technical analysis of Russian reactor technology and on the development of safety requirements for a new generation of nuclear power plants in Europe. Another step towards closer relationship with Eastern Europe was done when the GRS together with its French partner IPSN opened an office in Moscow in 1993. That office does not only serve to establish direct contacts with Russian institutions, but also with other East European states. Inspite of all those new challenges, the continuous monitoring of nuclear facilities in operation and the evaluation of operational experience at home and abroad will remain a priority of GRS activities. (orig./HSCH) [de

Modifications to ORIGEN2 for generating N Reactor source terms. Volume 3: ORIGEN2 N-Reactor output files

International Nuclear Information System (INIS)

1997-01-01

This text is intended to be a brief outline of the ORIGEN2 computer code which is a revised and updated version of the ORIGEN documented in report ORNL-4628 (May 1973). Included here are: a brief description of the functions of ORIGEN2; a listing of the major data sources; a listing of the published documentation concerning ORIGEN2; and an outline of the ORIGEN2 output organization. ORIGEN2 is available from the ORNL Radiation Shielding Information Center (RSIC). Past experience has indicated that many users encounter considerable difficulty in finding the desired information in a ORIGEN2 output which is sometimes rather massive. This section is intended as a brief outline of the organization of ORIGEN2 output

Multi-Wavelength Monitoring of GRS 1915+105

Science.gov (United States)

Bandyopadhyay, R.; Martini, P.; Gerard, E.; Charles, P. A.; Wagner, R. M.; Shrader, C.; Shahbaz, T.; Mirabel, I. F.

1997-01-01

Since its discovery in 1992, the superluminal X-ray transient GRS 1915+105 has been extensively observed in an attempt to understand its behaviour. We present here first results from a multi-wavelength campaign undertaken from July to September 1996. This study includes X-ray data from the RXTE All Sky Monitor and BATSE, two-frequency data from the Nancay radio telescope, and infrared photometry from the 1.8 m Perkins telescope at Lowell Observatory. The first long-term well-sampled IR light curve of GRS 1915+105 is presented herein and is consistent with the interpretation of this source as a long-period binary. We compare the various light curves, searching for correlations in the behaviour of the source at differing wavelengths and for possible periodicities.

ZZ ORIGEN2.2-UPJ, A complete package of ORIGEN2 libraries based on JENDL-3.2 and JENDL-3.3

International Nuclear Information System (INIS)

Ishikawa, Makoto; Kataoka, Masaharu; Ohkawachi, Yasushi; Ohki, Shigeo; JIN, Tomoyuki; Katakura, Jun-ich; Suyama, Kenya; Yanagisawa, Hiroshi; Matsumoto, Hideki; ONOUE, Akira; Sasahara, Akihiro

2006-01-01

1 - Description: ORLIBJ32 is a package of the libraries for ORIGEN2 code based on JENDL-3.2(NEA-1642). The one grouped cross section data for PWR and BWR were compiled using the burnup calculation results by SWAT code. The FBR libraries were compiled by the analysis system used at JNC for FBR core calculation. The fission yield and decay constants data were also updated using the second version of the JNDC FP library. In ORLIBJ32, not only one-grouped cross section data but also variable actinide cross section data are prepared, using a code written in FORTRAN77. The routines should be linked to the Original ORIGEN2.1 program. The LWR Libraries are prepared based on the current PWR fuel assembly specification, and the FBR libraries are based on the request by the Japanese FBR researchers. Before compiling the libraries, the specification of fuel assembly was completely reviewed and evaluated by the members of Working Group in the Japanese Nuclear Data Committee, 'working group on the evaluation of the amount of isotope generation'. ORLIBJ33 is a new libraries based on JENDL-3.3 following the release of JENDL-3.3. The parameters used to prepare the library are the same as those of ORLIBJ32. The Original version or ORLIBJ33 is coupled with ORIGEN2.1. But after the release of ORIGEN2.2 from ORNL as CCC-0371 through RSICC, several requests for a combination with ORLIBJ33 and ORIGEN2.2 were received. During the development of ORLIBJ33, released as NEA-1642, authors found a problem in the library maker for FBR libraries, and consequently it was revised and tested in JNC-Oarai. This package 'ORIGEN2.2-UPJ' contains: - updated source code of ORIGEN2.2 of CCC-0371 to use ORLIBJ32 and ORLIBJ33, - all Original libraries in CCC-0371, - ORLIBJ32 in NEA-164/03 (but libraries for FBR are revised), - and ORLIBJ33. In this package, decay data based on the second version of the JNDC FP library and, photon and decay data libraries based on JENDL-3.3 are also included. NLB and NLIB

Application of NASVD method in the CE1-GRS spectra analysis

International Nuclear Information System (INIS)

Yang Jia; Ge Liangquan; Xiong Shengqing

2010-01-01

From the spectral shape features of the Chang'e-1 gamma-ray spectrometer (CE1-GRS) spectra data of level 3, it is difficult to identify elemental composition of the lunar surface. The paper proposes using Noise Adjusted Singular Value Decomposition (NASVD) method to qualitative analysis of CE1-GRS spectra. The result shows that a number of possible elements such as U, Th, K, Fe, Ti, Si, O, Al, Mg and Ca are qualitatively determined by this method.On the other hand, for each measured spectrum, the absolute value of the amplitude corresponding to the first spectral component indicates the total activity of its corresponding lunar surface region's radioactivity. (authors)

Effect of burn-up on the radioactivation behavior of cladding hull materials studied using the ORIGEN-S code

International Nuclear Information System (INIS)

Min Ku Jeon; Chang Hwa Lee; Jung Hoon Choi; In Hak Cho; Kweon Ho Kang; Hwan-Seo Park; Geun Il Park; Chang Je Park

2013-01-01

The effect of fuel burn-up on the radioactivation behavior of cladding hull materials was investigated using the ORIGEN-S code for various materials of Zircaloy-4, Zirlo, HANA-4, and HANA-6 and for various fuel burn-ups of 30, 45, 60, and 75 GWD/MTU. The Zircaloy-4 material is the only one that does not contain Nb as an alloy constituent, and it was revealed that 125 Sb, 125m Te, and 55 Fe are the major sources of radioactivity. On the other hand, 93m Nb was identified as the most radioactive nuclide for the other materials although minor radioactive nuclides varied owing to their different initial constituents. The radioactivity of 94 Nb was of particular focus owing to its acceptance limit against a Korean intermediate-/low-level waste repository. The radioactivation calculation results revealed that only Zircaloy-4 is acceptable for the Korean repository, while the other materials required at least 4,900 of Nb decontamination factor owing to the high radioactivity of 94 Nb regardless of the fuel burn-up. A discussion was also made on the feasibility of Zr recovery methods (chlorination and electrorefining) for selective recovery of Zr so that it can be disposed of in the Korean repository. (author)

Mapping jet-ISM interactions in X-ray binaries with ALMA: a GRS 1915+105 case study

Science.gov (United States)

Tetarenko, A. J.; Freeman, P.; Rosolowsky, E. W.; Miller-Jones, J. C. A.; Sivakoff, G. R.

2018-03-01

We present Atacama Large Millimetre/Sub-Millimetre Array (ALMA) observations of IRAS 19132+1035, a candidate jet-interstellar medium (ISM) interaction zone near the black hole X-ray binary (BHXB) GRS 1915+105. With these ALMA observations (combining data from the 12 m array and the Atacama Compact Array), we map the molecular line emission across the IRAS 19132+1035 region. We detect emission from the 12CO [J = 2 - 1], 13CO [ν = 0, J = 2 - 1], C18O [J = 2 - 1], H2CO [J = 30, 3 - 20, 2], H2CO [J = 32, 2 - 22, 1], H2CO [J = 32, 1 - 22, 0], SiO [ν = 0, J = 5 - 4], CH3OH [J = 42, 2 - 31, 2], and CS [ν = 0, J = 5 - 4] transitions. Given the morphological, spectral, and kinematic properties of this molecular emission, we present several lines of evidence that support the presence of a jet-ISM interaction at this site, including a jet-blown cavity in the molecular gas. This compelling new evidence identifies this site as a jet-ISM interaction zone, making GRS 1915+105, the third Galactic BHXB with at least one conclusive jet-ISM interaction zone. However, we find that this interaction occurs on much smaller scales than was postulated by previous work, where the BHXB jet does not appear to be dominantly powering the entire IRAS 19132+1035 region. Using estimates of the ISM conditions in the region, we utilize the detected cavity as a calorimeter to estimate the time-averaged power carried in the GRS 1915+105 jets of (8.4^{+7.7}_{-8.1})× 10^{32} erg s^{-1}. Overall, our analysis demonstrates that molecular lines are excellent diagnostic tools to identify and probe jet-ISM interaction zones near Galactic BHXBs.

OTUS - Reactor inventory management system based on ORIGEN2

Energy Technology Data Exchange (ETDEWEB)

Poellaenen, R; Toivonen, H; Lahtinen, J; Ilander, T

1995-10-01

ORIGEN2 is a computer code that calculates nuclide composition and other characteristics of nuclear fuel. The use of ORIGEN2 requires good knowledge in reactor physics. However, once the input has been defined for a particular reactor type, the calculations can be easily repeated for any burnup and decay time. This procedure produces large output files that are difficult to handle manually. A new computer code, known as OTUS, was designed to facilitate the postprocessing of the data. OTUS makes use of the inventory files precalculated with ORIGEN2 in a way that enables their versatile treatment for different safety analysis purposes. A data base is created containing a comprehensive set of ORIGEN2 calculations as a function of fuel burnup and decay time. OTUS is a reactor inventory management system for a microcomputer with Windows interface. Four major data operations are available: (1) Build data modifies ORIGEN2 output data into a suitable format, (2) View data enables flexible presentation of the data as such, (3) Different calculations, such as nuclide ratios and hot particle characteristics, can be performed for severe accident analyses, consequence analyses and research purposes, (4) Summary files contain both burnup dependent and decay time dependent inventory information related to the nuclide and the reactor specified. These files can be used for safeguards, radiation monitoring and safety assessment. (orig.) (22 refs., 29 figs.).

Application of coupled codes for safety analysis and licensing issues

International Nuclear Information System (INIS)

Langenbuch, S.; Velkov, K.

2006-01-01

An overview is given on the development and the advantages of coupled codes which integrate 3D neutron kinetics into thermal-hydraulic system codes. The work performed within GRS by coupling the thermal-hydraulic system code ATHLET and the 3D neutronics code QUABOX/CUBBOX is described as an example. The application of the coupled codes as best-estimate simulation tools for safety analysis is discussed. Some examples from German licensing practices are given which demonstrate how the improved analytical methods of coupled codes have contributed to solve licensing issues related to optimized and more economical use of fuel. (authors)

The integral analysis of 40 mm diameter pipe rupture in cooling system of fusion facility W7-X with ASTEC code

Energy Technology Data Exchange (ETDEWEB)

Kačegavičius, Tomas, E-mail: Tomas.Kacegavicius@lei.lt; Povilaitis, Mantas, E-mail: Mantas.Povilaitis@lei.lt

2015-12-15

Highlights: • The analysis of loss-of-coolant accident (LOCA) in W7-X facility. • Burst disc is sufficient to prevent pressure inside the plasma vessel exceeding 110 kPa. • Developed model of the cooling system adequately represents the expected phenomena. - Abstract: Fusion is the energy production technology, which could potentially solve problems with growing energy demand of population in the future. Wendelstein 7-X (W7-X) is an experimental facility of stellarator type, which is currently being built at the Max-Planck-Institute for Plasmaphysics located in Greifswald, Germany. W7-X shall demonstrate that in future the energy could be produced in such type of fusion reactors. The safety analysis is required before the operation of the facility could be started. A rupture of 40 mm diameter pipe, which is connected to the divertor unit (module for plasma cooling) to ensure heat removal from the vacuum vessel in case of no-plasma operation mode “baking” is one of the design basis accidents to be investigated. During “baking” mode the vacuum vessel structures and working fluid – water are heated to the temperature 160 °C. This accident was selected for the detailed analysis using integral code ASTEC, which is developed by IRSN (France) and GRS mbH (Germany). This paper presents the integral analysis of W7-X response to a selected accident scenario. The model of the main cooling circuit and “baking” circuit was developed for ASTEC code. There were analysed two cases: (1) rupture of a pipe connected to the upper divertor unit and (2) rupture of a pipe connected to the lower divertor unit. The results of analysis showed that in both cases the water is almost completely released from the units into the plasma vessel. In both cases the pressure in the plasma vessel rapidly increases and in 28 s the set point for burst disc opening is reached preventing further pressurisation.

Generation and validation of ORIGEN-S libraries for depletion and transmutation calculations based on JEF2.2 and EAF3 basic data

Energy Technology Data Exchange (ETDEWEB)

Hoogenboom, J.E. [Delft Univ. of Technol. (Netherlands). Interfaculty Reactor Inst.; Kloosterman, J.L. [Netherlands Energy Research Foundation ECN, P.O. Box 1, 1755 ZG Petten (Netherlands)

1997-07-01

The data libraries for light elements, actinides and fission products of the ORIGEN-S code for depletion and transmutation calculations in the SCALE4.1 computer code system have been updated with respect to cross-section data, radioactive-decay data and fission-product yield data using JEF2.2 as the basic data source and EAF3 as an additional source. This required the fission-product library to be extended with 201 new fission-product nuclides or isomeric states. The effect of the update of different quantities involved is evaluated with a burn-up benchmark. When ORIGEN-S is used as a stand-alone code, i.e. without regular update of cross-sections of the major nuclides due to changes in the neutron spectrum during burn-up, the results show appreciable differences in actinide and fission-product densities due to the cross-section update. The effects of updates of decay data and fission-product yields are generally small, but with noticeable exceptions. The update of fission and capture reaction energies gives a small but systematic change in actinide and fission-product concentration. The new ORIGEN-S libraries have also been converted for use with the SCALE4.2 package. (orig.)

Generation and validation of ORIGEN-S libraries for depletion and transmutation calculations based on JEF2.2 and EAF3 basic data

International Nuclear Information System (INIS)

Hoogenboom, J.E.

1997-01-01

The data libraries for light elements, actinides and fission products of the ORIGEN-S code for depletion and transmutation calculations in the SCALE4.1 computer code system have been updated with respect to cross-section data, radioactive-decay data and fission-product yield data using JEF2.2 as the basic data source and EAF3 as an additional source. This required the fission-product library to be extended with 201 new fission-product nuclides or isomeric states. The effect of the update of different quantities involved is evaluated with a burn-up benchmark. When ORIGEN-S is used as a stand-alone code, i.e. without regular update of cross-sections of the major nuclides due to changes in the neutron spectrum during burn-up, the results show appreciable differences in actinide and fission-product densities due to the cross-section update. The effects of updates of decay data and fission-product yields are generally small, but with noticeable exceptions. The update of fission and capture reaction energies gives a small but systematic change in actinide and fission-product concentration. The new ORIGEN-S libraries have also been converted for use with the SCALE4.2 package. (orig.)

Detection of X-ray spectral state transitions in mini-outbursts of black hole transient GRS 1739-278

Science.gov (United States)

Yan, Zhen; Yu, Wenfei

2017-10-01

We report the detection of the state transitions and hysteresis effect in the two mini-outbursts of the black hole (BH) transient GRS 1739-278 following its 2014 major outburst. The X-ray spectral evolutions in these two mini-outbursts are similar to the major outburst in spite of their peak luminosities and the outburst durations are one order of magnitude lower. We found L_hard{-to-soft} and Lpeak,soft of the mini-outbursts also follow the correlation previously found in other X-ray binaries. L_hard{-to-soft} of the mini-outbursts is still higher than that of the persistent BH binary Cyg X-1, which supports that there is a link between the maximum luminosity a source can reach in the hard state and the corresponding non-stationary accretion represented by substantial rate of change in the mass accretion rate during flares/outbursts. The detected luminosity range of these two mini-outbursts is roughly in 3.5 × 10-5 to 0.015 (D/7.5 kpc)2(M/8M⊙) LEdd. The X-ray spectra of other BH transients at such low luminosities are usually dominated by a power-law component, and an anti-correlation is observed between the photon index and the X-ray luminosity below 1 per cent LEdd. So, the detection of X-ray spectral state transitions indicates that the accretion flow evolution in these two mini-outbursts of GRS 1739-278 are different from other BH systems at such low-luminosity regime.

Comparison of (alpha, n) thick-target neutron yields and spectra from ORIGEN-S and SOURCES

International Nuclear Information System (INIS)

Brown, T.H.; Wilson, W.B.; Perry, R.T.; Charlton, W.S.

1998-01-01

Both ORIGEN-S and SOURCES generate thick-target neutron yields and energy spectra from (α,n) reactions in homogeneous materials. SOURCES calculates yield and spectra for any material containing α-emitting and (α,n) target elements by simulating reaction physics, using α-emission energy spectra, elemental stopping cross sections, (α,n) cross sections for target nuclei, and branching fractions to produce-nuclide energy levels. This methodology results in accurate yield and spectra. ORIGEN-S has two options for calculating yields and spectra. The UO 2 option (default) estimates yields and spectra assuming the input α-emitters to be infinitely dilute in UO 2 . The borosilicate-glass option estimates yields from the total input material composition and generates spectra purportedly representative of spectra generated by 238 Pu, 241 Am, 242 Cm, and 244 Cm infinitely dilute in borosilicate glass, even if none of these four α-emitters are present in the input material composition. Because yields from the borosilicate-glass option in ORIGEN-S are based on entire input material composition and are reasonably accurate, the same is often assumed to be true for spectra. The input/output functionality of the borosilicate-glass option, along with ambiguity in ORIGEN-S documentation, gives the incorrect impression that spectra representative of input compositions are generated. This impression is reinforced by wide usage of the SCALE code system and its ORIGEN-S module and their sponsorship by the US Nuclear Regulatory Commission

PHASE-RESOLVED TIMING ANALYSIS OF GRS 1915+105 IN ITS ρ STATE

International Nuclear Information System (INIS)

Yan, Shu-Ping; Wang, Na; Ding, Guo-Qiang; Qu, Jin-Lu

2013-01-01

We made a phase-resolved timing analysis of GRS 1915+105 in its ρ state and obtained detailed ρ cycle evolutions of the frequency, amplitude, and coherence of the low-frequency quasi-periodic oscillation (LFQPO). We combined our timing results with the spectral study by Neilsen et al. to perform an elaborate comparison analysis. Our analyses show that the LFQPO frequency does not scale with the inner disk radius, but it is related to the spectral index, indicating a possible correlation between the LFQPO and the corona. The LFQPO amplitude spectrum and other results are naturally explained by tying the LFQPO to the corona. The similarities of the spectra of variability parameters between the LFQPOs from ρ state and those from more steady states indicate that the LFQPOs of GRS 1915+105 in very different states seem to share the same origin.

PHASE-RESOLVED TIMING ANALYSIS OF GRS 1915+105 IN ITS {rho} STATE

Energy Technology Data Exchange (ETDEWEB)

Yan, Shu-Ping; Wang, Na; Ding, Guo-Qiang [Xinjiang Astronomical Observatory, Chinese Academy of Sciences, 150 Science 1-Street, Urumqi, Xinjiang 830011 (China); Qu, Jin-Lu, E-mail: yanshup@xao.ac.cn, E-mail: na.wang@xao.ac.cn [Key Laboratory for Particle Astrophysics, Institute of High Energy Physics, Chinese Academy of Sciences, 19B Yuquan Road, Beijing 100049 (China)

2013-04-10

We made a phase-resolved timing analysis of GRS 1915+105 in its {rho} state and obtained detailed {rho} cycle evolutions of the frequency, amplitude, and coherence of the low-frequency quasi-periodic oscillation (LFQPO). We combined our timing results with the spectral study by Neilsen et al. to perform an elaborate comparison analysis. Our analyses show that the LFQPO frequency does not scale with the inner disk radius, but it is related to the spectral index, indicating a possible correlation between the LFQPO and the corona. The LFQPO amplitude spectrum and other results are naturally explained by tying the LFQPO to the corona. The similarities of the spectra of variability parameters between the LFQPOs from {rho} state and those from more steady states indicate that the LFQPOs of GRS 1915+105 in very different states seem to share the same origin.

ORIGAMI Automator Primer. Automated ORIGEN Source Terms and Spent Fuel Storage Pool Analysis

Energy Technology Data Exchange (ETDEWEB)

Wieselquist, William A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Thompson, Adam B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Peterson, Joshua L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-04-01

Source terms and spent nuclear fuel (SNF) storage pool decay heat load analyses for operating nuclear power plants require a large number of Oak Ridge Isotope Generation and Depletion (ORIGEN) calculations. SNF source term calculations also require a significant amount of bookkeeping to track quantities such as core and assembly operating histories, spent fuel pool (SFP) residence times, heavy metal masses, and enrichments. The ORIGEN Assembly Isotopics (ORIGAMI) module in the SCALE code system provides a simple scheme for entering these data. However, given the large scope of the analysis, extensive scripting is necessary to convert formats and process data to create thousands of ORIGAMI input files (one per assembly) and to process the results into formats readily usable by follow-on analysis tools. This primer describes a project within the SCALE Fulcrum graphical user interface (GUI) called ORIGAMI Automator that was developed to automate the scripting and bookkeeping in large-scale source term analyses. The ORIGAMI Automator enables the analyst to (1) easily create, view, and edit the reactor site and assembly information, (2) automatically create and run ORIGAMI inputs, and (3) analyze the results from ORIGAMI. ORIGAMI Automator uses the standard ORIGEN binary concentrations files produced by ORIGAMI, with concentrations available at all time points in each assembly’s life. The GUI plots results such as mass, concentration, activity, and decay heat using a powerful new ORIGEN Post-Processing Utility for SCALE (OPUS) GUI component. This document includes a description and user guide for the GUI, a step-by-step tutorial for a simplified scenario, and appendices that document the file structures used.

MOX Cross-Section Libraries for ORIGEN-ARP

International Nuclear Information System (INIS)

Gauld, I.C.

2003-01-01

The use of mixed-oxide (MOX) fuel in commercial nuclear power reactors operated in Europe has expanded rapidly over the past decade. The predicted characteristics of MOX fuel such as the nuclide inventories, thermal power from decay heat, and radiation sources are required for design and safety evaluations, and can provide valuable information for non-destructive safeguards verification activities. This report describes the development of computational methods and cross-section libraries suitable for the analysis of irradiated MOX fuel with the widely-used and recognized ORIGEN-ARP isotope generation and depletion code of the SCALE (Standardized Computer Analyses for Licensing Evaluation) code system. The MOX libraries are designed to be used with the Automatic Rapid Processing (ARP) module of SCALE that interpolates appropriate values of the cross sections from a database of parameterized cross-section libraries to create a problem-dependent library for the burnup analysis. The methods in ORIGEN-ARP, originally designed for uranium-based fuels only, have been significantly upgraded to handle the larger number of interpolation parameters associated with MOX fuels. The new methods have been incorporated in a new version of the ARP code that can generate libraries for low-enriched uranium (LEU) and MOX fuel types. The MOX data libraries and interpolation algorithms in ORIGEN-ARP have been verified using a database of declared isotopic concentrations for 1042 European MOX fuel assemblies. The methods and data are validated using a numerical MOX fuel benchmark established by the Organization for Economic Cooperation and Development (OECD) Working Group on burnup credit and nuclide assay measurements for irradiated MOX fuel performed as part of the Belgonucleaire ARIANE International Program

iBEST: a program for burnup history estimation of spent fuels based on ORIGEN-S

International Nuclear Information System (INIS)

Kim, Do Yeon; Hong, Ser Gi; Ahn, Gil Hoon

2015-01-01

In this paper, we describe a computer program, iBEST (inverse Burnup ESTimator), that we developed to accurately estimate the burnup histories of spent nuclear fuels based on sample measurement data. The burnup history parameters include initial uranium enrichment, burnup, cooling time after discharge from reactor, and reactor type. The program uses algebraic equations derived using the simplified burnup chains of major actinides for initial estimations of burnup and uranium enrichment, and it uses the ORIGEN-S code to correct its initial estimations for improved accuracy. In addition, we newly developed a stable bisection method coupled with ORIGEN-S to correct burnup and enrichment values and implemented it in iBEST in order to fully take advantage of the new capabilities of ORIGEN-S for improving accuracy. The iBEST program was tested using several problems for verification and well-known realistic problems with measurement data from spent fuel samples from the Mihama-3 reactor for validation. The test results show that iBEST accurately estimates the burnup history parameters for the test problems and gives an acceptable level of accuracy for the realistic Mihama-3 problems

MOCUP, MCNP/ORIGEN Coupling Utility Programs

International Nuclear Information System (INIS)

SEIDL, Marcus

2003-01-01

1 - Description of program or function: MOCUP is a series of utility and data manipulation programs to solve time and space-dependent coupled neutronics/isotopics problems. 2 - Methods: The neutronics calculation is performed by the Los Alamos National Laboratory code system, version 4a or later (CCC-200 or CCC-660),and the depletion and isotopics calculation is performed by CCC-371/ORIGEN2.1 developed at Oak Ridge National Laboratory. MCNP and ORIGEN2.1 are NOT included in this package. MOCUP consists of three utility programs (mcnpPRO, origenPRO, compPRO) to, respectively, search the MCNP output and tally files for relevant cell and tally parameters, prepare ORIGEN2.1 input files and execute the ORIGEN2.1 runs, and search ORIGEN2.1 punch files for relevant isotope concentrations and produce new MCNP input files. A graphical user interface is provided for execution convenience. 3 - Restrictions on the complexity of the problem: At present, no mechanism exists for automatic serial execution of the program modules. The user must interface with the GUI to run each of the modules

Nuclide identifier and grat data reader application for ORIGEN output file

International Nuclear Information System (INIS)

Arif Isnaeni

2011-01-01

ORIGEN is a one-group depletion and radioactive decay computer code developed at the Oak Ridge National Laboratory (ORNL). ORIGEN takes one-group neutronics calculation providing various nuclear material characteristics (the buildup, decay and processing of radioactive materials). ORIGEN output is a text-based file, ORIGEN output file contains only numbers in the form of group data nuclide, nuclide identifier and grat. This application was created to facilitate data collection nuclide identifier and grat, this application also has a function to acquire mass number data and calculate mass (gram) for each nuclide. Output from these applications can be used for computer code data input for neutronic calculations such as MCNP. (author)

Integrated burnup calculation code system SWAT

International Nuclear Information System (INIS)

Suyama, Kenya; Hirakawa, Naohiro; Iwasaki, Tomohiko.

1997-11-01

SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. It enables us to analyze the burnup problem using neutron spectrum depending on environment of irradiation, combining SRAC which is Japanese standard thermal reactor analysis code system and ORIGEN2 which is burnup code widely used all over the world. SWAT makes effective cross section library based on results by SRAC, and performs the burnup analysis with ORIGEN2 using that library. SRAC and ORIGEN2 can be called as external module. SWAT has original cross section library on based JENDL-3.2 and libraries of fission yield and decay data prepared from JNDC FP Library second version. Using these libraries, user can use latest data in the calculation of SWAT besides the effective cross section prepared by SRAC. Also, User can make original ORIGEN2 library using the output file of SWAT. This report presents concept and user's manual of SWAT. (author)

QPO detection in superluminal black hole GRS 1915+105

Science.gov (United States)

Bhulla, Yashpal; Jaaffrey, S. N. A.

2018-05-01

We report on the first superluminal Black Hole GRS 1915+105 observed by the Rossi X-ray Timing Explorer - Proportion Counter Array (RXTE/PCA). We detect the Quasi Periodic Oscillations (QPOs) in the Power Density Spectrum (PDS) of source which have luminosity very near to Eddington limit and long variability in X-ray light curve. In power density spectrum, we deal with the study of highly variability amplitude, time evolution of the characteristic timescale, Quality Factor and Full Width at Half Maximum (FWHM). We find significant QPOs in 15 different observation IDs with frequency around 67 Hz although quality factor nearly 20 but in two IDs frequency is found just double. Typical fractional rms for GRS 1915+105 is dominating the hard band increasing steeply with energy more than 13% at 20-40 keV band.

An alternative to the SAS2H/ORIGEN-S sequence to account for water-density effects in BWR systems

International Nuclear Information System (INIS)

Leal, L.C.; Hermann, O.W.; Ryman, J.C.; Broadhead, B.L.

1996-01-01

A scheme to generate one-group problem-dependent cross-section libraries for point-depletion calculations with the ORIGEN-S code was developed as an alternative to the SAS2H sequence of the SCALE code system. The methodology, named Automatic Rapid Processing (ARP), generates libraries by interpolating in SAS2H precomputed cross section libraries. The method has been used to generate ORIGEN-S cross section libraries on a personal computer resulting in a great reduction of computer time without a sacrifice of accuracy over that required by corresponding SAS2H calculations. The ARP scheme generates ORIGEN-S libraries by interpolating in burnup and enrichment for PWR assemblies. The intent of this work is to describe a procedure which extends the application of the ARP methodology to BWR assemblies by including the axial water-density effects in the generation of the ORIGEN-S cross-section libraries. The axial liquid- to-steam change of state in BWR systems leads to a variation in the water density and significant cross-section changes as a function of the water density. To account for the axial water-density changes in a SAS2H calculation, the water density is entered explicitly in the generation of the one-group ORIGEN-S cross-section libraries generated from the SCALE 27-group library. In its original version, ARP does not account for the effects of water-density variation in ORIGEN-S cross-section library generation, and, therefore, its application is restricted to systems for which the impact of this parameter is negligible. To update the ARP methodology to account for the water-density effect, a detailed study of the cross-section change with this parameter was performed with an 8 x 8 (General Electric) BWR assembly

Validation and uncertainty analysis of the Athlet thermal-hydraulic computer code

International Nuclear Information System (INIS)

Glaeser, H.

1995-01-01

The computer code ATHLET is being developed by GRS as an advanced best-estimate code for the simulation of breaks and transients in Pressurized Water Reactor (PWRs) and Boiling Water Reactor (BWRs) including beyond design basis accidents. A systematic validation of ATHLET is based on a well balanced set of integral and separate effects tests emphasizing the German combined Emergency Core Cooling (ECC) injection system. When using best estimate codes for predictions of reactor plant states during assumed accidents, qualification of the uncertainty in these calculations is highly desirable. A method for uncertainty and sensitivity evaluation has been developed by GRS where the computational effort is independent of the number of uncertain parameters. (author)

MOCUP: MCNP-ORIGEN2 coupled utility program

International Nuclear Information System (INIS)

Moore, R.L.; Schnitzler, B.G.; Wemple, C.A.

1995-01-01

MOCUP is a system of external processors that allow for a limited treatment of the temporal composition of the user-selected MCNP cells in a time-dependent flux environment. The ORIGEN2 code computes the time-dependent compositions of these individually selected MCNP cells. All data communication between the two codes is accomplished through the MCNP and ORIGEN2 input/output files, the MOCUP Processor Output files, and two user supplied tables. MOCUP is either command line or interactively driven. The interactive interface is based on the portable XII window environment and the Motif tool kit. MOCUP was constructed so that no modifications to either MCNP or ORIGEN2 were necessary. Section 4 of the writeup contains the input instructions needed to set up the MOCUP run. MOCUP is extremely useful for analysts who perform isotope production, material transformation, and depletion and isotope analyses on complex, non-lattice geometries, and uniform and non-uniform lattices

GRS annual report 1985

International Nuclear Information System (INIS)

1985-01-01

The report summarizes the tasks and objectives of the GRS and gives an account of the activities in the year under review, including a report of the managing board and the organizational tree. The various projects performed and activities accompanying research work are discussed in detail. Information is given on activities in cooperation with foreign organisations, and on publications and conference papers by members of the GRS. (DG) [de

Code of Practice for Protection Against Ionizing Radiation Emitted from X-ray Analysis Equipment (1984)

International Nuclear Information System (INIS)

1984-01-01

Appropriate working rules, safety features and monitoring requirements for general X-ray analysis units or equipment are laid down in this Code which is intended for users of such equipment. The Code advises that establishments draw up their own working procedures based on appropriate legislation and on the recommendations contained in this Code. The Code also describes the requirements for X-ray analysis equipment necessary to ensure safety. (NEA) [fr

Validation of MCNP and ORIGEN-S 3-D computational model for reactivity predictions during BR2 operation

International Nuclear Information System (INIS)

Kalcheva, S.; Koonen, E.; Ponsard, B.

2005-01-01

The Belgian Material Test Reactor (MTR) BR2 is strongly heterogeneous high flux engineering test reactor at SCK-CEN (Centre d'Etude de l'energie Nucleaire) in Mol at a thermal power 60 to 100 MW. It deploys highly enriched uranium, water cooled concentric plate fuel elements, positioned inside a beryllium reflector with complex hyperboloid arrangement of test holes. The objective of this paper is the validation of a MCNP and ORIGEN-S 3D model for reactivity predictions of the entire BR2 core during reactor operation. We employ the Monte Carlo code MCNP-4C for evaluating the effective multiplication factor k eff and 3D space dependent specific power distribution. The 1D code ORIGEN-S is used for calculation of isotopic fuel depletion versus burn up and preparation of a database (DB) with depleted fuel compositions. The approach taken is to evaluate the 3D power distribution at each time step and along with DB to evaluate the 3D isotopic fuel depletion at the next step and to deduce the corresponding shim rods positions of the reactor operation. The capabilities of the both codes are fully exploited without constraints on the number of involved isotope depletion chains or increase of the computational time. The reactor has a complex operation, with important shutdowns between cycles, and its reactivity is strongly influenced by poisons, mainly 3 He and 6 Li from the beryllium reflector, and burnable absorbers 149 Sm and 10 B in the fresh UAlx fuel. Our computational predictions for the shim rods position at various restarts are within 0.5$ (β eff =0.0072). (author)

Accretion States of the Galactic Micro Quasar GRS 1758-258

Science.gov (United States)

Soria, Roberto; Mehdipour, Missagh; Broderick, Jess W.; Hao, JingFang; Hannikainen, Diana C.; Pottschmidt, Katja; Zhang, Shuang-Nan

2011-01-01

We present the results of a radio and X-ray study of the Galactic micro quasar GRS 1758-258, using unpublished archival data and new observations. We focus in particular on the 2000-2002 state transitions, and on its more quiet behaviour in 2008-2009. Our spectral and timing analysis of the XMM-Newton data shows that the source was in the canonical intermediate, soft and hard states in 2000 September 19,2001 March 22 and 2002 September 28, respectively. We estimate the disk size, luminosity and temperature, which are consistent with a black hole mass approx.10 Solar Mass, There is much overlap between the range of total X-ray luminosities (on average approx. 0.02L(sub Edd)) in the hard and soft states, and probably between the corresponding mass accretion rates; in fact, the hard state is often more luminous. The extended radio lobes seen in 1992 and 1997 are still present in 2008-2009. The 5-GHz radio core flux density has shown variability between approx. 0.1-0.5 mJy over the last two decades. This firmly places GRS 1758-258 in the radio-quiet sequence of Galactic black holes, in the radio/X-ray plane. We note that this dichotomy is similar to the dichotomy between the radio/X-ray sequences of Seyfert and radio galaxies. We propose that the different radio efficiency of the two sequences is due to relativistic electron/positron jets in radio-loud black holes, and sub-relativistic, thermally dominated outflows in radio-quiet sources.

ORIGEN2.1 Cycle Specific Calculation of Krsko Nuclear Power Plant Decay Heat and Core Inventory

International Nuclear Information System (INIS)

Vukovic, J.; Grgic, D.; Konjarek, D.

2010-01-01

This paper presents ORIGEN2.1 computer code calculation of Krsko Nuclear Power Plant core for Cycle 24. The isotopic inventory, core activity and decay heat are calculated in one run for the entire core using explicit depletion and decay of each fuel assembly. Separate pre-ori application which was developed is utilized to prepare corresponding ORIGEN2.1 inputs. This application uses information on core loading pattern to determine fuel assembly specific depletion history using 3D burnup which is obtained from related PARCS computer code calculation. That way both detailed single assembly calculations as well as whole core inventory calculations are possible. Because of the immense output of the ORIGEN2.1, another application called post-ori is used to retrieve and plot any calculated property on the basis of nuclide, element, summary isotope or group of elements for activation products, actinides and fission products segments. As one additional possibility, with the post-ori application it is able to calculate radiotoxicity from calculated ORIGEN2.1 inventory. The results which are obtained using the calculation model of ORIGEN2.1 computer code are successfully compared against corresponding ORIGEN-S computer code results.(author).

Different Types of X-Ray Bursts from GRS 1915+105 and Their Origin

Science.gov (United States)

Yadav, J. S.; Rao, A. R.; Agrawal, P. C.; Paul, B.; Seetha, S.; Kasturirangan, K.

1999-06-01

We report X-ray observations of the Galactic X-ray transient source GRS 1915+105 with the pointed proportional counters of the Indian X-ray Astronomy Experiment (IXAE) onboard the Indian satellite IRS-P3, which show remarkable richness in temporal variability. The observations were carried out on 1997 June 12-29 and August 7-10, in the energy range of 2-18 keV and revealed the presence of very intense X-ray bursts. All the observed bursts have a slow exponential rise, a sharp linear decay, and broadly can be put in two classes: irregular and quasi-regular bursts in one class, and regular bursts in the other. The regular bursts are found to have two distinct timescales and to persist over extended durations. There is a strong correlation between the preceding quiescent time and the burst duration for the quasi-regular and irregular bursts. No such correlation is found for the regular bursts. The ratio of average flux during the burst time to the average flux during the quiescent phase is high and variable for the quasi-regular and irregular bursts, while it is low and constant for the regular bursts. We present a comprehensive picture of the various types of bursts observed in GRS 1915+105 in the light of the recent theories of advective accretion disks. We suggest that the peculiar bursts that we have seen are characteristic of the change of state of the source. The source can switch back and forth between the low-hard state and the high-soft state near critical accretion rates in a very short timescale, giving rise to the irregular and quasi-regular bursts. The fast timescale for the transition of the state is explained by invoking the appearance and disappearance of the advective disk in its viscous timescale. The periodicity of the regular bursts is explained by matching the viscous timescale with the cooling timescale of the postshock region. A test of the model is presented using the publicly available 13-60 keV RXTE/PCA data for irregular and regular bursts

Modification Of Origen2's Libraries For RSG-Gas

International Nuclear Information System (INIS)

Purwadi, Mohammad Dhandhang

2000-01-01

The Origen2 code is a versatile computer code for analixing nuclide characteristics in the nuclear materials.the code is used broadly, and the libraries have been adjusted in the conventional reactor types.Today, the code has become the main tool which support nuclear material safeguard programs in PPTRR-BATAN but its library is not adjusted to the conventional reactor types. Today, the code has become the main tool which supports nuclear material safeguard Programs in PPTR-BATAN but its library is not adjusted for RSG-Gas yet. In this work effort to adjust the libraries to RSG-Gas condition was carried out. The nuclides in the libraries are grouped as actinide, fission products, activation product and structure material. Totally, there are 1700 nuclides of ORIGEN2 libraries that must be adjusted.. In this work so far only 80 nuclides have been adjusted

THE RELATIONSHIP BETWEEN X-RAY LUMINOSITY AND MAJOR FLARE LAUNCHING IN GRS 1915+105

International Nuclear Information System (INIS)

Punsly, Brian; Rodriguez, Jérôme

2013-01-01

We perform the most detailed analysis to date of the X-ray state of the Galactic black hole candidate GRS 1915+105 just prior to (0-4 hr) and during the brief (1-7 hr) ejection of major (superluminal) radio flares. A very strong model independent correlation is found between the 1.2 keV-12 keV X-ray flux 0-4 hr before flare ejections with the peak optically thin 2.3 GHz emission of the flares. This suggests a direct physical connection between the energy in the ejection and the luminosity of the accretion flow preceding the ejection. In order to quantify this concept, we develop techniques to estimate the intrinsic (unabsorbed) X-ray luminosity, L intrinsic , from RXTE All Sky Monitor data and to implement known methods to estimate the time-averaged power required to launch the radio emitting plasmoids, Q (sometimes called jet power). We find that the distribution of intrinsic luminosity from 1.2 keV-50 keV, L intrinsic (1.2-50), is systematically elevated just before ejections compared to arbitrary times when there are no major ejections. The estimated Q is strongly correlated with L intrinsic (1.2-50) 0-4 hr before the ejection, the increase in L intrinsic (1.2-50) in the hours preceding the ejection and the time-averaged L intrinsic (1.2-50) during the flare rise. Furthermore, the total time-averaged power during the ejection (Q + the time average of L intrinsic (1.2-50) during ejection) is strongly correlated with L intrinsic (1.2-50) just before launch with near equality if the distance to the source is ≈10.5 kpc.

Assessment of the system code DRUFAN/ATHLET using results of LOBI tests

International Nuclear Information System (INIS)

Burwell, J.M.; Kirmse, R.E.; Kyncl, M.; Malhotra, P.K.

1989-09-01

Four post-test analyses have been performed by GRS within the Shared Cost Action Programme (SCAP) sponsored by the Commission of the European Communities (contract 3015-86-07 EL ISP D) and by the Bundesminister fuer Forschung und Technologie of the Federal Republic of Germany (Research project RS 739). The four tests were mutually selected by the contractors (CEA, GRS, IKE, Univ. Pisa) of activity No. 3 and by the project organizer. Some of the tests were selected to be analyzed by more than one participant in order to allow comparison between analytical results obtained with different codes or obtained by different code-users. DRUFAN/ATHLET verification analyses were performed by IKE too. The four tests selected for the GRS activity are: - A2-77A (Natural Circulation Test), Analysis with ATHLET - A1-76 (Steam Generator Performance Test), Analysis with DRUFAN - BL-01 (Intermediate Leak), Analysis with ATHLET - A2-81 (Small Leak), Analysis with ATHLET. This final report contains the results of the four post test analysis including the comparison between measured and calculated quantities and the description of the applied codes, the selected model of the LOBI facility and the conclusions drawn for the improvement of the codes models

Version of ORIGEN2 with automated sensitivity-calculation capability

International Nuclear Information System (INIS)

Worley, B.A.; Wright, R.Q.; Pin, F.G.

1986-01-01

ORIGEN2 is a widely used point-depletion and radioactive-decay computer code for use in simulating nuclear fuel cycles and/or spent fuel characteristics. The code calculates the amount of each nuclide being considered in the problem at a specified number of times, and upon request, a database of conversion factors relating mass compositions to specific material characteristics is used to calculate and print the total nuclide-dependent radioactivity, thermal power, and toxicity, as well as absorption, fission, neutron emission, and photon emission rates. The purpose of this paper is to report on the availability of a version of ORIGEN2 that will calculate, on option the derivative of all responses with respect to any variable used in the code

Development of a one-group cross section data base of the ORIGEN2 computer code for research reactor applications

Energy Technology Data Exchange (ETDEWEB)

Kim, Jung Do; Gil, Choong Sub; Lee, Jong Tai [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Hwang, Won Guk [Kyung Hee University, Seoul (Korea, Republic of)

1992-03-01

A one-group cross section data base of the ORIGEN2 computer code was developed for research reactor applications. For this, ENDF/B-IV and -V data were processed using the NJOY code system into 69-group data. The burnup dependent weighting spectra for KMRR were calculated with the WIMS-KAERI computer code, and then the 69-group data were collapsed to one-group using the spectra. The ORlGEN2-predicted burnup-dependent actinide compositions of KMRR spent fuel using the newly developed data base show a good agreement with the results of detailed multigroup transport calculation. In addition, the burnup characteristics of KMRR spent fuel was analyzed with the new data base. (Author).

Development of a one-group cross section data base of the ORIGEN2 computer code for research reactor applications

International Nuclear Information System (INIS)

Kim, Jung Do; Gil, Choong Sub; Lee, Jong Tai; Hwang, Won Guk

1992-01-01

A one-group cross section data base of the ORIGEN2 computer code was developed for research reactor applications. For this, ENDF/B-IV and -V data were processed using the NJOY code system into 69-group data. The burnup dependent weighting spectra for KMRR were calculated with the WIMS-KAERI computer code, and then the 69-group data were collapsed to one-group using the spectra. The ORlGEN2-predicted burnup-dependent actinide compositions of KMRR spent fuel using the newly developed data base show a good agreement with the results of detailed multigroup transport calculation. In addition, the burnup characteristics of KMRR spent fuel was analyzed with the new data base. (Author)

Uncertainty and sensitivity analysis applied to coupled code calculations for a VVER plant transient

International Nuclear Information System (INIS)

Langenbuch, S.; Krzykacz-Hausmann, B.; Schmidt, K. D.

2004-01-01

The development of coupled codes, combining thermal-hydraulic system codes and 3D neutron kinetics, is an important step to perform best-estimate plant transient calculations. It is generally agreed that the application of best-estimate methods should be supplemented by an uncertainty and sensitivity analysis to quantify the uncertainty of the results. The paper presents results from the application of the GRS uncertainty and sensitivity method for a VVER-440 plant transient, which was already studied earlier for the validation of coupled codes. For this application, the main steps of the uncertainty method are described. Typical results of the method applied to the analysis of the plant transient by several working groups using different coupled codes are presented and discussed The results demonstrate the capability of an uncertainty and sensitivity analysis. (authors)

THE VARIABLE NEAR-INFRARED COUNTERPART OF THE MICROQUASAR GRS 1758–258

International Nuclear Information System (INIS)

Luque-Escamilla, Pedro L.; Martí, Josep; Muñoz-Arjonilla, Álvaro J.

2014-01-01

We present a new study of the microquasar system GRS 1758–258 in the near-infrared domain based on archival observations with the Hubble Space Telescope and the NICMOS camera. In addition to confirming the near-infrared counterpart pointed out by Muñoz-Arjonilla et al., we show that this object displays significant photometric variability. From its average magnitudes, we also find that GRS 1758–258 fits well within the correlation between the optical/near-infrared and X-ray luminosity known to exist for low-mass, black-hole candidate X-ray binaries in a hard state. Moreover, the spectral energy distribution built using all radio, near-infrared, and X-ray data available closest in time to the NICMOS observations can be reasonably interpreted in terms of a self-absorbed radio jet and an irradiated accretion disk model around a stellar-mass black hole. All these facts match the expected behavior of a compact binary system and strengthen our confidence in the counterpart identification

How well does ORIGEN predict spent LWR fuel characteristics

International Nuclear Information System (INIS)

Mailen, J.C.; Roddy, J.W.

1987-01-01

The ORIGEN computer code is widely used to estimate the radionuclide content (actinides, activation and fission products) of irradiated reactor fuel and the resultant heat generation and radiation levels associated with such fuel. These estimates are used as source terms in safety evaluations of operating reactors, for evaluation of fuel behavior and regulation of the at-reactor storage, for transportation studies, and for evaluation of the ultimate geologic storage of the fuel. This survey summarizes the fuel data available in the open literature and, where given, the calculated values by ORIGEN. Plans for additional analyses of well-characterized reactor fuel samples to improve the validation of ORIGEN2 are discussed

GRS 1758–258: RXTE Monitoring of a Rare Persistent Hard State Black Hole

Directory of Open Access Journals (Sweden)

M. Obst

2011-01-01

Full Text Available GRS 1758–258 is the least studied of the three persistent black hole X-ray binaries in our Galaxy. It is also one of only two known black hole candidates, including all black hole transients, which shows a decrease of its 3-10 keV flux when entering the thermally dominated soft state, rather than an increase.We present the spectral evolution of GRS 1758–258 from RXTE-PCA observations spanning a time of about 11 years from 1996 to 2007. During this time, seven dim soft states are detected. We also consider INTEGRAL monitoring observations of the source and compare the long-term behavior to that of the bright persistent black hole X-ray binary Cygnus X-1. We discuss the observed state transitions in the light of physical scenarios for black hole transitions.

NuSTAR OBSERVATION OF A TYPE I X-RAY BURST FROM GRS 1741.9-2853

International Nuclear Information System (INIS)

Barrière, Nicolas M.; Krivonos, Roman; Tomsick, John A.; Boggs, Steven E.; Craig, William W.; Bachetti, Matteo; Chakrabarty, Deepto; Christensen, Finn E.; Hailey, Charles J.; Mori, Kaya; Harrison, Fiona A.; Hong, Jaesub; Stern, Daniel; Zhang, William W.

2015-01-01

We report on two NuSTAR observations of GRS 1741.9-2853, a faint neutron star (NS) low-mass X-ray binary burster located 10' away from the Galactic center. NuSTAR detected the source serendipitously as it was emerging from quiescence: its luminosity was 6 × 10 34  erg s –1 on 2013 July 31 and 5 × 10 35  erg s –1 in a second observation on 2013 August 3. A bright, 800 s long, H-triggered mixed H/He thermonuclear Type I burst with mild photospheric radius expansion (PRE) was present during the second observation. Assuming that the luminosity during the PRE was at the Eddington level, an H mass fraction X = 0.7 in the atmosphere, and an NS mass M = 1.4 M ☉ , we determine a new lower limit on the distance for this source of 6.3 ± 0.5 kpc. Combining with previous upper limits, this places GRS 1741.9-2853 at a distance of 7 kpc. Energy independent (achromatic) variability is observed during the cooling of the NS, which could result from the disturbance of the inner accretion disk by the burst. The large dynamic range of this burst reveals a long power-law decay tail. We also detect, at a 95.6% confidence level (1.7σ), a narrow absorption line at 5.46 ± 0.10 keV during the PRE phase of the burst, reminiscent of the detection by Waki et al. We propose that the line, if real, is formed in the wind above the photosphere of the NS by a resonant Kα transition from H-like Cr gravitationally redshifted by a factor 1 + z = 1.09, corresponding to a radius range of 29.0-41.4 km for a mass range of 1.4-2.0 M ☉

Benchmarking and application of the state-of-the-art uncertainty analysis methods XSUSA and SHARK-X

International Nuclear Information System (INIS)

Aures, A.; Bostelmann, F.; Hursin, M.; Leray, O.

2017-01-01

Highlights: • Application of the uncertainty analysis methods XSUSA and SHARK-X. • Propagation of nuclear data uncertainty through PWR pin cell depletion calculation. • Uncertainty quantification of eigenvalue, nuclide densities and Doppler coefficient. • Top contributor to overall output uncertainty by sensitivity analysis. • Comparison with SAMPLER and TSUNAMI of the SCALE code package. - Abstract: This study presents collaborative work performed between GRS and PSI on benchmarking and application of the state-of-the-art uncertainty analysis methods XSUSA and SHARK-X. Applied to a PWR pin cell depletion calculation, both methods propagate input uncertainty from nuclear data to output uncertainty. The uncertainty of the multiplication factors, nuclide densities, and fuel temperature coefficients derived by both methods are compared at various burnup steps. Comparisons of these quantities are furthermore performed with the SAMPLER module of SCALE 6.2. The perturbation theory based TSUNAMI module of both SCALE 6.1 and SCALE 6.2 is additionally applied for comparisons of the reactivity coefficient.

Development of the point-depletion code DEPTH

International Nuclear Information System (INIS)

She, Ding; Wang, Kan; Yu, Ganglin

2013-01-01

Highlights: ► The DEPTH code has been developed for the large-scale depletion system. ► DEPTH uses the data library which is convenient to couple with MC codes. ► TTA and matrix exponential methods are implemented and compared. ► DEPTH is able to calculate integral quantities based on the matrix inverse. ► Code-to-code comparisons prove the accuracy and efficiency of DEPTH. -- Abstract: The burnup analysis is an important aspect in reactor physics, which is generally done by coupling of transport calculations and point-depletion calculations. DEPTH is a newly-developed point-depletion code of handling large burnup depletion systems and detailed depletion chains. For better coupling with Monte Carlo transport codes, DEPTH uses data libraries based on the combination of ORIGEN-2 and ORIGEN-S and allows users to assign problem-dependent libraries for each depletion step. DEPTH implements various algorithms of treating the stiff depletion systems, including the Transmutation trajectory analysis (TTA), the Chebyshev Rational Approximation Method (CRAM), the Quadrature-based Rational Approximation Method (QRAM) and the Laguerre Polynomial Approximation Method (LPAM). Three different modes are supported by DEPTH to execute the decay, constant flux and constant power calculations. In addition to obtaining the instantaneous quantities of the radioactivity, decay heats and reaction rates, DEPTH is able to calculate the integral quantities by a time-integrated solver. Through calculations compared with ORIGEN-2, the validity of DEPTH in point-depletion calculations is proved. The accuracy and efficiency of depletion algorithms are also discussed. In addition, an actual pin-cell burnup case is calculated to illustrate the DEPTH code performance in coupling with the RMC Monte Carlo code

Fission product inventory calculation by a CASMO/ORIGEN coupling program

Energy Technology Data Exchange (ETDEWEB)

Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong; Jung, In Ha [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

A CASMO/ORIGEN coupling utility program was developed to predict the composition of all the fission products in spent PWR fuels. The coupling program reads the CASMO output file, modifies the ORIGEN cross section library and reconstructs the ORIGEN input file at each depletion step. In ORIGEN, the burnup equation is solved for actinides and fission products based on the fission reaction rates and depletion flux of CASMO. A sample calculation has been performed using a 14 x 14 PWR fuel assembly and the results are given in this paper. 3 refs., 1 fig., 1 tab. (Author)

Fission product inventory calculation by a CASMO/ORIGEN coupling program

Energy Technology Data Exchange (ETDEWEB)

Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong; Jung, In Ha [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-31

A CASMO/ORIGEN coupling utility program was developed to predict the composition of all the fission products in spent PWR fuels. The coupling program reads the CASMO output file, modifies the ORIGEN cross section library and reconstructs the ORIGEN input file at each depletion step. In ORIGEN, the burnup equation is solved for actinides and fission products based on the fission reaction rates and depletion flux of CASMO. A sample calculation has been performed using a 14 x 14 PWR fuel assembly and the results are given in this paper. 3 refs., 1 fig., 1 tab. (Author)

ORIGEN-S Decay Data Library and Half-Life Uncertainties

International Nuclear Information System (INIS)

Hermann, O.W.

1998-01-01

The results of an extensive update of the decay data of the ORIGEN-S library are presented in this report. The updated decay data were provided for both the ORIGEN-S and ORIGEN2 libraries in the same project. A complete edit of the decay data plus the available half-life uncertainties are included in Appendix A. A detailed description of the types of data contained in the library, the format of the library, and the data sources are also presented. Approximately 24% of the library nuclides are stable, 66% were updated from ENDF/B-VI, about 8% were updated from ENSDF, and the remaining 2% were not updated. Appendix B presents a listing of percentage changes in decay heat from the old to the updated library for all nuclides containing a difference exceeding 1% in any parameter

Coupled code analysis of uncertainty and sensitivity of Kalinin-3 benchmark

Energy Technology Data Exchange (ETDEWEB)

Pasichnyk, Ihor; Zwermann, Winfried; Velkov, Kiril [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany); Nikonov, Sergey [VNIIAES, Moscow (Russian Federation)

2016-09-15

An uncertainty and sensitivity analysis is performed for the OECD/NEA coolant transient Benchmark (K-3) on measured data at Kalinin-3 Nuclear Power Plant (NPP). A switch off of one main coolant pump (MCP) at nominal reactor power is calculated using a coupled thermohydraulic and neutron-kinetic ATHLET-PARCS code. The objectives are to study uncertainty of total reactor power and to identify the main sources of reactor power uncertainty. The GRS uncertainty and sensitivity software package XSUSA is applied to propagate uncertainties in nuclear data libraries to the full core coupled transient calculations. A set of most important thermal-hydraulic parameters of the primary circuit is identified and a total of 23 thermohydraulic parameters are statistically varied using GRS code SUSA. The ATHLET model contains also a balance-of-plant (BOP) model which is simulated using ATHLET GCSM module. In particular the operation of the main steam generator regulators is modelled in detail. A set of 200 varied coupled ATHLET-PARCS calculations is analyzed. The results obtained show a clustering effect in the behavior of global reactor parameters. It is found that the GCSM system together with varied input parameters strongly influence the overall nuclear power plant behavior and can even lead to a new scenario. Possible reasons of the clustering effect are discussed in the paper. This work is a step forward in establishing a ''best-estimate calculations in combination with performing uncertainty analysis'' methodology for coupled full core calculations.

Evidence of elevated X-ray absorption before and during major flare ejections in GRS 1915+105

Energy Technology Data Exchange (ETDEWEB)

Punsly, Brian [1415 Granvia Altamira, Palos Verdes Estates, CA 90274 (United States); Rodriguez, Jérôme [Laboratoire AIM, CEA/DSM-CNRS-Université Paris Diderot, IRFU SAp, F-91191 Gif-sur-Yvette (France); Trushkin, Sergei A., E-mail: brian.punsly1@verizon.net, E-mail: brian.punsly@comdev-usa.com [Special Astrophysical Observatory RAS, Nizhnij Arkhyz, 369167 (Russian Federation)

2014-03-10

We present time-resolved X-ray spectroscopy of the microquasar GRS 1915+105 with the MAXI observatory in order to study the accretion state just before and during the ejections associated with its major flares. Radio monitoring with the RATAN-600 radio telescope from 4.8-11.2 GHz has revealed two large, steep-spectrum major flares in the first eight months of 2013. Since the RATAN has received one measurement per day, we cannot determine the jet-forming time without more information. Fortunately, this is possible since a distinct X-ray light curve signature that occurs preceding and during major ejections has been determined in an earlier study. The X-ray luminosity spikes to very high levels in the hours before ejection, then becomes variable (with a nearly equal X-ray luminosity when averaged over the duration of the ejection) during a brief 3-8 hr ejection process. By comparing this X-ray behavior with MAXI light curves, we can estimate the beginning and end of the ejection episode of the strong 2013 flares to within ∼3 hr. Using this estimate in conjunction with time-resolved spectroscopy from the data in the MAXI archives allows us to deduce that the X-ray absorbing hydrogen column density increases significantly in the hours preceding the ejections and remains elevated during the ejections responsible for the major flares. This finding is consistent with an outflowing wind or enhanced accretion at high latitudes.

Energy scaling of the "heartbeat" pulse width of GRS 1915+105, IGR J17091-3624, and MXB 1730-335 from Rossi-XTE observations

Science.gov (United States)

Maselli, A.; Capitanio, F.; Feroci, M.; Massa, F.; Massaro, E.; Mineo, T.

2018-04-01

We investigate some key aspects of the "heartbeat" variability consisting of series of bursts with a slow rise and a fast decay, thus far detected only in GRS 1915+105, IGR J17091-3624, and MXB 1730-335. A previous analysis based on BeppoSAX data of GRS 1915+105 revealed a hard-X delay (HXD), that is a lag of the burst rise at higher energies with respect to lower ones; this leads to narrower pulse widths, w, at higher energies. We here use some light curves of Rossi-XTE observations of GRS 1915+105 for a deeper analysis of this effect and search for its presence in those extracted from some IGR J17091-3624 and MXB 1730-335 observations performed with the same satellite. Our results show that, at variance with GRS 1915+105, no HXD is evident in the light curves of MXB 1730-335 and only a marginal HXD may be argued for IGR J17091-3624. For GRS 1915+105 we find a decreasing trend of the pulse width with energy following a power law w = A ṡ E-s with an index s ≈ 0.8. Furthermore, we confirm the increase of the HXD with the recurrence time Trec of the bursts in each series that was already found in previous works using BeppoSAX data. Based on a spectral analysis of these three sources we conclude that the differences highlighted in the properties of the "heartbeat" variability are probably related to the different accreting compact object and the eventual presence of a corona in these binary interacting systems.

On Error Analysis of ORIGEN Decay Data Library Based on ENDF/B-VII.1 via Decay Heat Estimation after a Fission Event

Energy Technology Data Exchange (ETDEWEB)

Kim, Do Heon; Gil, Choong-Sup; Lee, Young-Ouk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

The method is strongly dependent on the available nuclear structure data, i.e., fission product yield data and decay data. Consequently, the improvements in the nuclear structure data could have guaranteed more reliable decay heat estimation for short cooling times after fission. The SCALE-6.1.3 code package includes the ENDF/B-VII.0-based fission product yield data and ENDF/B-VII.1-based decay data libraries for the ORIGEN-S code. The generation and validation of the new ORIGEN-S yield data libraries based on the recently available fission product yield data such as ENDF/B-VII.1, JEFF-3.1.1, JENDL/FPY-2011, and JENDL-4.0 have been presented in the previous study. According to the study, the yield data library in the SCALE-6.1.3 could be regarded as the latest one because it resulted in almost the same outcomes as the ENDF/B-VII.1. A research project on the production of the nuclear structure data for decay heat estimation of nuclear fuel has been carried out in Korea Atomic Energy Research Institute (KAERI). The data errors contained in the ORIGEN-S decay data library of SCALE-6.1.3 have been clearly identified by their changing variables. Also, the impacts of the decay data errors have been analyzed by estimating the decay heats for the fission product nuclides and their daughters after {sup 235}U thermal-neutron fission. Although the impacts of decay data errors are quite small, it reminds us the possible importance of decay data when estimating the decay heat for short cooling times after a fission event.

Fast Transition between High-soft and Low-soft States in GRS 1915 ...

Indian Academy of Sciences (India)

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

for extended periods. ... have been made to connect the observed radio characteristics in GRS 1915 + 105 like .... with low value of hard X-ray photon flux with BATSE in detail. ..... estimation and response matrix generation have been applied. ..... readjustment of the accretion disk at the outer edge to create a significant ...

Nuclear knowledge management: The GRS realisation

International Nuclear Information System (INIS)

Beraha, D.; Westerheide, P.

2004-01-01

Full text: Knowledge has become a crucial factor in our information society in deciding whether companies will last into the future. The aim of knowledge management is to promote systematically the acquisition, use, and distribution of knowledge in an organisation and to initiate measures for improving the knowledge processes. It therefore has a strategic significance for corporate decisions. The GRS faces the challenge of helping to shape the status and use of science and technology even in conditions in which many experts are retiring for age reasons. Knowledge management makes an essential contribution here. Methods and tools of knowledge management that support these activities will be outlined in the final paper. It will show also the ways of how systematic knowledge management in the GRS and its integration into projects is being performed. The approach of knowledge management for strengthening this factor consists in systematically analysing the knowledge-related processes in the identification, acquisition, sharing, distribution, use, retention, and growth of knowledge and in developing methods with the aim of managing knowledge processes better in the company. Advanced information technology enables knowledge-orientated methods ('enabling technologies') to be supported and put to good practical use. The acute lack of newcomers in the field of nuclear techniques and the narrowing financial situation are strengthening the GRS approach. To show the operational actions of knowledge management in a complete form a knowledge model has been introduced at GRS that puts the different fields of action into context with the normative and strategic goals. That model will be presented. A further section will illuminate the supporting means for practical knowledge management, whether these are procedures for the daily work or tools like the GRS Portal, the Document Management System or tools for online collaboration. Under its caption 'project oriented knowledge

Software testing and source code for the calculation of clearance values. Final report; Erprobung von Software und Quellcode zur Berechnung von Freigabewerten. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Artmann, Andreas; Meyering, Henrich

2016-11-15

The GRS research project was aimed to the test the appropriateness of the software package ''residual radioactivity'' (RESRAD) for the calculation of clearance values according to German and European regulations. Comparative evaluations were performed with RESRAD-OFFSITE and the code SiWa-PRO DSS used by GRS and the GRS program code ARTM. It is recommended to use RESRAD-OFFSITE for comparative calculations. The dose relevant air-path dispersion of radionuclides should not be modeled using RESRAD-OFFSITE, the use of ARTM is recommended. The sensitivity analysis integrated into RESRAD-OFFSITE allows a fast identification of crucial parameters.

Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS). Annual report 2010/2011; Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS). Jahresbericht 2010/2011

Energy Technology Data Exchange (ETDEWEB)

NONE

2011-07-01

The annual report 2010/11 of the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) covers the following topics: organization, corporate development and public relations; reactor accident in Fukushima; Projects and work priorities (reactor safety; waste management; radiation and environmental protection; international projects); project management and project support; equity participations and subsidiaries (RISKAUDIT IRSN/GRS International; institute for safety technologies (ISTec), European Nuclear Safety Training and Tutoring Institute (ENSTTI)).

Annual report 1986 of Society for Reactor Safety (GRS)

International Nuclear Information System (INIS)

1987-01-01

The overall development of Society for Reactor Safety (GRS) is indicated by its annual volume of achievement. After initially high growth rates that achievement has come to stand at some 62 million marks or so over the last years. GRS today employs some 450 staff, of which 325 are scientific-technical experts representing altogether nearly 4000 man-years of experience in nuclear technology. The report presents essential results from the main points of activity of GRS over 10 years. In accordance with the aim of the institution they primarily cover projects in the sectors safety and radiation protection for nuclear facilities. Since 1985, since the extension of its contract of associates, GRS has also increasingly been concerned with safety questions in non-nuclear technology, concentrating mainly on the transfer of existing methods. (orig./DG) [de

Photon and decay data libraries for ORIGEN2 code based on JENDL FP decay data file 2000

CERN Document Server

Katakura, J I

2002-01-01

Photon and decay data libraries for the ORIGEN2 code have been updated by using JENDL FP Decay Data File 2000 (JENDL/FPD-00). As for the decay data, half-lives, branching ratios and recoverable energy values have been replaced with those of the JENDL/FPD-00 file. The data of the photon library has been also replaced with those of the JENDL/FPD-00 file in which photon data of the nuclides without measured data are calculated with a theoretical method. Using the updated photon library, the calculation of photon spectrum at a short time after fission event is able to be made.

Sensibility analysis of fuel depletion using different nuclear fuel depletion codes

Energy Technology Data Exchange (ETDEWEB)

Martins, F.; Velasquez, C.E.; Castro, V.F.; Pereira, C.; Silva, C. A. Mello da, E-mail: felipmartins94@gmail.com, E-mail: carlosvelcab@hotmail.com, E-mail: victorfariascastro@gmail.com, E-mail: claubia@nuclear.ufmg.br, E-mail: clarysson@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-07-01

Nowadays, the utilization of different nuclear codes to perform the depletion and criticality calculations has been used to simulated nuclear reactors problems. Therefore, the goal is to analyze the sensibility of the fuel depletion of a PWR assembly using three different nuclear fuel depletion codes. The burnup calculations are performed using the codes MCNP5/ORIGEN2.1 (MONTEBURNS), KENO-VI/ORIGEN-S (TRITONSCALE6.0) and MCNPX (MCNPX/CINDER90). Each nuclear code performs the burnup using different depletion codes. Each depletion code works with collapsed energies from a master library in 1, 3 and 63 groups, respectively. Besides, each code uses different ways to obtain neutron flux that influences the depletions calculation. The results present a comparison of the neutronic parameters and isotopes composition such as criticality and nuclides build-up, the deviation in results are going to be assigned to features of the depletion code in use, such as the different radioactive decay internal libraries and the numerical method involved in solving the coupled differential depletion equations. It is also seen that the longer the period is and the more time steps are chosen, the larger the deviation become. (author)

Sensibility analysis of fuel depletion using different nuclear fuel depletion codes

International Nuclear Information System (INIS)

Martins, F.; Velasquez, C.E.; Castro, V.F.; Pereira, C.; Silva, C. A. Mello da

2017-01-01

Nowadays, the utilization of different nuclear codes to perform the depletion and criticality calculations has been used to simulated nuclear reactors problems. Therefore, the goal is to analyze the sensibility of the fuel depletion of a PWR assembly using three different nuclear fuel depletion codes. The burnup calculations are performed using the codes MCNP5/ORIGEN2.1 (MONTEBURNS), KENO-VI/ORIGEN-S (TRITONSCALE6.0) and MCNPX (MCNPX/CINDER90). Each nuclear code performs the burnup using different depletion codes. Each depletion code works with collapsed energies from a master library in 1, 3 and 63 groups, respectively. Besides, each code uses different ways to obtain neutron flux that influences the depletions calculation. The results present a comparison of the neutronic parameters and isotopes composition such as criticality and nuclides build-up, the deviation in results are going to be assigned to features of the depletion code in use, such as the different radioactive decay internal libraries and the numerical method involved in solving the coupled differential depletion equations. It is also seen that the longer the period is and the more time steps are chosen, the larger the deviation become. (author)

GRS current activities in East-European countries overview

International Nuclear Information System (INIS)

Teske, H.; Janke, R.

2001-01-01

This report presents the following items: agreement between the Federal Government and Utility Companies about the future utilization of the existing nuclear power plants in Germany, June 2001; GRS current activities in the Eastern European countries; science-technical co-operation with Russia; assessment of safety significance of the events for German NPP; Leningrad in-depth safety assessment, co-operation SEC NRS and GRS

How well does ORIGEN predict spent LWR [Light Water Reactor] fuel characteristics

International Nuclear Information System (INIS)

Mailen, J.C.; Roddy, J.W.

1987-01-01

The ORIGEN computer code is widely used to estimate the radionuclide content (actinides, activation and fission products) of irradiated reactor fuel and the resultant heat generation and radiation levels associated with such fuel. These estimates are used as source terms in safety evaluations of operating reactors, for evaluation of fuel behavior and regulation of the at-reactor storage, for transportation studies, and for evaluation of the ultimate geologic storage of the fuel. Calculated values determined using several variations of ORIGEN have been compared with experimentally determined values for actual fuel for many, but not all, of the parameters desired. In most cases, the comparisons did not use the most recent ORIGEN2 program, the most recent data libraries, or currently required quality assurance (QA) procedures. Comparisons of fuel composition data with ORIGEN2 are very limited, and the only data with proper QA are currently being acquired by Battelle Pacific Northwest Laboratory. This survey summarizes the fuel data available in the open literature and, where given, the calculated values by ORIGEN. Plans for additional analyses of well-characterized reactor fuel samples to improve the validation of ORIGEN2 are discussed

First steps towards a validation of the new burnup and depletion code TNT

Energy Technology Data Exchange (ETDEWEB)

Herber, S.C.; Allelein, H.J. [RWTH Aachen (Germany). Inst. for Reactor Safety and Reactor Technology; Research Center Juelich (Germany). Inst. for Energy and Climate Research - Nuclear Waste Disposal and Reactor Safety (IEK-6); Friege, N. [RWTH Aachen (Germany). Inst. for Reactor Safety and Reactor Technology; Kasselmann, S. [Research Center Juelich (Germany). Inst. for Energy and Climate Research - Nuclear Waste Disposal and Reactor Safety (IEK-6)

2012-11-01

In the frame of the fusion of the core design calculation capabilities, represented by V.S.O.P., and the accident calculation capabilities, represented by MGT(-3D), the successor of the TINTE code, difficulties were observed in defining an interface between a program backbone and the ORIGEN code respectively the ORIGENJUEL code. The estimation of the effort of refactoring the ORIGEN code or to write a new burnup code from scratch, led to the decision that it would be more efficient writing a new code, which could benefit from existing programming and software engineering tools from the computer code side and which can use the latest knowledge of nuclear reactions, e.g. consider all documented reaction channels. Therefore a new code with an object-oriented approach was developed at IEK-6. Object-oriented programming is currently state of the art and provides mostly an improved extensibility and maintainability. The new code was named TNT which stands for Topological Nuclide Transformation, since the code makes use of the real topology of the nuclear reactions. Here we want to present some first validation results from code to code benchmarks with the codes ORIGEN V2.2 and FISPACT2005 and whenever possible analytical results also used for the comparison. The 2 reference codes were chosen due to their high reputation in the field of fission reactor analysis (ORIGEN) and fusion facilities (FISPACT). (orig.)

The Management System of a TSO, the Example of GRS, Germany. Annex V

International Nuclear Information System (INIS)

2018-01-01

The Integrated Management System of GRS (Gesellschaft für Anlagen- und Reaktorsicherheit gGmbH) as well as its quality policy and the principles are described in the Management Manual with the current version being certified according to DIN EN ISO 9001:2015 by TÜV Rheinland Cert (Germany) and in the Organization Manual of GRS. The Management Manual of Quality (QMM) comprises process descriptions, technical instructions, and standard documents. The Management Manual is accessible via the intranet of GRS. The QMM as well as other information and support services can be delivered upon request to the following email-address: scc@grs.de. The central GRS intranet website of the Integrated Management System gives all staff at GRS quick access to all internal management documents. Through the internal online news bulletin the staff of GRS is informed by the Key Quality Manager (KQM) about the release of new versions of the QMM which is structured according to the table of contents provided.

ORIGEN-based Nuclear Fuel Inventory Module for Fuel Cycle Assessment: Final Project Report

Energy Technology Data Exchange (ETDEWEB)

Skutnik, Steven E. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering

2017-06-19

processing, and depletion/decay solvers) can be self-contained into a single executable sequence. Further, to embed this capability into other software environments (such as the Cyclus fuel cycle simulator) requires that Origen’s capabilities be encapsulated into a portable, self-contained library which other codes can then call directly through function calls, thereby directly accessing the solver and data processing capabilities of Origen. Additional components relevant to this work include modernization of the reactor data libraries used by Origen for conducting nuclear fuel depletion calculations. This work has included the development of new fuel assembly lattices not previously available (such as for CANDU heavy-water reactor assemblies) as well as validation of updated lattices for light-water reactors updated to employ modern nuclear data evaluations. The CyBORG reactor analysis module as-developed under this workscope is fully capable of dynamic calculation of depleted fuel compositions from all commercial U.S. reactor assembly types as well as a number of international fuel types, including MOX, VVER, MAGNOX, and PHWR CANDU fuel assemblies. In addition, the Origen-based depletion engine allows for CyBORG to evaluate novel fuel assembly and reactor design types via creation of Origen reactor data libraries via SCALE. The establishment of this new modeling capability affords fuel cycle modelers a substantially improved ability to model dynamically-changing fuel cycle and reactor conditions, including recycled fuel compositions from fuel cycle scenarios involving material recycle into thermal-spectrum systems.

On Rational Interpolation-Based List-Decoding and List-Decoding Binary Goppa Codes

DEFF Research Database (Denmark)

Beelen, Peter; Høholdt, Tom; Nielsen, Johan Sebastian Rosenkilde

2013-01-01

We derive the Wu list-decoding algorithm for generalized Reed–Solomon (GRS) codes by using Gröbner bases over modules and the Euclidean algorithm as the initial algorithm instead of the Berlekamp–Massey algorithm. We present a novel method for constructing the interpolation polynomial fast. We gi...... and a duality in the choice of parameters needed for decoding, both in the case of GRS codes and in the case of Goppa codes....

Development of a system of programs HAMMER - ORIGEN - 2 system

International Nuclear Information System (INIS)

Guimaraes, L.N.F.

1984-07-01

The development of a system of programs (Hamor-2) to calculate precision inventory of LWR reactors is presented. The Hamor-2 code generates with the Hammer-technion code the four (4) energy groupe effective cross section of the actinides: U-235, U-236, U-238, Pu-239, Pu-240, Pu-241, Pu-242. Hamor-2 collapses the four group energy effective cross section to one group energy effective cross section. Hamor-2 calculates with origen-2 the new concentrations for the actinides using the one group energy effective cross section. Two PWR reactors were analised. (HBR, KWO). The discrepancy between the final actinides conentration results calculated by Hamor-2 and the results of measures finded in the references were calculated. That discrepancies have been compared with the discrepancy between : the final actinide concentration results calculated by Origen-2 and some measured results finded in the references. (E.G.) [pt

New approach to Fork measurements data analysis by RADAR-CRISP and ORIGEN integration

International Nuclear Information System (INIS)

Vaccaro, S.; Svedkauskaite, J.; Smejkal, A.; Schwalbach, P.; De Baere, P.; Hu, J.; Gauld, I.C.

2013-06-01

Currently, in the EU, activities related to interim storage of spent fuel are constantly increasing. This is particularly true in Finland and Sweden, where final geological repository sites are planned to be operational in 2023 and 2026 respectively, but also in several other countries where fuel is moved from wet ponds to dry storage (Germany, Belgium, Spain, Czech, Bulgaria, etc). The required verification activities present a considerable challenge to the EURATOM Safeguards authority.. Both EURATOM and IAEA safeguards need to know what is in the storage casks and keep continuity of knowledge of the spent fuel. A frequently-used tool for the verification of the nuclear material during loading is the 'Fork' detectors for gross gamma and neutron counting. The IT applications RADAR (Remote Acquisition of Data and Review) and CRISP (Central RADAR Inspection Support Package), developed by EURATOM, are used to acquire safeguards measurement data and to analyze them in order to verify the declarations of the nuclear plant operators. Under the framework of the U.S. DOE-EURATOM Agreement on nuclear safeguards and security, a module for automated analysis of spent fuel measurement data using the ORIGEN (Oak Ridge Isotope Generation) code, part of the SCALE nuclear systems modeling and simulation package, has been integrated into CRISP. Measurement data are collected in an unattended mode by RADAR and then processed by CRISP, which outputs, for each fuel assembly, the measured gamma and neutron count rates. Simultaneously, ORIGEN performs burn-up calculations based on operator declarations previously entered into CRISP and calculates the expected neutron and gamma count rates for each assembly. These calculations also used response functions, developed using Monte Carlo modeling, to account for the detection probabilities of both neutron and photon particles that originated in each fuel pin. Finally, CRISP correlates and compares the expected (calculated) gamma

Development of ORIGEN libraries for mixed oxide (MOX) fuel assembly designs

International Nuclear Information System (INIS)

Mertyurek, Ugur; Gauld, Ian C.

2016-01-01

Highlights: • ORIGEN MOX library generation process is described. • SCALE burnup calculations are validated against measured MOX fuel samples from the MALIBU program. • ORIGEN MOX libraries are verified using the OECD Phase IV-B benchmark. • There is good agreement for calculated-to-measured isotopic distributions. - Abstract: ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup. The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. The nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.

Status of the ASTEC integral code

International Nuclear Information System (INIS)

Van Dorsselaere, J.P.; Jacq, F.; Allelein, H.J.

2000-01-01

The ASTEC (Accident Source Term Evaluation Code) integrated code is developed since 1997 in close collaboration by IPSN and GRS to predict an entire LWR severe accident sequence from the initiating event up to Fission Product (FP) release out of the containment. The applications of such a code are source term determination studies, scenario evaluations, accident management studies and Probabilistic Safety Assessment level 2 (PSA-2) studies. The version V0 of ASTEC is based on the RCS modules of the ESCADRE integrated code (IPSN) and on the upgraded RALOC and FIPLOC codes (GRS) for containment thermalhydraulics and aerosol behaviour. The latest version V0.2 includes the general feed-back from the overall validation performed in 1998 (25 separate-effect experiments, PHEBUS.FP FPT1 integrated experiment), some modelling improvements (i.e. silver-iodine reactions in the containment sump), and the implementation of the main safety systems for Severe Accident Management. Several reactor-applications are under way on French and German PWR, and on VVER-1000, all with a multi-compartment configuration of the containment. The total IPSN-GRS manpower involved in ASTEC project is today about 20 men/year. The main evolution of the next version V1, foreseen end of 2001, concerns the integration of the front-end phase and the improvement of the in-vessel degradation late-phase modelling. (author)

A SAS2H/KENO-V methodology for 3D fuel burnup analysis

International Nuclear Information System (INIS)

Milosevic, M.; Greenspan, E.; Vujic, J.

2002-01-01

An efficient methodology for 3D fuel burnup analysis of LWR reactors is described in this paper. This methodology is founded on coupling Monte Carlo method for 3D calculation of node power distribution, and transport method for depletion calculation in ID Wigner-Seitz equivalent cell for each node independently. The proposed fuel burnup modeling, based on application of SCALE-4.4a control modules SAS2H and KENO-V.a is verified for the case of 2D x-y model of IRIS 15 x 15 fuel assembly (with reflective boundary condition) by using two well benchmarked code systems. The one is MOCUP, a coupled MCNP-4C and ORIGEN2.1 utility code, and the second is KENO-V.a/ORIGEN2.1 code system recently developed by authors of this paper. The proposed SAS2H/KENO-V.a methodology was applied for 3D burnup analysis of IRIS-1000 benchmark.44 core. Detailed k sub e sub f sub f and power density evolution with burnup are reported. (author)

Development of the tool for generating ORIGEN2 library based on JENDL-3.2 for FBR

International Nuclear Information System (INIS)

Ohkawachi, Yasushi; Fukushima, Manabu

1999-05-01

ORIGEN2 is one of the most widely-used burnup analysis code in the world. This code has one-grouped cross section libraries compiled for various types of reactors. However, these libraries have some problems. One is that these libraries were developed from old nuclear data libraries (ENDF/B-IV,V) and the other is that core and fuel designs from which these libraries are generated do not match the current analysis. In order to solve the problems, analysis tool is developed for generating ORIGEN2 library from JENDL-3.2 considering multi-energy neutron spectrum. And eight new libraries are prepared using this tool for analysis of sodium-cooled FBR. These new libraries are prepared for eight kinds of cores in total. Seven of them are made by changing core size (small core - large core), fuel type (oxide, nitride, metal) and Pu vector as a parameter. The eighth one is a Pu burner core. Burnup calculation using both new and original libraries, shows large difference in buildup or depletion numbers of nuclides among the libraries. It is estimated that the analysis result is greatly influenced by the neutron spectrum which is used in collapse of cross section. By using this tool or new libraries, it seems to improve evaluation accuracy of buildup or depletion numbers of nuclides in transmutation research on FBR fuel cycle. (author)

SITA version 0. A simulation and code testing assistant for TOUGH2 and MARNIE

Energy Technology Data Exchange (ETDEWEB)

Seher, Holger; Navarro, Martin

2016-06-15

High quality standards have to be met by those numerical codes that are applied in long-term safety assessments for deep geological repositories for radioactive waste. The software environment SITA (''a simulation and code testing assistant for TOUGH2 and MARNIE'') has been developed by GRS in order to perform automated regression testing for the flow and transport simulators TOUGH2 and MARNIE. GRS uses the codes TOUGH2 and MARNIE in order to assess the performance of deep geological repositories for radioactive waste. With SITA, simulation results of TOUGH2 and MARNIE can be compared to analytical solutions and simulations results of other code versions. SITA uses data interfaces to operate with codes whose input and output depends on the code version. The present report is part of a wider GRS programme to assure and improve the quality of TOUGH2 and MARNIE. It addresses users as well as administrators of SITA.

DKPRO: A radionuclide decay and reprocessing code

International Nuclear Information System (INIS)

Wootan, D.; Schmittroth, F.A.

1997-01-01

The DKPRO code solves the general problem of modeling complex nuclear wastes streams using ORIGEN2 radionuclide production files. There is a continuing need for estimates of Hanford radionuclides. Physical measurements are one basis; calculational estimates, the approach represented here, are another. Given a known nuclear fuel history, it is relatively straightforward to calculate radionuclide inventories with codes such as the widely-used Oak Ridge National Laboratory code ORIGEN2

Determination of the radioactive inventory of a fuel assembly from a U3O8 design core using ORIGEN 2.1 code

International Nuclear Information System (INIS)

Castro, Jose; Ticona, Braulio; Madariaga, Marcelo

2014-01-01

This paper shows a methodology to determine the radioactive inventory of a fuel assembly of the RP-10 design core, which was proposed in 1988, using the ORIGEN 2.1 code, which allows to determine the activity of the 52 most characteristic fission products, its growth in activity during reactor operation under the terms of the design and evolution of decay of the fission products after 4 hours after the reactor shutdown, which conservatively, a fuel element represents an average fraction of the considered power in the radioactive inventory assessment. (authors).

The octopus burnup and criticality code system

Energy Technology Data Exchange (ETDEWEB)

Kloosterman, J.L.; Kuijper, J.C. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Leege, P.F.A. de

1996-09-01

The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional geometries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (author)

The OCTOPUS burnup and criticality code system

Energy Technology Data Exchange (ETDEWEB)

Kloosterman, J.L. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Kuijper, J.C. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.

1996-06-01

The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional goemetries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (orig.).

The octopus burnup and criticality code system

International Nuclear Information System (INIS)

Kloosterman, J.L.; Kuijper, J.C.; Leege, P.F.A. de.

1996-01-01

The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional geometries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (author)

The OCTOPUS burnup and criticality code system

International Nuclear Information System (INIS)

Kloosterman, J.L.; Kuijper, J.C.; Leege, P.F.A. de

1996-06-01

The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional goemetries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (orig.)

Infrared spectroscopy of the superluminal Galactic source GRS 1915+105 during the 1994 September outburst

DEFF Research Database (Denmark)

CastroTirado, A.J.; Geballe, T.R.; Lund, Niels

1996-01-01

observed in SS 433, suggesting that the ionized regions in the new source are not related to the twin beams of energetic particles that are believed to be responsible for the observed radio lobes. In contrast to Cygnus X-3, where the companion is likely to be a Wolf-Rayet star, we suggest that GRS 1915...

Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS). Annual report 2010/2011

International Nuclear Information System (INIS)

2011-01-01

The annual report 2010/11 of the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) covers the following topics: organization, corporate development and public relations; reactor accident in Fukushima; Projects and work priorities (reactor safety; waste management; radiation and environmental protection; international projects); project management and project support; equity participations and subsidiaries (RISKAUDIT IRSN/GRS International; institute for safety technologies (ISTec), European Nuclear Safety Training and Tutoring Institute (ENSTTI)).

Severe damage analysis of VVER 1000 following large break LOCA using Astec code

International Nuclear Information System (INIS)

Chatterjee, B.; Mukhopadhyay, D.; Lele, H.G.; Ghosh, A.K.; Kushwaha, H.S.

2007-01-01

Severe accident analysis of a reactor is an important aspect in the evaluation of source term. This in turn helps in emergency planning. An analysis has been carried out for VVER-1000 (V320) reactor following Large Break LOCA (loss of coolant accident) along with Station Blackout (SBO). Computer code ASTEC (jointly developed by IRSN, France, and GRS, Germany) is used for analyzing the transient. This integral code has been designed to be used as reference code for PSA2 studies. Severe accident analysis is carried out for an accident initiated by Large break LOCA along with SBO. Two cases have been analysed with the version ASTEC V1.2-rev1. In the first case hydro-accumulators are considered not available while the second case has been analysed with hydro accumulators. In this paper, ASTEC predictions have been studied for the in-vessel phase of the accident till vessel failure. The vessel failure was observed at 6979 s when accumulators were assumed not available. The vessel failure was quite delayed (19294 s) with operating accumulators. The hydrogen production was found to be very large (22% of total Zr inventory) in the case with accumulators compared to the case without accumulators (1.5% of total Zr inventory)

Use of WIMS-E lattice code for prediction of the transuranic source term for spent fuel dose estimation

International Nuclear Information System (INIS)

Schwinkendorf, K.N.

1996-01-01

A recent source term analysis has shown a discrepancy between ORIGEN2 transuranic isotopic production estimates and those produced with the WIMS-E lattice physics code. Excellent agreement between relevant experimental measurements and WIMS-E was shown, thus exposing an error in the cross section library used by ORIGEN2

Uncertainty Analysis of RBMK-Related Experimental Data

International Nuclear Information System (INIS)

Urbonas, Rolandas; Kaliatka, Algirdas; Liaukonis, Mindaugas

2002-01-01

An attempt to validate state-of-the-art thermal hydraulic code ATHLET (GRS, Germany) on the basis of E-108 test facility was made. Originally this code was developed and validated for different type reactors than RBMK. Since state-of-art thermal hydraulic codes are widely used for simulation of RBMK reactors, further codes' implementation and validation is required. The phenomena associated with channel type flow instabilities and CHF were found to be an important step in the frame of the overall effort of state-of-the-art validation and application for RBMK reactors. In the paper one-channel approach analysis is presented. Thus, the oscillatory behaviour of the system was not detected. The results show dependence on the nodalization used in the heated channels, initial and boundary conditions and code selected models. It is shown that the code is able to predict a sudden heat structure temperature excursion, when critical heat flux is approached. GRS developed uncertainty and sensitivity methodology was employed in the analysis. (authors)

Avances en investigación de la discapacidad intelectual de origen genético: tratamientos experimentales en el síndrome X frágil

Directory of Open Access Journals (Sweden)

Yolanda de Diego Otero

2017-06-01

Full Text Available De Diego Otero, Y. et al. (2017: “Avances en investigación de la discapacidad intelectual de origen genético: tratamientos experimentales en el síndrome X frágil”. Revista Española de Discapacidad, 5 (I: 217-227.

SWAT3.1 - the integrated burnup code system driving continuous energy Monte Carlo codes MVP and MCNP

International Nuclear Information System (INIS)

Suyama, Kenya; Mochizuki, Hiroki; Takada, Tomoyuki; Ryufuku, Susumu; Okuno, Hiroshi; Murazaki, Minoru; Ohkubo, Kiyoshi

2009-05-01

Integrated burnup calculation code system SWAT is a system that combines neutronics calculation code SRAC,which is widely used in Japan, and point burnup calculation code ORIGEN2. It has been used to evaluate the composition of the uranium, plutonium, minor actinides and the fission products in the spent nuclear fuel. Based on this idea, the integrated burnup calculation code system SWAT3.1 was developed by combining the continuous energy Monte Carlo code MVP and MCNP, and ORIGEN2. This enables us to treat the arbitrary fuel geometry and to generate the effective cross section data to be used in the burnup calculation with few approximations. This report describes the outline, input data instruction and several examples of the calculation. (author)

PRIMA-X Final Report

Energy Technology Data Exchange (ETDEWEB)

Lorenz, Daniel [German Research School for Simulation Sciences GmbH, Aachen (Germany); Wolf, Felix [German Research School for Simulation Sciences GmbH, Aachen (Germany)

2016-02-17

The PRIMA-X (Performance Retargeting of Instrumentation, Measurement, and Analysis Technologies for Exascale Computing) project is the successor of the DOE PRIMA (Performance Refactoring of Instrumentation, Measurement, and Analysis Technologies for Petascale Computing) project, which addressed the challenge of creating a core measurement infrastructure that would serve as a common platform for both integrating leading parallel performance systems (notably TAU and Scalasca) and developing next-generation scalable performance tools. The PRIMA-X project shifts the focus away from refactorization of robust performance tools towards a re-targeting of the parallel performance measurement and analysis architecture for extreme scales. The massive concurrency, asynchronous execution dynamics, hardware heterogeneity, and multi-objective prerequisites (performance, power, resilience) that identify exascale systems introduce fundamental constraints on the ability to carry forward existing performance methodologies. In particular, there must be a deemphasis of per-thread observation techniques to significantly reduce the otherwise unsustainable flood of redundant performance data. Instead, it will be necessary to assimilate multi-level resource observations into macroscopic performance views, from which resilient performance metrics can be attributed to the computational features of the application. This requires a scalable framework for node-level and system-wide monitoring and runtime analyses of dynamic performance information. Also, the interest in optimizing parallelism parameters with respect to performance and energy drives the integration of tool capabilities in the exascale environment further. Initially, PRIMA-X was a collaborative project between the University of Oregon (lead institution) and the German Research School for Simulation Sciences (GRS). Because Prof. Wolf, the PI at GRS, accepted a position as full professor at Technische Universität Darmstadt (TU

Methodology, status and plans for development and assessment of the code ATHLET

Energy Technology Data Exchange (ETDEWEB)

Teschendorff, V.; Austregesilo, H.; Lerchl, G. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH Forschungsgelaende, Garching (Germany)

1997-07-01

The thermal-hydraulic computer code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients) is being developed by the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) for the analysis of anticipated and abnormal plant transients, small and intermediate leaks as well as large breaks in light water reactors. The aim of the code development is to cover the whole spectrum of design basis and beyond design basis accidents (without core degradation) for PWRs and BWRs with only one code. The main code features are: advanced thermal-hydraulics; modular code architecture; separation between physical models and numerical methods; pre- and post-processing tools; portability. The code has features that are of special interest for applications to small leaks and transients with accident management, e.g. initialization by a steady-state calculation, full-range drift-flux model, dynamic mixture level tracking. The General Control Simulation Module of ATHLET is a flexible tool for the simulation of the balance-of-plant and control systems including the various operator actions in the course of accident sequences with AM measures. The code development is accompained by a systematic and comprehensive validation program. A large number of integral experiments and separate effect tests, including the major International Standard Problems, have been calculated by GRS and by independent organizations. The ATHLET validation matrix is a well balanced set of integral and separate effects tests derived from the CSNI proposal emphasizing, however, the German combined ECC injection system which was investigated in the UPTF, PKL and LOBI test facilities.

Methodology, status and plans for development and assessment of the code ATHLET

International Nuclear Information System (INIS)

Teschendorff, V.; Austregesilo, H.; Lerchl, G.

1997-01-01

The thermal-hydraulic computer code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients) is being developed by the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) for the analysis of anticipated and abnormal plant transients, small and intermediate leaks as well as large breaks in light water reactors. The aim of the code development is to cover the whole spectrum of design basis and beyond design basis accidents (without core degradation) for PWRs and BWRs with only one code. The main code features are: advanced thermal-hydraulics; modular code architecture; separation between physical models and numerical methods; pre- and post-processing tools; portability. The code has features that are of special interest for applications to small leaks and transients with accident management, e.g. initialization by a steady-state calculation, full-range drift-flux model, dynamic mixture level tracking. The General Control Simulation Module of ATHLET is a flexible tool for the simulation of the balance-of-plant and control systems including the various operator actions in the course of accident sequences with AM measures. The code development is accompained by a systematic and comprehensive validation program. A large number of integral experiments and separate effect tests, including the major International Standard Problems, have been calculated by GRS and by independent organizations. The ATHLET validation matrix is a well balanced set of integral and separate effects tests derived from the CSNI proposal emphasizing, however, the German combined ECC injection system which was investigated in the UPTF, PKL and LOBI test facilities

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.

Comparative analysis of a LOCA for a German PWR with ASTEC and ATHLET-CD

International Nuclear Information System (INIS)

Reinke, N.; Chan, H.W.; Sonnenkalb, M.

2013-01-01

This paper presents the results of a comparative analysis performed with ASTEC V2.02 and a coupled ATHLET-CD V2.2c /COCOSYS V2.4 calculation for a German 1300 MWe KONVOI type PWR. The purpose of this analysis is mainly to assess the ASTEC code behaviour in modelling of both the thermal-hydraulic phenomena in the coolant circuit arising during a hypothetical severe accident and the early phase of the core degradation versus the more mechanistic code system ATHLET-CD/COCOSYS. The performed analyses cover a loss of coolant accident sequence (LOCA). Such comparison has been done for the first time. The integral code ASTEC (Accident Source Term Evaluation Code) commonly developed since 1996 by IRSN and GRS is a fast running programme, which allows the calculation of entire sequences of severe accidents (SA) in light water reactors from the initiating event up to the release of fission products into the environment, thereby covering all important in-vessel and containment phenomena. The thermal-hydraulic mechanistic system code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients) is being developed by GRS for the analysis of the whole spectrum of leaks and transients in PWRs and BWRs. For modeling of core degradation processes the CD part (Core Degradation) of ATHLET can be activated. For analyses of the containment behavior, ATHLET-CD has been coupled to the GRS code COCOSYS (COntainment COde SYStem). (orig.)

X-ray image coding

International Nuclear Information System (INIS)

1974-01-01

The invention aims at decreasing the effect of stray radiation in X-ray images. This is achieved by putting a plate between source and object with parallel zones of alternating high and low absorption coefficients for X-radiation. The image is scanned with the help of electronic circuits which decode the signal space coded by the plate, thus removing the stray radiation

International activities of the GRS

International Nuclear Information System (INIS)

Antoni, R.; Hachenberg, G.; Schally, P.

1990-01-01

The individual activities fall into three main subject areas, each with differing objectives and functions: Cooperation in multilateral and bilateral bodies and organizations, cooperation in international research projects, performance of safety analyses. In addition to describing the international activities of GRS, the paper also shows how complexly interwoven the safety of nuclear facilities is at the international level, and how varied initiatives are required in order to ensure effective cooperation in the various fields. (orig.) [de

Application of the NJOY code for unresolved resonance treatment in the MCNP utility code

International Nuclear Information System (INIS)

Milosevic, M.; Greenspan, E.; Vujic, J. . E-mail addresses of corresponding authors: mmilos@vin.bg.ac.yu , vujic@nuc.berkeley.edu ,; Milosevic, M.; Vujic, J.)

2005-01-01

There are numerous uncertainties in the prediction of neutronic characteristics of reactor cores, particularly in the case of innovative reactor designs, arising from approximations used in the solution of the transport equation, and in nuclear data processing and cross section libraries generation. This paper describes the problems encountered in the analysis of the Encapsulated Nuclear Heat Source (ENHS) benchmark core and the new procedures and cross section libraries developed to overcome these problems. The ENHS is a new lead-bismuth or lead cooled novel reactor concept that is fuelled with metallic alloy of Pu, U and Zr, and it is designed to operate for 20 effective full power years without refuelling and with very small burnup reactivity swing. The computational tools benchmarked include: MOCUP - a coupled MCNP-4C and ORIGEN2.1 utility codes with MCNP data libraries based on the ENDF/B-VI evaluations; and KWO2 - a coupled KENO-V.a and ORIGEN2.1 code with ENDFB-V.2 based 238 group library. Calculations made for the ENHS benchmark have shown that the differences between the results obtained using different code systems and cross section libraries are significant and should be taken into account in assessing the quality of nuclear data libraries. (author)

Comparative analysis of the results obtained by computer code ASTEC V2 and RELAP 5.3.2 for small leak ID 80 for VVER 1000

International Nuclear Information System (INIS)

Atanasova, B.; Grudev, P.

2011-01-01

The purpose of this report is to present the results obtained by simulation and subsequent analysis of emergency mode for small leak with ID 80 for WWER 1000/B320 - Kozloduy NPP Units 5 and 6. Calculations were performed with the ASTEC v2 computer code used for calculation of severe accident, which was designed by French and German groups - IRSN and GRS. Integral RELAP5 computer code is used as a reference for comparison of results. The analyzes are focused on the processes occurring in reactor internals phase of emergency mode with significant core damage. The main thermohydraulic parameters, start of reactor core degradation and subsequent fuel relocalization till reactor vessel failure are evaluated in the analysis. RELAP5 computer code is used as a reference code to compare the results obtained till early core degradation that occurs after core stripping and excising of fuel temperature above 1200 0 C

Monitoring the Black Hole Binary GRS 1758-258 with INTEGRAL and RXTE

Science.gov (United States)

Pottschmidt, Katja; Chernyakova, Masha; Lubinski, Piotr; Migliari, Simone; Smith, David M.; Zdziarski, Andrzej A.; Tomsick, John A.; Bezayiff, N.; Kreykenbohm, Ingo; Kretschmar, Peter;

Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS). Annual report 2013/14; Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS). Jahresbericht 2013/14

Energy Technology Data Exchange (ETDEWEB)

NONE

2016-02-15

The annual report 2013/14 of the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) includes projects and key activities on the following issues: reactor safety, waste disposal, radiation protection, environmental protection and the support of foreign authorities and TSO. The project executing organizations, the governmental funding, share holding and subsidiaries are summarized.

NuSTAR SPECTROSCOPY OF GRS 1915+105: DISK REFLECTION, SPIN, AND CONNECTIONS TO JETS

Energy Technology Data Exchange (ETDEWEB)

Miller, J. M.; King, A. L. [Department of Astronomy, The University of Michigan, 500 Church Street, Ann Arbor, MI 48109-1046 (United States); Parker, M. L.; Fabian, A. C. [Institute of Astronomy, The University of Cambridge, Madingley Road, Cambridge CB3 OHA (United Kingdom); Fuerst, F.; Walton, D. J. [Cahill Center for Astronomy and Astrophysics, California Institute of Technology, Pasadena, CA, 91125 (United States); Bachetti, M.; Harrison, F. A.; Barret, D.; Grefenstette, B. W. [Universite de Toulouse, UPS-OMP, F-31400 Toulouse (France); Boggs, S. E.; Tomsick, J. A. [Space Sciences Laboratory, University of California, Berkeley, 7 Gauss Way, Berkeley, CA 94720-7450 (United States); Chakrabarty, D. [Kavli Institute for Astrophysics and Space Research, Massachusetts Institute of Technology, 70 Vassar Street, Cambridge, MA 02139 (United States); Christensen, F. E. [Danish Technical University, DK-2800 Lyngby (Denmark); Craig, W. W. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Hailey, C. J. [Columbia University, New York, NY 10027 (United States); Stern, D. K. [Jet Propulsion Laboratory, California Institute of Technology, 4800 Oak Grove Drive, Pasadena, CA 91109 (United States); Zhang, W. W., E-mail: jonmm@umich.edu [NASA Goddard Space Flight Center, Greenbelt, MD 20771 (United States)

2013-10-01

We report on the results of spectral fits made to a NuSTAR observation of the black hole GRS 1915+105 in a 'plateau' state. This state is of special interest because it is similar to the 'low/hard' state seen in other black holes, especially in that compact, steady jets are launched in this phase. The 3-79 keV bandpass of NuSTAR, and its ability to obtain moderate-resolution spectra free from distortions such as photon pile-up, are extremely well suited to studies of disk reflection in X-ray binaries. In only 15 ks of net exposure, an extraordinarily sensitive spectrum of GRS 1915+105 was measured across the full bandpass. Ionized reflection from a disk around a rapidly spinning black hole is clearly required to fit the spectra; even hybrid Comptonization models including ionized reflection from a disk around a Schwarzschild black hole proved inadequate. A spin parameter of a = 0.98 ± 0.01 (1σ statistical error) is measured via the best-fit model; low spins are ruled out at a high level of confidence. This result suggests that jets can be launched from a disk extending to the innermost stable circular orbit. A very steep inner disk emissivity profile is also measured, consistent with models of compact coronae above Kerr black holes. These results support an emerging association between the hard X-ray corona and the base of the relativistic jet.

Origen and the Platonic Tradition

Directory of Open Access Journals (Sweden)

Ilaria L.E. Ramelli

2017-02-01

Full Text Available This study situates Origen of Alexandria within the Platonic tradition, presenting Origenas a Christian philosopher who taught and studied philosophy, of which theology was part and parcel. More specifically, Origen can be described as a Christian Platonist. He criticized “false philosophies” as well as “heresies,” but not the philosophy of Plato. Against the background of recent scholarly debates, the thorny issue of the possible identity between Origen the Christian Platonist and Origen the Neoplatonist is partially addressed (although it requires a much more extensive discussion; it is also discussed in the light of Origen’s formation at Ammonius’s school and the reception of his works and ideas in “pagan” Platonism. As a consequence, and against scholarly perspectives that tend to see Christianity as anti-Platonism, the final section of this paper asks the question of what is imperial and late antique Platonism and, on the basis of rich evidence ,suggests that this was not only “pagan” institutional Platonism.

xLPR Scenario Analysis Report.

Energy Technology Data Exchange (ETDEWEB)

Eckert-Gallup, Aubrey Celia [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lewis, John R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Brooks, Dusty Marie [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Martin, Nevin [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hund, Lauren [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Clark, Andrew Jordan [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Mariner, Paul [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2017-03-01

This report describes the methods, results, and conclusions of the analysis of 11 scenarios defined to exercise various options available in the xLPR (Extremely Low Probability of Rupture) Version 2 .0 code. The scope of the scenario analysis is three - fold: (i) exercise the various options and components comprising xLPR v2.0 and defining each scenario; (ii) develop and exercise methods for analyzing and interpreting xLPR v2.0 outputs ; and (iii) exercise the various sampling options available in xLPR v2.0. The simulation workflow template developed during the course of this effort helps to form a basis for the application of the xLPR code to problems with similar inputs and probabilistic requirements and address in a systematic manner the three points covered by the scope.

MONTEBURNS 2.0: An Automated, Multi-Step Monte Carlo Burnup Code System

International Nuclear Information System (INIS)

2007-01-01

A - Description of program or function: MONTEBURNS Version 2 calculates coupled neutronic/isotopic results for nuclear systems and produces a large number of criticality and burnup results based on various material feed/removal specifications, power(s), and time intervals. MONTEBURNS is a fully automated tool that links the LANL MCNP Monte Carlo transport code with a radioactive decay and burnup code. Highlights on changes to Version 2 are listed in the transmittal letter. Along with other minor improvements in MONTEBURNS Version 2, the option was added to use CINDER90 instead of ORIGEN2 as the depletion/decay part of the system. CINDER90 is a multi-group depletion code developed at LANL and is not currently available from RSICC, nor from the NEA Databank. This MONTEBURNS release was tested with various combinations of CCC-715/MCNPX 2.4.0, CCC-710/MCNP5, CCC-700/MCNP4C, CCC-371/ORIGEN2.2, ORIGEN2.1 and CINDER90. Perl is required software and is not included in this distribution. MCNP, ORIGEN2, and CINDER90 are not included. The following changes have been made: 1) An increase in the number of removal group information that must be provided for each material in each step in the feed input file. 2) The capability to use CINDER90 instead of ORIGEN2.1 as the depletion/decay part of the code. 3) ORIGEN2.2 can also be used instead of ORIGEN2.1 in Monteburns. 4) The correction of including the capture cross sections to metastable as well as ground states if applicable for an isotope (i.e. Am-241 and Am-243 in particular). 5) The ability to use a MCNP input file that has a title card starting with 'm' (this was a bug in the first version of Monteburns). 6) A decrease in run time for cases involving decay-only steps (power of 0.0). Monteburns does not run MCNP to calculate cross sections for a step unless it is an irradiation step. 7) The ability to change the cross section libraries used each step. If different cross section libraries are desired for multiple steps. 8

Coded diffraction system in X-ray crystallography using a boolean phase coded aperture approximation

Science.gov (United States)

Pinilla, Samuel; Poveda, Juan; Arguello, Henry

2018-03-01

Phase retrieval is a problem present in many applications such as optics, astronomical imaging, computational biology and X-ray crystallography. Recent work has shown that the phase can be better recovered when the acquisition architecture includes a coded aperture, which modulates the signal before diffraction, such that the underlying signal is recovered from coded diffraction patterns. Moreover, this type of modulation effect, before the diffraction operation, can be obtained using a phase coded aperture, just after the sample under study. However, a practical implementation of a phase coded aperture in an X-ray application is not feasible, because it is computationally modeled as a matrix with complex entries which requires changing the phase of the diffracted beams. In fact, changing the phase implies finding a material that allows to deviate the direction of an X-ray beam, which can considerably increase the implementation costs. Hence, this paper describes a low cost coded X-ray diffraction system based on block-unblock coded apertures that enables phase reconstruction. The proposed system approximates the phase coded aperture with a block-unblock coded aperture by using the detour-phase method. Moreover, the SAXS/WAXS X-ray crystallography software was used to simulate the diffraction patterns of a real crystal structure called Rhombic Dodecahedron. Additionally, several simulations were carried out to analyze the performance of block-unblock approximations in recovering the phase, using the simulated diffraction patterns. Furthermore, the quality of the reconstructions was measured in terms of the Peak Signal to Noise Ratio (PSNR). Results show that the performance of the block-unblock phase coded apertures approximation decreases at most 12.5% compared with the phase coded apertures. Moreover, the quality of the reconstructions using the boolean approximations is up to 2.5 dB of PSNR less with respect to the phase coded aperture reconstructions.

Determination of the decay power for a U3O8 designed core using the ORIGEN 2.1 code

International Nuclear Information System (INIS)

Castro, Jose; Gallardo, Alberto; Madariaga, Marcelo

2014-01-01

After the operation of a nuclear research reactor at a higher power (more than 300 kW), a cooling time is required to remove the residual heat from the core due to the heat produced by the energy emitted by fission products, this fact is common in reactors. There is a short time where the heat output falls to 6 % after the reactor shutdown, the importance of knowing this power is because of the accidental events that this power could cause and affect the fuel after a sudden shutdown in the cooling system of the reactor and there is any other refrigeration system, only that one surrounding the reactor core. This report shows the results of the calculation of the U 3 O 8 core residual power a for the RP-10, using the ORIGEN 2.1 calculation code, verifying the safety of the proposed core within the safety limits accepted for the reactor. (authors).

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.

Computer code validation by high temperature chemistry

International Nuclear Information System (INIS)

Alexander, C.A.; Ogden, J.S.

1988-01-01

At least five of the computer codes utilized in analysis of severe fuel damage-type events are directly dependent upon or can be verified by high temperature chemistry. These codes are ORIGEN, CORSOR, CORCON, VICTORIA, and VANESA. With the exemption of CORCON and VANESA, it is necessary that verification experiments be performed on real irradiated fuel. For ORIGEN, the familiar knudsen effusion cell is the best choice and a small piece of known mass and known burn-up is selected and volatilized completely into the mass spectrometer. The mass spectrometer is used in the integral mode to integrate the entire signal from preselected radionuclides, and from this integrated signal the total mass of the respective nuclides can be determined. For CORSOR and VICTORIA, experiments with flowing high pressure hydrogen/steam must flow over the irradiated fuel and then enter the mass spectrometer. For these experiments, a high pressure-high temperature molecular beam inlet must be employed. Finally, in support of VANESA-CORCON, the very highest temperature and molten fuels must be contained and analyzed. Results from all types of experiments will be discussed and their applicability to present and future code development will also be covered

GRS annual report 1991/92

International Nuclear Information System (INIS)

1992-01-01

Even in the year under report, the publications of the Society for Reactor Safety (GRS) - a consulting firm in matters of science and technology - were a substantial contribution to the further optimization of safety in nuclear installations. The basic principle in this respect is that statements made with regard to safety should be backed by comprehensive research and development (R+D) done by the society itself. The centralized registration, assessment and utilization of operationa experience, incidents and accidents consitute the essential element of the safety-relevant evaluation of nuclear power stations. (orig./HP) [de

Severe accident analysis in a two-loop PWR nuclear power plant with the ASTEC code

International Nuclear Information System (INIS)

Sadek, Sinisa; Amizic, Milan; Grgic, Davor

2013-01-01

The ASTEC/V2.0 computer code was used to simulate a hypothetical severe accident sequence in the nuclear power plant Krsko, a 2-loop pressurized water reactor (PWR) plant. ASTEC is an integral code jointly developed by Institut de Radioprotection et de Surete Nucleaire (IRSN, France) and Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS, Germany) to assess nuclear power plant behaviour during a severe accident. The analysis was conducted in 2 steps. First, the steady state calculation was performed in order to confirm the applicability of the plant model and to obtain correct initial conditions for the accident analysis. The second step was the calculation of the station blackout accident with a leakage of the primary coolant through degraded reactor coolant pump seals, which was a small LOCA without makeup capability. Two scenarios were analyzed: one with and one without the auxiliary feedwater (AFW). The latter scenario, without the AFW, resulted in earlier core damage. In both cases, the accident ended with a core melt and a reactor pressure vessel failure with significant release of hydrogen. In addition, results of the ASTEC calculation were compared with results of the RELAP5/SCDAPSIM calculation for the same transient scenario. The results comparison showed a good agreement between predictions of those 2 codes. (orig.)

Parameters of the CGCS 2000 Ellipsoid and Comparisons with GRS 80 and WGS 84

Directory of Open Access Journals (Sweden)

CHENG Pengfei

2016-02-01

Full Text Available According to the definition of China Geodetic Coordinate System 2000(CGCS 2000 and defined constants of the ellipsoid adopted by CGCS 2000,the other geometrical and physical parameters of this ellipsoid are derived and compared with that from GRS 80 and WGS 84,respectively.Meanwhile the coordinates and normal gravity on the CGCS 2000 ellipsoid are compared with that on WGS 84 and GRS 80.The difference between the normal gravity on CGCS 2000 ellipsoid and that on GRS 80 is about -143.54×10-8 m/s2,while it is 0.02×10-8 m/s2 compared to WGS 84.The longitudes of a point on these three ellipsoids are the same,but the maximum difference of latitude between CGCS 2000 and GRS 80 is 8.26×10-11 arc seconds,which is about 2.5×10-6 mm,and the maximum difference of latitude between CGCS 2000 and WGS 84 is 3.6×10-6 arc seconds,which is about 0.11 mm.

GRS annual report 1981

International Nuclear Information System (INIS)

The Society and its staff have contributed to further improving the high degree of the achieved safety in nuclear engineering. On the national and international level the GRS keeps a steady position among the leading organizations for safety in nuclear engineering. Its main activities are: thermohydraulics, structural dynamics, reliability, man/machine communication and operating experiences. This technical orientation is supplemented by the overlapping effort of creating a balanced safety level in all fields of nuclear engineering. The technical control equipment has been considerably developed and improved during the past years. Emphasis has been laid on increased safety in the quality of the components and on the continued training and advanced training of the staff. (orig./HSCH) [de

Environmental measurement-while-drilling-gamma ray spectrometer (EMWD-GRS) system technology demonstration plan for use at the Savannah River Site F-Area Retention Basin

International Nuclear Information System (INIS)

Williams, C.V.; Lockwood, G.J.; Normann, R.A.; Gruebel, R.D.

1996-08-01

The Environmental Measurement-While-Drilling-Gamma Ray Spectrometer (EMWD-GRS) system represents an innovative blend of new and existing technology that provides the capability of producing real-time environmental and drillbit data during drilling operations. This demonstration plan presents information on the EMWD-GRS technology, demonstration design, Cs-137 contamination at the Savannah River Site F-Area Retention Basin, responsibilities of demonstration participants, and the policies and procedures for the demonstration to be conducted at the Savannah River Site F-Area Retention Basin. The EMWD-GRS technology demonstration will consist of continuously monitoring for gamma-radiation contamination while drilling two horizontal boreholes below the backfilled retention basin. These boreholes will pass near previously sampled vertical borehole locations where concentrations of contaminant levels are known. Contaminant levels continuously recorded by the EMWD-GRS system during drilling will be compared to contaminant levels previously determined through quantitative laboratory analysis of soil samples

Environmental measurement-while-drilling-gamma ray spectrometer (EMWD-GRS) system technology demonstration plan for use at the Savannah River Site F-Area Retention Basin

Energy Technology Data Exchange (ETDEWEB)

Williams, C.V.; Lockwood, G.J.; Normann, R.A. [Sandia National Labs., Albuquerque, NM (United States); Gruebel, R.D. [Tech Reps, Inc., Albuquerque, NM (United States)

1996-08-01

The Environmental Measurement-While-Drilling-Gamma Ray Spectrometer (EMWD-GRS) system represents an innovative blend of new and existing technology that provides the capability of producing real-time environmental and drillbit data during drilling operations. This demonstration plan presents information on the EMWD-GRS technology, demonstration design, Cs-137 contamination at the Savannah River Site F-Area Retention Basin, responsibilities of demonstration participants, and the policies and procedures for the demonstration to be conducted at the Savannah River Site F-Area Retention Basin. The EMWD-GRS technology demonstration will consist of continuously monitoring for gamma-radiation contamination while drilling two horizontal boreholes below the backfilled retention basin. These boreholes will pass near previously sampled vertical borehole locations where concentrations of contaminant levels are known. Contaminant levels continuously recorded by the EMWD-GRS system during drilling will be compared to contaminant levels previously determined through quantitative laboratory analysis of soil samples.

Radionuclide inventories : ORIGEN2.2 isotopic depletion calculation for high burnup low-enriched uranium and weapons-grade mixed-oxide pressurized-water reactor fuel assemblies.

Energy Technology Data Exchange (ETDEWEB)

Gauntt, Randall O.; Ross, Kyle W. (Los Alamos National Laboratory, Los Alamos, NM); Smith, James Dean; Longmire, Pamela

2010-04-01

The Oak Ridge National Laboratory computer code, ORIGEN2.2 (CCC-371, 2002), was used to obtain the elemental composition of irradiated low-enriched uranium (LEU)/mixed-oxide (MOX) pressurized-water reactor fuel assemblies. Described in this report are the input parameters for the ORIGEN2.2 calculations. The rationale for performing the ORIGEN2.2 calculation was to generate inventories to be used to populate MELCOR radionuclide classes. Therefore the ORIGEN2.2 output was subsequently manipulated. The procedures performed in this data reduction process are also described herein. A listing of the ORIGEN2.2 input deck for two-cycle MOX is provided in the appendix. The final output from this data reduction process was three tables containing the radionuclide inventories for LEU/MOX in elemental form. Masses, thermal powers, and activities were reported for each category.

System code improvements for modelling passive safety systems and their validation

Energy Technology Data Exchange (ETDEWEB)

Buchholz, Sebastian; Cron, Daniel von der; Schaffrath, Andreas [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

2016-11-15

GRS has been developing the system code ATHLET over many years. Because ATHLET, among other codes, is widely used in nuclear licensing and supervisory procedures, it has to represent the current state of science and technology. New reactor concepts such as Generation III+ and IV reactors and SMR are using passive safety systems intensively. The simulation of passive safety systems with the GRS system code ATHLET is still a big challenge, because of non-defined operation points and self-setting operation conditions. Additionally, the driving forces of passive safety systems are smaller and uncertainties of parameters have a larger impact than for active systems. This paper addresses the code validation and qualification work of ATHLET on the example of slightly inclined horizontal heat exchangers, which are e. g. used as emergency condensers (e. g. in the KERENA and the CAREM) or as heat exchanger in the passive auxiliary feed water systems (PAFS) of the APR+.

Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS). Annual report 1999

International Nuclear Information System (INIS)

Butz, H.P.; Monheimius, I.; Berberich, G.

2000-01-01

The aim is to place GRS in a position to be able to adapt to the changed tasks in the nuclear sector, remain competent in all relevant issues of nuclear safety, and maintain an adequate number of qualified staff in spite of the extremely difficult situation with regard to finding junior nuclear engineers. Key elements of this concept are - an adaptation to the expected medium-term reduction of nuclear tasks by socially acceptable cuts in the number of staff over the coming five years, - a concentration of our own competencies, - partnerships with other capable organisations at home and abroad. The major tasks of GRS will remain to perform safety analyses and assessments in the fields of reactor safety, fuel supply, spent fuel management and disposal as well as to carry out nuclear and final repository safety research. (orig.) [de

CFD and system analysis code investigations of the multidimensional flow mixing phenomena in the reactor pressure vessel

International Nuclear Information System (INIS)

Ceuca, S.C.; Herb, J.; Schoeffel, P.J.; Hollands, T.; Austregesilo, H.; Hristov, H.V.

2017-01-01

The realistic numerical prediction of transient fluid-dynamic scenarios including the complex, three-dimensional flow mixing phenomena occurring in the reactor pressure vessel (RPV) both in normal or abnormal operation are an important issue in today's reactor safety assessment studies. Both Computational Fluid Dynamics (CFD) tools as well as fluid-dynamic system analysis codes, each with its advantages and drawbacks, are commonly used to model such transients. Simulation results obtained with the open-source CFD tool-box OpenFOAM and the German thermal-hydraulic system code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients), the later developed by Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) for the analysis of the whole spectrum of operational transients, design-basis accidents and beyond design basis accidents anticipated for nuclear energy facilities, are compared against experimental data from the ROssendorf Coolant Mixing (ROCOM) test facility. In the case of the OpenFOAM CFD simulations the influence of various turbulence models and numerical schemes has been assessed while in the case of the system analysis code ATHLET a multidimensional nodalization recommended for real power plant applications has been employed. The simulation results show a good agreement with the experimental data, indicating that both OpenFOAM and ATHLET can capture the key flow features of the mixing processes in the Reactor Pressure Vessel (RPV). (author)

Female-biased expression of long non-coding RNAs in domains that escape X-inactivation in mouse

Directory of Open Access Journals (Sweden)

Lu Lu

2010-11-01

Full Text Available Abstract Background Sexual dimorphism in brain gene expression has been recognized in several animal species. However, the relevant regulatory mechanisms remain poorly understood. To investigate whether sex-biased gene expression in mammalian brain is globally regulated or locally regulated in diverse brain structures, and to study the genomic organisation of brain-expressed sex-biased genes, we performed a large scale gene expression analysis of distinct brain regions in adult male and female mice. Results This study revealed spatial specificity in sex-biased transcription in the mouse brain, and identified 173 sex-biased genes in the striatum; 19 in the neocortex; 12 in the hippocampus and 31 in the eye. Genes located on sex chromosomes were consistently over-represented in all brain regions. Analysis on a subset of genes with sex-bias in more than one tissue revealed Y-encoded male-biased transcripts and X-encoded female-biased transcripts known to escape X-inactivation. In addition, we identified novel coding and non-coding X-linked genes with female-biased expression in multiple tissues. Interestingly, the chromosomal positions of all of the female-biased non-coding genes are in close proximity to protein-coding genes that escape X-inactivation. This defines X-chromosome domains each of which contains a coding and a non-coding female-biased gene. Lack of repressive chromatin marks in non-coding transcribed loci supports the possibility that they escape X-inactivation. Moreover, RNA-DNA combined FISH experiments confirmed the biallelic expression of one such novel domain. Conclusion This study demonstrated that the amount of genes with sex-biased expression varies between individual brain regions in mouse. The sex-biased genes identified are localized on many chromosomes. At the same time, sexually dimorphic gene expression that is common to several parts of the brain is mostly restricted to the sex chromosomes. Moreover, the study uncovered

Validation of CBZ code system for post-irradiation examination analysis and sensitivity analysis of (n,γ) branching ratio

International Nuclear Information System (INIS)

Kawamoto, Yosuke; Chiba, Go; Tsuji, Masashi; Narabayashi, Tadashi

2013-01-01

A code system CBZ is being developed in Hokkaido University. In order to validate it, PIE data, which are nuclide composition data of a spent fuel, have been analyzed with CBZ. The validity is evaluated as ratios of the calculation values to the experimental ones (C/E ratios). Differences between experimental values and calculation ones are smaller than 20% except some nuclides. Thus this code system is validated. Additionally, we evaluate influence of change of (n,γ) branching ratio on inventories of fission products and actinides. As a result, branching ratios of Sb-121, Pm-147, and Am-241 influence inventories of several nuclides. We perform PIE analysis using different (n,γ) branching ratio data from the ORIGEN-2 library, JNDC-Ver.2, and JEFF-3.1A, and find that differences in (n,γ) branching ratios between different nuclear libraries have a non-negligible influence on inventories of several nuclides. (author)

Coding aperture applied to X-ray imaging

International Nuclear Information System (INIS)

Brunol, J.; Sauneuf, R.; Gex, J.P.

1980-05-01

We present some X-ray images of grids and plasmas. These images were obtained by using a single circular slit (annular code) as coding aperture and a computer decoding process. The experimental resolution is better than 10μm and it is expected to be in the order of 2 or 3 μm with the same code and an improved decoding process

Preservation of competence within the framework of knowledge management of GRS

International Nuclear Information System (INIS)

Beraha, D.; Erven, U.; Puhr-Westerheide, P.

2006-01-01

As a consequence of the loss of competence threatening to arise from the retirement of many experts, countermeasures in the form of expanded and improved training schemes and the development of knowledge management have been introduced by the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH. The new modules of basic as well as specialized training are described. A brief outline is given of the GRS Intranet portal with its integrated document management. The development of a portal for each project allows all project-related information to be combined. An account is presented of initial experience in using methods of knowledge representation and display, and an outlook is given into expected further developments. (orig.)

Comparison of Time/Phase Lags in the Hard State and Plateau State of GRS 1915+105

NARCIS (Netherlands)

Pahari, M.; Neilsen, J.; Yadav, J.S.; Misra, R.; Uttley, P.

2013-01-01

We investigate the complex behavior of energy- and frequency-dependent time/phase lags in the plateau state and the radio-quiet hard (χ) state of GRS 1915+105. In our timing analysis, we find that when the source is faint in the radio, quasi-periodic oscillations (QPOs) are observed above 2 Hz and

2017 Laser Diagnostics in Combustion GRC/GRS

Science.gov (United States)

2017-10-06

scientists and engineers . The goal of GRC was to bring together scientists and engineers with a strong interest in laser-based combustion diagnostics to...processes. The GRS, held the weekend before the GRC, was organized to bring together emerging leaders in the field at an early stage in their careers ...coupled multi-scale combustion processes. The GRC on Laser Diagnostics in Combustion was designed to bring together scientists and engineers working at

Single Use Letter Report for the Verification and Validation of the RADNUC-2A and ORIGEN2 S.2 Computer Codes

International Nuclear Information System (INIS)

PACKER, M.J.

2000-01-01

This report documents the verification and validation (VandV) activities undertaken to support the use of the RADNUC2-A and ORIGEN2 S.2 computer codes for the specific application of calculating isotopic inventories and decay heat loadings for Spent Nuclear Fuel Project (SNFP) activities as described herein. Two recent applications include the reports HNF-SD-SNF-TI-009, 105-K Basin Material Design Basis Feed Description for Spent Nuclear Fuel Project Facilities, Volume 1, Fuel (Praga, 1998), and HNF-3035, Rev. 0B, MCO Gas Composition for Low Reactive Surface Areas (Packer, 1998). Representative calculations documented in these two reports were repeated using RADNUC2-A, and the results were identical to the documented results. This serves as verification that version 2A of Radnuc was used for the applications noted above; the same version was tested herein, and perfect agreement was shown. Comprehensive VandV is demonstrated for RADNUC2-A in Appendix A

Modification of an x-ray diffraction unit to comply with the NH and MRC code of practice

International Nuclear Information System (INIS)

Ibbetson, V.J.; Young, J.G.

2004-01-01

X-ray analysis units are commonly used in research and industrial laboratories throughout Australia. Despite a well-established Code of Practice and working protocols for the safe use of such units, there are all too many stories of users by-passing safety features significantly increasing the risk of accidental exposure to the primary X-ray beam. Since the output of such units may be as high as 300 Gy x s 1 , such accidental exposures could have very serious consequences. Australian Radiation Services Pty Ltd undertook a compliance audit of an X-ray diffraction unit with respect to the NH and MRC Code of Practice for protection against ionising radiation emitted from X-ray analysis equipment. This paper discusses the findings from the initial inspection and the modifications recommended for the XRD unit to ensure compliance with the Code, without unnecessarily restricting its use. Copyright (2004) Australasian Radiation Protection Society Inc

Application of ASTEC, MELCOR, and MAAP Computer Codes for Thermal Hydraulic Analysis of a PWR Containment Equipped with the PCFV and PAR Systems

Directory of Open Access Journals (Sweden)

Siniša Šadek

2017-01-01

Full Text Available The integrity of the containment will be challenged during a severe accident due to pressurization caused by the accumulation of steam and other gases and possible ignition of hydrogen and carbon monoxide. Installation of a passive filtered venting system and passive autocatalytic recombiners allows control of the pressure, radioactive releases, and concentration of flammable gases. Thermal hydraulic analysis of the containment equipped with dedicated passive safety systems after a hypothetical station blackout event is performed for a two-loop pressurized water reactor NPP with three integral severe accident codes: ASTEC, MELCOR, and MAAP. MELCOR and MAAP are two major US codes for severe accident analyses, and the ASTEC code is the European code, joint property of Institut de Radioprotection et de Sûreté Nucléaire (IRSN, France and Gesellschaft für Anlagen und Reaktorsicherheit (GRS, Germany. Codes’ overall characteristics, physics models, and the analysis results are compared herein. Despite considerable differences between the codes’ modelling features, the general trends of the NPP behaviour are found to be similar, although discrepancies related to simulation of the processes in the containment cavity are also observed and discussed in the paper.

A first accident simulation for Angra-1 power plant using the ALMOD computer code

International Nuclear Information System (INIS)

Camargo, C.T.M.

1981-02-01

The acquisition of the Almod computer code from GRS-Munich to CNEN has permited doing calculations of transients in PWR nuclear power plants, in which doesn't occur loss of coolant. The implementation of the german computer code Almod and its application in the calculation of Angra-1, a nuclear power plant different from the KWU power plants, demanded study and models adaptation; and due to economic reasons simplifications and optimizations were necessary. The first results define the analytical potential of the computer code, confirm the adequacy of the adaptations done and provide relevant conclusions about the Angra-1 safety analysis, showing at the same time areas in which the model can be applied or simply improved. (Author) [pt

Analysis of burnup of Angra 2 PWR nuclear with addition of thorium dioxide fuel using ORIGEN-ARP

Energy Technology Data Exchange (ETDEWEB)

Goncalves, Isadora C.; Wichrowski, Caio C.; Oliveira, Claudio L. de; Vellozo, Sergio O.; Baptista, Camila O., E-mail: isadora.goncalves@ime.eb.br, E-mail: wichrowski@ime.eb.br, E-mail: d7luiz@yahoo.com.br, E-mail: vellozo@ime.eb.br, E-mail: camila.oliv.baptista@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Secao de Engenharia Nuclear

2017-11-01

It is known that isotope {sup 232}thorium is a fertile nuclide with the ability to convert into {sup 233}uranium, a potentially fissile isotope, after absorbing a neutron. As there is a large stock of available thorium in the world, this element shows great promise in mitigate the world energy crisis, more particularly in the problem of uranium scarcity, besides being an alternative nuclear fuel for those currently used in reactors, and yet presenting advantages as an option for the non-proliferation movement, among others. In this study, the analysis of the remaining nuclides of burnup was carried out for the core configuration of a PWR (pressurized water reactor) reactor, specifically the Angra 2 reactor, using only uranium dioxide, its current configuration, and in different configurations including a mixed oxide of uranium and thorium in three concentrations, allowing a preliminary assessment of the feasibility of the modification of the fuel, the resulting production of {sup 233}uranium, the emergence of {sup 231}protactinium (an isotope that only occurs as a fission product of {sup 232}Th) resulting from burning. The study was carried out using data obtained from FSAR (Final Safety Analysis Report) of Angra 2, using the SCALE 6.1, a modeling and simulation nuclear code, especially its ORIGEN-ARP module, which analyzes the depletion of isotopes presents in a reactor. (author)

Analysis of burnup of Angra 2 PWR nuclear with addition of thorium dioxide fuel using ORIGEN-ARP

International Nuclear Information System (INIS)

Goncalves, Isadora C.; Wichrowski, Caio C.; Oliveira, Claudio L. de; Vellozo, Sergio O.; Baptista, Camila O.

2017-01-01

It is known that isotope "2"3"2thorium is a fertile nuclide with the ability to convert into "2"3"3uranium, a potentially fissile isotope, after absorbing a neutron. As there is a large stock of available thorium in the world, this element shows great promise in mitigate the world energy crisis, more particularly in the problem of uranium scarcity, besides being an alternative nuclear fuel for those currently used in reactors, and yet presenting advantages as an option for the non-proliferation movement, among others. In this study, the analysis of the remaining nuclides of burnup was carried out for the core configuration of a PWR (pressurized water reactor) reactor, specifically the Angra 2 reactor, using only uranium dioxide, its current configuration, and in different configurations including a mixed oxide of uranium and thorium in three concentrations, allowing a preliminary assessment of the feasibility of the modification of the fuel, the resulting production of "2"3"3uranium, the emergence of "2"3"1protactinium (an isotope that only occurs as a fission product of "2"3"2Th) resulting from burning. The study was carried out using data obtained from FSAR (Final Safety Analysis Report) of Angra 2, using the SCALE 6.1, a modeling and simulation nuclear code, especially its ORIGEN-ARP module, which analyzes the depletion of isotopes presents in a reactor. (author)

Development of new ORIGEN2 data library sets for research reactors with light water cooled oxide and silicide LEU (20 w/o) fuels based on JENDL-3.3 nuclear data

International Nuclear Information System (INIS)

Liem, Peng Hong; Sembiring, Tagor Malem

2013-01-01

Highlights: • We developed new ORIGEN2 data library sets for research reactors based on JENDL-3.3. • The sets cover oxide and silicide LEU fuels with meat density up to 4.74 g U/cm 3 . • Two kinds of data library sets are available: fuel region and non-fuel regions. • We verified the new data library sets with other codes. • We validated the new data library against a non-destructive test. -- Abstract: New sets of ORIGEN2 data library dedicated to research/testing reactors with light water cooled oxide and silicide LEU fuel plates based on JENDL-3.3 nuclear data were developed, verified and validated. The new sets are considered to be an extension of the most recent release of ORIGEN2.2UPJ code, i.e. the ORLIBJ33 library sets. The newly generated ORIGEN2 data library sets cover both oxide and silicide LEU fuels with fuel meat density range from 2.96 to 4.74 g U/cm 3 used in the present and future operation of the Indonesian 30 MWth RSG GAS research reactor. The new sets are expected applicable also for other research/testing reactors which utilize similar fuels or have similar neutron spectral indices. In addition to the traditional ORIGEN2 library sets for fuel depletion analyses in fuel regions, in the new data library sets, new ORIGEN2 library sets for irradiation/activation analyses were also prepared which cover all representative non-fuel regions of RSG GAS such as reflector elements, irradiation facilities, etc. whose neutron spectra are significantly softer than fuel regions. Verification with other codes as well as validation with a non-destructive test result showed promising results where a good agreement was confirmed

X447 EBR-II Experiment Benchmark for Verification of Audit Code of SFR Metal Fuel

Energy Technology Data Exchange (ETDEWEB)

Choi, Yong Won; Bae, Moo-Hoon; Shin, Andong; Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2016-10-15

In KINS (Korea Institute of Nuclear Safety), to prepare audit calculation of PGSFR licensing review, the project has been started to develop the regulatory technology for SFR system including a fuel area. To evaluate the fuel integrity and safety during an irradiation, the fuel performance code must be used for audit calculation. In this study, to verify the new code system, the benchmark analysis is performed. In the benchmark, X447 EBR-II experiment data are used. Additionally, the sensitivity analysis according to mass flux change of coolant is performed. In case of LWR fuel performance modeling, various and advanced models have been proposed and validated based on sufficient in-reactor test results. However, due to the lack of experience of SFR operation, the current understanding of SFR fuel behavior is limited. In this study, X447 EBR-II Experiment data are used for benchmark. The fuel composition of X447 assembly is U-10Zr and PGSFR also uses this composition in initial phase. So we select X447 EBR-II experiment for benchmark analysis. Due to the lack of experience of SFR operation and data, the current understanding of SFR fuel behavior is limited. However, in order to prepare the licensing of PGSFR, regulatory audit technologies of SFR must be secured. So, in this study, to verify the new audit fuel performance analysis code, the benchmark analysis is performed using X447 EBR-II experiment data. Also, the sensitivity analysis with mass flux change of coolant is performed. In terms of verification, it is considered that the results of benchmark and sensitivity analysis are reasonable.

X447 EBR-II Experiment Benchmark for Verification of Audit Code of SFR Metal Fuel

International Nuclear Information System (INIS)

Choi, Yong Won; Bae, Moo-Hoon; Shin, Andong; Suh, Namduk

2016-01-01

In KINS (Korea Institute of Nuclear Safety), to prepare audit calculation of PGSFR licensing review, the project has been started to develop the regulatory technology for SFR system including a fuel area. To evaluate the fuel integrity and safety during an irradiation, the fuel performance code must be used for audit calculation. In this study, to verify the new code system, the benchmark analysis is performed. In the benchmark, X447 EBR-II experiment data are used. Additionally, the sensitivity analysis according to mass flux change of coolant is performed. In case of LWR fuel performance modeling, various and advanced models have been proposed and validated based on sufficient in-reactor test results. However, due to the lack of experience of SFR operation, the current understanding of SFR fuel behavior is limited. In this study, X447 EBR-II Experiment data are used for benchmark. The fuel composition of X447 assembly is U-10Zr and PGSFR also uses this composition in initial phase. So we select X447 EBR-II experiment for benchmark analysis. Due to the lack of experience of SFR operation and data, the current understanding of SFR fuel behavior is limited. However, in order to prepare the licensing of PGSFR, regulatory audit technologies of SFR must be secured. So, in this study, to verify the new audit fuel performance analysis code, the benchmark analysis is performed using X447 EBR-II experiment data. Also, the sensitivity analysis with mass flux change of coolant is performed. In terms of verification, it is considered that the results of benchmark and sensitivity analysis are reasonable

Application of the integral code MELCOR for German NPPs and use within accident management and PSA projects

International Nuclear Information System (INIS)

Sonnenkalb, Martin

2006-01-01

The paper summarizes the application of MELCOR to German NPPS with PWR and BWR. A development of different code systems like ATHLET/ATHLET-CD, COCOSYS and ASTEC is done as well at GRS but it is not discussed in this paper. GRS has been using MELCOR since 1990 for real plant calculations. The results of MELCOR analyses are used mainly in PSA level 2 studies and in Accident Management projects for both types of NPPs. MELCOR has been a very useful and robust tool for these analyses. The calculations performed within the PSA level 2 studies for both types of German NPPs have shown that typical severe accident scenarios are characterized by several phases and that the consideration of plant specifics are important not only for realistic source term calculations. An overview of typically severe accident phases together with main accident management measures installed in German NPPs is presented in the paper. Several severe accident sequences have been calculated for both reactor types and some detailed nodalisation studies and code to code comparisons have been prepared in the past, to prove the developed core, reactor circuit and containment/building nodalisation schemes. Together with the compilation of the MELCOR data set, the qualification of the nodalisation schemes has been pursued with comparative calculations with detailed GRS codes for selected phases of severe accidents. The results of these comparative analyses showed in most of the areas a good agreement of essential parameters and of the general description of the plant behaviour during the accident progression. The in general detail of the German plant nodalisation schemes developed for MELCOR contributes significantly to this good agreement between integral and detailed code results. The implementation of MELCOR into the GRS simulator ATLAS was very important for the assessment of the results, not only due to the great detail of the nodalisation schemes used. It is used for training of severe accident

Quantification of the computational accuracy of code systems on the burn-up credit using experimental re-calculations; Quantifizierung der Rechengenauigkeit von Codesystemen zum Abbrandkredit durch Experimentnachrechnungen

Energy Technology Data Exchange (ETDEWEB)

Behler, Matthias; Hannstein, Volker; Kilger, Robert; Moser, Franz-Eberhard; Pfeiffer, Arndt; Stuke, Maik

2014-06-15

In order to account for the reactivity-reducing effect of burn-up in the criticality safety analysis for systems with irradiated nuclear fuel (''burnup credit''), numerical methods to determine the enrichment and burnup dependent nuclide inventory (''burnup code'') and its resulting multiplication factor k{sub eff} (''criticality code'') are applied. To allow for reliable conclusions, for both calculation systems the systematic deviations of the calculation results from the respective true values, the bias and its uncertainty, are being quantified by calculation and analysis of a sufficient number of suitable experiments. This quantification is specific for the application case under scope and is also called validation. GRS has developed a methodology to validate a calculation system for the application of burnup credit in the criticality safety analysis for irradiated fuel assemblies from pressurized water reactors. This methodology was demonstrated by applying the GRS home-built KENOREST burnup code and the criticality calculation sequence CSAS5 from SCALE code package. It comprises a bounding approach and alternatively a stochastic, which both have been exemplarily demonstrated by use of a generic spent fuel pool rack and a generic dry storage cask, respectively. Based on publicly available post irradiation examination and criticality experiments, currently the isotopes of uranium and plutonium elements can be regarded for.

El origen no humano de la música

OpenAIRE

Juan David Leongómez

2008-01-01

Resumen: Si la música tiene un origen cultural -como suele asumirse- ¿cómo puede explicarse su universalidad? Este artículo parte de este problema para plantear la hipótesis de que el fenómeno musical tenga un origen anterior al hombre moderno, o incluso anterior al hombre mismo y sea el resultado de una serie de adaptaciones al medio; de que tenga un origen biológico y no cultural. Tras analizar diversos estudios en áreas tan diversas como la semiótica, la neurología, la etología y la evoluc...

Activation calculation for the dismantling and decommissioning of a light water reactor using MCNP™ with ADVANTG and ORIGEN-S

Directory of Open Access Journals (Sweden)

Schlömer Luc

2017-01-01

Full Text Available The decommissioning of a light water reactor (LWR, which is licensed under § 7 of the German Atomic Energy Act, following the post-operational phase requires a comprehensive licensing procedure including in particular radiation protection aspects and possible impacts to the environment. Decommissioning includes essential changes in requirements for the systems and components and will mainly lead to the direct dismantling. In this context, neutron induced activation calculations for the structural components have to be carried out to predict activities in structures and to estimate future costs for conditioning and packaging. To avoid an overestimation of the radioactive inventory and to calculate the expenses for decommissioning as accurate as possible, modern state-of-the-art Monte-Carlo-Techniques (MCNP™ are applied and coupled with present-day activation and decay codes (ORIGEN-S. In this context ADVANTG is used as weight window generator for MCNP™ i. e. as variance reduction tool to speed up the calculation in deep penetration problems. In this paper the calculation procedure is described and the obtained results are presented with a validation along with measured activities and photon dose rates measured in the post-operational phase. The validation shows that the applied calculation procedure is suitable for the determination of the radioactive inventory of a nuclear power plant. Even the measured gamma dose rates in the post-operational phase at different positions in the reactor building agree within a factor of 2 to 3 with the calculation results. The obtained results are accurate and suitable to support effectively the decommissioning planning process.

Activation calculation for the dismantling and decommissioning of a light water reactor using MCNP™ with ADVANTG and ORIGEN-S

Science.gov (United States)

Schlömer, Luc; Phlippen, Peter-W.; Lukas, Bernard

2017-09-01

The decommissioning of a light water reactor (LWR), which is licensed under § 7 of the German Atomic Energy Act, following the post-operational phase requires a comprehensive licensing procedure including in particular radiation protection aspects and possible impacts to the environment. Decommissioning includes essential changes in requirements for the systems and components and will mainly lead to the direct dismantling. In this context, neutron induced activation calculations for the structural components have to be carried out to predict activities in structures and to estimate future costs for conditioning and packaging. To avoid an overestimation of the radioactive inventory and to calculate the expenses for decommissioning as accurate as possible, modern state-of-the-art Monte-Carlo-Techniques (MCNP™) are applied and coupled with present-day activation and decay codes (ORIGEN-S). In this context ADVANTG is used as weight window generator for MCNP™ i. e. as variance reduction tool to speed up the calculation in deep penetration problems. In this paper the calculation procedure is described and the obtained results are presented with a validation along with measured activities and photon dose rates measured in the post-operational phase. The validation shows that the applied calculation procedure is suitable for the determination of the radioactive inventory of a nuclear power plant. Even the measured gamma dose rates in the post-operational phase at different positions in the reactor building agree within a factor of 2 to 3 with the calculation results. The obtained results are accurate and suitable to support effectively the decommissioning planning process.

Improvements of MCOR: A Monte Carlo depletion code system for fuel assembly reference calculations

Energy Technology Data Exchange (ETDEWEB)

Tippayakul, C.; Ivanov, K. [Pennsylvania State Univ., Univ. Park (United States); Misu, S. [AREVA NP GmbH, An AREVA and SIEMENS Company, Erlangen (Germany)

2006-07-01

This paper presents the improvements of MCOR, a Monte Carlo depletion code system for fuel assembly reference calculations. The improvements of MCOR were initiated by the cooperation between the Penn State Univ. and AREVA NP to enhance the original Penn State Univ. MCOR version in order to be used as a new Monte Carlo depletion analysis tool. Essentially, a new depletion module using KORIGEN is utilized to replace the existing ORIGEN-S depletion module in MCOR. Furthermore, the online burnup cross section generation by the Monte Carlo calculation is implemented in the improved version instead of using the burnup cross section library pre-generated by a transport code. Other code features have also been added to make the new MCOR version easier to use. This paper, in addition, presents the result comparisons of the original and the improved MCOR versions against CASMO-4 and OCTOPUS. It was observed in the comparisons that there were quite significant improvements of the results in terms of k{sub inf}, fission rate distributions and isotopic contents. (authors)

The scale analysis sequence for LWR fuel depletion

International Nuclear Information System (INIS)

Hermann, O.W.; Parks, C.V.

1991-01-01

The SCALE (Standardized Computer Analyses for Licensing Evaluation) code system is used extensively to perform away-from-reactor safety analysis (particularly criticality safety, shielding, heat transfer analyses) for spent light water reactor (LWR) fuel. Spent fuel characteristics such as radiation sources, heat generation sources, and isotopic concentrations can be computed within SCALE using the SAS2 control module. A significantly enhanced version of the SAS2 control module, which is denoted as SAS2H, has been made available with the release of SCALE-4. For each time-dependent fuel composition, SAS2H performs one-dimensional (1-D) neutron transport analyses (via XSDRNPM-S) of the reactor fuel assembly using a two-part procedure with two separate unit-cell-lattice models. The cross sections derived from a transport analysis at each time step are used in a point-depletion computation (via ORIGEN-S) that produces the burnup-dependent fuel composition to be used in the next spectral calculation. A final ORIGEN-S case is used to perform the complete depletion/decay analysis using the burnup-dependent cross sections. The techniques used by SAS2H and two recent applications of the code are reviewed in this paper. 17 refs., 5 figs., 5 tabs

Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS). Annual report 2013/14

International Nuclear Information System (INIS)

2016-02-01

The annual report 2013/14 of the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) includes projects and key activities on the following issues: reactor safety, waste disposal, radiation protection, environmental protection and the support of foreign authorities and TSO. The project executing organizations, the governmental funding, share holding and subsidiaries are summarized.

Electromagnetic field and mechanical stress analysis code

International Nuclear Information System (INIS)

1978-01-01

Analysis TEXMAGST is a two stage linear finite element code for the analysis of static magnetic fields in three dimensional structures and associated mechanical stresses produced by the anti J x anti B forces within these structures. The electromagnetic problem is solved in terms of magnetic vector potential A for a given current density anti J as curl 1/μ curl anti A = anti J considering the magnetic permeability as constant. The Coulombian gauge (div anti A = o) was chosen and was implemented through the use of Lagrange multipliers. The second stage of the problem - the calculation of mechanical stresses in the same three dimensional structure is solved by using the same code with few modifications - through a restart card. Body forces anti J x anti B within each element are calculated from the solution of the first stage run and represent the input to the second stage run which will give the solution for the stress problem

Validation of the ASSERT subchannel code for MAPLE-X10 reactor conditions

International Nuclear Information System (INIS)

Carver, M.B.; Kiteley, J.C.; Junop, S.V.; Wasilewicz, J.F.

1993-01-01

The ASSERT subchannel analysis code has been developed specifically to model flow and phase distributions within CANDU fuel channels. Recently, ASSERT has been adapted for use in simulating the MAPLE-X10 reactor. ASSERT uses an advanced drift-flux model, which permits the phases to have unequal velocities and unequal temperatures (UVUT), and thus can model non-equilibrium effects such as phase separation tendencies and subcooled boiling. Modelling subcooled boiling accurately is particularly important for MAPLE-X10. This paper briefly summarizes the non-equilibrium model used in the ASSERT code, the equations used to represent these models, and the algorithms used to solve the equations numerically. Very few modifications to the ASSERT models were needed to address MAPLE conditions. These centered on the manner in which finned fuel rods are treated, and they are discussed in the paper. The paper also gives results from validation exercises, in which the ASSERT code predictions of subcooled boiling void-fraction and critical heat flux were compared to experiments using MAPLE-X10 finned fuel elements in annuli and various bundles. 18 refs., 13 figs., 3 tabs

MANIA (276-3/4/5). Nuclear analysis

International Nuclear Information System (INIS)

Sciolla, C.M.

1993-11-01

This report contains the results of the nuclear calculations performed for the MANIA-276 experiment, sample holders 3, 4 and 5. The codes MICROFLUX-2, GAM, HFR-TEDDI and ORIGEN-S have been used for this analysis. Nuclear constants, dpa, reactivity effect and activity of the samples and of the structural materials have been calculated. The results are given in the tables and appendices of the present report. (orig.)

GRS - Guarantee of results for grid-connected solar photovoltaic systems; GRS - Garantierte Resultate von Solarstromanlagen

Energy Technology Data Exchange (ETDEWEB)

Meier, Ch.; Frei, R.

2001-07-01

This final report for the Swiss Federal Office of Energy describes the development of a performance guarantee for the power delivered by grid-connected photovoltaic (PV) power plant. This is important for investors, who need to be able to calculate the price of the energy produced by the PV plant. The report examines the results of various case studies and experience gained with GRS contracts developed in different countries including Switzerland, the Netherlands, France and Germany. The problems encountered, which provided feedback for the further development of the contract presented in the report, are discussed. The model contract, which proposes a framework which provides many possible variations and adaptations, is presented in detail. Various problem areas such as guaranteed energy yield, meteorological references, correction methods, failure detection, dispute settlement and many other points are covered.

Analysis of top flooding during molten corium concrete interaction (MCCI) with the code MEDICIS using a simplified approach for the combined effect of crust formation and boiling

International Nuclear Information System (INIS)

Spengler, C.

2012-01-01

The objective of this work is to provide adequate models in the code MEDICIS for the molten corium concrete interaction (MCCI) phase in a severe accident. Here, the multidimensional distribution of heat fluxes from the molten pool of corium to the sidewall and bottom wall concrete structures in the reactor pit and to the top surface is a persistent subject of international research activities on MCCI. In recent experi-ments with internally heated oxide melts it was observed that the erosion progress may be anisotropic - with an apparent preference of the sidewall compared to the bottom wall - or isotropic, in dependence of the type of concrete with which the cori-um interacts. The lumped parameter code MEDICIS, which is part of the severe accident codes ASTEC and COCOSYS - developed and used at IRSN/GRS respectively GRS for the latter one -, is dedicated to simulate the phenomenology during MCCI. In this work a simplified modelling in MEDICIS is tested to account for the observed ablation behaviour during MCCI, with focus on the heat transfer to the top surface under flooded conditions. This approach is assessed by calculations for selected MCCI experiments involving the top flooding of the melt. (orig.)

Validation of integrated burnup code system SWAT2 by the analyses of isotopic composition of spent nuclear fuel

International Nuclear Information System (INIS)

Suyama, K.; Mochizuki, H.; Okuno, H.; Miyoshi, Y.

2004-01-01

This paper provides validation results of SWAT2, the revised version of SWAT, which is a code system combining point burnup code ORIGEN2 and continuous energy Monte Carlo code MVP, by the analysis of post irradiation examinations (PIEs). Some isotopes show differences of calculation results between SWAT and SWAT2. However, generally, the differences are smaller than the error of PIE analysis that was reported in previous SWAT validation activity, and improved results are obtained for several important fission product nuclides. This study also includes comparison between an assembly and a single pin cell geometry models. (authors)

Obtaining of primary rays of spectrum X codes Penelope and MCNP5; Obtencion del espectro primario de Rayos X con los codigos Penelope y MCNP5

Energy Technology Data Exchange (ETDEWEB)

Pozuelo, F.; Querol, A.; Gallardo, S.; Rodenas, J.; Verdu, G.

2012-07-01

In this case, used codes PENELOPE MCNP5, based on the Monte Carlo method for x-ray spectrum taking into account the characteristics of the x-ray tube. In order to achieve a greater fit of simulated by the theoretical spectrum. It carried out a sensitivity analysis of the parameters available in both codes. The obtaining of the simulated spectrum could lead to an improvement in quality control of the x-ray tube to incorporate it as a method complementary to techniques.

Turbo Pascal Computer Code for PIXE Analysis

International Nuclear Information System (INIS)

Darsono

2002-01-01

To optimal utilization of the 150 kV ion accelerator facilities and to govern the analysis technique using ion accelerator, the research and development of low energy PIXE technology has been done. The R and D for hardware of the low energy PIXE installation in P3TM have been carried on since year 2000. To support the R and D of PIXE accelerator facilities in harmonize with the R and D of the PIXE hardware, the development of PIXE software for analysis is also needed. The development of database of PIXE software for analysis using turbo Pascal computer code is reported in this paper. This computer code computes the ionization cross-section, the fluorescence yield, and the stopping power of elements also it computes the coefficient attenuation of X- rays energy. The computer code is named PIXEDASIS and it is part of big computer code planed for PIXE analysis that will be constructed in the near future. PIXEDASIS is designed to be communicative with the user. It has the input from the keyboard. The output shows in the PC monitor, which also can be printed. The performance test of the PIXEDASIS shows that it can be operated well and it can provide data agreement with data form other literatures. (author)

ORIGEN-2 libraries based on JENDL-3.2 for PWR-MOX fuel

Energy Technology Data Exchange (ETDEWEB)

Matsumoto, Hideki; Onoue, Masaaki; Tahara, Yoshihisa [Mitsubishi Heavy Industries Ltd., Tokyo (Japan)

2001-08-01

A set of ORIGEN-2 libraries for PWR MOX fuel was developed based on JENDL-3.2 in the Working Group on Evaluation of Nuclide Production, Japanese Nuclear Data Committee. The calculational model generating ORIGEN-2 libraries of PWR MOX is explained here in detail. The ORIGEN-2 calculation with the new ORIGEN-2 MOX library can predict the nuclides contents within 10% for U and Pu isotopes and 20% for both minor actinides and main FPs. (author)

Synthesis of the ASTEC integral code activities in SARNET – Focus on ASTEC V2 plant applications

International Nuclear Information System (INIS)

Chatelard, P.; Reinke, N.; Ezzidi, A.; Lombard, V.; Barnak, M.; Lajtha, G.; Slaby, J.; Constantin, M.; Majumdar, P.

2014-01-01

Highlights: • Independent assessment of the ASTEC severe accident code vs. experiments is summarised. • Main remaining modelling issues and development perspectives are identified. • Independent assessment of ASTEC code at full scale conditions is described. • Main requirements to address BWR and PHWR types of reactors are identified. - Abstract: Among the 43 organisations which joined the SARNET2 FP7 project from 2009 to 2013, 31 have been involved in the activities on the ASTEC code. This paper presents a synthesis of the main achievements that have been obtained on the ASTEC V2 integral code, jointly developed by IRSN (France) and GRS (Germany), on development, validation vs. experimental data and applications at full scale conditions for both Gen.II and Gen.III plants. As to code development, while the current V2.0 series of ASTEC versions was continuously improved (elaboration and release by IRSN and GRS of three successive V2.0 revisions), IRSN and GRS have also intensively continued in parallel the elaboration of the second ASTEC V2 major version (version V2.1) to be delivered end of 2014. Regarding code validation vs. experiments, the partners have assessed the V2.0 version and subsequent revisions vs. more than 50 experiments; this extended assessment notably confirmed that most models are today close to the State of the Art, while it also corroborated the yet known key-topics on which modelling efforts should focus in priority. As to plant applications, the comparison of ASTEC results with other codes allows concluding on a globally good agreement for in-vessel and ex-vessel severe accident progression. As to ASTEC adaptations to BWR and PHWR, significant achievements have been obtained through the elaboration and integration in the future V2.1 version of dedicated core degradation models, notably to account for multi coolant flows

Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH, Germany. Annex I [Example of TSOs and their Interaction with Key Stakeholders

International Nuclear Information System (INIS)

2018-01-01

The State, Federal States and TÜVs signed the Shareholders Agreement for the Company for Reactor Safety (GRS) mbH on 26th of May 1976. GRS is a private non-profit, non-governmental organization, established as a private limited company with shareholders. All services of GRS to the regulatory authorities are provided through projects based on both long term and task specific contracts within framework agreements.

The GRS Emergency Centre during the Fukushima NPS Accident: Communicating Radiological Information to the Public

International Nuclear Information System (INIS)

Sentuc, F.N.; Dokter, S.

2012-01-01

During Fukushima 2011, the GRS Emergency Centre continuously analysed the situation and provided detailed information on the current plant status and the radiological consequences, not only for the German government but also for the worried public. The expert team primarily informed the competent authority about on-going developments in Japan via status reports. The status reports provided technical and radiological information promptly and were updated several times a day at the beginning and towards the end twice a week until July 1 2011. Additionally, these reports were translated into English and published on the web-site. In the course of the event, approximately 200 reports were issued by the GRS-EC. It became clear through media and public enquires that information, even at a very basic level, was needed in the field of nuclear technology and radiation protection. Therefore, technical experts and public relation team members worked together to prepare explanatory and background material. Owing to this strong interest, GRS established a special Fukushima portal on its web-site, to meet the need for information. As well as the technical aspects, which dominated the expert's work in the GRS Emergency Centre, one challenge was communicating the radiological information to the public in a comprehensible yet professional way. The media and the public appear to be very interested in getting information about the exposure of workers at the plant, the dose limits, the release of radioactive material in the atmosphere and the long-term effects of the contamination. The article is followed by the slides of the presentation

2 years of integral monitoring of GRS 1915+105. II. X-ray spectro-temporal analysis

DEFF Research Database (Denmark)

Rodriguez, J.; Shaw, S.E.; Hannikainen, D.C.

2008-01-01

-ray emission. In the steady state observation, the X-ray spectrum is indicative of the hard-intermediate state, with the presence of a relatively strong emission at 15 GHz. The X-ray spectrum is the sum of a Comptonized component and an extra power law extending to energies > 200 keV without any evidence...

European Validation of the Integral Code ASTEC (EVITA)

International Nuclear Information System (INIS)

Allelein, H.-J.; Neu, K.; Dorsselaere, J.P. Van

2005-01-01

The main objective of the European Validation of the Integral Code ASTEC (EVITA) project is to distribute the severe accident integral code ASTEC to European partners in order to apply the validation strategy issued from the VASA project (4th EC FWP). Partners evaluate the code capability through validation on reference experiments and plant applications accounting for severe accident management measures, and compare results with reference codes. The basis version V0 of ASTEC (Accident Source Term Evaluation Code)-commonly developed and basically validated by GRS and IRSN-was made available in late 2000 for the EVITA partners on their individual platforms. Users' training was performed by IRSN and GRS. The code portability on different computers was checked to be correct. A 'hot line' assistance was installed continuously available for EVITA code users. The actual version V1 has been released to the EVITA partners end of June 2002. It allows to simulate the front-end phase by two new modules:- for reactor coolant system 2-phase simplified thermal hydraulics (5-equation approach) during both front-end and core degradation phases; - for core degradation, based on structure and main models of ICARE2 (IRSN) reference mechanistic code for core degradation and on other simplified models. Next priorities are clearly identified: code consolidation in order to increase the robustness, extension of all plant applications beyond the vessel lower head failure and coupling with fission product modules, and continuous improvements of users' tools. As EVITA has very successfully made the first step into the intention to provide end-users (like utilities, vendors and licensing authorities) with a well validated European integral code for the simulation of severe accidents in NPPs, the EVITA partners strongly recommend to continue validation, benchmarking and application of ASTEC. This work will continue in Severe Accident Research Network (SARNET) in the 6th Framework Programme

Radionuclide composition in nuclear fuel waste. Calculations performed by ORIGEN2; Radionuklidinnehaall i utbraent kaernbraensle. Beraekningar med ORIGEN2

Energy Technology Data Exchange (ETDEWEB)

Lyckman, C

1996-01-01

The report accounts for results from calculations on the content of radionuclides in nuclear fuel waste. It also accounts for the results from calculations on the neutron flow from spent fuel, which is very important during transports. The calculations have been performed using the ORIGEN2 software. The results have been compared to other results from earlier versions of ORIGEN and some differences have been discovered. This is due to the updating of the software. 7 refs, 10 figs, 15 tabs.

LWR spent-fuel radiochemical measurements and comparison with ORIGEN2 predictions

International Nuclear Information System (INIS)

Blahnik, D.E.; Jenquin, U.P.; Guenther, R.J.

1988-01-01

The Materials Characterization Center (MCC) at Pacific Northwest Laboratory is responsible for providing characterized spent fuel, designated approved testing material (ATMs) for subsequent use in the investigation of nuclear waste disposal forms by the US Department of Energy geologic repository project. The ATMs are selected to assure that test material is available that has a representative range of characteristics important to spent-fuel behavior in a geologic repository. Burnup and fission gas release were the primary criteria for selecting the ATMs. The five spent-fuel ATMs (ATM-101, -103, -104, -105, and -106) currently being characterized by the MCC have rod average burnups ranging from 20 to 43 MWd/kg M, fission gas releases ranging from 0.2 to 11.2%, and are from both boiling water reactors and pressurized water reactors. Radiochemical analyses of the fuel included measurements of 148 Nd (for fuel burnup), the isotopes of uranium and plutonium, and nuclides of importance to repository performance. Cladding samples were analyzed for 14 C. The measured values of selected nuclides were compared with values obtained from calculations with the ORIGEN2 code that was used to predict isotopic quantities for all of the ATMS. Ratios of the ORIGEN2 calculated values to the measured values for ATM-103 and ATM-106 fuel are given

Code comparison for accelerator design and analysis

International Nuclear Information System (INIS)

Parsa, Z.

1988-01-01

We present a comparison between results obtained from standard accelerator physics codes used for the design and analysis of synchrotrons and storage rings, with programs SYNCH, MAD, HARMON, PATRICIA, PATPET, BETA, DIMAD, MARYLIE and RACE-TRACK. In our analysis we have considered 5 (various size) lattices with large and small angles including AGS Booster (10/degree/ bend), RHIC (2.24/degree/), SXLS, XLS (XUV ring with 45/degree/ bend) and X-RAY rings. The differences in the integration methods used and the treatment of the fringe fields in these codes could lead to different results. The inclusion of nonlinear (e.g., dipole) terms may be necessary in these calculations specially for a small ring. 12 refs., 6 figs., 10 tabs

The European source-term evaluation code ASTEC: status and applications, including CANDU plant applications

International Nuclear Information System (INIS)

Van Dorsselaere, J.P.; Giordano, P.; Kissane, M.P.; Montanelli, T.; Schwinges, B.; Ganju, S.; Dickson, L.

2004-01-01

Research on light-water reactor severe accidents (SA) is still required in a limited number of areas in order to confirm accident-management plans. Thus, 49 European organizations have linked their SA research in a durable way through SARNET (Severe Accident Research and management NETwork), part of the European 6th Framework Programme. One goal of SARNET is to consolidate the integral code ASTEC (Accident Source Term Evaluation Code, developed by IRSN and GRS) as the European reference tool for safety studies; SARNET efforts include extending the application scope to reactor types other than PWR (including VVER) such as BWR and CANDU. ASTEC is used in IRSN's Probabilistic Safety Analysis level 2 of 900 MWe French PWRs. An earlier version of ASTEC's SOPHAEROS module, including improvements by AECL, is being validated as the Canadian Industry Standard Toolset code for FP-transport analysis in the CANDU Heat Transport System. Work with ASTEC has also been performed by Bhabha Atomic Research Centre, Mumbai, on IPHWR containment thermal hydraulics. (author)

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

Code of safe practice for the use of x-rays in podiatry

International Nuclear Information System (INIS)

1990-02-01

The purpose of this code which applies to podiatric diagnostic use of x-rays is to provide criteria for working procedures, x-ray equipment and protective materials necessary for the use of x-rays in podiatric diagnosis according to currently accepted standards of safety. Conformity with this code may be taken as a primary indication of compliance with radiation protection legislation. 5 refs

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

International Nuclear Information System (INIS)

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

2001-08-01

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

Inner disc obscuration in GRS 1915+105 based on relativistic slim disc model

Czech Academy of Sciences Publication Activity Database

Vierdayanti, K.; Sądowski, A.; Mineshige, L.S.; Bursa, Michal

2013-01-01

Roč. 436, č. 1 (2013), s. 71-81 ISSN 0035-8711 Institutional support: RVO:67985815 Keywords : accretion discs * black hole physics * GRS 1915+105 Subject RIV: BN - Astronomy, Celestial Mechanics, Astrophysics Impact factor: 5.226, year: 2013

Familia de Origen y Violencia Conyugal en Mujeres

OpenAIRE

Collazo Muñoz, Grissell Jasveth

2008-01-01

Se presenta un estudio sobre las diferencias y relaciones entre Violencia en la Pareja y percepción de la Familia de Origen. Los instrumentos empleados fueron: la Escala de la familia de origen (FOS) validada por Eisenberg (1994), la Escala de Violencia en la Pareja (Cienfuegos, 2004) e Historias de vida. La muestra estuvo conformada por disponibilidad de 68 mujeres; 22 MCV y 46 MSV. Mujeres que experimentaron violencia frente al grupo de las que no, presentaron diferencias ...

El papa Urbano II y el origen de la Garcineida

Directory of Open Access Journals (Sweden)

González García, Alberto

2013-12-01

Full Text Available This paper aims to re-examine the most recent scholarly work on the eleventh-century Latin satire known as Garcineida, in order to determine its origin, authorship and credibility. The analysis of this work, as well as other primary and secondary sources, leads us to two interesting conclusions. First, we reject the theory of a German original, and we suggest the plausibility of a Spanish-Cluniac origin. In second place, the existence of facts that contradict the image of Urban II as a faultless reformer is outlined. They allow to verify that, despite its virulence, the aforesaid satire was based on reality.Este trabajo pretende revisar las más recientes investigaciones académicas sobre la sátira latina del siglo XI llamada Garcineida, a fin de determinar su origen, autoría y credibilidad. El análisis de esta obra, así como de otras fuentes primarias y secundarias, nos lleva a dos interesantes conclusiones. Primero, a rechazar la teoría de un origen alemán y apreciar lo plausible de un origen hispánico cluniacense. En segundo lugar, planteamos la existencia de hechos que contrarían la imagen de Urbano II como un reformador inmaculado y permiten constatar que, a pesar de su virulencia, la susodicha sátira se basaba en una realidad.

Code of safe practice for the use of x-rays in dentistry

International Nuclear Information System (INIS)

1991-06-01

The purpose of this Code of Safe Practice for the Use of X-rays in Dentistry is to provide criteria for working procedures, x-ray equipment and protective materials necessary for the use of x-rays in dental diagnosis according to currently accepted standards of safety. Conformity with this code may be taken as a primary indication of compliance with radiation protection legislation. 7 refs., 1 ill

A Shielding Analysis of Hot Cell for a 10 MW Research Reactor

International Nuclear Information System (INIS)

Alnajjar, Alaaddin; Park, Chang Je; Roh, Gyuhong; Lee, Byunchul

2013-01-01

In this paper, a shielding analysis has been performed for the hot cell in a 10 MW research reactor. Two kinds of shielding analysis code systems are used such as MCNPX2.7 and M-Shield7. The first one is Monte Carlo stochastic code and the second one is a deterministic point kernel code. The results are compared in this study. In order to obtain source term, the ORIGEN-S code is used for different kinds of source. Four kinds of sources are taken into consideration. From the simulation, it is also proposed that the proper thickness of shielding material and the maximum source capacity in the hot cell. This study shows preliminary analysis results of hot cell shielding for 10MW research reactor. Total four different source terms are considered such as spent fuel assembly, Ir-192, Mo-99, and I-131. For shielding material, general concrete, heavy concrete, and lead are used. MCNPX code is mainly used for a simplified hot cell model and the result are nearly consistent when compared with M-Shield code. Required shielding thickness and the hot cell capacity are also obtained for various criterion of surface dose rates

Introduction of thermal-hydraulic analysis code and system analysis code for HTGR

International Nuclear Information System (INIS)

Tanaka, Mitsuhiro; Izaki, Makoto; Koike, Hiroyuki; Tokumitsu, Masashi

1984-01-01

Kawasaki Heavy Industries Ltd. has advanced the development and systematization of analysis codes, aiming at lining up the analysis codes for heat transferring flow and control characteristics, taking up HTGR plants as the main object. In order to make the model of flow when shock waves propagate to heating tubes, SALE-3D which can analyze a complex system was developed, therefore, it is reported in this paper. Concerning the analysis code for control characteristics, the method of sensitivity analysis in a topological space including an example of application is reported. The flow analysis code SALE-3D is that for analyzing the flow of compressible viscous fluid in a three-dimensional system over the velocity range from incompressibility limit to supersonic velocity. The fundamental equations and fundamental algorithm of the SALE-3D, the calculation of cell volume, the plotting of perspective drawings and the analysis of the three-dimensional behavior of shock waves propagating in heating tubes after their rupture accident are described. The method of sensitivity analysis was added to the analysis code for control characteristics in a topological space, and blow-down phenomena was analyzed by its application. (Kako, I.)

NORTICA - a new code for cyclotron analysis

International Nuclear Information System (INIS)

Gorelov, D.; Johnson, D.; Marti, F.

2001-01-01

The new package NORTICA (Numerical ORbit Tracking In Cyclotrons with Analysis) of computer codes for beam dynamics simulations is under development at NSCL. The package was started as a replacement for the code MONSTER developed in the laboratory in the past. The new codes are capable of beam dynamics simulations in both CCF (Coupled Cyclotron Facility) accelerators, the K500 and K1200 superconducting cyclotrons. The general purpose of this package is assisting in setting and tuning the cyclotrons taking into account the main field and extraction channel imperfections. The computer platform for the package is Alpha Station with UNIX operating system and X-Windows graphic interface. A multiple programming language approach was used in order to combine the reliability of the numerical algorithms developed over the long period of time in the laboratory and the friendliness of modern style user interface. This paper describes the capability and features of the codes in the present state

NORTICA—a new code for cyclotron analysis

Science.gov (United States)

Gorelov, D.; Johnson, D.; Marti, F.

2001-12-01

The new package NORTICA (Numerical ORbit Tracking In Cyclotrons with Analysis) of computer codes for beam dynamics simulations is under development at NSCL. The package was started as a replacement for the code MONSTER [1] developed in the laboratory in the past. The new codes are capable of beam dynamics simulations in both CCF (Coupled Cyclotron Facility) accelerators, the K500 and K1200 superconducting cyclotrons. The general purpose of this package is assisting in setting and tuning the cyclotrons taking into account the main field and extraction channel imperfections. The computer platform for the package is Alpha Station with UNIX operating system and X-Windows graphic interface. A multiple programming language approach was used in order to combine the reliability of the numerical algorithms developed over the long period of time in the laboratory and the friendliness of modern style user interface. This paper describes the capability and features of the codes in the present state.

Rate-adaptive BCH codes for distributed source coding

DEFF Research Database (Denmark)

Salmistraro, Matteo; Larsen, Knud J.; Forchhammer, Søren

2013-01-01

This paper considers Bose-Chaudhuri-Hocquenghem (BCH) codes for distributed source coding. A feedback channel is employed to adapt the rate of the code during the decoding process. The focus is on codes with short block lengths for independently coding a binary source X and decoding it given its...... strategies for improving the reliability of the decoded result are analyzed, and methods for estimating the performance are proposed. In the analysis, noiseless feedback and noiseless communication are assumed. Simulation results show that rate-adaptive BCH codes achieve better performance than low...... correlated side information Y. The proposed codes have been analyzed in a high-correlation scenario, where the marginal probability of each symbol, Xi in X, given Y is highly skewed (unbalanced). Rate-adaptive BCH codes are presented and applied to distributed source coding. Adaptive and fixed checking...

Evaluation of spent fuel isotopics, radiation spectra and decay heat using the scale computational system

International Nuclear Information System (INIS)

Parks, C.V.; Hermann, O.W.; Ryman, J.C.

1986-01-01

In order to be a self-sufficient system for transport/storage cask shielding and heat transfer analysis, the SCALE system developers included modules to evaluate spent fuel radiation spectra and decay heat. The primary module developed for these analyses is ORIGEN-S which is an updated verision of the original ORIGEN code. The COUPLE module was also developed to enable ORIGEN-S to easily utilize multigroup cross sections and neutron flux data during a depletion analysis. Finally, the SAS2 control module was developed for automating the depletion and decay via ORIGEN-S while using burnup-dependent neutronic data based on a user-specified fuel assembly and reactor history. The ORIGEN-S data libraries available for depletion and decay have also been significantly updated from that developed with the original ORIGEN code

C2x: A tool for visualisation and input preparation for CASTEP and other electronic structure codes

Science.gov (United States)

Rutter, M. J.

2018-04-01

The c2x code fills two distinct roles. Its first role is in acting as a converter between the binary format .check files from the widely-used CASTEP [1] electronic structure code and various visualisation programs. Its second role is to manipulate and analyse the input and output files from a variety of electronic structure codes, including CASTEP, ONETEP and VASP, as well as the widely-used 'Gaussian cube' file format. Analysis includes symmetry analysis, and manipulation arbitrary cell transformations. It continues to be under development, with growing functionality, and is written in a form which would make it easy to extend it to working directly with files from other electronic structure codes. Data which c2x is capable of extracting from CASTEP's binary checkpoint files include charge densities, spin densities, wavefunctions, relaxed atomic positions, forces, the Fermi level, the total energy, and symmetry operations. It can recreate .cell input files from checkpoint files. Volumetric data can be output in formats useable by many common visualisation programs, and c2x will itself calculate integrals, expand data into supercells, and interpolate data via combinations of Fourier and trilinear interpolation. It can extract data along arbitrary lines (such as lines between atoms) as 1D output. C2x is able to convert between several common formats for describing molecules and crystals, including the .cell format of CASTEP. It can construct supercells, reduce cells to their primitive form, and add specified k-point meshes. It uses the spglib library [2] to report symmetry information, which it can add to .cell files. C2x is a command-line utility, so is readily included in scripts. It is available under the GPL and can be obtained from http://www.c2x.org.uk. It is believed to be the only open-source code which can read CASTEP's .check files, so it will have utility in other projects.

REVISITING EVIDENCE OF CHAOS IN X-RAY LIGHT CURVES: THE CASE OF GRS 1915+105

Energy Technology Data Exchange (ETDEWEB)

Mannattil, Manu; Gupta, Himanshu; Chakraborty, Sagar, E-mail: mmanu@iitk.ac.in, E-mail: hiugupta@iitk.ac.in, E-mail: sagarc@iitk.ac.in [Department of Physics, Indian Institute of Technology Kanpur, Uttar Pradesh 208016 (India)

2016-12-20

Nonlinear time series analysis has been widely used to search for signatures of low-dimensional chaos in light curves emanating from astrophysical bodies. A particularly popular example is the microquasar GRS 1915+105, whose irregular but systematic X-ray variability has been well studied using data acquired by the Rossi X-ray Timing Explorer . With a view to building simpler models of X-ray variability, attempts have been made to classify the light curves of GRS 1915+105 as chaotic or stochastic. Contrary to some of the earlier suggestions, after careful analysis, we find no evidence for chaos or determinism in any of the GRS 1915+105 classes. The dearth of long and stationary data sets representing all the different variability classes of GRS 1915+105 makes it a poor candidate for analysis using nonlinear time series techniques. We conclude that either very exhaustive data analysis with sufficiently long and stationary light curves should be performed, keeping all the pitfalls of nonlinear time series analysis in mind, or alternative schemes of classifying the light curves should be adopted. The generic limitations of the techniques that we point out in the context of GRS 1915+105 affect all similar investigations of light curves from other astrophysical sources.

Development of a code for the isotopic analysis of Uranium

Energy Technology Data Exchange (ETDEWEB)

Kim, J. H.; Kang, M. Y.; Kim, Jinhyeong; Choi, H. D. [Seoul National Univ., Seoul (Korea, Republic of)

2013-05-15

To strengthen the national nuclear nonproliferation regime by an establishment of nuclear forensic system, the techniques for nuclear material analysis and the categorization of important domestic nuclear materials are being developed. MGAU and FRAM are commercial software for the isotopic analysis of Uranium by using γ-spectroscopy, but the diversity of detection geometry and some effects - self attenuation, coincidence summing, etc. - suggest an analysis tool under continual improvement and modification. Hence, developing another code for HPGe γ- and x-ray spectrum analysis is started in this study. The analysis of the 87-101 keV region of Uranium spectrum is attempted based on the isotopic responses similar to those developed in MGAU. The code for isotopic analysis of Uranium is started from a fitting.

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

International Nuclear Information System (INIS)

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

2000-10-01

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

Origen del castillo y coto de Aranga, siglos X-XII

Directory of Open Access Journals (Sweden)

Barros, Carlos

2009-12-01

Full Text Available The importance of the fortress of the Higth Middle Age in the constitution of the feudal jurisdictions around the year 1000 is illustrated in this work which is the origin of the castle and a lordship of Aranga (now municipality of Galicia from the family of Traba. Th king Alfonso IX of Leon and Galicia transferred in 1193, the domain jurisdiction from a Aranga to Monastery of Sobrado, against the opinion of their vassals and peasants who, in 1168, preferring to continue to belong to the heirs of the counts of Traba and Aranga.

La importancia de las fortalezas altomedievales en la constitución de las jurisdicciones feudales, en los alrededores del año 1000, se ilustra en este trabajo que trata del origen del castillo y coto de Aranga (actualmente, municipio de Galicia de la casa de Traba. El rey Alfonso IX de León y Galicia traspasa, en 1193, la jurisdicción señorial de Aranga al Monasterio de Sobrado, contra la opinión de sus campesinos y vasallos que, en 1168, preferían seguir perteneciendo a los herederos de los condes de Traba y Aranga.

Easy web interfaces to IDL code for NSTX Data Analysis

International Nuclear Information System (INIS)

Davis, W.M.

2012-01-01

Highlights: ► Web interfaces to IDL code can be developed quickly. ► Dozens of Web Tools are used effectively on NSTX for Data Analysis. ► Web interfaces are easier to use than X-window applications. - Abstract: Reusing code is a well-known Software Engineering practice to substantially increase the efficiency of code production, as well as to reduce errors and debugging time. A variety of “Web Tools” for the analysis and display of raw and analyzed physics data are in use on NSTX [1], and new ones can be produced quickly from existing IDL [2] code. A Web Tool with only a few inputs, and which calls an IDL routine written in the proper style, can be created in less than an hour; more typical Web Tools with dozens of inputs, and the need for some adaptation of existing IDL code, can be working in a day or so. Efficiency is also increased for users of Web Tools because of the familiar interface of the web browser, and not needing X-windows, or accounts and passwords, when used within our firewall. Web Tools were adapted for use by PPPL physicists accessing EAST data stored in MDSplus with only a few man-weeks of effort; adapting to additional sites should now be even easier. An overview of Web Tools in use on NSTX, and a list of the most useful features, is also presented.

Critical review of the >>Stellungnahme zur 'deutschen Risikostudie' der Gesellschaft fuer Reaktorsicherheit (GRS)<< by the Bremer Arbeits- und Umweltschutz-Zentrum

International Nuclear Information System (INIS)

Danzmann, H.J.

1981-08-01

In August 1979, the German Risk Study was presented to the public by the Federal Minister of Research and Technology. The main contractor to prepare the German Risk Study is the Gesellschaft fuer Reaktorsicherheit (GRS). The objective of the Study is to investigate the collective risk involved in accidents at nuclear power plants. The essential results and statements of the Study are comprised in a Summary submitted to the public in August 1979, too. In September 1979, the 'Bremer Arbeits- und Umweltschutz-Zentrum' ('BAUZ') published a 'Stellungnahme zur 'deutschen Risikostudie' der Gesellschaft fuer Reaktorsicherheit (GRS)' (Review of the German Risk Study of the Gesellschaft fuer Reaktorsicherheit), dealing only with the summary of this study, however. The BAUZ review is extremely critical of the German Risk Study and ends in condemning the peaceful utilization of nuclear energy as well as the competent authorities. The GRS review critique in hand scrutinizes the BAUZ paper according to the run of the text. Deliberately, one has renounced a comment in such cases where the reproaches slip in the political or ideological field. The object of the BAUZ review as well as of the GRS critique is the matter to be investigated, aim and methods of the investigations and the results of the German Risk Study. The revision of the review acknowledges that in some cases the tendency of criticism is justified. Often, however, the critique is only a sharpened repetition of what the GRS itself has identified as restrictions of accuracy and inherent significance of the results. In most items, however, the critique is rejected as to be wrong or not proof. (orig./HP) [de

Short-Duration X-ray Transients Observed with WATCH on Granat

DEFF Research Database (Denmark)

Castro-Tirado, Alberto J.; Brandt, Søren; Lund, Niels

1995-01-01

During 1990–92, the WATCH all-sky X-ray monitor on GRANAT has discovered 6 short-duration X-ray transients. We discuss their possible relationship to peculiar stars. Only one source, GRS 1100-77 seems to be related to a T Tauri star....

Songs in the desert : The Rabbis and Origen on spiritual ascent

NARCIS (Netherlands)

Poorthuis, Marcel; van Geest, Paul

The songs in the desert serve both the Rabbis and Origen to sketch a development of spiritual growth. By comparing the songs chosen it is possible to assess the different perspectives: the Rabbis view Moses as the pinnacle, Origen avoids Moses and instead highlights Solomon and his Song of Songs as

SCALE: A modular code system for performing standardized computer analyses for licensing evaluation: Functional modules F1-F8

International Nuclear Information System (INIS)

1997-03-01

This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U.S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This volume consists of the section of the manual dealing with eight of the functional modules in the code. Those are: BONAMI - resonance self-shielding by the Bondarenko method; NITAWL-II - SCALE system module for performing resonance shielding and working library production; XSDRNPM - a one-dimensional discrete-ordinates code for transport analysis; XSDOSE - a module for calculating fluxes and dose rates at points outside a shield; KENO IV/S - an improved monte carlo criticality program; COUPLE; ORIGEN-S - SCALE system module to calculate fuel depletion, actinide transmutation, fission product buildup and decay, and associated radiation source terms; ICE

SCALE: A modular code system for performing standardized computer analyses for licensing evaluation: Functional modules F1-F8

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-03-01

This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U.S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This volume consists of the section of the manual dealing with eight of the functional modules in the code. Those are: BONAMI - resonance self-shielding by the Bondarenko method; NITAWL-II - SCALE system module for performing resonance shielding and working library production; XSDRNPM - a one-dimensional discrete-ordinates code for transport analysis; XSDOSE - a module for calculating fluxes and dose rates at points outside a shield; KENO IV/S - an improved monte carlo criticality program; COUPLE; ORIGEN-S - SCALE system module to calculate fuel depletion, actinide transmutation, fission product buildup and decay, and associated radiation source terms; ICE.

DISCERNMENT IN ORIGEN OF ALEXANDRIA

African Journals Online (AJOL)

“intelligence” is created after the image of God, the Logos, its fulfilment lies in a ... 1 A number of studies have appeared recently on the influence of Origen on ... Produced by SUN MeDIA Bloemfontein ... The issue of freedom and discernment is the .... reality of their bodies and strive for the pleasures of the body; some focus.

Hamor-2: a computer code for LWR inventory calculation

International Nuclear Information System (INIS)

Guimaraes, L.N.F.; Marzo, M.A.S.

1985-01-01

A method for calculating the accuracy inventory of LWR reactors is presented. This method uses the Hamor-2 computer code. Hamor-2 is obtained from the coupling of two other computer codes Hammer-Techion and Origen-2 for testing Hamor-2, its results were compared to concentration values measured from activides of two PWR reactors; Kernkraftwerk Obrighein (KWO) and H.B. Robinson (HBR). These actinides are U 235 , U 236 , U 238 , Pu 239 , Pu 241 and PU 242 . The computer code Hammor-2 shows better results than the computer code Origem-2, when both are compared with experimental results. (E.G.) [pt

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

Origen y filogenia de los primeros homínidos de Europa

Directory of Open Access Journals (Sweden)

A. Gómez

2005-01-01

Full Text Available Se discuten el origen y relaciones filogenéticas de la especie Homo antecesor, Bermudez de Castro et al.., 1997, con especial atención a los datos aportados por la nueva mandíbula recuperada en el nivel TD6 del yacimiento de Gran Dolina, en la Sierra de Atapuerca, que refuerzan la identidad taxonómica de esta especie y su relación con el origen de H. sapiens. Se constata la hipótesis de que H. antecessor represente el último antecesor común de los neandertales y de las poblaciones modernas y se considera la posibilidad de un origen asiático para el primer poblamiento de Europa

X-ray Measurements of Black Hole X-ray Binary Source GRS 1915+ ...

Indian Academy of Sciences (India)

tribpo

March 30th, 1997 during a quiescent phase of the source. .... The field of view ... tagged with a 25µsec resolution and transmitted to ground on a 40 Kbit PCM/FM ... only composite model fits for the soft and hard X ray band are used and the ...

German (GRS) approach to accident analysis (part I). German licensing basis for accident analyses. Applicants accident analyses in second part license for Konvoi-plants. Appendix 1. Assessor accident analyses in second part license for Konvoi-plants. Appendix 2. Reference list of DBA to be considered in the safety status analysis of a PSR. Appendix 3a. Reference list of special very rare and BDB plant conditions to be considered in the safety status analysis of a PSE. Appendix 3b

International Nuclear Information System (INIS)

Velkov, K.

2002-01-01

Appendix 1: The Safety Analysis Report (S.A.R.) is presented from 3 Handbooks - ECC Handbook (LOCA), Plant Dynamics Handbook (Transients incl. ATWS), and Core Design Handbook. The first one Conceived as Living handbook, Basis for design, catalogue of transients, specifications and licensing. Handbook contains LOCA in primary system, it contains also core damage analysis, and description of codes, description of essential plant data and code input data. The second one consists of Basis for design, commissioning, operation, and catalogue of transients, specifications and licensing, as well as specified operation, disturbed operation, incidents, non-LOCA, SS-procedures and Code description. The third book consists of Reactivity balance and reactivity coefficients, efficiency of shutdown systems. Calculation of burn up cycle, power density distribution, and critical boron concentration. Also Codes used, as SAV79A standard analysis methodology including FASER for nuclear data generation, MEDIUM and PANBOX for static and transient core calculations. Appendix 2: The three TUEV (Technical Inspection Agencies) responsible for the three individual plants of type KONVOI: TUEV Bayern for ISAR-2, TUV-Hanover for KKE, TUEV-Stuttgart for GKN-2 and GRS performed the safety assessment. TUV-Bayern for disturbance and failure of secondary heat sink without loss of coolant (failure of main heat sink, erroneous operation of valves in MS and in FW system, failure of MFW supply), long term LONOP, performance of selected SBLOCA analyses. TUV Hanover for disturbances due to failure of MCPs, short term LONOP, damages of SG tubes incl. SGTR, performance of selected LOCA analyses (blowdown phase of LBLOCA). TUV-Stuttgart for breaks and leaks in MS and FW system with and without leaks in SG tubes. GRS for ATWS, sub-cooling transients due to disturbances on secondary side, initial and boundary conditions for transients with opening of pressurizer valves with and without stuck-open, most of the

Rate-adaptive BCH coding for Slepian-Wolf coding of highly correlated sources

DEFF Research Database (Denmark)

Forchhammer, Søren; Salmistraro, Matteo; Larsen, Knud J.

2012-01-01

This paper considers using BCH codes for distributed source coding using feedback. The focus is on coding using short block lengths for a binary source, X, having a high correlation between each symbol to be coded and a side information, Y, such that the marginal probability of each symbol, Xi in X......, given Y is highly skewed. In the analysis, noiseless feedback and noiseless communication are assumed. A rate-adaptive BCH code is presented and applied to distributed source coding. Simulation results for a fixed error probability show that rate-adaptive BCH achieves better performance than LDPCA (Low......-Density Parity-Check Accumulate) codes for high correlation between source symbols and the side information....

Analysis methods for structure reliability of piping components

International Nuclear Information System (INIS)

Schimpfke, T.; Grebner, H.; Sievers, J.

2004-01-01

In the frame of the German reactor safety research program of the Federal Ministry of Economics and Labour (BMWA) GRS has started to develop an analysis code named PROST (PRObabilistic STructure analysis) for estimating the leak and break probabilities of piping systems in nuclear power plants. The long-term objective of this development is to provide failure probabilities of passive components for probabilistic safety analysis of nuclear power plants. Up to now the code can be used for calculating fatigue problems. The paper mentions the main capabilities and theoretical background of the present PROST development and presents some of the results of a benchmark analysis in the frame of the European project NURBIM (Nuclear Risk Based Inspection Methodologies for Passive Components). (orig.)

Simulations of X-ray synchrotron beams using the EGS4 code system in medical applications

International Nuclear Information System (INIS)

Orion, I.; Henn, A.; Sagi, I.; Dilmanian, F.A.; Pena, L.; Rosenfeld, A.B.

2001-01-01

X-ray synchrotron beams are commonly used in biological and medical research. The availability of intense, polarized low-energy photons from the synchrotron beams provides a high dose transfer to biological materials. The EGS4 code system, which includes the photoelectron angular distribution, electron motion inside a magnetic field, and the LSCAT package, found to be the appropriate Monte Carlo code for synchrotron-produced X-ray simulations. The LSCAT package was developed in 1995 for the EGS4 code to contain the routines to simulate the linear polarization, the bound Compton, and the incoherent scattering functions. Three medical applications were demonstrated using the EGS4 Monte Carlo code as a proficient simulation code system for the synchrotron low-energy X-ray source. (orig.)

FORIG: a computer code for calculating radionuclide generation and depletion in fusion and fission reactors. User's manual

International Nuclear Information System (INIS)

Blink, J.A.

1985-03-01

In this manual we describe the use of the FORIG computer code to solve isotope-generation and depletion problems in fusion and fission reactors. FORIG runs on a Cray-1 computer and accepts more extensive activation cross sections than ORIGEN2 from which it was adapted. This report is an updated and a combined version of the previous ORIGEN2 and FORIG manuals. 7 refs., 15 figs., 13 tabs

The Shielding Analysis for the I and C System(GSF-2002) Development of an Instrumented Fuel Capsule

Energy Technology Data Exchange (ETDEWEB)

Lee, Dong Su; Kang, Young Hwan; Kim, Bong Goo; Park, Chang Je

2006-09-15

The Origen-2 code have been used to calculate the amount of nuclides generated in this study. It was validated by a comparison with the values calculated by this code and the measured ones by gathering the fission gases from two spent fuel rod (burned PWR fuel rods in Kori and Yungkwang). Error rate the between two values was estimated to be less than 10% in this study. The quantities of the nuclides generated by fission, calculated by the Origen-2, were larger than that by Origen-ARP. These differences of the two values are 1Ci to 40Ci. If the fuel is not intact, even if operated in a closed system, radioactive gases will be released to the gas flow tube from the instrumented capsule and the I and C system (GSF-2002) In a previous case study after 27 nuclides of Kr, Xe, I and Br series have been chosen, source term was generated. R/B ratio, used in this case, was calculated to serve this analysis with a reference to the ANS-5.4 model. And the leakage ratio adopted the result of the leak test in the PLUTO reactor. Specific points in the work place were determined and the radioactivity dose rates were calculated by running the Microshield 6.20. code at these points. Equivalent dose rates were 4.82X10-2{approx}3.74 {mu}Sv/hr at each point. These were distributed within the safety guide(6.25{mu}Sv/hr). In the failure case, almost all of the fission gases in the flow tube will be sucked into the delay tank, thus the condition of this tank will be a high radioactive area. Therefore the delay tank should be shielded by heavy materials and the thicknesses of them should be calculated with 9 cm and 20.5cm depending on performing conditions.

Verification calculations for the WWER version of the TRANSURANUS code

International Nuclear Information System (INIS)

Elenkov, D.; Boneva, S.; Georgieva, M.; Georgiev, S.; Schubert, A.; Van Uffelen, P.

2006-01-01

The paper presents part of the work performed in the study project 'Research and Development for Licensing of Nuclear Fuel in Bulgaria'. The main objective of the project is to provide assistance for solving technical questions of the fuel licensing process in Bulgaria. One important issue is the extension of the predictive capabilities of fuel performance codes for Russian-type WWER reactors. In the last decade, a series of international projects has been based on the TRANSURANUS fuel performance code: Specific models for WWER fuel have been developed and implemented in the code in the late 90's. In 2000-2003, basic verification work was done by using experimental data of nuclear fuel irradiated in WWER-440 reactors. While the present paper focuses on the analysis of WWER-1000 standard fuel under normal operating conditions, the above study project covers additional tasks: 1) Post-irradiation calculations of ramp tests performed in the DR3 test reactor of the Risoe National Laboratory (five instrumented fuel rods of the Risoe 3 dataset contained in the IFPE database) using the TRANSURANUS code; 2) Compilation of cross-section libraries for isotope evolution calculations in WWER-440 and WWER-1000 fuel assemblies using the ORIGEN-S code; 3) Analysis of current situation and needs for an extension of the curriculum in Nuclear Engineering at the Technical University of Sofia. In this paper the post-irradiation calculations of steady-state irradiation experiments with nuclear fuel for Russian-type WWER-1000 reactors, using the latest release of the TRANSURANUS code (v1m1j03)are presented. Regarding a comprehensive verification of modern fuel performance codes, the burn-up region above 40 MWd/kgU is of increasing importance. A number of new phenomena emerge at high fuel burn-up, implying the need for enlarged databases of postirradiation examinations (PIE). For one fuel assembly irradiated in a WWER-1000 reactor with a rod discharge burn-up between 50 and 55 MWd

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

International Nuclear Information System (INIS)

Konstantin Mikityuk; Sandro Pelloni; Paul Coddington

2005-01-01

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

Fuel performance analysis code 'FAIR'

International Nuclear Information System (INIS)

Swami Prasad, P.; Dutta, B.K.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

1994-01-01

For modelling nuclear reactor fuel rod behaviour of water cooled reactors under severe power maneuvering and high burnups, a mechanistic fuel performance analysis code FAIR has been developed. The code incorporates finite element based thermomechanical module, physically based fission gas release module and relevant models for modelling fuel related phenomena, such as, pellet cracking, densification and swelling, radial flux redistribution across the pellet due to the build up of plutonium near the pellet surface, pellet clad mechanical interaction/stress corrosion cracking (PCMI/SSC) failure of sheath etc. The code follows the established principles of fuel rod analysis programmes, such as coupling of thermal and mechanical solutions along with the fission gas release calculations, analysing different axial segments of fuel rod simultaneously, providing means for performing local analysis such as clad ridging analysis etc. The modular nature of the code offers flexibility in affecting modifications easily to the code for modelling MOX fuels and thorium based fuels. For performing analysis of fuel rods subjected to very long power histories within a reasonable amount of time, the code has been parallelised and is commissioned on the ANUPAM parallel processing system developed at Bhabha Atomic Research Centre (BARC). (author). 37 refs

Burnup dependent core neutronic calculations for research and training reactors via SCALE4.4

International Nuclear Information System (INIS)

Tombakoglu, M.; Cecen, Y.

2001-01-01

In this work, the full core modelling is performed to improve neutronic analyses capability for nuclear research reactors using SCALE4.4 code system. KENOV.a module of SCALE4.4 code system is utilized for full core neutronic analysis. The ORIGEN-S module is coupled with the KENOV.a module to perform burnup dependent neutronic analyses. Results of neutronic calculations for 1 st cycle of Cekmece TR-2 research reactor are presented. In particular, coupling of KENOV.a and ORIGEN-S modules of SCALE4.4 is discussed. The preliminary results of 2-D burnup dependent neutronic calculations are also given. These results are extended to burnup dependent core calculations of TRIGA Mark-II research reactors. The code system developed here is similar to the code system that couples MCNP and ORIGEN2.(author)

SCALE Code System

Energy Technology Data Exchange (ETDEWEB)

Jessee, Matthew Anderson [ORNL

2016-04-01

depletion with TRITON (T5-DEPL/T6-DEPL),• CE sensitivity/uncertainty analysis with TSUNAMI-3D,• Simplified and efficient LWR lattice physics with Polaris,• Large scale detailed spent fuel characterization with ORIGAMI and ORIGAMI Automator,• Advanced fission source convergence acceleration capabilities with Sourcerer,• Nuclear data library generation with AMPX, and• Integrated user interface with Fulcrum.Enhanced capabilities include:• Accurate and efficient CE Monte Carlo methods for eigenvalue and fixed source calculations,• Improved MG resonance self-shielding methodologies and data,• Resonance self-shielding with modernized and efficient XSProc integrated into most sequences,• Accelerated calculations with TRITON/NEWT (generally 4x faster than SCALE 6.1),• Spent fuel characterization with 1470 new reactor-specific libraries for ORIGEN,• Modernization of ORIGEN (Chebyshev Rational Approximation Method [CRAM] solver, API for high-performance depletion, new keyword input format)• Extension of the maximum mixture number to values well beyond the previous limit of 2147 to ~2 billion,• Nuclear data formats enabling the use of more than 999 energy groups,• Updated standard composition library to provide more accurate use of natural abundances, andvi• Numerous other enhancements for improved usability and stability.

FERRET data analysis code

International Nuclear Information System (INIS)

Schmittroth, F.

1979-09-01

A documentation of the FERRET data analysis code is given. The code provides a way to combine related measurements and calculations in a consistent evaluation. Basically a very general least-squares code, it is oriented towards problems frequently encountered in nuclear data and reactor physics. A strong emphasis is on the proper treatment of uncertainties and correlations and in providing quantitative uncertainty estimates. Documentation includes a review of the method, structure of the code, input formats, and examples

Analysis of a large break LOCA in the cold leg of the WWER-440/W-213 plant Griefswald, Unit 5

International Nuclear Information System (INIS)

Horche, W.

1993-01-01

The Gessellschaft fur Anlagen und Reaktorsicherheit (GRS) has performed a safety evaluation of the nuclear power plant (NPP) Greifswald, unit 5, of the Soviet type WWER-440/W-213, in cooperation with the French Institute de Protection of de Surete Nucleaire (IPSN) and other partners. Within this project an independent accident analysis is performed by GRS in order to assess the results of existing analysis and to supplement them. In this paper the analysis of the double-ended guillotine break (DEGB) of one cold leg of the main circulation pipe is described. The major objective of the calculation was the investigation of the accident sequence with reduced availability of the emergency core cooling system (single failure criterion). In addition, the simultaneous loss of onsite and offsite power and the failure of scram were assumed. The thermal-hydraulic system code ATHLET/FLUT, developed at GRS and already applied for the safety analysis of several WWER plants, was chosen again. The pressure in the confinement, the back pressure for the discharge model, was calculated as a function of time for this accident separately with GRS-Code RALOC. Furthermore, it was necessary to model the local concentration of direct accumulator injection into the reactor vessel with the help of a special two-channel model of the core and upper plenum. For this model, results were considered obtained from the 1:1 scaled test facility UPTF. It was assumed that only 25% of the upper plenum and core volume is directly penetrated by the injected water. The DEGB was defined in that loop, which is connected with one of three low-pressure injection subsystems. This means that this injected water flows towards the leak without passing the core. As single failure the failure of one of three diesel generators was assumed. The full paper will contain nodalization schemes, which are generated by the ATHLET-Input-Grafic

A study of Venus surface elemental composition from 14 MeV neutron induced gamma ray spectroscopy: Activation analysis

International Nuclear Information System (INIS)

Jun, I.; Kim, W.; Smith, M.; Mitrofanov, I.; Litvak, M.

2011-01-01

The surface elemental composition of Venus can be determined using an artificially pulsed 14 MeV neutron generator (PNG) combined with a gamma ray spectrometer (GRS). The 14 MeV neutrons will interact with the surface materials and generate gamma rays, characteristic of specific elements, whose energy spectrum will be measured by GRS. These characteristic gamma rays are produced mainly through 3 different neutron interaction mechanisms: capture, inelastic, and activation reactions. Each reaction type has a different neutron energy dependency and different time scale for gamma ray production and transport. Certain elements are more easily identified through one reaction type over the others. Thus, careful analysis of the gamma ray spectra during and after the neutron pulse provides a comprehensive understanding of the surface elemental composition. In this paper, we use a well-tested neutron/gamma transport code, called Monte Carlo N-Particles (MCNP), to investigate the measurement capability of a PNG-GRS detection system through the neutron activation reactions. An activation analysis was performed for a representative soil composition of Venus with a notional operational scenario of PNG and GRS. The analysis shows that the proposed instrument concept can identify most of the modeled surface elements at Venus with sufficient accuracy through the activation mode. Specifically, U, Th, K, Si can be measured to within 1%, Fe within 2%, Al within 10%, Ca within 5%, Mg with 15%, Mn with 20%, and Cl within 6%. Although modeled in the analysis, it is shown that the activation mode alone cannot distinguish the S and Ti peaks.

Reactor safety in Eastern Europe

International Nuclear Information System (INIS)

1995-02-01

The papers presented to the GRS colloquium refer to the cooperative activities for reactor accident analysis and modification of the GRS computer codes for their application to reactors of the Russian design types of WWER or RBMK. Another topic is the safety of RBMK reactors in particular, and the current status of investigations and studies addressing the containment of unit 4 of the Chernobyl reactor station. All papers are indexed separately in report GRS--117. (HP)

NuSTAR spectroscopy of GRS 1915+105: Disk reflection, spin, and connections to jets

DEFF Research Database (Denmark)

Miller, J. M.; Parker, M. L.; Fuerst, F.

2013-01-01

We report on the results of spectral fits made to a NuSTAR observation of the black hole GRS 1915+105 in a “plateau” state. This state is of special interest because it is similar to the “low/hard” state seen in other black holes, especially in that compact, steady jets are launched in this phase...

El origen no humano de la música

Directory of Open Access Journals (Sweden)

Juan David Leongómez

2008-09-01

Full Text Available Resumen: Si la música tiene un origen cultural -como suele asumirse- ¿cómo puede explicarse su universalidad? Este artículo parte de este problema para plantear la hipótesis de que el fenómeno musical tenga un origen anterior al hombre moderno, o incluso anterior al hombre mismo y sea el resultado de una serie de adaptaciones al medio; de que tenga un origen biológico y no cultural. Tras analizar diversos estudios en áreas tan diversas como la semiótica, la neurología, la etología y la evolución, en los que se ve que muchas especies animales compartimos la capacidad de reconocer y entender signos, las estructuras y capacidades necesarias para su percepción, el placer y potencialmente la producción de memorias e imágenes metafóricas que se evocan cuando escuchamos música, los procesos culturales implicados, y que en muchos casos nos comunicamos acústicamente con los mismos objetivos e incluso sobre los mismos patrones universales que los humanos usamos al hacer música, se llega a la conclusión de que es posible que la música tenga un origen no humano y que sea producto de adaptaciones evolutivas, y a la sugestiva idea de que el origen de este fenómeno -es decir el desarrollo de las estructuras básicas implicadas en los procesos musicales- pueda rastrearse teóricamente incluso hasta los reptiles. Abstract: If music has a cultural origin -as is often assumed- how can one explain its universality? This paper is based on this question to raise the hypothesis that the musical phenomenon has its origin before the arising of modern man, or even before the evolution of man himself and, therefore, is the result of a series of adaptations to the environment, having a biological rather than a cultural origin. After considering further studies in areas as diverse as Semiotics, Neuroscience, Ethology and Evolution, showing that the physiological, behavioural, cognitive and perceptual characteristics related to the musical production

Radionuclide composition in nuclear fuel waste. Calculations performed by ORIGEN2

International Nuclear Information System (INIS)

Lyckman, C.

1996-01-01

The report accounts for results from calculations on the content of radionuclides in nuclear fuel waste. It also accounts for the results from calculations on the neutron flow from spent fuel, which is very important during transports. The calculations have been performed using the ORIGEN2 software. The results have been compared to other results from earlier versions of ORIGEN and some differences have been discovered. This is due to the updating of the software. 7 refs, 10 figs, 15 tabs

Status of SPACE Safety Analysis Code Development

International Nuclear Information System (INIS)

Lee, Dong Hyuk; Yang, Chang Keun; Kim, Se Yun; Ha, Sang Jun

2009-01-01

In 2006, the Korean the Korean nuclear industry started developing a thermal-hydraulic analysis code for safety analysis of PWR(Pressurized Water Reactor). The new code is named as SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). The SPACE code can solve two-fluid, three-field governing equations in one dimensional or three dimensional geometry. The SPACE code has many component models required for modeling a PWR, such as reactor coolant pump, safety injection tank, etc. The programming language used in the new code is C++, for new generation of engineers who are more comfortable with C/C++ than old FORTRAN language. This paper describes general characteristics of SPACE code and current status of SPACE code development

Fragile X mental retardation protein participates in non-coding RNA pathways.

Science.gov (United States)

Li, En-Hui; Zhao, Xin; Zhang, Ce; Liu, Wei

2018-02-20

Fragile X syndrome is one of the most common forms of inherited intellectual disability. It is caused by mutations of the Fragile X mental retardation 1(FMR1) gene, resulting in either the loss or abnormal expression of the Fragile X mental retardation protein (FMRP). Recent research showed that FMRP participates in non-coding RNA pathways and plays various important roles in physiology, thereby extending our knowledge of the pathogenesis of the Fragile X syndrome. Initial studies showed that the Drosophila FMRP participates in siRNA and miRNA pathways by interacting with Dicer, Ago1 and Ago2, involved in neural activity and the fate determination of the germline stem cells. Subsequent studies showed that the Drosophila FMRP participates in piRNA pathway by interacting with Aub, Ago1 and Piwi in the maintenance of normal chromatin structures and genomic stability. More recent studies showed that FMRP is associated with lncRNA pathway, suggesting a potential role for the involvement in the clinical manifestations. In this review, we summarize the novel findings and explore the relationship between FMRP and non-coding RNA pathways, particularly the piRNA pathway, thereby providing critical insights on the molecular pathogenesis of Fragile X syndrome, and potential translational applications in clinical management of the disease.

Progress report within the series of GRS-F progress reports on reactor safety, sponsored by the Federal Ministry of Economics and Labour. Period: 1 January - 30 June 2003

International Nuclear Information System (INIS)

2003-01-01

The Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, by order of the BMWi, continuously issues information on the status of the investigations into the safety of nuclear power plants by publishing semi-annual and annual progress reports within the series of GRS-F-Fortschrittsberichte (GRS-F-Progress Reports). Each progress report represents a compilation of individual reports about the objectives, work performed, results achieved, next steps of the work etc. The individual reports are prepared in a standard form by the research organisations themselves as documentation of their progress in work and are published by the Research Management Division of GRS within the framework of general information on the progress in reactor safety research. The compilation of the reports is classified according to general topics related to reactor safety research. Further, use is made of the classification system ''Joint Safety Research Index'' of the CEC (commission of the european communities). The reports are arranged in sequence of their project numbers. (orig.) [de

Progress report within the series of GRS-F progress reports on reactor safety, sponsored by the Federal Ministry of Economics and Labour. Period: 1 July - 31 December 2003

International Nuclear Information System (INIS)

2004-01-01

The Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, by order of the BMWi, continuously issues information on the status of the investigations into the safety of nuclear power plants by publishing semi-annual and annual progress reports within the series of GRS-F-Fortschrittsberichte (GRS-F-Progress Reports). Each progress report represents a compilation of individual reports about the objectives, work performed, results achieved, next steps of the work etc. The individual reports are prepared in a standard form by the research organisations themselves as documentation of their progress in work and are published by the Research Management Division of GRS within the framework of general information on the progress in reactor safety research. The compilation of the reports is classified according to general topics related to reactor safety research. Further, use is made of the classification system ''Joint Safety Research Index'' of the CEC (commission of the european communities). The reports are arranged in sequence of their project numbers. (orig.) [de

Safety code 19: recommended safety procedures for the selection, installation and use of x-ray diffraction equipment

International Nuclear Information System (INIS)

1984-01-01

This document is one of a series of Safety Codes prepared by the Radiation Protection Bureau to set out requirements for the safe use of radiation emitting devices. The equipment and installation guidelines and safety procedures detailed in this Code are primarily for the instruction and guidance of persons employed in Federal Public Service Departments and Agencies, as well as those coming under the jurisdiction of the Canada Labour Code. This Safety Code is also intended to assist other users of X-ray diffraction equipment to select safe equipment and to install and use it so that the radiation hazard to the operator and other persons in its vicinity is negligible. It should be noted that facilities under provincial jurisdiction may be subject to requirements specified under provincial statutes. This Code supersedes Safety Code RPD-SC-7, entitled 'Requirements For Non-Medical X-Ray Equipment, Use and Installation', insofar as X-ray diffraction equipment is concerned, and it is intended to complement X-ray equipment design, construction and performance standards promulgated under the Radiation Emitting Devices Act

Development of a reference spent fuel library of 17x17 PWR fuel assemblies

International Nuclear Information System (INIS)

Rossa, Riccardo; Borella, Alessandro; Van der Meer, Klaas

2013-01-01

One of the most common ways to investigate new Non-Destructive Assays (NDA) for the spent fuel assemblies are Monte Carlo simulations. In order to build realistic models the user must define in an accurate way the material compositions and the source terms in the system. This information can be obtained using burnup codes such as ORIGEN-ARP and ALEPH2.2, developed at SCK-CEN. These software applications allow the user to select the irradiation history of the fuel assembly and to calculate the corresponding isotopic composition and neutron/gamma emissions as a function of time. In the framework of the development of an innovative NDA for spent fuel verifications, SCK•CEN built an extensive fuel library for 17x17 PWR assemblies, using both ORIGEN-ARP and ALEPH2.2. The parameters considered in the calculations were initial enrichment, discharge burnup, and cooling time. The combination of these variables allows to obtain more than 1500 test cases. Considering the broad range of the parameters, the fuel library can be used for other purposes apart from spent fuel verifications, for instance for the direct disposal in geological repositories. In addition to the isotopic composition of the spent fuel, the neutron and photon emissions were also calculated and compared between the two codes. The comparison of the isotopic composition showed a good agreement between the codes for most of the relevant isotopes in the spent fuel. However, specific isotopes as well as neutron and gamma spectra still need to be investigated in detail.

Self-complementary circular codes in coding theory.

Science.gov (United States)

Fimmel, Elena; Michel, Christian J; Starman, Martin; Strüngmann, Lutz

2018-04-01

Self-complementary circular codes are involved in pairing genetic processes. A maximal [Formula: see text] self-complementary circular code X of trinucleotides was identified in genes of bacteria, archaea, eukaryotes, plasmids and viruses (Michel in Life 7(20):1-16 2017, J Theor Biol 380:156-177, 2015; Arquès and Michel in J Theor Biol 182:45-58 1996). In this paper, self-complementary circular codes are investigated using the graph theory approach recently formulated in Fimmel et al. (Philos Trans R Soc A 374:20150058, 2016). A directed graph [Formula: see text] associated with any code X mirrors the properties of the code. In the present paper, we demonstrate a necessary condition for the self-complementarity of an arbitrary code X in terms of the graph theory. The same condition has been proven to be sufficient for codes which are circular and of large size [Formula: see text] trinucleotides, in particular for maximal circular codes ([Formula: see text] trinucleotides). For codes of small-size [Formula: see text] trinucleotides, some very rare counterexamples have been constructed. Furthermore, the length and the structure of the longest paths in the graphs associated with the self-complementary circular codes are investigated. It has been proven that the longest paths in such graphs determine the reading frame for the self-complementary circular codes. By applying this result, the reading frame in any arbitrary sequence of trinucleotides is retrieved after at most 15 nucleotides, i.e., 5 consecutive trinucleotides, from the circular code X identified in genes. Thus, an X motif of a length of at least 15 nucleotides in an arbitrary sequence of trinucleotides (not necessarily all of them belonging to X) uniquely defines the reading (correct) frame, an important criterion for analyzing the X motifs in genes in the future.

The Mystery of History in Origen: a Conflict of Interpretations

Directory of Open Access Journals (Sweden)

P. MIKHAYLOV

2014-02-01

Full Text Available The proposed issue represents an analytical review of the most striking as well as opposite interpretations of theology of history of Origen. They belong to the Christian theologians and researchers in the fi eld of patristic thought — archpriest George Florovsky and cardinal Henri de Lubac. Both approaches are characterized by large and at the same time contextual vision of their subject. As has been shown behind each explanation is lying certain theological position of author determined by his biographical and historical situation. For example an impulse for Florovsky in turning to the historical thought of Origen is an ideological confl ict of his own time, as for de Lubac such a reason became a demand of contemporary catholic theology in hermeneutical resource of ancient Christian tradition. The comparison of diff erent approaches is completed by short characteristic of an essence of historical and hermeneutical position of Origen. As has been shown by the latest researches of B. Studer history as such for Origen was completely enclosed in the Scripture. The Greek word Ñστορία is related to the idea of narration. Some kind of “bibliocentrism” is typical for his thought, and the only possible history is a narration of Holy Scripture as well as its interpretation is a spiritual explanation of its history. Scripture includes the fi rst moments of being of the world, the creation of manhood, the fall and a long way of humanity towards Incarnation, that collects in itself all the dimensions of biblical testimony and conducts believers to its fi nal fulfi llment in the future age

Analysis of Elektrogorsk 108 test facility experimental data

International Nuclear Information System (INIS)

Urbonas, R.

2001-01-01

In the paper an evaluation of experimental data obtained at Russian Elektrogorsk 108 (E-108) test facility is presented. E-108 facility is a scaled model of Russian RBMK design reactor. An attempt to validate state-of-the-art thermal hydraulic codes on the basis of E-108 test facility was made. Originally these codes were developed and validated for BWRs and PWRs. Since state-of-art thermal hydraulic codes are widely used for simulation of RBMK reactors further codes' implementation and validation is required. The facility was modelled by employing RELAP5 (INEEL, USA) thermal hydraulic system analysis best estimate code. The results show dependence from number of nodes used in the heated channels, frictional and form losses employed. The obtained oscillatory behaviour is resulted by density wave and critical heat flux. It is shown that codes are able to predict thermal hydraulic instability and sudden heat structure temperature excursion, when critical heat flux is approached, well. In addition, an uncertainty analysis of one of the experiments was performed by employing GRS developed System for Uncertainty and Sensitivity Analysis (SUSA). It was one of the first attempts to use this statistic-based methodology in Lithuania.(author)

Ondas T negativas gigantes de origen indeterminado

Directory of Open Access Journals (Sweden)

Eduardo Eyheremendy

2009-01-01

Full Text Available Se define onda T negativa gigante o inversión masiva de la onda T a la aparición de ondas T negativas con una amplitud = 1 mV en por lo menos dos derivaciones contiguas del ECG. Se presenta el caso de una paciente de 66 años con antecedentes de dislipidemia tipo IV e hipertensión arterial leve, a la que en buen estado de salud y totalmente asintomática en un examen de rutina se le detecta en el ECG de superficie un hemibloqueo anterior y ondas T negativas gigantes en las derivaciones I, II, aVL, aVF y de V1 a V6. La paciente fue internada en la UCI, donde se descartó que fueran de origen coronario. Durante el seguimiento, las ondas T se tornaron menos negativas hasta que al cabo de un tiempo (45 días aproximadamente se normalizaron. A pesar de los estudios cardiológicos y no cardiológicos realizados, no se pudo establecer su origen.REV ARGENT CARDIOL 2009;77:131-134.

A Study on MCNPX-CINDER90 System for Activation Analysis

International Nuclear Information System (INIS)

Kim, Sung-Min; Kim, Myung Hyun

2014-01-01

Neutron spectrum at every cell is different depending on the geometrical characteristics. For the problem of neutron activation to the reactor containment wall, neutron spectra are varying from the center of core to the containment wall. Therefore, process of one group cross section library for all relevant isotopes need an extensive works for all locations. On the other hand, CINDER can concerns 3 dimensional geometry effects and handles up to 3,400 nuclides. It is believed that CINDER is more reliable and accurate compared to ORIGEN because it treats 63-group cross section. In this paper, a new coupling of MCNP-CINDER was tested and compared with MCNP-ORIGEN and MCNPX 2.6.0. MCNPX is a coupled code of MCNP with CINDER90 for fuel depletion chain only. The simple UO2 single pin was modelled in order to compare and evaluate the fission product densities for fuel depletion chains. The simple reactor pressure vessel (RPV) and concrete wall were modelled for the comparison of isotopic inventory chains for activation products simulating the RPV boundaries. The UO2 single pin, simple RPV, and concrete wall model were modeled in order to compare inventory change and radioactivity with MCNPX 2.6..0, ORIGEN 2.1, and CINDER90. It is regarded that CINDER90 is more reliable and accurate compared to ORIGEN 2.1 because it has 63-group multi cross section library. In addition, several error by the approximation of model description and the difference of the cross section library, fission yield data, and et cetera in each code result in the relative error between each code. Also as the decay chain is the simple nuclide, the difference of the result is little between the code and it is the complicated nuclide, the difference of the result is large between the codes. So the error by the difference of the decay chain has to be considered between each code. As a result of this, the practicality of MCNP-CINDER system was verified and it is expected to be used for the study on the

geográfico del origen de nombres y apellidos

Directory of Open Access Journals (Sweden)

Pablo Mateos Rodríguez

2006-01-01

Full Text Available El estudio geodemográfico de la composición y distribución de comunidades de minorías étnicas o extranjeros se ha convertido en una necesidad prioritaria para las autoridades públicas en los principales países o regiones receptoras. Esta comunicación presenta una nueva metodología, inicialmente desarrollada para el Reino Unido, que permite clasificar a la población en grupos étnicos de acuerdo con el origen geográfico, lingüístico, religioso o cultural de sus nombres y apellidos (origen étnico. Este método permite la actualización continua de datos sobre el origen étnico de la población y una mayor resolución nominal y espacial que el Censo. Se presentan los resultados de la aplicación de dicha metodología al estudio de la distribución geográfica y segregación residencial de poblaciones de minorías étnicas en España, utilizando un directorio telefónico como registro de nombres de la población total. Se utilizan Sistemas de Información Geográfica a una escala nacional y regional para medir la desigual distribución espacial de dichas minorías, indicando las limitaciones y potencial de este método.

Nuclear data evaluation and group constant generation for reactor analysis

Energy Technology Data Exchange (ETDEWEB)

Kim, Jung Do; Gil, Choong Sub; Lee, Jong Tai; Hwang, Won Guk [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of)

1990-12-01

A one-group cross section data base used by the ORIGEN2 computer code to simulate the depletion, buildup and decay of radionuclides in research reactor was developed. For this, ENDF/B-IV or -V data was processed using the NJOY code system into 69-group data. The burn-up-dependent weighting spectra were calculated with the WIMS-KAERI code, and then the 69-group data were collapsed to one-group using the spectra. The ORIGEN2 depletion calculations for the KMRR fuel were performed using an old PWR and the new data base. By comparing these results to the WIMS-KAERI calculations, it is seen that the results of actinide composition calculated by the ORIGEN2 with the new data base turn out to be in an excellent agreement with the WIMS-KAERI results in the range up to 120 GWD/MTIHM burnup. (Author).

Polyacrylonitrile Fibers Anchored Cobalt/Graphene Sheet Nanocomposite: A Low-Cost, High-Performance and Reusable Catalyst for Hydrogen Generation.

Science.gov (United States)

Zhang, Fei; Huang, Guoji; Hou, Chengyi; Wang, Hongzhi; Zhang, Qinghong; Li, Yaogang

2016-06-01

Cobalt and its composites are known to be active and inexpensive catalysts in sodium borohydride (NaBH4) hydrolysis to generate clean and renewable hydrogen energy. A novel fiber catalyst, cobalt/graphene sheet nanocomposite anchored on polyacrylonitrile fibers (Co/GRs-PANFs), which can be easily recycled and used in any reactor with different shapes, were synthesized by anchoring cobalt/graphene (Co/GRs) on polyacrylonitrile fibers coated with graphene (GRs-PANFs) at low temperature. The unique structure design effectively prevents the inter-sheet restacking of Co/GRs and fully exploits the large surface area of novel hybrid material for generate hydrogen. And the extra electron transfer path supplied by GRs on the surface of GRs-PANFs can also enhance their catalysis performances. The catalytic activity of the catalyst was investigated by the hydrolysis of NaBH4 in aqueous solution with GRs-PANFs. GRs powders and Co powders were used as control groups. It was found that both GRs and fiber contributed to the hydrogen generation rate of Co/GRs-PANFs (3222 mL x min(-1) x g(-1)), which is much higher than that of cobalt powders (915 mL x min(-1) x g(-1)) and Co/GRs (995 mL x min(-1) x g(-1)). The improved hydrogen generation rate, low cost and uncomplicated recycling make the Co/GRs-PANFs promising candidate as catalysts for hydrogen generation.

Participation of the GRS in the 'Degraded piping program' of the USNRC

International Nuclear Information System (INIS)

Azodi, D.; Hoefler, A.; Sievers, J.

1989-01-01

During the course of the Degraded Piping Program, Battelle was the organizer of three round-robin activities for advanced elasto-plastic fracture mechanics calculations (e.g. finite element method and J-estimation scheme). GRS participated in all of them and submitted finite element and J-estimation scheme results. As a main result of round-robin on elasto-plastic fracture mechanics calculations: Based on finite element calculations, the J-integral method (energy release rate) provided the ability to describe the fracture behaviour of flawed piping even in a very ductile material. (orig./HP) [de

Adsorbentes sólidos de origen natural para metales pesados en solución acuosa

OpenAIRE

Ángeles Villón, Luis Rosas

2013-01-01

En la presente investigación se plantea el resumir y ordenar las investigaciones respecto a los a los adsorbentes sólidos de origen natural para metales pesados en solución acuosa, a través de la clasificación de los mismos por su origen natural, el presentar los parámetros más utilizados en la caracterización de (los adsorbentes, así como en los procesos de adsorción e identificar los adsorbentes de origen natural con capacidad de ser reutilizados. Es así que se ha efectuado esta clasific...

Computer codes for safety analysis

International Nuclear Information System (INIS)

Holland, D.F.

1986-11-01

Computer codes for fusion safety analysis have been under development in the United States for about a decade. This paper will discuss five codes that are currently under development by the Fusion Safety Program. The purpose and capability of each code will be presented, a sample given, followed by a discussion of the present status and future development plans

Automation of ORIGEN2 calculations for the transuranic waste baseline inventory database using a pre-processor and a post-processor

International Nuclear Information System (INIS)

Liscum-Powell, J.

1997-06-01

The purpose of the work described in this report was to automate ORIGEN2 calculations for the Waste Isolation Pilot Plant (WIPP) Transuranic Waste Baseline Inventory Database (WTWBID); this was done by developing a pre-processor to generate ORIGEN2 input files from WWBID inventory files and a post-processor to remove excess information from the ORIGEN2 output files. The calculations performed with ORIGEN2 estimate the radioactive decay and buildup of various radionuclides in the waste streams identified in the WTWBID. The resulting radionuclide inventories are needed for performance assessment calculations for the WIPP site. The work resulted in the development of PreORG, which requires interaction with the user to generate ORIGEN2 input files on a site-by-site basis, and PostORG, which processes ORIGEN2 output into more manageable files. Both programs are written in the FORTRAN 77 computer language. After running PreORG, the user will run ORIGEN2 to generate the desired data; upon completion of ORIGEN2 calculations, the user can run PostORG to process the output to make it more manageable. All the programs run on a 386 PC or higher with a math co-processor or a computer platform running under VMS operating system. The pre- and post-processors for ORIGEN2 were generated for use with Rev. 1 data of the WTWBID and can also be used with Rev. 2 and 3 data of the TWBID (Transuranic Waste Baseline Inventory Database)

Extension of the code COCOSYS to a dispersion code for smoke and carbon monoxide

International Nuclear Information System (INIS)

Sdouz, Gert; Mayrhofer, Robert

2009-01-01

The code COCOSYS (Containment Code SYStem) was developed by GRS in Germany to simulate processes and nuclear plant states during severe accidents in the containments of light water reactors. It contains several physical models, especially a module for aerosol behaviour. The goal of this work was to extend COCOSYS for applications in more general geometries mainly for complex public buildings. For the application in public buildings models for air condition systems and different boundary conditions according to different environments were developed. The principal application of the extended code COCOSYS is in the area of emergency situations especially in the simulation for carbon monoxide and smoke dispersion. After developing and implementing the new models several test calculations were performed to evaluate the functionality of the extended code. The comparison of the results with those of the original COCOSYS code showed no discrepancies. For the first realistic application several fire emergency scenarios in the Vienna General Hospital (AKH) were selected in agreement with the fire department of the hospital. One of the scenarios addresses the danger of carbon monoxide (CO) and smoke leaking into a fire protection section through a damaged fire protection flap. As a result of the dispersion simulation the CO-concentration in all of the rooms is obtained. Together with additional results as deposition and smoke dispersion the outcome of the simulation can be used for training. Among the next steps are the validation of the new models and the selection of critical scenarios. (author)

Performance of asynchronous fiber-optic code division multiple access system based on three-dimensional wavelength/time/space codes and its link analysis.

Science.gov (United States)

Singh, Jaswinder

2010-03-10

A novel family of three-dimensional (3-D) wavelength/time/space codes for asynchronous optical code-division-multiple-access (CDMA) systems with "zero" off-peak autocorrelation and "unity" cross correlation is reported. Antipodal signaling and differential detection is employed in the system. A maximum of [(W x T+1) x W] codes are generated for unity cross correlation, where W and T are the number of wavelengths and time chips used in the code and are prime. The conditions for violation of the cross-correlation constraint are discussed. The expressions for number of generated codes are determined for various code dimensions. It is found that the maximum number of codes are generated for S systems. The codes have a code-set-size to code-size ratio greater than W/S. For instance, with a code size of 2065 (59 x 7 x 5), a total of 12,213 users can be supported, and 130 simultaneous users at a bit-error rate (BER) of 10(-9). An arrayed-waveguide-grating-based reconfigurable encoder/decoder design for 2-D implementation for the 3-D codes is presented so that the need for multiple star couplers and fiber ribbons is eliminated. The hardware requirements of the coders used for various modulation/detection schemes are given. The effect of insertion loss in the coders is shown to be significantly reduced with loss compensation by using an amplifier after encoding. An optical CDMA system for four users is simulated and the results presented show the improvement in performance with the use of loss compensation.

Non-coding RNAs and epigenome: de novo DNA methylation, allelic exclusion and X-inactivation

Directory of Open Access Journals (Sweden)

V. A. Halytskiy

2013-12-01

Full Text Available Non-coding RNAs are widespread class of cell RNAs. They participate in many important processes in cells – signaling, posttranscriptional silencing, protein biosynthesis, splicing, maintenance of genome stability, telomere lengthening, X-inactivation. Nevertheless, activity of these RNAs is not restricted to posttranscriptional sphere, but cover also processes that change or maintain the epigenetic information. Non-coding RNAs can directly bind to the DNA targets and cause their repression through recruitment of DNA methyltransferases as well as chromatin modifying enzymes. Such events constitute molecular mechanism of the RNA-dependent DNA methylation. It is possible, that the RNA-DNA interaction is universal mechanism triggering DNA methylation de novo. Allelic exclusion can be also based on described mechanism. This phenomenon takes place, when non-coding RNA, which precursor is transcribed from one allele, triggers DNA methylation in all other alleles present in the cell. Note, that miRNA-mediated transcriptional silencing resembles allelic exclusion, because both miRNA gene and genes, which can be targeted by this miRNA, contain elements with the same sequences. It can be assumed that RNA-dependent DNA methylation and allelic exclusion originated with the purpose of counteracting the activity of mobile genetic elements. Probably, thinning and deregulation of the cellular non-coding RNA pattern allows reactivation of silent mobile genetic elements resulting in genome instability that leads to ageing and carcinogenesis. In the course of X-inactivation, DNA methylation and subsequent hete­rochromatinization of X chromosome can be triggered by direct hybridization of 5′-end of large non-coding RNA Xist with DNA targets in remote regions of the X chromosome.

Light-water reactor safety analysis codes

International Nuclear Information System (INIS)

Jackson, J.F.; Ransom, V.H.; Ybarrondo, L.J.; Liles, D.R.

1980-01-01

A brief review of the evolution of light-water reactor safety analysis codes is presented. Included is a summary comparison of the technical capabilities of major system codes. Three recent codes are described in more detail to serve as examples of currently used techniques. Example comparisons between calculated results using these codes and experimental data are given. Finally, a brief evaluation of current code capability and future development trends is presented

SWAT2: The improved SWAT code system by incorporating the continuous energy Monte Carlo code MVP

International Nuclear Information System (INIS)

Mochizuki, Hiroki; Suyama, Kenya; Okuno, Hiroshi

2003-01-01

SWAT is a code system, which performs the burnup calculation by the combination of the neutronics calculation code, SRAC95 and the one group burnup calculation code, ORIGEN2.1. The SWAT code system can deal with the cell geometry in SRAC95. However, a precise treatment of resonance absorptions by the SRAC95 code using the ultra-fine group cross section library is not directly applicable to two- or three-dimensional geometry models, because of restrictions in SRAC95. To overcome this problem, SWAT2 which newly introduced the continuous energy Monte Carlo code, MVP into SWAT was developed. Thereby, the burnup calculation by the continuous energy in any geometry became possible. Moreover, using the 147 group cross section library called SWAT library, the reactions which are not dealt with by SRAC95 and MVP can be treated. OECD/NEA burnup credit criticality safety benchmark problems Phase-IB (PWR, a single pin cell model) and Phase-IIIB (BWR, fuel assembly model) were calculated as a verification of SWAT2, and the results were compared with the average values of calculation results of burnup calculation code of each organization. Through two benchmark problems, it was confirmed that SWAT2 was applicable to the burnup calculation of the complicated geometry. (author)

Application of X-ray imagery with coded aperture to laser-produced plasmas

International Nuclear Information System (INIS)

Fleurot, N.; Gex, J.P.; Sauneuf, R.; Beaucoudray, N. de.

1981-05-01

The analysis of the phenomena occurring inside microplasmas generated by laser techniques is often performed with images obtained by the X-rays or alpha particles emitted. Pinhole chambers are extensively used for this type of diagnostic. The number of photons or particles collected, however, decreases with increasing resolution as does the signal-to-noise (S/B) ratio. The coding schema shows that the coder and detector are both flat and parallel to one another. Each object point projects the shadow of the coder onto the detector. The coder shadow is a homothetic figure of the coder. The dimensions and the position of this shadow depend on the position of the point in space, which gives three dimensional information [fr

Developments of HTGR thermofluid dynamic analysis codes and HTGR plant dynamic simulation code

International Nuclear Information System (INIS)

Tanaka, Mitsuhiro; Izaki, Makoto; Koike, Hiroyuki; Tokumitsu, Masashi

1983-01-01

In nuclear power plants as well as high temperature gas-cooled reactor plants, the design is mostly performed on the basis of the results after their characteristics have been grasped by carrying out the numerical simulation using the analysis code. Also in Kawasaki Heavy Industries Ltd., on the basis of the system engineering accumulated with gas-cooled reactors since several years ago, the preparation and systematization of analysis codes have been advanced, aiming at lining up the analysis codes for heat transferring flow and control characteristics, taking up HTGR plants as the main object. In this report, a part of the results is described. The example of the analysis applying the two-dimensional compressible flow analysis codes SOLA-VOF and SALE-2D, which were developed by Los Alamos National Laboratory in USA and modified for use in Kawasaki, to HTGR system is reported. Besides, Kawasaki has developed the control characteristics analyzing code DYSCO by which the change of system composition is easy and high versatility is available. The outline, fundamental equations, fundamental algorithms and examples of application of the SOLA-VOF and SALE-2D, the present status of system characteristic simulation codes and the outline of the DYSCO are described. (Kako, I.)

A method for including external feed in depletion calculations with CRAM and implementation into ORIGEN

International Nuclear Information System (INIS)

Isotalo, A.E.; Wieselquist, W.A.

2015-01-01

Highlights: • A method for handling external feed in depletion calculations with CRAM. • Source term can have polynomial or exponentially decaying time-dependence. • CRAM with source term and adjoint capability implemented to ORIGEN in SCALE. • The new solver is faster and more accurate than the original solver of ORIGEN. - Abstract: A method for including external feed with polynomial time dependence in depletion calculations with the Chebyshev Rational Approximation Method (CRAM) is presented and the implementation of CRAM to the ORIGEN module of the SCALE suite is described. In addition to being able to handle time-dependent feed rates, the new solver also adds the capability to perform adjoint calculations. Results obtained with the new CRAM solver and the original depletion solver of ORIGEN are compared to high precision reference calculations, which shows the new solver to be orders of magnitude more accurate. Furthermore, in most cases, the new solver is up to several times faster due to not requiring similar substepping as the original one

A study of different approaches for multi-scale sensitivity analysis of the TALL-3D experiment using thermal-hydraulic computer codes

International Nuclear Information System (INIS)

Geffray, Clotaire; Macian-Juan, Rafael

2014-01-01

In the context of the FP7 European THINS Project, complex thermal-hydraulic phenomena relevant for the Generation IV of nuclear reactors are investigated. KTH (Sweden) built the TALL-3D facility to investigate the transition from forced to natural circulation of the Lead-Bismuth Eutectic (LBE) in a pool connected to a 3-leg primary circuit with two heaters and a heat exchanger. The simulation of such 3D phenomena is a challenging task. GRS (Germany) developed the coupling between the Computational Fluid Dynamics (CFD) code ANSYS CFX and the System Analysis code ATHLET. Such coupled codes combine the advantages of CFD, which allow a fine resolution of 3D phenomena, and of System Analysis codes, which are fast running. TUM (Germany) is responsible for the Uncertainty and Sensitivity Analysis of the coupled ATHLET-CFX model in the THINS Project. The influence of modeling uncertainty on simulation results needs to be assessed to characterize and to improve the model and, eventually, to assess its performance against experimental data. TUM has developed a computational framework capable of propagating model input uncertainty through coupled codes. This framework can also be used to apply different approaches for the assessment of the influence of the uncertain input parameters on the model output (Sensitivity Analysis). The work reported in this paper focuses on three methods for the assessment of the sensitivity of the results to the modeling uncertainty. The first method (Morris) allows for the computation of the Elementary Effects resulting from the input parameters. This method is widely used to perform Screening Analysis. The second method (Spearman's rank correlation) relies on regression-based non-parametric measures. This method is suitable if the relation between the input and the output variables is at least monotonic, with the advantage of a low computational cost. The last method (Sobol') computes so-called total effect indices which account for

R-matrix analysis code (RAC)

International Nuclear Information System (INIS)

Chen Zhenpeng; Qi Huiquan

1990-01-01

A comprehensive R-matrix analysis code has been developed. It is based on the multichannel and multilevel R-matrix theory and runs in VAX computer with FORTRAN-77. With this code many kinds of experimental data for one nuclear system can be fitted simultaneously. The comparisions between code RAC and code EDA of LANL are made. The data show both codes produced the same calculation results when one set of R-matrix parameters was used. The differential cross section of 10 B (n, α) 7 Li for E n = 0.4 MeV and the polarization of 16 O (n,n) 16 O for E n = 2.56 MeV are presented

Development of realistic thermal hydraulic system analysis code

Energy Technology Data Exchange (ETDEWEB)

Lee, Won Jae; Chung, B. D; Kim, K. D. [and others

2002-05-01

The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others.

Development of realistic thermal hydraulic system analysis code

International Nuclear Information System (INIS)

Lee, Won Jae; Chung, B. D; Kim, K. D.

2002-05-01

The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others

Further development and test of the calculation code PROST for the probabilistic and deterministic evaluation of piping and vessels in the frame of leak-before-break confirmation; Weiterentwicklung und Erprobung des Rechenprogramms PROST zur probabilistischen und deterministischen Bewertung von Rohrleitungen und Behaeltern im Rahmen von Leck-vor-Bruch Nachweisen

Energy Technology Data Exchange (ETDEWEB)

Heckmann, Klaus; Sievers, Juergen; Arndt, Jens; Blaesius, Christoph

2016-07-15

The structural mechanics computer code PROST of GRS was extended in the project RS1516 for more applications and improved usability. The fracture mechanical features were extended by supplementary structure types, additional crack geometries, refined failure criteria, and temperature-dependent parameters. The consideration of in-service inspections was revised, while upgraded fatigue analyses enhance the quality of the computed results. The modeling of leakage rates was ameliorated by coupling the WinLeck computer code of GRS for leak rate computation, leading to better predictions of these features. Additional probabilistic techniques were implemented in PROST, which allow a more accurate quantification of risks. PROST's validation was done within in the project by two international benchmark studies. Analytical techniques for fracture mechanical parameters and in-service inspections in probabilistic crack-growth analyses were in the focus of the comparative studies. In addition, diverse test cases are analyzed, allowing the comparison with published studies. The available methodology for leak rate analysis was used for the planning of upcoming measurements. The PROST code has a graphical user interface, is accompanied by a detailed documentation and is made available for external users. Thus, it is a qualified tool for fracture mechanical studies, taking into account nuclear safety standards.

Parallel processing of structural integrity analysis codes

International Nuclear Information System (INIS)

Swami Prasad, P.; Dutta, B.K.; Kushwaha, H.S.

1996-01-01

Structural integrity analysis forms an important role in assessing and demonstrating the safety of nuclear reactor components. This analysis is performed using analytical tools such as Finite Element Method (FEM) with the help of digital computers. The complexity of the problems involved in nuclear engineering demands high speed computation facilities to obtain solutions in reasonable amount of time. Parallel processing systems such as ANUPAM provide an efficient platform for realising the high speed computation. The development and implementation of software on parallel processing systems is an interesting and challenging task. The data and algorithm structure of the codes plays an important role in exploiting the parallel processing system capabilities. Structural analysis codes based on FEM can be divided into two categories with respect to their implementation on parallel processing systems. The first category codes such as those used for harmonic analysis, mechanistic fuel performance codes need not require the parallelisation of individual modules of the codes. The second category of codes such as conventional FEM codes require parallelisation of individual modules. In this category, parallelisation of equation solution module poses major difficulties. Different solution schemes such as domain decomposition method (DDM), parallel active column solver and substructuring method are currently used on parallel processing systems. Two codes, FAIR and TABS belonging to each of these categories have been implemented on ANUPAM. The implementation details of these codes and the performance of different equation solvers are highlighted. (author). 5 refs., 12 figs., 1 tab

Safety analysis of spent fuel packaging

International Nuclear Information System (INIS)

Akamatsu, Hiroshi; Taniuchi, Hiroaki; Tai, Hideto

1987-01-01

Many types of spent fuel packagings have been manufactured and been used for transport of spent fuels discharged from nuclear power plant. These spent fuel packagings need to be assesed thoroughly about safety transportation because spent fuels loaded into the packaging have high radioactivity and generation of heat. This paper explains the outline of safety analysis of a packaging, Safety analysis is performed for structural, thermal, containment, shielding and criticality factors, and MARC-CDC, TRUMP, ORIGEN, QAD, ANISN, KENO, etc computer codes are used for such analysis. (author)

OPR1000 RCP Flow Coastdown Analysis using SPACE Code

Energy Technology Data Exchange (ETDEWEB)

Lee, Dong-Hyuk; Kim, Seyun [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

The Korean nuclear industry developed a thermal-hydraulic analysis code for the safety analysis of PWRs, named SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). Current loss of flow transient analysis of OPR1000 uses COAST code to calculate transient RCS(Reactor Coolant System) flow. The COAST code calculates RCS loop flow using pump performance curves and RCP(Reactor Coolant Pump) inertia. In this paper, SPACE code is used to reproduce RCS flowrates calculated by COAST code. The loss of flow transient is transient initiated by reduction of forced reactor coolant circulation. Typical loss of flow transients are complete loss of flow(CLOF) and locked rotor(LR). OPR1000 RCP flow coastdown analysis was performed using SPACE using simplified nodalization. Complete loss of flow(4 RCP trip) was analyzed. The results show good agreement with those from COAST code, which is CE code for calculating RCS flow during loss of flow transients. Through this study, we confirmed that SPACE code can be used instead of COAST code for RCP flow coastdown analysis.

ALFITeX. A new code for the deconvolution of complex alpha-particle spectra

International Nuclear Information System (INIS)

Caro Marroyo, B.; Martin Sanchez, A.; Jurado Vargas, M.

2013-01-01

A new code for the deconvolution of complex alpha-particle spectra has been developed. The ALFITeX code is written in Visual Basic for Microsoft Office Excel 2010 spreadsheets, incorporating several features aimed at making it a fast, robust and useful tool with a user-friendly interface. The deconvolution procedure is based on the Levenberg-Marquardt algorithm, with the curve fitting the experimental data being the mathematical function formed by the convolution of a Gaussian with two left-handed exponentials in the low-energy-tail region. The code also includes the capability of fitting a possible constant background contribution. The application of the singular value decomposition method for matrix inversion permits the fit of any kind of alpha-particle spectra, even those presenting singularities or an ill-conditioned curvature matrix. ALFITeX has been checked with its application to the deconvolution and the calculation of the alpha-particle emission probabilities of 239 Pu, 241 Am and 235 U. (author)

Z₂-double cyclic codes

OpenAIRE

Borges, J.

2014-01-01

A binary linear code C is a Z2-double cyclic code if the set of coordinates can be partitioned into two subsets such that any cyclic shift of the coordinates of both subsets leaves invariant the code. These codes can be identified as submodules of the Z2[x]-module Z2[x]/(x^r − 1) × Z2[x]/(x^s − 1). We determine the structure of Z2-double cyclic codes giving the generator polynomials of these codes. The related polynomial representation of Z2-double cyclic codes and its duals, and the relation...

Use of computer codes for system reliability analysis

International Nuclear Information System (INIS)

Sabek, M.; Gaafar, M.; Poucet, A.

1988-01-01

This paper gives a collective summary of the studies performed at the JRC, ISPRA on the use of computer codes for complex systems analysis. The computer codes dealt with are: CAFTS-SALP software package, FRANTIC, FTAP, computer code package RALLY, and BOUNDS codes. Two reference study cases were executed by each code. The results obtained logic/probabilistic analysis as well as computation time are compared

An X-ray imager based on silicon microstrip detector and coded mask

International Nuclear Information System (INIS)

Del Monte, E.; Costa, E.; Di Persio, G.; Donnarumma, I.; Evangelista, Y.; Feroci, M.; Frutti, M.; Lapshov, I.; Lazzarotto, F.; Mastropietro, M.; Morelli, E.; Pacciani, L.; Porrovecchio, G.; Rapisarda, M.; Rubini, A.; Soffitta, P.; Tavani, M.; Argan, A.

2007-01-01

SuperAGILE is the X-ray monitor of AGILE, a satellite mission for gamma-ray astronomy, and it is the first X-ray imaging instrument based on the technology of the silicon microstrip detectors combined with a coded aperture imaging technique. The SuperAGILE detection plane is composed of four 1-D silicon microstrip detector modules, mechanically coupled to tungsten coded mask units. The detector strips are separately and individually connected to the input analogue channels of the front-end electronics, composed of low-noise and low-power consumption VLSI ASIC chips. SuperAGILE can produce 1-D images with 6 arcmin angular resolution and ∼2-3 arcmin localisation capability, for intense sources, in a field of view composed of two orthogonal areas of 107 deg. x 68 deg. The time resolution is 2 μs, the overall dead time is ∼5 μs and the electronic noise is ∼7.5 keV full-width at half-maximum. The resulting instrument is very compact (40x40x14 cm 3 ), light (10 kg) and has low power consumption (12 W). AGILE is a mission of the Agenzia Spaziale Italiana and its launch is planned in 2007 in a low equatorial Earth orbit. In this contribution we present SuperAGILE and discuss its performance and scientific objectives

No time for dead time: timing analysis of bright black hole binaries with NuSTAR

DEFF Research Database (Denmark)

Bachetti, Matteo; Harrison, Fiona A.; Cook, Rick

2015-01-01

Timing of high-count-rate sources with the NuSTAR Small Explorer Mission requires specialized analysis techniques. NuSTAR was primarily designed for spectroscopic observations of sources with relatively low count rates rather than for timing analysis of bright objects. The instrumental dead time ...... techniques. We apply this technique to NuSTAR observations of the black hole binaries GX 339-4, Cyg X-1, and GRS 1915+105....

ORIGEN2 calculations supporting TRIGA irradiated fuel data package

Energy Technology Data Exchange (ETDEWEB)

Schmittroth, F.A.

1996-09-20

ORIGEN2 calculations were performed for TRIGA spent fuel elements from the Hanford Neutron Radiography Facility. The calculations support storage and disposal and results include mass, activity,and decay heat. Comparisons with underwater dose-rate measurements were used to confirm and adjust the calculations.

Validation of the RALOC-mod.4 thermal-hydraulics code on evaporation transients in the Phebus containment

International Nuclear Information System (INIS)

Spitz, P.B.; Lemoine, F.; Tirini, S.

1997-01-01

IPSN (Nuclear Protection and Safety Institute) and GRS (Gesellschaft fur Anlagen und Reaktorsicherheit Schwertnergasse 1) are developing the ESCADRE-ASTEC systems of codes devoted to the prediction of the behaviour of water-cooled reactors during a severe accident. The RALOC-mod 4 code belongs to this system and is specifically devoted to containment thermal-hydraulics studies. IPSN has designed a Thermal Hydraulic Containment Test Program in support to the Phebus Fission Product Test Program/2/. Evaporation tests have been recently performed in the Phebus containment test facility. The objective of this work is to assess against these tests the capability of the RALOC -mod 4 code to capture the phenomena observed in these experiments and more particularly the evaporation heat transfer and wall heat transfers. (DM)

Origen y evolución de la responsabilidad por productos defectuosos.

Directory of Open Access Journals (Sweden)

Línea de Investigación “Responsabilidad por productos defectuosos ”

2006-12-01

Full Text Available El presente escrito, que constituye el segundo documento del proyecto de Responsabilidad por Productos Defectuosos, continua con su estudio adelantado en el seno del Departamento de Derecho Comercial de la Universidad Externado de Colombia y concretamente, precisa cuál es el origen y la evolución de este tipo de responsabilidad especial en el common law, identificando aspectos fundamentales como el concepto de responsabilidad civil en general, su clasificación tradicional, sus antecedentes y el estado actual de la materia; y por último analiza en forma detallada su origen y evolución en la Comunidad Europea, haciendo énfasis en las disposiciones de la Directiva de 1985.

Error-correction coding and decoding bounds, codes, decoders, analysis and applications

CERN Document Server

Tomlinson, Martin; Ambroze, Marcel A; Ahmed, Mohammed; Jibril, Mubarak

2017-01-01

This book discusses both the theory and practical applications of self-correcting data, commonly known as error-correcting codes. The applications included demonstrate the importance of these codes in a wide range of everyday technologies, from smartphones to secure communications and transactions. Written in a readily understandable style, the book presents the authors’ twenty-five years of research organized into five parts: Part I is concerned with the theoretical performance attainable by using error correcting codes to achieve communications efficiency in digital communications systems. Part II explores the construction of error-correcting codes and explains the different families of codes and how they are designed. Techniques are described for producing the very best codes. Part III addresses the analysis of low-density parity-check (LDPC) codes, primarily to calculate their stopping sets and low-weight codeword spectrum which determines the performance of these codes. Part IV deals with decoders desi...

Manometer Behavior Analysis using CATHENA, RELAP and GOTHIC Codes

Energy Technology Data Exchange (ETDEWEB)

Kim, Yang Hoon; Han, Kee Soo; Moon, Bok Ja; Jang, Misuk [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

2014-05-15

In this presentation, simple thermal hydraulic behavior is analyzed using three codes to show the possibility of using alternative codes. We established three models of simple u-tube manometer using three different codes. CATHENA (Canadian Algorithm for Thermal hydraulic Network Analysis), RELAP (Reactor Excursion and Leak Analysis Program), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are used for this analysis. CATHENA and RELAP are widely used codes for the analysis of system behavior of CANDU and PWR. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. In this paper, the internal behavior of u-tube manometer was analyzed using 3 codes, CATHENA, RELAP and GOTHIC. The general transient behavior is similar among 3 codes. However, the behavior simulated using GOTHIC shows some different trend compared with the results from the other 2 codes at the end of the transient. It would be resulted from the use of different physical model in GOTHIC, which is specialized for the multi-phase thermal hydraulic behavior analysis of containment system unlike the other two codes.

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

Energy Technology Data Exchange (ETDEWEB)

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

2003-03-01

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

Performance Tuning of x86 OpenMP Codes with MAQAO

Science.gov (United States)

Barthou, Denis; Charif Rubial, Andres; Jalby, William; Koliai, Souad; Valensi, Cédric

Failing to find the best optimization sequence for a given application code can lead to compiler generated codes with poor performances or inappropriate code. It is necessary to analyze performances from the assembly generated code to improve over the compilation process. This paper presents a tool for the performance analysis of multithreaded codes (OpenMP programs support at the moment). MAQAO relies on static performance evaluation to identify compiler optimizations and assess performance of loops. It exploits static binary rewriting for reading and instrumenting object files or executables. Static binary instrumentation allows the insertion of probes at instruction level. Memory accesses can be captured to help tune the code, but such traces require to be compressed. MAQAO can analyze the results and provide hints for tuning the code. We show on some examples how this can help users improve their OpenMP applications.

Utilization of DRUFAN 01/MOD 02 computer code for the depressurization phase analysis of a postulated loss of coolant accident in Angra 2/3 Nuclear Power Plants

International Nuclear Information System (INIS)

Austregesilo Filho, H.; Figueiredo, P.J.M.

1985-08-01

The DRUFAN 01/Mod 2 developed by Gesellschaft fur Reaktorsicherheit (GRS) mbh to simulate thermohydraulic behavior of the primary circuit of PWR reactors, during the despressurization phase and initial refilling phase of loss of coolant accidents by great ruptures, is presented. The program simulates the system to be analysed by control volumes-concentrated parameters model - and it is based on numerical solution of conservation equations for mass of water, mass of vapor, quantities of motion and energy, and on the control volume homogeneity hypothesis. The possibilities of thermodynamic disequilibrium, determining mass transfer between liquid and vapor phases assuming that one saturated phase, are considered. The process of computer code implantation in the Honeywell Bull 64 DPS 7 system at CNEN, the modifications done into the program and the application to the despressurization phase analysis of a loss of coolant accident at Angra-2 and Angra-3 reactors are considered. (M.C.K.) [pt

Depletion methodology in the 3-D whole core transport code DeCART

Energy Technology Data Exchange (ETDEWEB)

Kim, Kang Seog; Cho, Jin Young; Zee, Sung Quun

2005-02-01

Three dimensional whole-core transport code DeCART has been developed to include a characteristics of the numerical reactor to replace partly the experiment. This code adopts the deterministic method in simulating the neutron behavior with the least assumption and approximation. This neutronic code is also coupled with the thermal hydraulic code CFD and the thermo mechanical code to simulate the combined effects. Depletion module has been implemented in DeCART code to predict the depleted composition in the fuel. The exponential matrix method of ORIGEN-2 has been used for the depletion calculation. The library of including decay constants, yield matrix and others has been used and greatly simplified for the calculation efficiency. This report summarizes the theoretical backgrounds and includes the verification of the depletion module in DeCART by performing the benchmark calculations.

Reactor safety in Eastern Europe. Proceedings

International Nuclear Information System (INIS)

1995-02-01

The papers presented to the GRS colloquium refer to the cooperative activities for reactor accident analysis and modification of the GRS computer codes for their application to reactors of the Russian design types of WWER or RBMK. Another topic is the safety of RBMK reactors in particular, and the current status of investigations and studies addressing the containment of unit 4 of the Chernobyl reactor station. (HP) [de

Computer code for shielding calculations of x-rays rooms

International Nuclear Information System (INIS)

Affonso, R.R.W.; Borges, D. da S.; Lava, D.D.; Moreira, M. de L.; Guimarães, A.C.F.

2015-01-01

The building an effective barrier against ionizing radiation present in radiographic rooms requires consideration of many variables. The methodology used for thickness specification of primary and secondary, barrier of a traditional radiographic room, considers the following factors: Use Factor, Occupational Factor, distance between the source and the wall, Workload, Kerma in the air and distance between the patient and the source. With these data it was possible to develop a computer code, which aims to identify and use variables in functions obtained through graphics regressions provided by NCRP-147 (Structural Shielding Design for Medical X-Ray Imaging Facilities) report, for shielding calculation of room walls, and the walls of the dark room and adjacent areas. With the implemented methodology, it was made a code validation by comparison of results with a study case provided by the report. The obtained values for thickness comprise different materials such as concrete, lead and glass. After validation it was made a case study of an arbitrary radiographic room.The development of the code resulted in a user-friendly tool for planning radiographic rooms to comply with the limits established by CNEN-NN-3:01 published in september/2011. (authors)

INTEGRAL/JEM-X detection of a possible new outburst from GRS 1747-312

DEFF Research Database (Denmark)

Chenevez, Jérôme; Bozzo, E.; Bordas, P.

2013-01-01

During the observations performed during INTEGRAL monitoring of the Galactic Bulge region (see ATel #438) on 2013 March 11, we detected enhanced emission from the direction of the globular cluster Terzan 6 using the INTEGRAL/JEM-X instrument. This likely indicates that the bursting low-mass X-ray...

Coincidence: Fortran code for calculation of (e, e'x) differential cross-sections, nuclear structure functions and polarization asymmetry in self-consistent random phase approximation with Skyrme interaction

Energy Technology Data Exchange (ETDEWEB)

Cavinato, M.; Marangoni, M.; Saruis, A.M.

1990-10-01

This report describes the COINCIDENCE code written for the IBM 3090/300E computer in Fortran 77 language. The output data of this code are the (e, e'x) threefold differential cross-sections, the nuclear structure functions, the polarization asymmetry and the angular correlation coefficients. In the real photon limit, the output data are the angular distributions for plane polarized incident photons. The code reads from tape the transition matrix elements previously calculated, by in continuum self-consistent RPA (random phase approximation) theory with Skyrme interactions. This code has been used to perform a numerical analysis of coincidence (e, e'x) reactions with polarized electrons on the /sup 16/O nucleous.

Static Code Analysis with Gitlab-CI

CERN Document Server

Datko, Szymon Tomasz

2016-01-01

Static Code Analysis is a simple but efficient way to ensure that application’s source code is free from known flaws and security vulnerabilities. Although such analysis tools are often coming with more advanced code editors, there are a lot of people who prefer less complicated environments. The easiest solution would involve education – where to get and how to use the aforementioned tools. However, counting on the manual usage of such tools still does not guarantee their actual usage. On the other hand, reducing the required effort, according to the idea “setup once, use anytime without sweat” seems like a more promising approach. In this paper, the approach to automate code scanning, within the existing CERN’s Gitlab installation, is described. For realization of that project, the Gitlab-CI service (the “CI” stands for "Continuous Integration"), with Docker assistance, was employed to provide a variety of static code analysers for different programming languages. This document covers the gene...

Seismic Analysis Code (SAC): Development, porting, and maintenance within a legacy code base

Science.gov (United States)

Savage, B.; Snoke, J. A.

2017-12-01

The Seismic Analysis Code (SAC) is the result of toil of many developers over almost a 40-year history. Initially a Fortran-based code, it has undergone major transitions in underlying bit size from 16 to 32, in the 1980s, and 32 to 64 in 2009; as well as a change in language from Fortran to C in the late 1990s. Maintenance of SAC, the program and its associated libraries, have tracked changes in hardware and operating systems including the advent of Linux in the early 1990, the emergence and demise of Sun/Solaris, variants of OSX processors (PowerPC and x86), and Windows (Cygwin). Traces of these systems are still visible in source code and associated comments. A major concern while improving and maintaining a routinely used, legacy code is a fear of introducing bugs or inadvertently removing favorite features of long-time users. Prior to 2004, SAC was maintained and distributed by LLNL (Lawrence Livermore National Lab). In that year, the license was transferred from LLNL to IRIS (Incorporated Research Institutions for Seismology), but the license is not open source. However, there have been thousands of downloads a year of the package, either source code or binaries for specific system. Starting in 2004, the co-authors have maintained the SAC package for IRIS. In our updates, we fixed bugs, incorporated newly introduced seismic analysis procedures (such as EVALRESP), added new, accessible features (plotting and parsing), and improved the documentation (now in HTML and PDF formats). Moreover, we have added modern software engineering practices to the development of SAC including use of recent source control systems, high-level tests, and scripted, virtualized environments for rapid testing and building. Finally, a "sac-help" listserv (administered by IRIS) was setup for SAC-related issues and is the primary avenue for users seeking advice and reporting bugs. Attempts are always made to respond to issues and bugs in a timely fashion. For the past thirty-plus years

Results of neutronic benchmark analysis for a high temperature reactor of the GT-MHR type - HTR2008-58107

International Nuclear Information System (INIS)

Boyarinov, V. F.; Bryzgalov, V. I.; Davidenko, V. D.; Fomichenko, P. A.; Glushkov, E. S.; Gomin, E. A.; Gurevich, M. I.; Kodochigov, N. G.; Marova, E. V.; Mitenkova, E. F.; Novikov, N. V.; Osipov, S. L.; Sukharev, Y. P.; Tsibulsky, V. F.; Yudkevich, M. S.

2008-01-01

The paper presents a description of benchmark cases, achieved results, analysis of possible reasons of differences of calculation results obtained by various neutronic codes. The comparative analysis is presented showing the benchmark-results obtained with reference and design codes by Russian specialists (WIMS-D, JAR-HTGR, UNK, MCU, MCNP5-MONTEBURNS1.0-ORIGEN2.0), by French specialists (AP0LL02, TRIP0LI4 codes), and by Korean specialists (HELIOS, MASTER, MCNP5 codes). The analysis of possible reasons for deviations was carried out, which was aimed at the decrease of uncertainties in calculated characteristics. This additional investigation was conducted with the use of 2D models of a fuel assembly cell and a reactor plane section. (authors)

Coding Partitions

Directory of Open Access Journals (Sweden)

Fabio Burderi

2007-05-01

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

Optimization and Validation of the Developed Uranium Isotopic Analysis Code

Energy Technology Data Exchange (ETDEWEB)

Kim, J. H.; Kang, M. Y.; Kim, Jinhyeong; Choi, H. D. [Seoul National Univ., Seoul (Korea, Republic of)

2014-10-15

γ-ray spectroscopy is a representative non-destructive assay for nuclear material, and less time-consuming and less expensive than the destructive analysis method. The destructive technique is more precise than NDA technique, however, there is some correction algorithm which can improve the performance of γ-spectroscopy. For this reason, an analysis code for uranium isotopic analysis is developed by Applied Nuclear Physics Group in Seoul National University. Overlapped γ- and x-ray peaks in the 89-101 keV X{sub α}-region are fitted with Gaussian and Lorentzian distribution peak functions, tail and background functions. In this study, optimizations for the full-energy peak efficiency calibration and fitting parameters of peak tail and background are performed, and validated with 24 hour acquisition of CRM uranium samples. The optimization of peak tail and background parameters are performed with the validation by using CRM uranium samples. The analysis performance is improved in HEU samples, but more optimization of fitting parameters is required in LEU sample analysis. In the future, the optimization research about the fitting parameters with various type of uranium samples will be performed. {sup 234}U isotopic analysis algorithms and correction algorithms (coincidence effect, self-attenuation effect) will be developed.

Module type plant system dynamics analysis code (MSG-COPD). Code manual

International Nuclear Information System (INIS)

Sakai, Takaaki

2002-11-01

MSG-COPD is a module type plant system dynamics analysis code which involves a multi-dimensional thermal-hydraulics calculation module to analyze pool type of fast breeder reactors. Explanations of each module and the methods for the input data are described in this code manual. (author)

76 FR 64931 - Building Energy Codes Cost Analysis

Science.gov (United States)

2011-10-19

...-0046] Building Energy Codes Cost Analysis AGENCY: Office of Energy Efficiency and Renewable Energy... reopening of the time period for submitting comments on the request for information on Building Energy Codes... the request for information on Building Energy Code Cost Analysis and provide docket number EERE-2011...

Parallelization for X-ray crystal structural analysis program

Energy Technology Data Exchange (ETDEWEB)

Watanabe, Hiroshi [Japan Atomic Energy Research Inst., Tokyo (Japan); Minami, Masayuki; Yamamoto, Akiji

1997-10-01

In this report we study vectorization and parallelization for X-ray crystal structural analysis program. The target machine is NEC SX-4 which is a distributed/shared memory type vector parallel supercomputer. X-ray crystal structural analysis is surveyed, and a new multi-dimensional discrete Fourier transform method is proposed. The new method is designed to have a very long vector length, so that it enables to obtain the 12.0 times higher performance result that the original code. Besides the above-mentioned vectorization, the parallelization by micro-task functions on SX-4 reaches 13.7 times acceleration in the part of multi-dimensional discrete Fourier transform with 14 CPUs, and 3.0 times acceleration in the whole program. Totally 35.9 times acceleration to the original 1CPU scalar version is achieved with vectorization and parallelization on SX-4. (author)

DynamiX, numerical tool for design of next-generation x-ray telescopes.

Science.gov (United States)

Chauvin, Maxime; Roques, Jean-Pierre

2010-07-20

We present a new code aimed at the simulation of grazing-incidence x-ray telescopes subject to deformations and demonstrate its ability with two test cases: the Simbol-X and the International X-ray Observatory (IXO) missions. The code, based on Monte Carlo ray tracing, computes the full photon trajectories up to the detector plane, accounting for the x-ray interactions and for the telescope motion and deformation. The simulation produces images and spectra for any telescope configuration using Wolter I mirrors and semiconductor detectors. This numerical tool allows us to study the telescope performance in terms of angular resolution, effective area, and detector efficiency, accounting for the telescope behavior. We have implemented an image reconstruction method based on the measurement of the detector drifts by an optical sensor metrology. Using an accurate metrology, this method allows us to recover the loss of angular resolution induced by the telescope instability. In the framework of the Simbol-X mission, this code was used to study the impacts of the parameters on the telescope performance. In this paper we present detailed performance analysis of Simbol-X, taking into account the satellite motions and the image reconstruction. To illustrate the versatility of the code, we present an additional performance analysis with a particular configuration of IXO.

DynamiX, numerical tool for design of next-generation x-ray telescopes

International Nuclear Information System (INIS)

Chauvin, Maxime; Roques, Jean-Pierre

2010-01-01

We present a new code aimed at the simulation of grazing-incidence x-ray telescopes subject to deformations and demonstrate its ability with two test cases: the Simbol-X and the International X-ray Observatory (IXO) missions. The code, based on Monte Carlo ray tracing, computes the full photon trajectories up to the detector plane, accounting for the x-ray interactions and for the telescope motion and deformation. The simulation produces images and spectra for any telescope configuration using Wolter I mirrors and semiconductor detectors. This numerical tool allows us to study the telescope performance in terms of angular resolution, effective area, and detector efficiency, accounting for the telescope behavior. We have implemented an image reconstruction method based on the measurement of the detector drifts by an optical sensor metrology. Using an accurate metrology, this method allows us to recover the loss of angular resolution induced by the telescope instability. In the framework of the Simbol-X mission, this code was used to study the impacts of the parameters on the telescope performance. In this paper we present detailed performance analysis of Simbol-X, taking into account the satellite motions and the image reconstruction. To illustrate the versatility of the code, we present an additional performance analysis with a particular configuration of IXO.

DynamiX, numerical tool for design of next-generation x-ray telescopes

Energy Technology Data Exchange (ETDEWEB)

Chauvin, Maxime; Roques, Jean-Pierre

2010-07-20

We present a new code aimed at the simulation of grazing-incidence x-ray telescopes subject to deformations and demonstrate its ability with two test cases: the Simbol-X and the International X-ray Observatory (IXO) missions. The code, based on Monte Carlo ray tracing, computes the full photon trajectories up to the detector plane, accounting for the x-ray interactions and for the telescope motion and deformation. The simulation produces images and spectra for any telescope configuration using Wolter I mirrors and semiconductor detectors. This numerical tool allows us to study the telescope performance in terms of angular resolution, effective area, and detector efficiency, accounting for the telescope behavior. We have implemented an image reconstruction method based on the measurement of the detector drifts by an optical sensor metrology. Using an accurate metrology, this method allows us to recover the loss of angular resolution induced by the telescope instability. In the framework of the Simbol-X mission, this code was used to study the impacts of the parameters on the telescope performance. In this paper we present detailed performance analysis of Simbol-X, taking into account the satellite motions and the image reconstruction. To illustrate the versatility of the code, we present an additional performance analysis with a particular configuration of IXO.

SARNET, a success story. Survey of major achievements on severe accidents and of knowledge capitalization within the ASTEC code

International Nuclear Information System (INIS)

Albiol, T.; Van Dorsselaere, J.P.; Reinke, N.

2013-01-01

51 organizations from Europe and Canada cooperated within SARNET (Severe Accident Research Network of Excellence) joining their capacities of research in order to resolve the most important pending issues for enhancing, in regard to Severe Accidents (SA), the safety of existing and future Nuclear Power Plants (NPPs). SARNET defines common research programmes and develops common computer codes and methodologies for safety assessment. The ASTEC integral code, jointly developed by IRSN (France) and GRS (Germany) for Light Water Reactor (LWR) source term SA evaluation, Probabilistic Safety Assessment (PSA) level-2 studies and SA management evaluation, is the main integrating component of SARNET. The scientific knowledge generated in the Corium, Source Term and Containment Topics has been integrated into the code through improved or new physical models. ASTEC constitutes now the reference European SA integral code. During the 4 and half years of SARNET, 30 partners have assessed the successive versions of the ASTEC V1 code through validation. More than 60 scientists have been trained on the code use. Validation tasks on about 65 experiments were performed to cover all physical phenomena occurring in a severe accident: circuit thermalhydraulic, core degradation, fission products (FP) release and transport, Molten-Corium-Concrete-Interaction (MCCI), and in the containment, thermalhydraulic, aerosol and iodine as well as hydrogen behaviour. The overall status of validation can be considered as good, with results often close to results of mechanistic codes. Some reach the limits of present knowledge, for instance on MCCI, and, like in most codes, an adequate model for reflooding of a degraded core is still missing. IRSN and GRS are currently preparing the new series of ASTEC V2 versions that will account for most of the needs of evolution expressed by the SARNET partners. The first version V2.0, planned for March 09, will be applicable to EPR and will include the ICARE2

Two-dimensional disruption thermal analysis code DREAM

International Nuclear Information System (INIS)

Yamazaki, Seiichiro; Kobayashi, Takeshi; Seki, Masahiro.

1988-08-01

When a plasma disruption takes place in a tokamak type fusion reactor, plasma facing components such as first wall and divertor/limiter are subjected to an intense heat load with very high heat flux and short duration. At the surface of the wall, temperature rapidly rises, and melting and evaporation occurs, it causes reduction of wall thickness and crack initiation/propagation. As lifetime of the components is significantly affected by them, the transient analysis in consideration of phase changes (melting/evaporation) and radiation heat loss is required in the design of these components. This paper describes the computer code DREAM developed to perform the two-dimensional transient thermal analysis that takes phase changes and radiation into account. The input and output of the code and a sample analysis on a disruption simulation experiment are also reported. The user's input manual is added as an appendix. The profiles and time variations of temperature, and melting and evaporated thicknesses of the material subjected to intense heat load can be obtained, using this computer code. This code also gives the temperature data for elastoplastic analysis with FEM structural analysis codes (ADINA, MARC, etc.) to evaluate the thermal stress and crack propagation behavior within the wall materials. (author)

Subchannel analysis code development for CANDU fuel channel

International Nuclear Information System (INIS)

Park, J. H.; Suk, H. C.; Jun, J. S.; Oh, D. J.; Hwang, D. H.; Yoo, Y. J.

1998-07-01

Since there are several subchannel codes such as COBRA and TORC codes for a PWR fuel channel but not for a CANDU fuel channel in our country, the subchannel analysis code for a CANDU fuel channel was developed for the prediction of flow conditions on the subchannels, for the accurate assessment of the thermal margin, the effect of appendages, and radial/axial power profile of fuel bundles on flow conditions and CHF and so on. In order to develop the subchannel analysis code for a CANDU fuel channel, subchannel analysis methodology and its applicability/pertinence for a fuel channel were reviewed from the CANDU fuel channel point of view. Several thermalhydraulic and numerical models for the subchannel analysis on a CANDU fuel channel were developed. The experimental data of the CANDU fuel channel were collected, analyzed and used for validation of a subchannel analysis code developed in this work. (author). 11 refs., 3 tabs., 50 figs

Experimental validation for combustion analysis of GOTHIC code in 2-dimensional combustion chamber

International Nuclear Information System (INIS)

Lee, J. W.; Yang, S. Y.; Park, K. C.; Jung, S. H.

2002-01-01

In this study, the prediction capability of GOTHIC code for hydrogen combustion phenomena was validated with the results of two-dimensional premixed hydrogen combustion experiment executed by Seoul National University. The experimental chamber has about 24 liter free volume (1x0.024x1 m 3 ) and 2-dimensional rectangular shape. The test were preformed with 10% hydrogen/air gas mixture and conducted with combination of two igniter positions (top center, top corner) and two boundary conditions (bottom full open, bottom right half open). Using the lumped parameter and mechanistic combustion model in GOTHIC code, the SNU experiments were simulated under the same conditions. The GOTHIC code prediction of the hydrogen combustion phenomena did not compare well with the experimental results. In case of lumped parameter simulation, the combustion time was predicted appropriately. But any other local information related combustion phenomena could not be obtained. In case of mechanistic combustion analysis, the physical combustion phenomena of gas mixture were not matched experimental ones. In boundary open cases, the GOTHIC predicted very long combustion time and the flame front propagation could not simulate appropriately. Though GOTHIC showed flame propagation phenomenon in adiabatic calculation, the induction time of combustion was still very long compare with experimental results. Also, it was found that the combustion model of GOTHIC code had some weak points in low concentration of hydrogen combustion simulation

Radiation Analysis for Skeleton of Spent Nuclear Fuel Assembly

International Nuclear Information System (INIS)

Park, Chang Je; Na, Sang Ho; Yang, Jae Hwan; Kang, Kweon Ho

2010-11-01

ORIGEN-S code was used in order to analyze the radioactive characteristics of skeleton of the spent nuclear fuel assembly. From the results, radioactivity, decay heat for various compositions in skeleton were obtained with a variation of cooling period and axial distribution of radioactivity was calculated, too. These data will be utilized later to process and dispose the skeleton of spent nuclear fuel assembly

Investigation on the anisotropic mechanical behaviour of the Callovo-Oxfordian clay rock within the framework of ANDRA/GRS cooperation programme. Final report

Energy Technology Data Exchange (ETDEWEB)

Zhang, Chun-Liang [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany); Armand, Gilles; Conil, Nathalie [French Agence Nationale Pour la Gestion de Dechets Radioactifs (ANDRA), Chatenay-Malabry (France)

2015-01-15

An underground repository for disposal of radioactive waste is planned to be constructed in the sedimentary Callovo-Oxfordian argillaceous formation (COX) in France /AND 05/. The clay rock exhibits inherent anisotropy with bedding structure, which leads to directional dependences of the rock properties (e.g. mineralogical, physical, mechanical, hydraulic, thermal, etc.) with respect to the bedding planes. For the design of the repository and the assessment of its safety during the operation and post-closure phases it is necessary to characterise and predict the anisotropic properties and processes in the host rock, particularly in the excavation damaged zone (EDZ) near the openings. Within the framework of the bilateral cooperation agreement between the French National Radioactive Waste Management Agency (ANDRA) and the German Federal Ministry of Economics and Technology (BMWi), concerning the research activities in the Meuse/Haute-Marne Underground Research Laboratory (MHM-URL), a joint research programme was initiated by ANDRA and GRS in 2013 to investigate mechanical anisotropy of the COX clay rock for the purpose of precise characterization, better understanding and reliable prediction of the development of EDZ around the repository. This programme was funded by ANDRA under contract number 059844 and performed by GRS during the time period of November 2013 to December 2014. GRS gratefully acknowledges the financial support from and the fruitful cooperation with ANDRA.

Immobilization of alpha-amylase produced by Bacillus circulans GRS 313

Directory of Open Access Journals (Sweden)

Gargi Dey

2003-03-01

Full Text Available A maltooligosaccharide-forming amylase from B circulans GRS 313 was immobilized by entrapment in calcium alginate beads. The immobilized activity was affected by the size of the bead and bead size of 2mm was found to be most effective for hydrolysis. Kinetics constants, Km and Vmax were estimated and were found to be affected by the bead size. The catalytic activity of the enzyme was studied in presence of various starchy residues and metal ions. HgCl2, CuSO4 and FeCl3 caused inhibition of the enzyme. The reaction conditions, pH and temperature, was optimized using response surface methodology. At the optimum pH and temperature of 4.9 and 57ºC, the apparent activity was 25.6U/g of beads, resulting in almost 2-fold increase in activity. The immobilized enzyme showed a high operational stability by retaining almost 85% of the initial activity after seventh use.Um maltooligossacarideo obtido a partir de amilase produzida por B. circulans GRS 313 foi imobilizada em alginato de sódio. A atividade enzimática foi afetada pelo tamanho da partícula. Partículas com 2mm foram as mais efetivas na hidrólise. Constantes cinética Km e Vmax foram estimadas e afetadas pelo tamanho das partículas. A atividade catalítica da enzima foi estuda na presença de diferentes tipos de amido e íons metálicos. HgCl2, CuSO4 e FeCl3 provocaram inibição na enzima. As condições de reação (temperatura e pH foram otimizadas utilizando a metodologia da superfície de resposta. Em pH ótimo de 4,9 e temperatura de 57 ºC, a atividade aparente foi de 25.6 U/g de partículas, resultando num acréscimo de mais de 2 vezes na atividade da enzima. A imobilização da enzima mostrou uma alta estabilidade operacional pela retenção de 85% de sua atividade inicial após sete ciclos de utilização.

GRS' research on clay rock in the Mont Terri underground laboratory

Energy Technology Data Exchange (ETDEWEB)

Wieczorek, Klaus; Czaikowski, Oliver [Gesellschaft fuer Anlagen- und Reaktorsicherheit gGmbH, Braunschweig (Germany)

2016-07-15

For constructing a nuclear waste repository and for ensuring the safety requirements are met over very long time periods, thorough knowledge about the safety-relevant processes occurring in the coupled system of waste containers, engineered barriers, and the host rock is indispensable. For respectively targeted research work, the Mont Terri rock laboratory is a unique facility where repository research is performed in a clay rock environment. It is run by 16 international partners, and a great variety of questions are investigated. Some of the work which GRS as one of the Mont Terri partners is involved in is presented in this article. The focus is on thermal, hydraulic and mechanical behaviour of host rock and/or engineered barriers.

Experimental transport analysis code system in JT-60

International Nuclear Information System (INIS)

Hirayama, Toshio; Shimizu, Katsuhiro; Tani, Keiji; Shirai, Hiroshi; Kikuchi, Mitsuru

1988-03-01

Transport analysis codes have been developed in order to study confinement properties related to particle and energy balance in ohmically and neutral beam heated plasmas of JT-60. The analysis procedure is divided into three steps as follows: 1) LOOK ; The shape of the plasma boundary is identified with a fast boundary identification code of FBI by using magnetic data, and flux surfaces are calculated with a MHD equilibrium code of SELENE. The diagnostic data are mapped to flux surfaces for neutral beam heating calculation and/or for radial transport analysis. 2) OFMC ; On the basis of transformed data, an orbit following Monte Carlo code of OFMC calculates both profiles of power deposition and particle source of neutral beam injected into a plasma. 3) SCOOP ; In the last stage, a one dimensional transport code of SCOOP solves particle and energy balance for electron and ion, in order to evaluate transport coefficients as well as global parameters such as energy confinement time and the stored energy. The analysis results are provided to a data bank of DARTS that is used to find an overview of important consideration on confinement with a regression analysis code of RAC. (author)

A simple in-surge pressure analysis using the SPACE code

International Nuclear Information System (INIS)

Youn, Bum Soo; Kim, Yo Han; Lee, Dong Hyuk; Yang, Chang Keun; Kim, Se Yun; Ha, Sang Jun

2010-01-01

Currently, nuclear safety analysis codes used in Korea are developed by all the overseas. These codes are paying huge fee and permission must be obtained for use in the country. In addition, orders for nuclear power plants must ensure the safety analysis code for independent domestic technology. Therefore, Korea Electric Power Research Institute(KEPRI) is developing the domestic nuclear power safety analysis, SPACE(Safety and Performance Analysis Code for nuclear power plants). To determine the computational power of pressurizer model in development SPACE code, it was compared with existing commercial nuclear power safety analysis code, RETRAN

Business Ethics: International Analysis of Codes of Ethics and Conduct

Directory of Open Access Journals (Sweden)

Josmar Andrade

2017-03-01

Full Text Available Codes of ethics and code of conduct formalize an ideal of expected behavior patterns to managers and employees of organizations, providing standards and orientation that states companies interactions with the community, through products /services, sales force, marketing communications, investments, and relationships with other stakeholders, influencing company reputation and overall Marketing performance. The objective of this study is to analyze the differences in codes of ethics of the largest companies based in Brazil and in Portugal, given their cultural and linguistic similarities. Findings show that the use of codes of ethics are more common in Brazil than in Portugal and that codes of ethics are substantially more extensive and cover a larger number of categories in Brazilian companies, reflecting the organizations’ mission and perception of stakeholders concerns and priorities. We conclude that ethical issues severely impact company reputation and, in a comprehensive sense, overall Marketing performance. Marketing professionals should be systematically aware of how company core values are transmitted to different audiences, including the use of code of ethics to communicate both with internal and external publics. 0 0 1 171 966 CASA DOS ANDRADES 23 14 1123 14.0 96 800x600 Normal 0 false false false EN-US JA X-NONE  

Automated uncertainty analysis methods in the FRAP computer codes

International Nuclear Information System (INIS)

Peck, S.O.

1980-01-01

A user oriented, automated uncertainty analysis capability has been incorporated in the Fuel Rod Analysis Program (FRAP) computer codes. The FRAP codes have been developed for the analysis of Light Water Reactor fuel rod behavior during steady state (FRAPCON) and transient (FRAP-T) conditions as part of the United States Nuclear Regulatory Commission's Water Reactor Safety Research Program. The objective of uncertainty analysis of these codes is to obtain estimates of the uncertainty in computed outputs of the codes is to obtain estimates of the uncertainty in computed outputs of the codes as a function of known uncertainties in input variables. This paper presents the methods used to generate an uncertainty analysis of a large computer code, discusses the assumptions that are made, and shows techniques for testing them. An uncertainty analysis of FRAP-T calculated fuel rod behavior during a hypothetical loss-of-coolant transient is presented as an example and carried through the discussion to illustrate the various concepts

Accelerator-driven transmutation reactor analysis code system (ATRAS)

Energy Technology Data Exchange (ETDEWEB)

Sasa, Toshinobu; Tsujimoto, Kazufumi; Takizuka, Takakazu; Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1999-03-01

JAERI is proceeding a design study of the hybrid type minor actinide transmutation system which mainly consist of an intense proton accelerator and a fast subcritical core. Neutronics and burnup characteristics of the accelerator-driven system is important from a view point of the maintenance of subcriticality and energy balance during the system operation. To determine those characteristics accurately, it is necessary to involve reactions at high-energy region, which are not treated on ordinary reactor analysis codes. The authors developed a code system named ATRAS to analyze the neutronics and burnup characteristics of accelerator-driven subcritical reactor systems. ATRAS has a function of burnup analysis taking account of the effect of spallation neutron source. ATRAS consists of a spallation analysis code, a neutron transport codes and a burnup analysis code. Utility programs for fuel exchange, pre-processing and post-processing are also incorporated. (author)

Research for enhancing reactor safety

International Nuclear Information System (INIS)

1989-05-01

Recent research for enhanced reactor safety covers extensive and numerous experiments and computed modelling activities designed to verify and to improve existing design requirements. The lectures presented at the meeting report GRS research results and the current status of reactor safety research in France. The GRS experts present results concerning expert systems and their perspectives in safety engineering, large-scale experiments and their significance in the development and verification of computer codes for thermohydraulic modelling of safety-related incidents, the advanced system code ATHLET for analysis of thermohydraulic processes of incidents, the analysis simulator which is a tool for fast evaluation of accident management measures, and investigations into event sequences and the required preventive emergency measures within the German Risk Study. (DG) [de

Optimum design of a coded mask X-ray telescope for rocket applications

International Nuclear Information System (INIS)

Gunson, J.; Polychronopulos, B.

1976-01-01

A review of the principles of current X-ray telescopes is made with particular emphasis on two-step imaging techniques involving coding masks. The merits and limitations of the various types of coding masks in use are examined in detail. The limitations are shown to arise from the finite nature of practical masks. By postulating periodicity, 'optimum masks' can be constructed with ideal imaging qualities. The theory for the design of such masks and the practical considerations involved in the design of a rocket-borne X-ray telescope system are discussed in full, with particular attention paid to resolution, field of view and image noise. The main emphasis throughout the paper is on one-dimensional masks but two-dimensional masks are also studied. It is concluded that optimum masks could prove very valuable in astronomical applications and also in other fields such as radiography, where high imaging quality coupled with high sensitivity and low cost are of utmost importance. (author)

Design and performance of coded aperture optical elements for the CESR-TA x-ray beam size monitor

Energy Technology Data Exchange (ETDEWEB)

Alexander, J.P.; Chatterjee, A.; Conolly, C.; Edwards, E.; Ehrlichman, M.P. [Cornell University, Ithaca, NY 14853 (United States); Flanagan, J.W. [High Energy Accelerator Research Organization (KEK), Tsukuba (Japan); Department of Accelerator Science, Graduate University for Advanced Studies (SOKENDAI), Tsukuba (Japan); Fontes, E. [Cornell University, Ithaca, NY 14853 (United States); Heltsley, B.K., E-mail: bkh2@cornell.edu [Cornell University, Ithaca, NY 14853 (United States); Lyndaker, A.; Peterson, D.P.; Rider, N.T.; Rubin, D.L.; Seeley, R.; Shanks, J. [Cornell University, Ithaca, NY 14853 (United States)

2014-12-11

We describe the design and performance of optical elements for an x-ray beam size monitor (xBSM), a device measuring e{sup +} and e{sup −} beam sizes in the CESR-TA storage ring. The device can measure vertical beam sizes of 10–100μm on a turn-by-turn, bunch-by-bunch basis at e{sup ±} beam energies of ∼2–5GeV. x-rays produced by a hard-bend magnet pass through a single- or multiple-slit (coded aperture) optical element onto a detector. The coded aperture slit pattern and thickness of masking material forming that pattern can both be tuned for optimal resolving power. We describe several such optical elements and show how well predictions of simple models track measured performances. - Highlights: • We characterize optical element performance of an e{sup ±} x-ray beam size monitor. • We standardize beam size resolving power measurements to reference conditions. • Standardized resolving power measurements compare favorably to model predictions. • Key model features include simulation of photon-counting statistics and image fitting. • Results validate a coded aperture design optimized for the x-ray spectrum encountered.

Test and validation of CFD codes for the simulation of accident-typical phenomena in the reactor containment; Erprobung und Validierung von CFD-Codes fuer die Simulation von unfalltypischen Phaenomenen im Sicherheitseinschluss

Energy Technology Data Exchange (ETDEWEB)

Schramm, Berthold; Stewering, Joern; Sonnenkalb, Martin

2014-03-15

CFD (Computational Fluid Dynamic) simulation techniques have a growing relevance for the simulation and assessment of accidents in nuclear reactor containments. Some fluid dynamic problems like the calculation of the flow resistances in a complex geometry, turbulence calculations or the calculation of deflagrations could only be solved exactly for very simple cases. These fluid dynamic problems could not be represented by lumped parameter models and must be approximated numerically. Therefore CFD techniques are discussed by a growing international community in conferences like the CFD4NRS-conference. Also the number of articles with a CFD topic is increasing in professional journals like Nuclear Engineering and Design. CFD tools like GASFLOW or GOTHIC are already in use in European nuclear site licensing processes for future nuclear power plants like EPR or AP1000 and the results of these CFD tools are accepted by the authorities. For these reasons it seems to be necessary to build up national competences in the field of CFD techniques and it is important to validate and assess the existing CFD tools. GRS continues the work for the validation and assessment of CFD codes for the simulation of accident scenarios in a nuclear reactor containment within the framework of the BMWi sponsored project RS1500. The focus of this report is on the following topics: - Further validation of condensation models from GRS, FZJ and ANSYS and development of a new condensate model. - Validation of a new turbulence model which was developed by the University of Stuttgart in cooperation with ANSYS. - The formation and dissolution of light gas stratifications are analyzed by large scale experiments. These experiments were simulated by GRS. - The AREVA correlations for hydrogen recombiners (PARs) could be improved by GRS after the analysis of experimental data. Relevant experiments were simulated with this improved recombiner correlation. - Analyses on the simulation of H{sub 2

Quantitative X-ray fluorescence analysis at the ESRF ID18F microprobe

International Nuclear Information System (INIS)

Vekemans, B.; Vincze, L.; Somogyi, A.; Drakopoulos, M.; Kempenaers, L.; Simionovici, A.; Adams, F.

2003-01-01

The new ID18F end-station at the European synchrotron radiation facility (ESRF) in Grenoble (France) is dedicated to sensitive and accurate quantitative micro-X-ray fluorescence (XRF) analysis at the ppm level with accuracy better than 10% for elements with atomic numbers above 18. For accurate quantitative analysis, given a high level of instrumental stability, major steps are the extraction and conversion of experimental X-ray line intensities into elemental concentrations. For this purpose a two-step quantification approach was adopted. In the first step, the collected XRF spectra are deconvoluted on the basis of a non-linear least-squares fitting algorithm (AXIL). The extracted characteristic line intensities are then used as input for a detailed Monte Carlo (MC) simulation code dedicated to XRF spectroscopy taking into account specific experimental conditions (excitation/detection) as well as sample characteristics (absorption and enhancement effects, sample topology, heterogeneity etc.). The iterative use of the MC code gives a 'no-compromise' solution for the quantification problem

Quantitative X-ray fluorescence analysis at the ESRF ID18F microprobe

Energy Technology Data Exchange (ETDEWEB)

Vekemans, B. E-mail: vekemans@uia.ua.ac.be; Vincze, L.; Somogyi, A.; Drakopoulos, M.; Kempenaers, L.; Simionovici, A.; Adams, F

2003-01-01

The new ID18F end-station at the European synchrotron radiation facility (ESRF) in Grenoble (France) is dedicated to sensitive and accurate quantitative micro-X-ray fluorescence (XRF) analysis at the ppm level with accuracy better than 10% for elements with atomic numbers above 18. For accurate quantitative analysis, given a high level of instrumental stability, major steps are the extraction and conversion of experimental X-ray line intensities into elemental concentrations. For this purpose a two-step quantification approach was adopted. In the first step, the collected XRF spectra are deconvoluted on the basis of a non-linear least-squares fitting algorithm (AXIL). The extracted characteristic line intensities are then used as input for a detailed Monte Carlo (MC) simulation code dedicated to XRF spectroscopy taking into account specific experimental conditions (excitation/detection) as well as sample characteristics (absorption and enhancement effects, sample topology, heterogeneity etc.). The iterative use of the MC code gives a 'no-compromise' solution for the quantification problem.

Safety analysis for boiling water reactors

International Nuclear Information System (INIS)

Kersting, E.; Linden, J. von; Mueller-Ecker, D.; Werner, W.

1993-07-01

This report is the translation of GRS-95 'Sicherheitsanalyse fuer Siedewasserreaktoren - Zusammenfassende Darstellung'. Recent analysis results -concerning the chapters on accident management, fire and earthquake - that were not included in the German text have been added to this translation. In cases of doubt, GRS-102 (main volume) is the factually correct version. (orig.)

Comparative study of Thermal Hydraulic Analysis Codes for Pressurized Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Kim, Yang Hoon; Jang, Mi Suk; Han, Kee Soo [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

2015-05-15

Various codes are used for the thermal hydraulic analysis of nuclear reactors. The use of some codes among these is limited by user and some codes are not even open to general person. Thus, the use of alternative code is considered for some analysis. In this study, simple thermal hydraulic behaviors are analyzed using three codes to show that alternative codes are possible for the analysis of nuclear reactors. We established three models of the simple u-tube manometer using three different codes. RELAP5 (Reactor Excursion and Leak Analysis Program), SPACE (Safety and Performance Analysis CodE for nuclear power Plants), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are selected for this analysis. RELAP5 is widely used codes for the analysis of system behavior of PWRs. SPACE has been developed based on RELAP5 for the analysis of system behavior of PWRs and licensing of the code is in progress. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. The internal behavior of u-tube manometer was analyzed by RELAP5, SPACE and GOTHIC codes. The general transient behavior was similar among 3 codes. However, the stabilized status of the transient period analyzed by REPAP5 was different from the other codes. It would be resulted from the different physical models used in the other codes, which is specialized for the multi-phase thermal hydraulic behavior analysis.

JEM-X: The X-ray monitor on INTEGRAL

DEFF Research Database (Denmark)

Lund, Niels; Budtz-Jørgensen, Carl; Westergaard, Niels Jørgen Stenfeldt

1999-01-01

and identification of gamma ray sources as well as in the analysis and scientific interpretation of the combined X-ray and gamma ray data. JEM-X is a coded aperture X-ray telescope consisting of two identical detectors. Each detector has a sensitive area of 500 cm(2), and views the sky (6.6 deg FOV, FWHM) through...

Improved hydrogen combustion model for multi-compartment analysis

International Nuclear Information System (INIS)

Ogino, Masao; Hashimoto, Takashi

2000-01-01

NUPEC has been improving a hydrogen combustion model in MELCOR code for severe accident analysis. In the proposed combustion model, the flame velocity in a node was predicted using six different flame front shapes of fireball, prism, bubble, spherical jet, plane jet, and parallelepiped. A verification study of the proposed model was carried out using the NUPEC large-scale combustion test results following the previous work in which the GRS/Battelle multi-compartment combustion test results had been used. The selected test cases for the study were the premixed test and the scenario-oriented test which simulated the severe accident sequences of an actual plant. The improved MELCOR code replaced by the proposed model could predict sufficiently both results of the premixed test and the scenario-oriented test of NUPEC large-scale test. The improved MELCOR code was confirmed to simulate the combustion behavior in the multi-compartment containment vessel during a severe accident with acceptable degree of accuracy. Application of the new model to the LWR severe accident analysis will be continued. (author)

Sensitivity and uncertainty analysis for Ignalina NPP confinement in case of loss of coolant accident

International Nuclear Information System (INIS)

Urbonavicius, E.; Babilas, E.; Rimkevicius, S.

2003-01-01

At present the best-estimate approach in the safety analysis of nuclear power plants is widely used around the world. The application of such approach requires to estimate the uncertainty of the calculated results. Various methodologies are applied in order to determine the uncertainty with the required accuracy. One of them is the statistical methodology developed at GRS mbH in Germany and integrated into the SUSA tool, which was applied for the sensitivity and uncertainty analysis of the thermal-hydraulic parameters inside the confinement (Accident Localisation System) of Ignalina NPP with RBMK-1500 reactor in case of Maximum Design Basis Accident (break of 900 mm diameter pipe). Several parameters that could potentially influence the calculated results were selected for the analysis. A set of input data with different initial values of the selected parameters was generated. In order to receive the results with 95 % probability and 95 % accuracy, 100 runs were performed with COCOSYS code developed at GRS mbH. The calculated results were processed with SUSA tool. The performed analysis showed a rather low dispersion of the results and only in the initial period of the accident. Besides, the analysis showed that there is no threat to the building structures of Ignalina NPP confinement in case of the considered accident scenario. (author)

Análisis de las Denominaciones de Origen y Marcas de Garantía de la provincia de Soria

OpenAIRE

García Manrique, Paula

2017-01-01

los objetivos que propusimos a la hora de comenzar esta investigación fueron los siguientes: - Analizar las figuras de marca de garantía y denominación de origen y definir su papel garante de calidad y origen de los productos. - Conocer en profundidad el proceso de obtención de una figura de calidad diferenciada (denominación de origen y marca de garantía) - Saber las razones y motivos por las que las figuras de calidad diferenciada de la provincia de Soria han llegado a adquirir tal...

Development and validation of ALEPH Monte Carlo burn-up code

International Nuclear Information System (INIS)

Stankovskiy, A.; Van den Eynde, G.; Vidmar, T.

2011-01-01

The Monte-Carlo burn-up code ALEPH is being developed in SCK-CEN since 2004. Belonging to the category of shells coupling Monte Carlo transport (MCNP or MCNPX) and 'deterministic' depletion codes (ORIGEN-2.2), ALEPH possess some unique features that distinguish it from other codes. The most important feature is full data consistency between steady-state Monte Carlo and time-dependent depletion calculations. Recent improvements of ALEPH concern full implementation of general-purpose nuclear data libraries (JEFF-3.1.1, ENDF/B-VII, JENDL-3.3). The upgraded version of the code is capable to treat isomeric branching ratios, neutron induced fission product yields, spontaneous fission yields and energy release per fission recorded in ENDF-formatted data files. The alternative algorithm for time evolution of nuclide concentrations is added. A predictor-corrector mechanism and the calculation of nuclear heating are available as well. The validation of the code on REBUS experimental programme results has been performed. The upgraded version of ALEPH has shown better agreement with measured data than other codes, including previous version of ALEPH. (authors)

Interface requirements to couple thermal-hydraulic codes to severe accident codes: ATHLET-CD

Energy Technology Data Exchange (ETDEWEB)

Trambauer, K. [GRS, Garching (Germany)

1997-07-01

The system code ATHLET-CD is being developed by GRS in cooperation with IKE and IPSN. Its field of application comprises the whole spectrum of leaks and large breaks, as well as operational and abnormal transients for LWRs and VVERs. At present the analyses cover the in-vessel thermal-hydraulics, the early phases of core degradation, as well as fission products and aerosol release from the core and their transport in the Reactor Coolant System. The aim of the code development is to extend the simulation of core degradation up to failure of the reactor pressure vessel and to cover all physically reasonable accident sequences for western and eastern LWRs including RMBKs. The ATHLET-CD structure is highly modular in order to include a manifold spectrum of models and to offer an optimum basis for further development. The code consists of four general modules to describe the reactor coolant system thermal-hydraulics, the core degradation, the fission product core release, and fission product and aerosol transport. Each general module consists of some basic modules which correspond to the process to be simulated or to its specific purpose. Besides the code structure based on the physical modelling, the code follows four strictly separated steps during the course of a calculation: (1) input of structure, geometrical data, initial and boundary condition, (2) initialization of derived quantities, (3) steady state calculation or input of restart data, and (4) transient calculation. In this paper, the transient solution method is briefly presented and the coupling methods are discussed. Three aspects have to be considered for the coupling of different modules in one code system. First is the conservation of masses and energy in the different subsystems as there are fluid, structures, and fission products and aerosols. Second is the convergence of the numerical solution and stability of the calculation. The third aspect is related to the code performance, and running time.

Impact testing and analysis for structural code benchmarking

International Nuclear Information System (INIS)

Glass, R.E.

1989-01-01

Sandia National Laboratories, in cooperation with industry and other national laboratories, has been benchmarking computer codes (''Structural Code Benchmarking for the Analysis of Impact Response of Nuclear Material Shipping Cask,'' R.E. Glass, Sandia National Laboratories, 1985; ''Sample Problem Manual for Benchmarking of Cask Analysis Codes,'' R.E. Glass, Sandia National Laboratories, 1988; ''Standard Thermal Problem Set for the Evaluation of Heat Transfer Codes Used in the Assessment of Transportation Packages, R.E. Glass, et al., Sandia National Laboratories, 1988) used to predict the structural, thermal, criticality, and shielding behavior of radioactive materials packages. The first step in the benchmarking of the codes was to develop standard problem sets and to compare the results from several codes and users. This step for structural analysis codes has been completed as described in ''Structural Code Benchmarking for the Analysis of Impact Response of Nuclear Material Shipping Casks,'' R.E. Glass, Sandia National Laboratories, 1985. The problem set is shown in Fig. 1. This problem set exercised the ability of the codes to predict the response to end (axisymmetric) and side (plane strain) impacts with both elastic and elastic/plastic materials. The results from these problems showed that there is good agreement in predicting elastic response. Significant differences occurred in predicting strains for the elastic/plastic models. An example of the variation in predicting plastic behavior is given, which shows the hoop strain as a function of time at the impacting end of Model B. These differences in predicting plastic strains demonstrated a need for benchmark data for a cask-like problem. 6 refs., 5 figs

MÉXICO: CENTRO DE ORIGEN DE LA DOMESTICACIÓN DEL GIRASOL

Directory of Open Access Journals (Sweden)

Robert Bye

2009-01-01

Full Text Available Hasta hace poco tiempo, el origen geográfico del girasol domesticado (Helianthus annuus había sido reportado en el área del sureste de los Estados Unidos de América. El análisis de documentos históricos de México y "semillas" arqueológicas recientemente descubiertas en Tabasco y Morelos, México, indican que los girasoles cultivados fueron importantes durante la época prehispánica y del virreinato en el centro de México. Cabe mencionar que los aquenios prehistóricos más grandes y más antiguos son de México. Aunque hoy en día las plantas silvestres son genéticamente distantes de los cultivares comerciales contemporáneos, las evidencias indican que México es el centro de origen más antiguo.

xLPR - a probabilistic approach to piping integrity analysis

International Nuclear Information System (INIS)

Harrington, C.; Rudland, D.; Fyfitch, S.

2015-01-01

The xLPR Code is a probabilistic fracture mechanics (PFM) computational tool that can be used to quantitatively determine a best-estimate probability of failure with well characterized uncertainties for reactor coolant system components, beginning with the piping systems and including the effects of relevant active degradation mechanisms. The initial application planned for xLPR is somewhat narrowly focused on validating LBB (leak-before-break) compliance in PWSCC-susceptible systems such as coolant systems of PWRs. The xLPR code incorporates a set of deterministic models that represent the full range of physical phenomena necessary to evaluate both fatigue and PWSCC degradation modes from crack initiation through failure. These models are each implemented in a modular form and linked together by a probabilistic framework that contains the logic for xLPR execution, exercises the individual modules as required, and performs necessary administrative and bookkeeping functions. The completion of the first production version of the xLPR code in a fully documented, releasable condition is presently planned for spring 2015

Genetic Code Analysis Toolkit: A novel tool to explore the coding properties of the genetic code and DNA sequences

Science.gov (United States)

Kraljić, K.; Strüngmann, L.; Fimmel, E.; Gumbel, M.

2018-01-01

The genetic code is degenerated and it is assumed that redundancy provides error detection and correction mechanisms in the translation process. However, the biological meaning of the code's structure is still under current research. This paper presents a Genetic Code Analysis Toolkit (GCAT) which provides workflows and algorithms for the analysis of the structure of nucleotide sequences. In particular, sets or sequences of codons can be transformed and tested for circularity, comma-freeness, dichotomic partitions and others. GCAT comes with a fertile editor custom-built to work with the genetic code and a batch mode for multi-sequence processing. With the ability to read FASTA files or load sequences from GenBank, the tool can be used for the mathematical and statistical analysis of existing sequence data. GCAT is Java-based and provides a plug-in concept for extensibility. Availability: Open source Homepage:http://www.gcat.bio/

Development of an international code of practice for dosimetry in X-ray diagnostic radiology

International Nuclear Information System (INIS)

Pernicka, F.; Carlsson, G.A.; Dance, D.R.; DeWerd, L.A.; Kramer, H.-M.; Ng, K.-H.

2001-01-01

Medical x-ray examinations contribute greatly to the population dose from man-made radiation sources. There is a need to control this dose and therefore to optimise the design and use of x-ray imaging systems. A key stage in this process is the standardisation of the procedures for dose measurement in the clinic. The Dosimetry and Medical Radiation Physics Section of the IAEA has a number of activities to further advance the standards for x-ray diagnostics. One of these activities is the coordination of a working group to develop a code of practice, which will facilitate the IAEA calibration activities, TLD intercomparisons and audits, educational activities, and technical assistance to Member States. The code of practice will aid in the standardisation of various dosimetric techniques in x-ray diagnostic radiology. The CoP working group has had an initial meeting to review the current status of dosimetry for conventional radiology, fluoroscopy, mammography, computed tomography and dental radiology. The CoP will include the establishment of standards and calibrations at the SSDLs, phantom and patient measurements and procedures for dosimetry in the clinic. (author)

Application of the severe accident code ATHLET-CD. Modelling and evaluation of accident management measures (Project WASA-BOSS)

Energy Technology Data Exchange (ETDEWEB)

Wilhelm, Polina; Jobst, Matthias; Kliem, Soeren; Kozmenkov, Yaroslav; Schaefer, Frank [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Div. Reactor Safety

2016-07-01

The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. Numerical analyses are used to investigate the accident progression and the complex physical phenomena during the core degradation phase, as well as to evaluate the effectiveness of possible countermeasures in the preventive and mitigative domain [1, 2]. The presented analyses have been performed with the computer code ATHLET-CD developed by GRS [3, 4].

Challenges on innovations of newly-developed safety analysis codes

International Nuclear Information System (INIS)

Yang, Yanhua; Zhang, Hao

2016-01-01

With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.

Challenges on innovations of newly-developed safety analysis codes

Energy Technology Data Exchange (ETDEWEB)

Yang, Yanhua [Shanghai Jiao Tong Univ. (China). School of Nuclear Science and Engineering; Zhang, Hao [State Nuclear Power Software Development Center, Beijing (China). Beijing Future Science and Technology City

2016-05-15

With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.

TORT-TD/ATTICA3D: a coupled neutron transport and thermal hydraulics code system for 3-D transient analysis of gas cooled high temperature reactors

International Nuclear Information System (INIS)

Lapins, J.; Seubert, A.; Buck, M.; Bader, J.; Laurien, E.

2011-01-01

Comprehensive safety studies of high temperature gas cooled reactors (HTR) require full three dimensional coupled treatments of both neutron kinetics and thermal-hydraulics. In a common effort, GRS and IKE developed the coupled code system TORT-TD/ATTICA3D for pebble bed type HTR that connects the 3-D transient discrete-ordinates transport code TORT-TD with the 3-D porous medium thermal-hydraulics code ATTICA3D. In this paper, the physical models and calculation capabilities of TORT-TD and ATTICA3D are presented, focusing on model improvements in ATTICA3D and extensions made in TORT-TD related to HTR application. For first applications, the OECD/NEA/NSC PBMR-400 benchmark has been chosen. Results obtained with TORT-TD/ATTICA3D will be shown for transient exercises, e.g. control rod withdrawal and a control rod ejection. Results are compared to other benchmark participants' solutions with special focus on fuel temperature modelling features of ATTICA3D. The provided “grey-curtain” nuclear cross section libraries have been used. First results on 3-D effects during a control rod withdrawal transient will be presented. (author)

Code Coupling for Multi-Dimensional Core Transient Analysis

International Nuclear Information System (INIS)

Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il

2015-01-01

After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident

Code Coupling for Multi-Dimensional Core Transient Analysis

Energy Technology Data Exchange (ETDEWEB)

Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il [KEPCO NF, Daejeon (Korea, Republic of)

2015-05-15

After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident.

A development and an application of Mixset-X computer code for simulating the Purex solvent extraction system

International Nuclear Information System (INIS)

Shida, M.; Naito, M.; Suto, T.; Omori, E.; Nojiri, T.

2001-01-01

MIXSET is a FORTRAN code developed to simulate the Purex solvent extraction system using mixer-settler extractors. Japan Nuclear Cycle Development Institute (JNC) has been developing the MIXSET code since the years 1970 to analyze the behavior of nuclides in the solvent extraction processes in Tokai Reprocessing Plant (TRP). This paper describes the history of MIXSET code development, the features of the latest version, called MIXSET-X and the application of the code for safety evaluation work. (author)

A comparative study of MONTEBURNS and MCNPX 2.6.0 codes in ADS simulations

International Nuclear Information System (INIS)

Barros, Graiciany P.; Pereira, Claubia; Veloso, Maria A.F.; Velasquez, Carlos E.; Costa, Antonella L.

2013-01-01

The possible use of the MONTEBURNS and MCNPX 2.6.0 codes in Accelerator-driven systems (ADSs) simulations for fuel evolution description is discussed. ADSs are investigated for fuel breeding and long-lived fission product transmutation so simulations of fuel evolution have a great relevance. The burnup/depletion capability is present in both studied codes. MONTEBURNS code links Monte Carlo N-Particle Transport Code (MCNP) to the radioactive decay burnup code ORIGEN2, whereas MCNPX depletion/ burnup capability is a linked process involving steady-state flux calculations by MCNPX and nuclide depletion calculations by CINDER90. A lead-cooled accelerator-driven system fueled with thorium was simulated and the results obtained using MONTEBURNS code and the results from MCNPX 2.6.0 code were compared. The system criticality and the variation of the actinide inventory during the burnup were evaluated and the results indicate a similar behavior between the results of each code. (author)

Time-dependent reliability sensitivity analysis of motion mechanisms

International Nuclear Information System (INIS)

Wei, Pengfei; Song, Jingwen; Lu, Zhenzhou; Yue, Zhufeng

2016-01-01

Reliability sensitivity analysis aims at identifying the source of structure/mechanism failure, and quantifying the effects of each random source or their distribution parameters on failure probability or reliability. In this paper, the time-dependent parametric reliability sensitivity (PRS) analysis as well as the global reliability sensitivity (GRS) analysis is introduced for the motion mechanisms. The PRS indices are defined as the partial derivatives of the time-dependent reliability w.r.t. the distribution parameters of each random input variable, and they quantify the effect of the small change of each distribution parameter on the time-dependent reliability. The GRS indices are defined for quantifying the individual, interaction and total contributions of the uncertainty in each random input variable to the time-dependent reliability. The envelope function method combined with the first order approximation of the motion error function is introduced for efficiently estimating the time-dependent PRS and GRS indices. Both the time-dependent PRS and GRS analysis techniques can be especially useful for reliability-based design. This significance of the proposed methods as well as the effectiveness of the envelope function method for estimating the time-dependent PRS and GRS indices are demonstrated with a four-bar mechanism and a car rack-and-pinion steering linkage. - Highlights: • Time-dependent parametric reliability sensitivity analysis is presented. • Time-dependent global reliability sensitivity analysis is presented for mechanisms. • The proposed method is especially useful for enhancing the kinematic reliability. • An envelope method is introduced for efficiently implementing the proposed methods. • The proposed method is demonstrated by two real planar mechanisms.

Modeling bubble condenser containment with computer code COCOSYS: post-test calculations of the main steam line break experiment at ELECTROGORSK BC V-213 test facility

International Nuclear Information System (INIS)

Lola, I.; Gromov, G.; Gumenyuk, D.; Pustovit, V.; Sholomitsky, S.; Wolff, H.; Arndt, S.; Blinkov, V.; Osokin, G.; Melikhov, O.; Melikhov, V.; Sokoline, A.

2005-01-01

Regulatory Authority using computer code COCOSYS. The code has been developed in GRS for the analysis of the containment response under design basis and severe accident conditions in light water reactors of different design, including WWERs. Comparison of the code calculations with experimental results has demonstrated reasonable accuracy of the code predictions. The paper describes the test facility and the SLB-G02 experiment, conducted by EREC, the COCOSYS model of the test facility, and compares the code predictions to the experimental results. (authors)

Development of the next generation reactor analysis code system, MARBLE

International Nuclear Information System (INIS)

Yokoyama, Kenji; Hazama, Taira; Nagaya, Yasunobu; Chiba, Go; Kugo, Teruhiko; Ishikawa, Makoto; Tatsumi, Masahiro; Hirai, Yasushi; Hyoudou, Hideaki; Numata, Kazuyuki; Iwai, Takehiko; Jin, Tomoyuki

2011-03-01

A next generation reactor analysis code system, MARBLE, has been developed. MARBLE is a successor of the fast reactor neutronics analysis code systems, JOINT-FR and SAGEP-FR (conventional systems), which were developed for so-called JUPITER standard analysis methods. MARBLE has the equivalent analysis capability to the conventional system because MARBLE can utilize sub-codes included in the conventional system without any change. On the other hand, burnup analysis functionality for power reactors is improved compared with the conventional system by introducing models on fuel exchange treatment and control rod operation and so on. In addition, MARBLE has newly developed solvers and some new features of burnup calculation by the Krylov sub-space method and nuclear design accuracy evaluation by the extended bias factor method. In the development of MARBLE, the object oriented technology was adopted from the view-point of improvement of the software quality such as flexibility, expansibility, facilitation of the verification by the modularization and assistance of co-development. And, software structure called the two-layer system consisting of scripting language and system development language was applied. As a result, MARBLE is not an independent analysis code system which simply receives input and returns output, but an assembly of components for building an analysis code system (i.e. framework). Furthermore, MARBLE provides some pre-built analysis code systems such as the fast reactor neutronics analysis code system. SCHEME, which corresponds to the conventional code and the fast reactor burnup analysis code system, ORPHEUS. (author)

Fusion PIC code performance analysis on the Cori KNL system

Energy Technology Data Exchange (ETDEWEB)

Koskela, Tuomas S. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). National Energy Research Scientific Computing Center (NERSC); Deslippe, Jack [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). National Energy Research Scientific Computing Center (NERSC); Friesen, Brian [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). National Energy Research Scientific Computing Center (NERSC); Raman, Karthic [INTEL Corp. (United States)

2017-05-25

We study the attainable performance of Particle-In-Cell codes on the Cori KNL system by analyzing a miniature particle push application based on the fusion PIC code XGC1. We start from the most basic building blocks of a PIC code and build up the complexity to identify the kernels that cost the most in performance and focus optimization efforts there. Particle push kernels operate at high AI and are not likely to be memory bandwidth or even cache bandwidth bound on KNL. Therefore, we see only minor benefits from the high bandwidth memory available on KNL, and achieving good vectorization is shown to be the most beneficial optimization path with theoretical yield of up to 8x speedup on KNL. In practice we are able to obtain up to a 4x gain from vectorization due to limitations set by the data layout and memory latency.

Simulation of the fuel rod bundle test QUENCH-03 using the system codes ASTEC and ATHLET-CD

International Nuclear Information System (INIS)

Kruse, P.; Koch, M.K.

2011-01-01

The QUENCH-03 test was performed on the 21. of January 1999 at FZK (Forschungszentrum Karlsruhe) to investigate the behaviour on reflood of PWR (Pressurized Water Reactor) fuel rods with little oxidation. This paper presents the results of the simulation of QUENCH-03 performed with the version V1.3 of the integral code ASTEC (Accident Source Term Evaluation Code) which is being developed by IRSN (France) in cooperation with GRS (Germany) and with the program version 2.1A of the mechanistic code ATHLET-CD (Analysis of Thermal-hydraulics of Leaks and Transients - Core Degradation) which is under development by GRS. At first the QUENCH test facility and the QUENCH test program in general are described. The test conduct of the test QUENCH-03 follows as well as a description of the used codes ASTEC and ATHLET-CD with the associated modeling of the test section. The results of this calculation show that during the heat-up and transient phase both codes can calculate bundle and shroud temperatures as well as the hydrogen production in good approximation to the experimental data. During the quench phase and up to the end of the test only the oxidation model PRATER of ASTEC simulates the hydrogen production very well, the other oxidation models of ASTEC cannot calculate to some extent the measured amount of hydrogen. ATHLET-CD underestimates the integral amount at the end of the test. In the ASTEC calculations the temperatures during the quench phase show qualitatively good results, only time delays on some elevations of the bundle could be noticed. ATHLET-CD reproduces the thermal behaviour up to the first temperature escalation very well, after that the temperatures are partly over-estimated. The time delay recognized in the ASTEC calculations are seen as well. The results of the integral code ASTEC emphasize that the calculation of QUENCH-03 is possible and leading to good results concerning hydrogen release and corresponding temperatures. Because the QUENCH-03 test was

Estructura de las Reglas de Origen del TLCAN: un Análisis Metodológico del Valor de Contenido Regional

Directory of Open Access Journals (Sweden)

Odette Virginia Delfín Ortega

2011-12-01

Full Text Available La dinámica actual dentro del comercio internacional pone en manifiesto la importancia del origen de las mercancías para poder llevar a cabo los principales instrumentos de política comercial. La aplicación de los regímenes de origen no es una tarea sencilla; ya que se requiere información de diversos actores que intervienen dentro de la operación de comercio exterior. Las normas de origen generan impactos muy representativos en los flujos de mercancías, servicios e inversiones. Se puede decir entonces que las normas de origen son junto con la valoración en aduana, la clasificación arancelaria, requisitos previos para la aplicación de los principales instrumentos de la política comercial (aranceles, restricciones al comercio, medidas de defensa comercial, la gestión de las estadísticas comerciales, protección al consumidor, entre otros (GATT, 1993. Sin embargo, el origen de la mercancía paralelo a los instrumentos de la política comercial, puede activar o afectar el flujo comercial de una nación. Según los objetivos de política comercial que vienen estipulados en el régimen del origen, se encuentran dos grandes tipos de normas, las preferenciales y las comunes o no preferenciales. La principal diferencia entre un tipo de norma y otro guarda relación con el instrumento comercial al cual obedece. Las preferenciales, como su nombre indica, sirven para ejecutar un esquema comercial preferencial (acuerdos comerciales regionales y sistemas preferencias generalizadas, y las comunes resultan aplicables a todos los demás supuestos (aranceles NMF, cuotas, salvaguardias, estadísticas, etcétera. El régimen jurídico del origen se basa en dos principios básicos: el de producto enteramente obtenido y el de transformación sustancial (Rodriguez & Rodrik, 1999. Las reglas de origen han empezado a ser analizadas recientemente, autores como: Krueger, López-de-Silanes, Rosellón, Wonnacott (Reza & Kowalsky,2010 han hecho diversos

Adaptation of Toodee-2 computer code for reflood analysis in Angra-1 reactor

International Nuclear Information System (INIS)

Praes, J.G.L.; Onusic Junior, J.

1981-01-01

A method of calculation the heat transfer coefficient used in Toodee-2 computer code for core reflood analysis in a loss of coolant accident, is presented. Preliminary results are presented with the use of heat transfer correlations based on FLECHT experiments adequate to a geometric arrangement such as 16 x 16 (Angra I). Optional calculations are suggested for the heat transfer coefficients when the cooling of fuel cladding by steam is used. (Author) [pt

Percepción del alumnado universitario sobre el origen de la violencia escolar

Directory of Open Access Journals (Sweden)

José J. Gázquez

2008-03-01

Full Text Available Son muchas las teorías y estudios que analizan cuál es el origen de las conductas violentas, pero no de forma concreta, en lo que respecta a la violencia que tiene lugar dentro del contexto educativo, sino más bien orientados a la conducta violenta juvenil. En el presente estudio se analiza la percepción de los universitarios (recientes estudiantes de ESO y Bachillerato a cerca de todos los aspectos relacionados con el origen de las conductas violentas. La muestra está compuesta por un total de 184 sujetos universitarios pertenecientes a las titulaciones de Psicología y Magisterio de la especialidad de Educación Infantil y de Primaria. Se ha utilizado el Cuestionario de factores predictores de la violencia escolar para universitarios. Su aplicación nos muestra que los/as universitarios/as consideran “La escasa educación en el respeto a los además y a las cosas” el aspecto que más influye en el origen de la violencia escolar. La opinión general coincide con la opinión mayoritaria de los hombres, mientras que las mujeres destacan “La pertenencia a un grupo de iguales conflictivo”. Por último, observamos que medida que aumenta la frecuencia con que se es víctima, agresor y observador, disminuye la importancia dada a “La falta de dedicación por parte de los padres/tutores”, “La falta de educación en valores como solidaridad, generosidad, bondad, etc.”, y “El consumo de sustancias como el alcohol, tabaco, drogas, etc.” en el origen de las conductas violentas dentro de la escuela.

La guirnalda de las letras: origen, naturaleza y poder del lenguaje

OpenAIRE

Vachot, Charles

2010-01-01

Acerca del origen y poder del lenguaje, ilustrado mediante el análisis de formulaciones mítico-religiosas.Conferencia celebrada en Lyon por los Amigos del Museo Guimet, el 8 de junio de 1958. Traducción de Julio Sánchez Reyes.

Uncertainty analysis of the FRAP code

International Nuclear Information System (INIS)

Peck, S.O.

1978-01-01

A user oriented, automated uncertainty analysis capability has been built into the FRAP code (Fuel Rod Analysis Program) and applied to a PWR fuel rod undergoing a LOCA. The method of uncertainty analysis is the Response Surface Method (RSM). (author)

Performance testing of thermal analysis codes for nuclear fuel casks

International Nuclear Information System (INIS)

Sanchez, L.C.

1987-01-01

In 1982 Sandia National Laboratories held the First Industry/Government Joint Thermal and Structural Codes Information Exchange and presented the initial stages of an investigation of thermal analysis computer codes for use in the design of nuclear fuel shipping casks. The objective of the investigation was to (1) document publicly available computer codes, (2) assess code capabilities as determined from their user's manuals, and (3) assess code performance on cask-like model problems. Computer codes are required to handle the thermal phenomena of conduction, convection and radiation. Several of the available thermal computer codes were tested on a set of model problems to assess performance on cask-like problems. Solutions obtained with the computer codes for steady-state thermal analysis were in good agreement and the solutions for transient thermal analysis differed slightly among the computer codes due to modeling differences

Simulation of pulse height analysis soft X-ray spectra expected from W7-X

Science.gov (United States)

Jabłoński, S.; Czarnecka, A.; Kubkowska, M.; Ryć, L.; Weller, A.; Biedermann, C.; König, R.; W7-X Team

2015-10-01

A computer code named RayX has been developed for checking the performance of a spectroscopy system and optimizing individual parts, like detectors and filters for the pulse height analysis (PHA) diagnostic system designed for the stellarator W7-X. Using the code, the intensity and shape of the X-ray spectra are simulated for different plasma scenarios characterized by varying the temperature and density profiles as well as the electron cyclotron resonance heating (ECRH) power over a wide range. In the simulations of the recorded spectra, the influence of geometrical configuration changes of the diagnostic system (pinhole size, detector size, location of each diagnostic component), of the timing of data collection, as well as of the type and thickness of filters are being investigated. The atomic processes of free-free (bremsstrahlung), free-bound (recombination radiation), and bound-bound emission (line radiation) are considered. For the impurities fractional abundancies of 3% carbon (C), 0.5% oxygen (O) and 0.002% iron (Fe) are taken into account. Information about the number of photons which reach the detector and the current generated inside the detector is given. It is shown that the distance between pinhole and detector has a larger impact on the registered spectra (intensity and total number of photons) than the distance between plasma and pinhole. Based on the results of the simulations, the expected optimal positions of the individual components (pinholes, detectors) were defined for the PHA W7-X diagnostic system. A shorter version of this contribution is due to be published in PoS at: 1st EPS conference on Plasma Diagnostics

PARÁMETROS HEMATOLÓGICOS EN POLLUELOS DE PSITÁCIDOS EN CAUTIVERIO DE ORIGEN SILVESTRE

Directory of Open Access Journals (Sweden)

J. A. Herrera

2013-01-01

Full Text Available El objetivo de este estudio fue describir los parámetros hematológicos en 69 polluelos en cautiverio de origen silvestre, clínicamente sanos, menores de tres meses de edad, en cuatro especies de psitácidos: Pionus senilis (n = 18, Amazona albifrons (n = 15, Amazona autumnalis (n = 25 y Amazona oratrix (n = 11, procedentes de la región central de México. Para determinar dichos parámetros, de cada ejemplar se obtuvieron 500 μl de sangre por punción de la vena braquial. En las cuatro especies citadas, el porcentaje del hematocrito fue de 56,4 ± 3,36; 51 ± 6,52; 47,6 ± 6,47 y 53,8 ± 9,26; el VCM en fl fue de 184,6 ± 13,4; 142,0 ± 25,4; 106,3 ± 11,7 y 162,3 ± 71,9; la cuenta eritrocítica fue de 4,7 ± 2,1 x10 6 /μl; 6,1 ± 8,2 x10 6 /μl; 6,9 ± 1,2 x10 6 /μl y 5,4 ± 1,1 x10 6 /μl; y la cuenta leucocitaria de 3,06 ± 0,15 x10 3 /μl; 3,62 ± 0,24 x10 3 /μl; 4,48 ± 0,36 x10 3 /μl y 3,56 ± 0,71 x10 3 /μl respectivamente para P. senilis , A. albifrons , A. autumnalis y A. oratrix . Se constató tendencia a la heterofilia en A. albifrons y A. oratrix , lo cual puede ser de gran utilidad para un diagnóstico cada vez más preciso en la clínica de aves en cautiverio.

CATHENA 4. A thermalhydraulics network analysis code

International Nuclear Information System (INIS)

Aydemir, N.U.; Hanna, B.N.

2009-01-01

Canadian Algorithm for THErmalhydraulic Network Analysis (CATHENA) is a one-dimensional, non-equilibrium, two-phase, two fluid network analysis code that has been in use for over two decades by various groups in Canada and around the world. The objective of the present paper is to describe the design, application and future development plans for the CATHENA 4 thermalhydraulics network analysis code, which is a modernized version of the present frozen CATHENA 3 code. The new code is designed in modular form, using the Fortran 95 (F95) programming language. The semi-implicit numerical integration scheme of CATHENA 3 is re-written to implement a fully-implicit methodology using Newton's iterative solution scheme suitable for nonlinear equations. The closure relations, as a first step, have been converted from the existing CATHENA 3 implementation to F95 but modularized to achieve ease of maintenance. The paper presents the field equations, followed by a description of the Newton's scheme used. The finite-difference form of the field equations is given, followed by a discussion of convergence criteria. Two applications of CATHENA 4 are presented to demonstrate the temporal and spatial convergence of the new code for problems with known solutions or available experimental data. (author)

Origen del concepto «deporte»

Directory of Open Access Journals (Sweden)

Saul GARCÍA BLANCO

2009-11-01

Full Text Available RESUMEN: Pretende este ensayo zanjar los equívocos existentes sobre el origen del concepto «deporte». Hoy en día y especialmente desde el campo de la sociología, son no pocos los autores que otorgan a la sociedad industrializada de finales del XIX y fundamentalmente a la inglesa, el ser los creadores del deporte. El presente trabajo plantea una tesis contraria al respecto. Basándose en estudios de antropología, educación e incluso semánticos, y con la interpretación de un profesional de la Educación Física, se demuestra cómo el origen del deporte se encuentra enraizado en las actividades lúdicas de las sociedades primitivas. Así, el autor otorga la paternidad de los actuales deportes, a excepción de los mecánicos, a juegos practicados por culturas que habían alcanzado el nivel urbano y disponían, por tanto, de tiempo libre.SUMMARY: The aim of this paper is to settle existing misunderstandings as to the origin of the concept «sport». At present, and especially from the field of sociology, a number of authors attribute the creation of sport to industrialised society at the end of the 19th century, and basically English society. This study proposes a counter thesis in this respect. Taking as a basis studies in anthropology, education and even semantics, together with an interpretation by a professional in Physical Education, it is shown how the origin of sport derives from activities relating to games in primitive societies. Thus the author attributes the source of present sports, with the exception of mechanical ones, to games practised by cultures which had reached the urban stage and therefore had free time.

VVER 1000 SBO calculations with pressuriser relief valve stuck open with ASTEC computer code

International Nuclear Information System (INIS)

Atanasova, B.P.; Stefanova, A.E.; Groudev, P.P.

2012-01-01

Highlights: ► We modelled the ASTEC input file for accident scenario (SBO) and focused analyses on the behaviour of core degradation. ► We assumed opening and stuck-open of pressurizer relief valve during performance of SBO scenario. ► ASTEC v1.3.2 has been used as a reference code for the comparison study with the new version of ASTEC code. - Abstract: The objective of this paper is to present the results obtained from performing the calculations with ASTEC computer code for the Source Term evaluation for specific severe accident transient. The calculations have been performed with the new version of ASTEC. The ASTEC V2 code version is released by the French IRSN (Institut de Radioprotection at de surete nucleaire) and Gesellschaft für Anlagen-und Reaktorsicherheit (GRS), Germany. This investigation has been performed in the framework of the SARNET2 project (under the Euratom 7th framework program) by Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Science (INRNE-BAS).

Modeling of severe accident sequences with the new modules CESAR and DIVA of ASTEC system code

International Nuclear Information System (INIS)

Pignet, Sophie; Guillard, Gaetan; Barre, Francois; Repetto, Georges

2003-01-01

Systems of computer codes, so-called 'integral' codes, are being developed to simulate the scenario of a hypothetical severe accident in a light water reactor, from the initial event until the possible radiological release of fission products out of the containment. They couple the predominant physical phenomena that occur in the different reactor zones and simulate the actuation of safety systems by procedures and by operators. In order to allow to study a great number of scenarios, a compromise must be found between precision of results and calculation time: one day of accident time should take less than one day of real time to simulate on a PC computer. This search of compromise is a real challenge for such integral codes. The development of the ASTEC integral code was initiated jointly by IRSN and GRS as an international reference code. The latest version 1.0 of ASTEC, including the new modules CESAR and DIVA which model the behaviour of the reactor cooling system and the core degradation, is presented here. Validation of the modules and one plant application are described

Numerical Analysis for IFM Grid Effect on 5x5 Rods Bundle

International Nuclear Information System (INIS)

Kim, Seong Jin; Cha, Jeong Hun; Seo, Kyong Won; Kim, Tae Woo; Kwon, Hyuk; Hwang, Dae Hyun

2011-01-01

Generally, the fuel assembly consists of fuel rods, bottom and top grids, spacer grids, mixing vane, etc. The mixing vane with spacer grid is used to increase the thermal mixing between subchannels and to increase CHF(Critical Heat Flux). IFM(Intermediate Flow Mixer) grids are used to induce lateral flow between adjacent channels and are well-known as improving CHF, also. A numerical analysis using CFD code(ANSYS CFX, version 12.1) and subchannel code(MATRA-S) was conducted to investigate the influence of IFM grid on the subchannel temperature in 5x5 rods bundle with and without the IFM grid, thermohydraulically. In this study, the quantitative improvement of the mixing effect of the IFM grid is presented from the results of CFX and MATRA-S code. Moreover, capacity of predicting subchannel temperature of MATRA-S code is compared with CFX result

ORIGEN2 libraries based on JENDL-3.2 for LWR-MOX fuels

Energy Technology Data Exchange (ETDEWEB)

Suyama, Kenya; Katakura, Jun-ichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Onoue, Masaaki; Matsumoto, Hideki [Mitsubishi Heavy Industries Ltd., Tokyo (Japan); Sasahara, Akihiro [Central Research Inst. of Electric Power Industry, Tokyo (Japan)

2000-11-01

A set of ORIGEN2 libraries for LWR MOX fuels was developed based on JENDL-3.2. The libraries were compiled with SWAT using the specification of MOX fuels that will be used in nuclear power reactors in Japan. The verification of the libraries were performed by the analyses of post irradiation examinations for the fuels from European PWR. By the analysis of PIE data from PWR in United States, the comparison was made between calculation and experimental results in the case of that parameters for making the libraries are different from irradiation conditions. These new libraries for LWR MOX fuels are packaged in ORLIBJ32, the libraries released in 1999. (author)

Monte Carlo Simulation of Complete X-Ray Spectra for Use in Scanning Electron Microscopy Analysis

International Nuclear Information System (INIS)

Roet, David; Van Espen, Piet

2003-01-01

Full Text: The interactions of keV electrons and photons with matter can be simulated accurately with the aid of the Monte Carlo (MC) technique. In scanning electron microscopy x-ray analysis (SEM-EDX) such simulations can be used to perform quantitative analysis using a Reverse Monte Carlo method even if the samples have irregular geometry. Alternatively the MC technique can generate spectra of standards for use in quantization with partial least squares regression. The feasibility of these alternatives to the more classical ZAF or phi-rho-Z quantification methods has been proven already. In order to be applicable for these purposes the MC-code needs to generate accurately only the characteristic K and L x-ray lines, but also the Bremsstrahlung continuum, i.e. the complete x-ray spectrum need to be simulated. Currently two types of MC simulation codes are available. Programs like Electron Flight Simulator and CASINO simulate characteristic x-rays due to electron interaction in a fast and efficient way but lack provision for the simulation of the continuum. On the other hand, programs like EGS4, MCNP4 and PENELOPE, originally developed for high energy (MeV- GeV) applications, are more complete but difficult to use and still slow, even on todays fastest computers. We therefore started the development of a dedicated MC simulation code for use in quantitative SEM-EDX work. The selection of the most appropriate cross section for the different interactions will be discussed and the results obtained will be compared with those obtained with existing MC programs. Examples of the application of MC simulations for quantitative analysis of samples with various composition will be given

Input/output manual of light water reactor fuel analysis code FEMAXI-7 and its related codes

Energy Technology Data Exchange (ETDEWEB)

Suzuki, Motoe; Udagawa, Yutaka; Nagase, Fumihisa [Japan Atomic Energy Agency, Nuclear Safety Research Center, Tokai, Ibaraki (Japan); Saitou, Hiroaki [ITOCHU Techno-Solutions Corporation, Tokyo (Japan)

2013-10-15

A light water reactor fuel analysis code FEMAXI-7 has been developed, as an extended version from the former version FEMAXI-6, for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which are fully disclosed in the code model description published in the form of another JAEA-Data/Code report. The present manual, which is the very counterpart of this description document, gives detailed explanations of files and operation method of FEMAXI-7 code and its related codes, methods of input/output, sample Input/Output, methods of source code modification, subroutine structure, and internal variables in a specific manner in order to facilitate users to perform fuel analysis by FEMAXI-7. (author)

Input/output manual of light water reactor fuel analysis code FEMAXI-7 and its related codes

International Nuclear Information System (INIS)

Suzuki, Motoe; Udagawa, Yutaka; Nagase, Fumihisa; Saitou, Hiroaki

2013-10-01

A light water reactor fuel analysis code FEMAXI-7 has been developed, as an extended version from the former version FEMAXI-6, for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which are fully disclosed in the code model description published in the form of another JAEA-Data/Code report. The present manual, which is the very counterpart of this description document, gives detailed explanations of files and operation method of FEMAXI-7 code and its related codes, methods of input/output, sample Input/Output, methods of source code modification, subroutine structure, and internal variables in a specific manner in order to facilitate users to perform fuel analysis by FEMAXI-7. (author)

Containment Modelling with the ASTEC Code

International Nuclear Information System (INIS)

Sadek, Sinisa; Grgic, Davor

2014-01-01

ASTEC is an integral computer code jointly developed by Institut de Radioprotection et de Surete Nucleaire (IRSN, France) and Gesellschaft fur Anlagen-und Reaktorsicherheit (GRS, Germany) to assess the nuclear power plant behaviour during a severe accident (SA). It consists of 13 coupled modules which compute various SA phenomena in primary and secondary circuits of the nuclear power plants (NPP), and in the containment. The ASTEC code was used to model and to simulate NPP behaviour during a postulated station blackout accident in the NPP Krsko, a two-loop pressurized water reactor (PWR) plant. The primary system of the plant was modelled with 110 thermal hydraulic (TH) volumes, 113 junctions and 128 heat structures. The secondary system was modelled with 76 TH volumes, 77 junctions and 87 heat structures. The containment was modelled with 10 TH volumes by taking into account containment representation as a set of distinctive compartments, connected with 23 junctions. A total of 79 heat structures were used to simulate outer containment walls and internal steel and concrete structures. Prior to the transient calculation, a steady state analysis was performed. In order to achieve correct plant initial conditions, the operation of regulation systems was modelled. Parameters which were subjected to regulation were the pressurizer pressure, the pressurizer narrow range level and steam mass flow rates in the steam lines. The accident analysis was focused on containment behaviour, however the complete integral NPP analysis was carried out in order to provide correct boundary conditions for the containment calculation. During the accident, the containment integrity was challenged by release of reactor system coolant through degraded coolant pump seals and, later in the accident following release of the corium out of the reactor pressure vessel, by the molten corium concrete interaction and direct containment heating mechanisms. Impact of those processes on relevant

Verification of the CONPAS (CONtainment Performance Analysis System) code package

International Nuclear Information System (INIS)

Kim, See Darl; Ahn, Kwang Il; Song, Yong Man; Choi, Young; Park, Soo Yong; Kim, Dong Ha; Jin, Young Ho.

1997-09-01

CONPAS is a computer code package to integrate the numerical, graphical, and results-oriented aspects of Level 2 probabilistic safety assessment (PSA) for nuclear power plants under a PC window environment automatically. For the integrated analysis of Level 2 PSA, the code utilizes four distinct, but closely related modules: (1) ET Editor, (2) Computer, (3) Text Editor, and (4) Mechanistic Code Plotter. Compared with other existing computer codes for Level 2 PSA, and CONPAS code provides several advanced features: computational aspects including systematic uncertainty analysis, importance analysis, sensitivity analysis and data interpretation, reporting aspects including tabling and graphic as well as user-friendly interface. The computational performance of CONPAS has been verified through a Level 2 PSA to a reference plant. The results of the CONPAS code was compared with an existing level 2 PSA code (NUCAP+) and the comparison proves that CONPAS is appropriate for Level 2 PSA. (author). 9 refs., 8 tabs., 14 figs

Origen of Alexandria: The study of the Scriptures as transformation ...

African Journals Online (AJOL)

Test

2011-06-07

Jun 7, 2011 ... ... of the Scriptures as transformation of the readers into images of the God of love ... He continues: 'it is clear that the authoritative text being explicated was not ..... to John as John's mother.19 Origen first explains this kind of identity by ... the Logos himself approaches the home of the readers and knocks in ...

A Semantic Analysis Method for Scientific and Engineering Code

Science.gov (United States)

Stewart, Mark E. M.

1998-01-01

This paper develops a procedure to statically analyze aspects of the meaning or semantics of scientific and engineering code. The analysis involves adding semantic declarations to a user's code and parsing this semantic knowledge with the original code using multiple expert parsers. These semantic parsers are designed to recognize formulae in different disciplines including physical and mathematical formulae and geometrical position in a numerical scheme. In practice, a user would submit code with semantic declarations of primitive variables to the analysis procedure, and its semantic parsers would automatically recognize and document some static, semantic concepts and locate some program semantic errors. A prototype implementation of this analysis procedure is demonstrated. Further, the relationship between the fundamental algebraic manipulations of equations and the parsing of expressions is explained. This ability to locate some semantic errors and document semantic concepts in scientific and engineering code should reduce the time, risk, and effort of developing and using these codes.

No pain: no gain: The complex art of soft x-ray laser target design and analysis

International Nuclear Information System (INIS)

Rosen, M.D.; London, R.A.; Hagelstein, P.L.

1988-01-01

We review our methodologies in the design and analysis of soft x-ray laser experiments. We convolve large scale 2-D hydro code output with detailed atomic data bases in a kinetics code with 1-D or 2-D line transfer. The time and space dependent level population data is then post processed further with a beam transport code, including refraction, to predict actual experimental results. While mysteries do remain, we present many examples that show how this complex modeling procedure is crucial in explaining experimental results

Content Analysis Coding Schemes for Online Asynchronous Discussion

Science.gov (United States)

Weltzer-Ward, Lisa

2011-01-01

Purpose: Researchers commonly utilize coding-based analysis of classroom asynchronous discussion contributions as part of studies of online learning and instruction. However, this analysis is inconsistent from study to study with over 50 coding schemes and procedures applied in the last eight years. The aim of this article is to provide a basis…

Production of analysis code for 'JOYO' dosimetry experiment

International Nuclear Information System (INIS)

Sasaki, Makoto; Nakazawa, Masaharu.

1981-01-01

As part of the measurement and analysis plan for the Dosimetry Experiment at the ''JOYO'' experimental fast reactor, neutron flux spectra analysis is performed using the NEUPAC (Neutron Unfolding Code Package) computer program. The code calculates the neutron flux spectra and other integral quantities from the activation data of the dosimeter foils. The NEUPAC code is based on the J1-type unfolding method, and the estimated neutron flux spectra is obtained as its solution. The program is able to determine the integral quantities and their sensitivities, together with an error estimate of the unfolded spectra and integral quantities. The code also performs a chi-square test of the input/output data, and contains many options for the calculational routines. This report presents the analytic theory, the program algorithms, and a description of the functions and use of the NEUPAC code. (author)

A posteriori error analysis for hydro-mechanical couplings and implementation in Code-Aster; Analyse d'erreur a posteriori pour les couplages hydro-mecaniques et mise en oeuvre dans Code-Aster

Energy Technology Data Exchange (ETDEWEB)

Meunier, S

2007-11-15

We analyse approximations by finite elements in space and finite differences in time of coupled Hydro-Mechanical (HM) problems related to the quasi-static linear poro-elasticity theory. The physical bases of this theory are briefly restated and an abstract setting is proposed to perform the mathematical study of the stationary and un-stationary versions of the HM problem. For the stationary version, the well-posedness of the continuous and discrete problems are established and the a priori error analysis is performed. Then, we propose the a posteriori error analysis by using two different techniques suited to estimate the displacement error and the pressure error, respectively, both in the H{sub x}{sup 1}-norm. The classical properties of reliability and optimality are proved for the associated error estimators. Some numerical experiments using Code-Aster illustrate the theoretical results. For the un-stationary version, we first establish a stability result for the continuous problem. Then, we present an optimal a priori error analysis using elliptic projection techniques. Finally, the a posteriori error analysis is performed by using two different approaches: a direct approach and an elliptic reconstruction approach. The first is suited to estimate the pressure error in the L{sub t}{sup 2}(H{sub x}{sup 1})-norm and the second is suited to estimate the displacement error in the L{sub t}{sup {infinity}}(H{sub x}{sup 1})-norm and the pressure error in the L{sub t}{sup {infinity}}(H{sub x}{sup 1})-norm. Numerical experiments using Code-Aster complete the theoretical results. (author)

Study of high energy emissions from stellar mass accreting holes

International Nuclear Information System (INIS)

Cadolle-Bel, Marion

2006-01-01

The present work is dedicated to the study of various X-ray binary Systems harbouring accreting stellar mass black holes (or candidates) associated in X-ray binary Systems mainly through the spectral and timing properties of the high energy 3 keV"-"1 MeV emission, sometimes completed by observations performed in radio, near-infrared and optical. The first part is devoted to accretion physics phenomena and the challenges of understanding the X-ray/gamma emission produced with the modeling of such high energy processes. Then I will define in a second part the instruments on board INTEGRAL and the way coded masked aperture is employed. In a third part, I will develop the standard data reduction analysis and my own contribution in improving the usual software before detailing the specific informatics tools I have developed for my own analysis. In the fourth part I will turn towards the deep analysis and interpretations I have performed on several black hole X-ray binary Systems chosen properly: the persistent black hole source Cygnus X-1 which has been studied since several years and surprised us by a high-energy excess detected; two new transient sources which provide interesting information, XTE J1720-318 located in the galactic bulge and SWIFT J1753.5-0127, probably situated in the halo. I will also detail my work on H 1743-322, recently identified by INTEGRAL as the HEAO source discovered in 1977, and on three (almost) persistent micro-quasars with superluminal jets, 1E 1740.7-2942, GRS 1758-258 and GRS 1915+105. I will analyze for each source spectral parameter evolutions and their links with each other during state transitions. I will then discuss the presence of two different X/gamma-ray emitting media with a relatively changing geometry. While establishing a cyclic order for the different variability classes of GRS 1915+105 observed during ten years, I will propose an interpretation for such behaviour, compatible with the theoretical predictions of the

Development of burnup methods and capabilities in Monte Carlo code RMC

International Nuclear Information System (INIS)

She, Ding; Liu, Yuxuan; Wang, Kan; Yu, Ganglin; Forget, Benoit; Romano, Paul K.; Smith, Kord

2013-01-01

Highlights: ► The RMC code has been developed aiming at large-scale burnup calculations. ► Matrix exponential methods are employed to solve the depletion equations. ► The Energy-Bin method reduces the time expense of treating ACE libraries. ► The Cell-Mapping method is efficient to handle massive amounts of tally cells. ► Parallelized depletion is necessary for massive amounts of burnup regions. -- Abstract: The Monte Carlo burnup calculation has always been a challenging problem because of its large time consumption when applied to full-scale assembly or core calculations, and thus its application in routine analysis is limited. Most existing MC burnup codes are usually external wrappers between a MC code, e.g. MCNP, and a depletion code, e.g. ORIGEN. The code RMC is a newly developed MC code with an embedded depletion module aimed at performing burnup calculations of large-scale problems with high efficiency. Several measures have been taken to strengthen the burnup capabilities of RMC. Firstly, an accurate and efficient depletion module called DEPTH has been developed and built in, which employs the rational approximation and polynomial approximation methods. Secondly, the Energy-Bin method and the Cell-Mapping method are implemented to speed up the transport calculations with large numbers of nuclides and tally cells. Thirdly, the batch tally method and the parallelized depletion module have been utilized to better handle cases with massive amounts of burnup regions in parallel calculations. Burnup cases including a PWR pin and a 5 × 5 assembly group are calculated, thereby demonstrating the burnup capabilities of the RMC code. In addition, the computational time and memory requirements of RMC are compared with other MC burnup codes.

Status of development and verification of the CTFD code FLUBOX

International Nuclear Information System (INIS)

Graf, U.; Paradimitriou, P.

2004-01-01

The Computational Two-Fluid Dynamics (CTFD) code FLUBOX is developed at GRS for the multidimensional simulation of two-phase flows. FLUBOX will also be used as a multidimensional module for the German system code ATHLET. The Benchmark test cases of the European ASTAR project were used to verify the ability of the code FLUBOX to calculate typical two-phase flow phenomena and conditions: void and pressure wave propagation, phase transitions, countercurrent flows, sharp interface movements, compressible (vapour) and nearly incompressible (water) conditions, thermal and mechanical non-equilibrium, stiff source terms due to mass and heat transfer between the phases. Realistic simulations of two-phase require beside the pure conservation equations additional transport equations for the interfacial area, turbulent energy and dissipation. A transport equation for the interfacial area density covering the whole two-phase flow range is in development. First validation calculations are presented in the paper. Turbulent shear stress for two-phase flows will be modelled by the development of transport equations for the turbulent kinetic energy and the turbulent dissipation rate. The development of the transport equations is mainly based on first principles on bubbles or drops and is largely free from empiricism. (author)

Sensitivity analysis of the RESRAD, a dose assessment code

International Nuclear Information System (INIS)

Yu, C.; Cheng, J.J.; Zielen, A.J.

1991-01-01

The RESRAD code is a pathway analysis code that is designed to calculate radiation doses and derive soil cleanup criteria for the US Department of Energy's environmental restoration and waste management program. the RESRAD code uses various pathway and consumption-rate parameters such as soil properties and food ingestion rates in performing such calculations and derivations. As with any predictive model, the accuracy of the predictions depends on the accuracy of the input parameters. This paper summarizes the results of a sensitivity analysis of RESRAD input parameters. Three methods were used to perform the sensitivity analysis: (1) Gradient Enhanced Software System (GRESS) sensitivity analysis software package developed at oak Ridge National Laboratory; (2) direct perturbation of input parameters; and (3) built-in graphic package that shows parameter sensitivities while the RESRAD code is operational

La importancia del <origen en la fase de evaluacion del consumidor de productos horticolas

OpenAIRE

Guerrero, Jose Felipe; Abad, Juan Carlos; Aguera, Emilia Del

2010-01-01

La percepcion sobre el pais de origen de un producto es uno de los principales determinates de la evaluacion y eleccion de dicho producto, sobre todo en los mercados foraneos. En la literatura de marketing existen multiples trabajos que han analizado este aspecto, si bien la amyor parte de ellos se han centrado en productos duraderos de alta implicacion. Este trabajo analiza la importancia que el atributo pais de origen tiene en el proceso de evaluacion del consumidor para el caso de la compr...

Analytical validation of the CACECO containment analysis code

International Nuclear Information System (INIS)

Peak, R.D.

1979-08-01

The CACECO containment analysis code was developed to predict the thermodynamic responses of LMFBR containment facilities to a variety of accidents. This report covers the verification of the CACECO code by problems that can be solved by hand calculations or by reference to textbook and literature examples. The verification concentrates on the accuracy of the material and energy balances maintained by the code and on the independence of the four cells analyzed by the code so that the user can be assured that the code analyses are numerically correct and independent of the organization of the input data submitted to the code

Development of a nuclear power plant system analysis code

International Nuclear Information System (INIS)

Sim, Suk K.; Jeong, J. J.; Ha, K. S.; Moon, S. K.; Park, J. W.; Yang, S. K.; Song, C. H.; Chun, S. Y.; Kim, H. C.; Chung, B. D.; Lee, W. J.; Kwon, T. S.

1997-07-01

During the period of this study, TASS 1.0 code has been prepared for the non-LOCA licensing and reload safety analyses of the Westinghouse and the Korean Standard Nuclear Power Plants (KSNPP) type reactors operating in Korea. TASS-NPA also has been developed for a real time simulation of the Kori-3/4 transients using on-line graphical interactions. TASS 2.0 code has been further developed to timely apply the TASS 2.0 code for the design certification of the KNGR. The COBRA/RELAP5 code, a multi-dimensional best estimate system code, has been developed by integrating the realistic three-dimensional reactor vessel model with the RELAP5 /MOD3.2 code, a one-dimensional system code. Also, a 3D turbulent two-phase flow analysis code, FEMOTH-TF, has been developed using finite element technique to analyze local thermal hydraulic phenomena in support of the detailed design analysis for the development of the advanced reactors. (author). 84 refs., 27 tabs., 83 figs

COMPUTATION FORMAT computer codes X4TOC4 and PLOTC4. Implementing and Testing on a Personal Computer

International Nuclear Information System (INIS)

McLaughlin, P.K.

1987-05-01

This document describes the contents of the diskette containing the COMPUTATION FORMAT codes X4TOC4 and PLOTC4 by D.E. Cullen, and example data for use in implementing and testing these codes on a Personal Computer of the type IBM-PC/AT. Upon request the codes are available from the IAEA Nuclear Data Section, free of charge, on a single diskette. (author)

Progress report within the series of GRS-F progress reports on reactor safety, sponsored by the Federal Ministry of Economics and Technology. Period: 1 January - 30 June, 2002

International Nuclear Information System (INIS)

2002-01-01

Within its competence for energy research, the Bundesministerium fuer Wirtschaft und Technologie (BMWi) (Federal Ministry of Economics and Technology) sponsors investigations into the safety of nuclear power plants. The objective of these investigations is to provide fundamental knowledge, procedures and methods to contribute to realistic safety assessments of nuclear installations, to the further development of safety technology and to make use of the potential of innovative safety-related approaches. The Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, by order to the BMWi, continuously issues information on the status of such investigations by publishing semi-annual and annual progress reports within the series of GRS-F-Fortschrittsberichte (GRS-F-Progress Reports). Each progress report represents a compilation of individual reports about the objectives, work performed, results achieved, next steps of the work etc. The individual reports are prepared in a standard form by the research organisations themselves as documentation of their progress in work and are published by the Research Managment Division of GRS within the framework of general information on the progress in reactor safety research. The compilation of the reports is classified according to general topics related to reactor safety research. Further, use is made of the classification system 'Joint Safety Research Index' of the CEC (Commission of the European Communities). The reports are arranged in sequence of their project numbers. It has to be pointed out that the authors of the reports are responsible for the contents of this compilation. The BMWi does not take any responsibility for the correctness, exactness and completeness of the information nor for the observance of private claims of third parties. (orig.) [de

Progress report within the series of GRS-F progress reports on reactor safety, sponsored by the Federal Ministry of Economics and Technology. Period: 1 July - 31 December 2001

International Nuclear Information System (INIS)

2002-01-01

Within its competence for energy research, the Bundesministerium fuer Wirtschaft und Technologie (BMWi) (Federal Ministry of Economics and Technology) sponsors investigations into the safety of nuclear power plants. The objective of these investigations is to provide fundamental knowledge, procedures and methods to contribute to realistic safety assessments of nuclear installations, to the further development of safety technology and to make use of the potential of innovative safety-related approaches. The Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, by order to the BMWi, continuously issues information on the status of such investigations by publishing semi-annual and annual progress reports within the series of GRS-F-Fortschrittsberichte (GRS-F-Progress Reports). Each progress report represents a compilation of individual reports about the objectives, work performed, results achieved, next steps of the work etc. The individual reports are prepared in a standard form by the research organisations themselves as documentation of their progress in work and are published by the Research Managment Division of GRS within the framework of general information on the progress in reactor safety research. The compilation of the reports is classified according to general topics related to reactor safety research. Further, use is made of the classification system 'Joint Safety Research Index' of the CEC (Commission of the European Communities). The reports are arranged in sequence of their project numbers. It has to be pointed out that the authors of the reports are responsible for the contents of this compilation. The BMWi does not take any responsibility for the correctness, exactness and completeness of the information nor for the observance of private claims of third parties. (orig.) [de

The Development of Severe Accident Codes at IRSN and Their Application to Support the Safety Assessment of EPR

International Nuclear Information System (INIS)

Caroli, Cataldo; Bleyer, Alexandre; Bentaib, Ahmed; Chatelard, Patrick; Cranga, Michel; Van Dorsselaere, Jean-Pierre

2006-01-01

IRSN uses a two-tier approach for development of codes analysing the course of a hypothetical severe accident (SA) in a Pressurized Water Reactor (PWR): on one hand, the integral code ASTEC, jointly developed by IRSN and GRS, for fast-running and complete analysis of a sequence; on the other hand, detailed codes for best-estimate analysis of some phenomena such as ICARE/CATHARE, MC3D (for steam explosion), CROCO and TONUS. They have been extensively used to support the level 2 Probabilistic Safety Assessment of the 900 MWe PWR and, in general, for the safety analysis of the French PWR. In particular the codes ICARE/CATHARE, CROCO, MEDICIS (module of ASTEC) and TONUS are used to support the safety assessment of the European Pressurized Reactor (EPR). The ICARE/CATHARE code system has been developed for the detailed evaluation of SA consequences in a PWR primary system. It is composed of the coupling of the core degradation IRSN code ICARE2 and of the thermal-hydraulics French code CATHARE2. The CFD code CROCO describes the corium flow in the spreading compartment. Heat transfer to the surrounding atmosphere and to the basemat, leading to the possible formation of an upper and lower crust, basemat ablation and gas sparging through the flow are modelled. CROCO has been validated against a wide experimental basis, including the CORINE, KATS and VULCANO programs. MEDICIS simulates MCCI (Molten-Corium-Concrete-Interaction) using a lumped-parameter approach. Its models are being continuously improved through the interpretation of most MCCI experiments (OECD-CCI, ACE...). The TONUS code has been developed by IRSN in collaboration with CEA for the analysis of the hydrogen risk (both distribution and combustion) in the reactor containment. The analyses carried out to support the EPR safety assessment are based on a CFD formulation. At this purpose a low-Mach number multi-component Navier-Stokes solver is used to analyse the hydrogen distribution. Presence of air, steam and

Immobilization of alpha-amylase from Bacillus circulans GRS 313 on coconut fiber.

Science.gov (United States)

Dey, Gargi; Nagpal, Varima; Banerjee, Rintu

2002-01-01

A simple and inexpensive method for immobilizing alpha-amylase from Bacillus circulans GRS 313 on coconut fiber was developed. The immobilization conditions for highest efficiency were optimized with respect to immobilization pH of 5.5, 30 degrees C, contact time of 4 h, and enzyme to support a ratio of 1:1 containing 0.12 mg/mL of protein. The catalytic properties of the immobilized enzyme were compared with that of the free enzyme. The activity of amylase adsorbed on coconut fiber was 38.7 U/g of fiber at its optimum pH of 5.7 and 48 degrees C, compared with the maximum activity of 40.2 U/mL of free enzyme at the optimum pH of 4.9 and 48 degrees C. The reutilization capacity of the immobilized enzyme was up to three cycles.

Extensions of the MCNP5 and TRIPOLI4 Monte Carlo Codes for Transient Reactor Analysis

Science.gov (United States)

Hoogenboom, J. Eduard; Sjenitzer, Bart L.

2014-06-01

To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branchless collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3x3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3x3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail.

No pain-no gain: The complex art of soft x-ray laser target design and analysis

International Nuclear Information System (INIS)

Rosen, M.D.; London, R.A.; Hagelstein, P.L.

1988-01-01

We review our methodologies in the design and analysis of soft x-ray laser experiments. We convolve large-scale 2-D hydro code output with detailed atomic data bases in a kinetics code with 1-D or 2-D line transfer. The time and space dependent level population data is then post processed further with a beam transport code, including refraction, to predict actual experimental results. While mysteries do remain, we present many examples that show how this complex modeling procedure is crucial in explaining experimental results. 23 refs., 8 figs., 1 tab

Las FARC: Origen y Evolución

Directory of Open Access Journals (Sweden)

Germán Nicolás Pataquiva García

2009-01-01

Full Text Available Las amenazas que sufren los Estados del llamado tercer mundo son similares en esencia, pero de características y evolución diferentes. En Colombia se viven conflictos sociales que afectan no solo este país, el narcotráfico, el terrorismo, el secuestro, la acción de grupos armados que afirman buscar objetivos políticos, crecieron de manera tal que terminaron afectando no solo la seguridad y la paz de los ciudadanos que se encontraban viviendo en apartadas regiones selváticas. El drama colombiano es ya conocido en casi todo el planeta y Las FARC, una de las guerrillas más antiguas del mundo son uno de los actores principales en ese drama. “Las FARC su origen y evolución” basado en documentos propios de esta organización, investigaciones y análisis de sociólogos e historiadores de diferentes nacionalidades, entrevistas a guerrilleros desmovilizados y a campesinos de las regiones afectadas e información difundida por internet presenta una visión de esa organización que permita un análisis un poco alejado de lo que pretende mostrar su propia propaganda y la que presentan sus aliados políticos o ideológicos, tampoco pretende atacarla. Dar una mirada a este grupo desde su origen, comprender su evolución y finalmente analizar sus posibles proyecciones, permiten entre otras cosas encontrar, elementos comunes con los dramas de otras naciones y evaluar posibles formulas de solución que necesariamente vendrán a la mente de quienes se interesen por estos temas.

Code development for eigenvalue total sensitivity analysis and total uncertainty analysis

International Nuclear Information System (INIS)

Wan, Chenghui; Cao, Liangzhi; Wu, Hongchun; Zu, Tiejun; Shen, Wei

2015-01-01

Highlights: • We develop a new code for total sensitivity and uncertainty analysis. • The implicit effects of cross sections can be considered. • The results of our code agree well with TSUNAMI-1D. • Detailed analysis for origins of implicit effects is performed. - Abstract: The uncertainties of multigroup cross sections notably impact eigenvalue of neutron-transport equation. We report on a total sensitivity analysis and total uncertainty analysis code named UNICORN that has been developed by applying the direct numerical perturbation method and statistical sampling method. In order to consider the contributions of various basic cross sections and the implicit effects which are indirect results of multigroup cross sections through resonance self-shielding calculation, an improved multigroup cross-section perturbation model is developed. The DRAGON 4.0 code, with application of WIMSD-4 format library, is used by UNICORN to carry out the resonance self-shielding and neutron-transport calculations. In addition, the bootstrap technique has been applied to the statistical sampling method in UNICORN to obtain much steadier and more reliable uncertainty results. The UNICORN code has been verified against TSUNAMI-1D by analyzing the case of TMI-1 pin-cell. The numerical results show that the total uncertainty of eigenvalue caused by cross sections can reach up to be about 0.72%. Therefore the contributions of the basic cross sections and their implicit effects are not negligible

Antioxidantes de origen vegetal. Efecto sobre el envejecimiento

OpenAIRE

Descalzo Perea, Virginia

2016-01-01

Este trabajo consiste en una revisión bibliográfica sobre diferentes tipos de antioxidantes de origen vegetal que podemos encontrar en los alimentos, así como sus características y efectos principales. Los radicales libres más frecuentes son los derivados del oxígeno y se producen como consecuencia de las reacciones de oxidación, que constituyen una parte esencial del metabolismo. Cuando la producción de radicales libres excede a la defensa antioxidante se da un desequilibrio y, como consecue...

Test and validation of CFD codes for the simulation of accident-typical phenomena in the reactor containment

International Nuclear Information System (INIS)

Schramm, Berthold; Stewering, Joern; Sonnenkalb, Martin

2014-03-01

CFD (Computational Fluid Dynamic) simulation techniques have a growing relevance for the simulation and assessment of accidents in nuclear reactor containments. Some fluid dynamic problems like the calculation of the flow resistances in a complex geometry, turbulence calculations or the calculation of deflagrations could only be solved exactly for very simple cases. These fluid dynamic problems could not be represented by lumped parameter models and must be approximated numerically. Therefore CFD techniques are discussed by a growing international community in conferences like the CFD4NRS-conference. Also the number of articles with a CFD topic is increasing in professional journals like Nuclear Engineering and Design. CFD tools like GASFLOW or GOTHIC are already in use in European nuclear site licensing processes for future nuclear power plants like EPR or AP1000 and the results of these CFD tools are accepted by the authorities. For these reasons it seems to be necessary to build up national competences in the field of CFD techniques and it is important to validate and assess the existing CFD tools. GRS continues the work for the validation and assessment of CFD codes for the simulation of accident scenarios in a nuclear reactor containment within the framework of the BMWi sponsored project RS1500. The focus of this report is on the following topics: - Further validation of condensation models from GRS, FZJ and ANSYS and development of a new condensate model. - Validation of a new turbulence model which was developed by the University of Stuttgart in cooperation with ANSYS. - The formation and dissolution of light gas stratifications are analyzed by large scale experiments. These experiments were simulated by GRS. - The AREVA correlations for hydrogen recombiners (PARs) could be improved by GRS after the analysis of experimental data. Relevant experiments were simulated with this improved recombiner correlation. - Analyses on the simulation of H_2 deflagration

Development of the integrated system reliability analysis code MODULE

International Nuclear Information System (INIS)

Han, S.H.; Yoo, K.J.; Kim, T.W.

1987-01-01

The major components in a system reliability analysis are the determination of cut sets, importance measure, and uncertainty analysis. Various computer codes have been used for these purposes. For example, SETS and FTAP are used to determine cut sets; Importance for importance calculations; and Sample, CONINT, and MOCUP for uncertainty analysis. There have been problems when the codes run each other and the input and output are not linked, which could result in errors when preparing input for each code. The code MODULE was developed to carry out the above calculations simultaneously without linking input and outputs to other codes. MODULE can also prepare input for SETS for the case of a large fault tree that cannot be handled by MODULE. The flow diagram of the MODULE code is shown. To verify the MODULE code, two examples are selected and the results and computation times are compared with those of SETS, FTAP, CONINT, and MOCUP on both Cyber 170-875 and IBM PC/AT. Two examples are fault trees of the auxiliary feedwater system (AFWS) of Korea Nuclear Units (KNU)-1 and -2, which have 54 gates and 115 events, 39 gates and 92 events, respectively. The MODULE code has the advantage that it can calculate the cut sets, importances, and uncertainties in a single run with little increase in computing time over other codes and that it can be used in personal computers

TURTLE 24.0 diffusion depletion code

International Nuclear Information System (INIS)

Altomare, S.; Barry, R.F.

1971-09-01

TURTLE is a two-group, two-dimensional (x-y, x-z, r-z) neutron diffusion code featuring a direct treatment of the nonlinear effects of xenon, enthalpy, and Doppler. Fuel depletion is allowed. TURTLE was written for the study of azimuthal xenon oscillations, but the code is useful for general analysis. The input is simple, fuel management is handled directly, and a boron criticality search is allowed. Ten thousand space points are allowed (over 20,000 with diagonal symmetry). TURTLE is written in FORTRAN IV and is tailored for the present CDC-6600. The program is core-contained. Provision is made to save data on tape for future reference. (auth)

Electronic manual of the nuclear characteristics analysis code-set for FBR

International Nuclear Information System (INIS)

Makino, Tohru

2001-03-01

Reactor Physics Gr., System Engineering Technology Division, O-arai Engineering Center has consolidated the nuclear design database to improve analytical methods and prediction accuracy for large fast breeder cores such as demonstration or commercial FBRs from the previous research. The up-to-date information about usage of the nuclear characteristics analysis code-set was compiled as a part of the improvement of basic design data base for FBR core. The outlines of the electronic manual are as follows; (1) The electronic manual includes explanations of following codes: JOINT : Code Interface Program. SLAROM, CASUP : Effective Cross Section Calculation Code. CITATION-FBR : Diffusion Analysis Code. PERKY : Perturbative Diffusion Analysis Code. SNPERT, SNPERT-3D : Perturbative Transport Analysis Code. SAGEP, SAGEP-3D : Sensitivity Coefficient Calculation Code. NSHEX : Transport Analysis Code using Nodal Method. ABLE : Cross Section Adjustment Calculation Code. ACCEPT : Predicting Accuracy Evaluation Code. (2) The electronic manual is described using HTML file format and PDF file for easy maintenance, updating and for easy referring through JNC Intranet. User can refer manual pages by usual Web browser software without any special setup. (3) Many of manual pages include link-tags to jump to related pages. String search is available in both HTML and PDF documents. (4) User can download source code, sample input data and shell script files to carry out each analysis from download page of each code (JNC inside only). (5) Usage of the electronic manual and maintenance/updating process are described in this report and it makes possible to enroll new code or new information in the electronic manual. Since the information has been taken into account about modifications and error fixings, added to each code after the last consolidation in 1994, the electronic manual would cover most recent status of the nuclear characteristics analysis code-set. One of other advantages of use

PACC information management code for common cause failures analysis

International Nuclear Information System (INIS)

Ortega Prieto, P.; Garcia Gay, J.; Mira McWilliams, J.

1987-01-01

The purpose of this paper is to present the PACC code, which, through an adequate data management, makes the task of computerized common-mode failure analysis easier. PACC processes and generates information in order to carry out the corresponding qualitative analysis, by means of the boolean technique of transformation of variables, and the quantitative analysis either using one of several parametric methods or a direct data-base. As far as the qualitative analysis is concerned, the code creates several functional forms for the transformation equations according to the user's choice. These equations are subsequently processed by boolean manipulation codes, such as SETS. The quantitative calculations of the code can be carried out in two different ways: either starting from a common cause data-base, or through parametric methods, such as the Binomial Failure Rate Method, the Basic Parameters Method or the Multiple Greek Letter Method, among others. (orig.)

Fault tree analysis. Implementation of the WAM-codes

International Nuclear Information System (INIS)

Bento, J.P.; Poern, K.

1979-07-01

The report describes work going on at Studsvik at the implementation of the WAM code package for fault tree analysis. These codes originally developed under EPRI contract by Sciences Applications Inc, allow, in contrast with other fault tree codes, all Boolean operations, thus allowing modeling of ''NOT'' conditions and dependent components. To concretize the implementation of these codes, the auxiliary feed-water system of the Swedish BWR Oskarshamn 2 was chosen for the reliability analysis. For this system, both the mean unavailability and the probability density function of the top event - undesired event - of the system fault tree were calculated, the latter using a Monte-Carlo simulation technique. The present study is the first part of a work performed under contract with the Swedish Nuclear Power Inspectorate. (author)

Developments of fuel performance analysis codes in KEPCO NF

International Nuclear Information System (INIS)

Han, H. T.; Choi, J. M.; Jung, C. D.; Yoo, J. S.

2012-01-01

The KEPCO NF has developed fuel performance analysis and design code named as ROPER, and utility codes of XGCOL and XDNB in order to perform fuel rod design evaluation for Korean nuclear power plants. The ROPER code intends to cover full range of fuel performance evaluation. The XGCOL code is for the clad flattening evaluation and the XDNB code is for the extensive DNB propagation evaluation. In addition to these, the KEPCO NF is now in the developing stage for 3-dimensional fuel performance analysis code, named as OPER3D, using 3-dimensional FEM for the nest generation within the joint project CANDU ENERGY in order to analyze PCMI behavior and fuel performance under load following operation. Of these, the ROPER code is now in the stage of licensing activities by Korean regulatory body and the other two are almost in the final developing stage. After finishing the developing, licensing activities are to be performed. These activities are intending to acquire competitiveness, originality, vendor-free ownership of fuel performance codes in the KEPCO NF

A compendium of computer codes in fault tree analysis

International Nuclear Information System (INIS)

Lydell, B.

1981-03-01

In the past ten years principles and methods for a unified system reliability and safety analysis have been developed. Fault tree techniques serve as a central feature of unified system analysis, and there exists a specific discipline within system reliability concerned with the theoretical aspects of fault tree evaluation. Ever since the fault tree concept was established, computer codes have been developed for qualitative and quantitative analyses. In particular the presentation of the kinetic tree theory and the PREP-KITT code package has influenced the present use of fault trees and the development of new computer codes. This report is a compilation of some of the better known fault tree codes in use in system reliability. Numerous codes are available and new codes are continuously being developed. The report is designed to address the specific characteristics of each code listed. A review of the theoretical aspects of fault tree evaluation is presented in an introductory chapter, the purpose of which is to give a framework for the validity of the different codes. (Auth.)

First insights from common GRS and IRSN evaluations of operational experiences for the European Clearinghouse

International Nuclear Information System (INIS)

Maqua, M.; Bertrand, R.

2012-01-01

The European Clearinghouse on Operating Experience Feedback (OEF) for NPPs was established in 2008 by several European Nuclear Safety Regulators. The technical work on common operational experience evaluation was launched at the Joint Research Center of the European Union. Main goals of the European Clearinghouse include promotion of collaboration on OEF, dissemination of the lessons learnt from NPP operational experience and promoting advanced event assessment approaches and methods. The extended knowledge of both IRSN and GRS in the evaluation of the national operating experiences in France and Germany, respectively, is used by the European Clearinghouse with the objective of receiving detailed insights on safety related topics. IRSN and GRS develop common reports on safety issues upon request of the Clearinghouse that involve a detailed review of their national event databases, and that are combined afterwards in a European Clearinghouse report including also the review carried out by the JRC on the basis of the Incident Report System (IRS) database and the US NRC Licensee Event Reports (LERs) database. The aim is to perform in-depth analyses (causes, root causes, contributing factors, actual and potential consequences, and lessons learnt) of specific OEF topics. 3 topics have been already analyzed: -) external events (Serious accidents caused by external hazards were not found in the German and French operating experience), -) ageing of components (it can be concluded that ageing management in French and German NPP is basically effective), and -) component supply (the main cause for events related to the supply of components was insufficiency during manufacturing or testing). This paper is followed by the slides of the presentation

Development of Nuclear Plant Specific Analysis Simulators with ATLAS

International Nuclear Information System (INIS)

Jakubowski, Z.; Draeger, P.; Horche, W.; Pointner, W.

2006-01-01

The simulation software ATLAS, based on the best-estimate code ATHLET, has been developed by the GRS for a range of applications in the field of nuclear plant safety analysis. Through application of versatile simulation tools and graphical interfaces the user should be able to analyse with ATLAS all essential accident scenarios. Detailed analysis simulators for several German and Russian NPPs are being constructed on the basis of ATLAS. An overview of the ATLAS is presented in the paper, describing its configuration, functions performed by main components and relationships among them. A significant part of any power plant simulator are the balance-of-plant (BOP) models, not only because all the plant transients and non-LOCA accidents can be initiated by operation of BOP systems, but also because the response of the plant to transients or accidents is strongly influenced by the automatic operation of BOP systems. Modelling aspects of BOP systems are shown in detail, also the interface between the process model and BOP systems. Special emphasis has been put on the BOP model builder based on the methodology developed in the GRS. The BOP modeler called GCSM-Generator is an object oriented tool which runs on the online expert system G2. It is equipped with utilities to edit the BOP models, to verification them and to generate a GCSM code, specific for the ATLAS. The communication system of ATLAS presents graphically the results of the simulation and allows interactively influencing the execution of the simulation process (malfunctions, manual control). Displays for communications with simulated processes and presentation of calculations results are also presented. In the framework of the verification of simulation models different tools are used e.g. the PC-codes MATHCAD for the calculation and documentation, ATLET-Input-Graphic for control of geometry data and the expert system G2 for development of BOP-Models. The validation procedure and selected analyses results